ML20065D050

Proposed Tech Specs Re Cycle 7 Reload Rept

ML20065D050
Person / Time
Site:	Catawba
Issue date:	03/24/1994
From:	DUKE POWER CO.
To:
Shared Package
ML20065D046	List: ML20065D043 ML20065D047 ML20065D050
References
NUDOCS 9404060191
Download: ML20065D050 (16)

Text

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Table 8-1 Technical Specification Changes Specification Descriotion of Chance 4.7.13.3 Increase spent fuel storage pool minimum boron concentration from 2000 ppm to 2175 ppm.

3.9.1 Increase RCS and refueling water canal minimum boron concentration from 2000 ppm to 2175 ppm.

6.9.1.9 Add Duke Topicals used to determine core operating limits, i

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8-2 9404060191 940324 ,

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l PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 92 days by verifying that the individual cell voltage is greater than or equal to 1.36 volts on float charge, and

c. At least once per 18 months by verifying that:

1) The batteries, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration, and

2) The battery-to-battery and terminal connections are clean, tight, and free of corrosion.

4.7.13.3 The Standby Makeup Pump water supply shall be demonstrated OPERABLE by:

a. Verifying at least once per 7 days:

1) That the requirements of Specification 3.9.10 are met and the l

boron concentration in the storage pool is greater than or equal to 200& ppm, or g l 2) That a contained borated water volume of at east 112,320 gallons

,g with minimum boron concentration of -2;000 ppm is available and capable of being aligned to the Standby Makeup Pump.

b. Verifying at least once per 92 days that the Standby Makeup Pump develops a flow of greater than or equal to 26 gpm at a pressure greater than or equal to 2488 psig.

4.7.13.4 The Standby Shutdown System 250/125-Volt Battery Bank and its

associated charger shall be demonstrated OPERABLE

a. At least once per 31 days by verifying:

l 1) That the electrolyte level of each battery is above the plates, I and

2) The total battery terminal voltage is greater than or equal to 258/129 volts on float charge.

b. At least once per 92 days by verifying that the specific gravity is appropriate for continued service of the battery, and i c. At least once per 18 months by verifying that:

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1) The batteries, cell plates, and battery racks show no visual ,

indications of physical damage or abnormal deterioration, and )

2) The battery-to-battery and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material. 1 CATAWBA - UNITS 1 &~2 3/4 7-41 Amendment No. 97 (Unit 1) I Amendment No. 91 (Unit 2)

3/4.9 REFUELING OPERATIONS

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/4.9.1 BORON CONCENTRATION g LIM TING CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled portions of he Reactor Coolant System a d the refueling canal shall be maintained uni rm and sufficient to ensure.th t the more restrictive of the following re tivity conditions is met either:

l a. A K ,,, f 0.95 or less, or I b. A boron ncentration of greater tha or equal to 2175 ppm. l APPLICABILITY: MODE . * (Unit 1) g l ACTION:

With the requirements of t above ecification not satisfied, immediately l

suspend all operations invol ing C E ALTERATIONS or positive reactivity l changes and initiate and cont u boration at greater than or equal to 30 gpm i of a solution containing greate than or equal to 7000 ppm baron or its equi ;'

valent until K.,, is reduced t i s than or equal to 0.95 or the baron con-centration is restored to gr ater han or equal to 2000 ppm, whichever is the more restrictive.

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SURVElllANCE RE0VIREMEM S 4.9.1.1 The more estrictive of the abov o reactivity conditions shall be determined prior o:

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a. Remov' g or unbolting the reactor vesse head, and

b. Wi drawal of any full-length control rod n excess of 3 feet from i s fully inserted position within the reac r vessel.

l- 4.9.1.2 The boron concentration of the Reactor Coolan System and the refuel ng canal shall be determined by chemical analysi at least once per 72 h rs.

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!, *The reactor shall be maintained in MODE 6 whenever fuel is in the actor t.

- vessel with the vessel head closure bolts less than fully tensioned r with the head removed.

CATAWBA - UNIT 1 3/4 A9-1 Amendment No. 112 k i

3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION

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LIMITING CONDITION FOR OPERATION 3.9.1. The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained. uniform and sufficient to ensure.that the more restrictive of the following reactivity conditions is met f either:

a. A K,,, of 0.95 or less, or .

b. A boron concentration of greater than or equal to -2000 ppm.

28 6 , g APPLICABILITY: MODE 6.* (L' nit 2; ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 30 gpa of a solution containing greater than or equal to 7000 ppm boron or its equi- '

valent until K,,, is reduced to less than or equal to 0.95 or the boron con-centration-is restored to greater than or equal to-2000 ppm, whichever is the more restrictive.

ZI '? f SURVEILLANCE REOUIREMENTS 4.' 9.1.1 The more restrictive of the above two reactivity conditions shall be ,

determined prior to:

a. Removing or unbolting the reactor vessel head, and

b. Withdrawal of any full-length control rod in excess of 3 feet from its fully inserted position within the reactor vessel.

4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

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• The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with

( the head removed.

Amendment No.106 i CATAWBA - UNIT E 3/4[9-1 UutTSI}2-l

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ADMINISTRATIVE CONTROt5 CORE OPERATING LIMITS REPORT (Continued) 9.

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DPC-NE-3000P-A, Rev.

Methodology," November 1991.1, " Thermal-Hydraulic Iransient Analysts (Modeling used in the system thermal-hydraulic nalyses) a 10.

DPC-NE-1004A, " Nuclear Design Methodology November 1992. - using CASM MULATE-3P,"

Coefficient.)(Methodology for Specification 3.1.1.3 emperature

- Moderator T ,

The core operating limits shall be determined so that (e.g., fuel thermal-mechanical limits, core thermal h all applicable limits I limits, nuclear limits such as shutdown margin

- ydraulic limits, ECCS analysis limits) of the safety analysis are met. and transient and accident Th'e CORE OPERATING LIMITS REPORT including any mid supplements thereto, shall be provided upon issuance-cycle revisions or to the NRC in accordance with 10 CFR 50 .. 4 each reload cycle,

, for 'I s

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>Pc- NE- 200+ f- A, " b4e Power Cs,,yny McGush ac/

i Ca w , h d u lear s+4us Gre. nermd- Hy rauCc Meth d '

using viru-ol

.b ecmf er i49I (DPC frofvlefary).

(/Hehdoloyy used ih 4e. core. +/,er J:J, de%;ue.

+1, e mol-4plradb awalyses core.

,pr&y Las) p, b PC - N E- 2001 P - A, Rev. t," fad Mahwlcd kl,J Anal sis '

(Mt$odology (or y Mab- S W fuel , Ocfobey 1940 [DfC fropriefaVf .

(Ikkdolog aseol <b +4e. (uel med,wica/ anal se.s clelevmiue 4he. corc.

y whkh ofeVab lm' l}.s CATAWBA - UNITS 1 & 2 6-19a Amendment No.107 Amendment No.101 (Unit 1) k (Unit 2)

- .- . = - - ~ - - . _ . - . ~ . . . - -

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9. STARTUP PHYSICS TESTING -

The standard scope of reload startup physics testing conducted at Duke ,

l Power Westinghouse units is. summarized below (Reference 1). The purpose of the test program is to provide assurance that the reactor core is loaded correctly and can be operated as designed.

i Zero Power Physics Testinc (ZPPT)

, - All Rods Out Critical Boron Concentration (AROCBC) l

- Isothermal Temperature Coefficient (ITC) i l - Control Rod Bank Worth (Reference 15) l Power Escalation Testina (PET)

- Flux Symmetry Check (Low Power, e.g. 30% FP)

L - Core Power Distribution - CPD (Intermediate Power)

[ - CPD (High Power) l

- All Rods Out Critical Boron Concentration'- AROCBC (High Power).

t All aspects of the existing program are acceptable with respect-to implementation of the Duke Power Company licensing analyses and a complete reload batch of Mark-BW fuel assemblies. Therefore, operation with either t a mixed Westinghouse and BWFC core or future cores with all BWFC fuel will l not require any changes to the current Duke startup physics testing program.

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I 10. REFERENCES

1. Catawba Nuclear Station, Final Safety Analysis Report, Docket Nos.

50 - 413/414.

2. RAW-10172-PA, Mark-BW Mechanical Design Report, Babcock & Wilcox, Lynchburg, Virginia, December 19, 1989.

3. DPC-NE-2001-PA, Rev. 1, Fuel Mechanical Reload Analysis Methodology for Mark-BW Fuel, Duke Power Company, October 1990.

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4. BAW-10084-A, Rev. 2, Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, Babcock &'Wilcox, October 1978,

5. SAW-10141-PA, Rev. 1, TACO 2 - Fuel Performance Analysis, Babcock &

Wilcox, June 1983.

6. DPC-NE-1004A, Nuclear Design Methodology Using CASMO-3/ SIMULATE-3P, ,

Duke Power Company, November 1992.

7. DPC-NE-2011-PA, Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors, Duke. Power Company, March 1990.

l l 8. DPC-NE-2004-PA, McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01, Duke Power Company, December 1991.

9. BAW-10159-PA, BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies, Babcock & Wilcox, July 1990,

10. BAW-10173-PA, Mark-BW Reload Safety Analysis for Catawba and McGuire, Babcock & Wilcox, Revision 2, February 20, 1991,

11. DPC-NE-3000P, Duke Power Company, Thermal-Hydraulic Transient Analysis Methodology, Revision 1, May 1989.

12. DPC-NE-3001-PA, Duke Power Company, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology, Revision 1, November 1991.

13. BAW-10174-A, Mark-BW Reload LOCA Analysis for the Catawba and McGuire Units, Babcock & Wilcox, May 1991.

14. BAW-10168-A, B&W Loss-of* Coolant Accident Evaluation Model For Recirculating Steam Generator Plants, Babcock & Wilcox, Lynchburg, Virginia, January 1991, *

15. DPC-NE-1003-A, Revision 1, McGuire Nuclear Station / Catawba Nuclear Station Rod Swap Methodology Report for Startup Physics Testing, December 1986.

16, DPC-NE-3002-A, McGuire Nuclear Station / Catawba Nuclear Station FSAR Chapter 15 System Transient Analysis Methodology, November 1991.

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EMERGENCY CORE COOLING SYSTEMS i LIMITING CONDITION FOR OPERATION (Continued) -(

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l ACTION: (Continued) l l

l- enter ACTION c.1 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of the low boron determination j' or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce Reactor l Coolant System pressure to less than 1000 psig within the fol- l lowing 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. i

3) The volume weighted average boron concentration of the accumula-tors 1800 ppm or less, return the volume weighted average baron concentration of the accumulators to greater than.1800 ppm and enter ACTION c.2 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the low boron determination or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce Reactor Coolant System pressure to less.than 1000 psig within the follow-

.ing 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS l

.A4rl- Each cold leg injection accumulator shall be demonstrated OPERABLE:

l g a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by:

l 1) Verifying, by the absence of. alarms, the contained borated water volume and nitrogen cover pressure in the tanks,.and

2) Verifying f. hat each cold leg injection accumulator isolation , I valve is open.

b. At least once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each solution volume increase of greater than or equ~al to 75 gallons by verifying the boron concentration of the accumulator solution;

c. At least once per 31 days when the Reactor Coolant System pressure .,

is above 2000.psig by verifying that power is removed from the l 1

isolation valve operators on Valves NI54A, NI65B, NI76A, and NI888 )

and that the respective circuit breakers are padlocked; and i

d. At'least once per 18 months by verifying that each cold leg injection accumulator isolation valve opens automatically under each of the following conditions:**- l l

1) When an actual or a simulated Reactor Coolant System pressure i signal exceeds the P-11 (Pressurizer Pressure Block of Safety I

. Injection) Setpoint, and j

2) ,Upon receipt of a Safety Injection test signal.

1 I erforeduftil riftfeferi[gfT[STAf0f

• • This urve llan nee no be foll wing he U it 1 efu.lin l

, CATAWBA - UNITS 1 & 2 3/4 5-2 Amendment No. 0 (Unit 1) l Amendment No. (Unit 2) l _.

EMERGEN'CY-CORE COOLING SYSTEMS ,

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- SURVEILLANCE REQUIREMENTS (Continued) l

' 5.1- Each cold leg injection' accumulator Water level'and pressure channel'shall be demonstrated OPERABLE: -l-

a. At least:once per 31 days by the performance of an ANALOG CHANNEL

.Y g I' 7 OPERATIONAL TEST, and I

b. At least once per 18 months by the performance of a CHANNEL' .

CALIBRATION,' .;

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l CATAWBA - UNITS 1.& 2 3/4 5-3 Amendment No. '(Unit'1)

(' Amendment No, (Unit 2)~

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PLANT SYSTEMS I

SURVEILLANCE REQUIREMENTS 4.7.12 The groundwater level shall be determined at the following frequencies by monitoring the water level and by verifying the absence of alarm in the six groundwater monitor wells as shown in FSAR Figure 2.1.10 14-installed around i the perimeter of the Reactor and Auxiliary Buildings: Ns 2 go I

a. At least once per 7 days when the groundwater level is at or below the top of the adjacent floor slab, and

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the groundwater level is above the top of the adjacent floor slab.

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CATAWBA - UNITS 1 & 2 3/4 7-39 Amendment No. - (Unit 1)

Amendment No. (Unit 2)

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ATTACIIMENT 2 TECIINICAL JUSTIFICATION l

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Pronosed Revision to Technical Snecification 4.7.13.3 This proposed Technical Specification revision redects an increase, from 2000 ppm to 2175 ppm, in the required spent fuel storage pool minimum boron concentration during Modes 1 - 3 operation.

Technical Justification This proposed revision is conservative, and is required only to maintain consistency between the boron concentration of the spent fuel storage pool and the boron concentration of the RWST (Catawba Unit 2 Cycle 7 COLR) during Modes 1 - 3 operation. Since the water source for the spent fuel storage pool is the RWST, the l water inventories are at the same boron concentrations, if Technical Specification 4.7.13.3 were to be used without regard to TS 3.1.2.6 when water is returned to the l RWST from the spent fuel storage pool, the RWST water inventory may be less than l the TS RWST minimum boron concentration limit in TS 3.1.2.6. While changes to the TS 4.7.13.3 boron concentration is not generally reload related, it should be changed to match that of TS 3.1.2.6, to provide consistency and prevent a possible occurrence of the above situation.

Proposed Revision to Technical Snecification 3.9.1 This proposed Technical Specification revision redects an increase, from 2000 ppm to 2175 ppm, in the required RCS and refueling canal minimum boron concentration during Mode 6 operation.

Technical Justification

! This proposed revision is conservative, and is required only to maintain consistency between the boron concentrations of the RCS and refueling canal and the boron concentration of the RWST (Catawba Unit 2 Cycle 7 COLR) during Mode 6 operation. In addition, the unit specific designation for TS 3.9.1 is removed as the concentrations at Units 1 & 2 are now identical. This change has already been approved for Unit I under Amendment 112 issued December 17,1993 for Catawba ,

Unit 1 Cycle 8.

Proposed Revision to Technical Specification 6,9.1.9 l This proposed Technical Specification revision is to include two reload related I

topicals, which have been previously reviewed and approved by the NRC, describing methodology used to determine core operating limits.

Technical Justification l l

l Duke Topical Reports DPC-NE-2004P-A and DPC-NE-200lP-A were inadvertently omitted from the list of topical reports that are used to determine the core operating limits. DPC-NE-2004P-A explains the methodology for performing the core thermal-hydraulic analyses that are used to determine the core operating limits. DPC-NE-200lP-A explains the methodology for performing the fuel rod mechanical analyses that are used to determine the core operating limits.

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l Proposed Revisions of an Administrative Nature to Correct Errors in Nomenclature and to l I

Remove Obsolete Footnotes l

These proposed Technical Specification revisions are considered administrative in nature. They are as follows:

1 (1) There are presently two different surveillance requirements pertaining to the cold I leg injection accumulators which have the same surveillance requirement number (SR )

4.5.1). The first of these is re. numbered to SR 4.5.1.1 and the second is renumbered to SR 4.5.1.2.

(2) Old SR 4.5.1 on page 3/4 5-2 contains a reference to an obsolete footnote. This footnote is being deleted.

(3) SR 4.7.12 references an FSAR figure which continues to utilize an obsolete -

numbering scheme. The correct FSAR figure number is 2-60.

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ATTACIIMENT 3 NO SIGNIFICANT IIAZARDS ANALYSIS AND ENVIRONMENTAL IMPACT STATEMENT

No Significant IIazards Analysis The following analysis, required by 10 CFR 50.91, concludes that the proposed amendment will not involve significant hazards consideration as defined by 10 CFR 50.92.

10 CFR 50.92 states that a proposed amendment involves no significant hazards consideration if opemtion in accordance with the proposed amendment would not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2) Create the possibility of a new or different kind of accident from any previously evaluated; or

3) Involve a significant reduction in the margin of safety.

INCREASE IN BORON CONCENTRATION LIMIT FOR TIIE SPENT FUEL STORAGE POOL (STANDBY AIAKEUP PUAIP WATER SUPPLY)

The required spent fuel storage pool minimum boron concentration was increased from 2000 ppm to 2175 ppm during Modes 1-3.

The proposed revision is conservative, and is required only to maintain consistency between the boron concentration of the spent fuel storage pool and the boron concentration of the RWST during Modes 1-3 operation. Therefore, there will be no adverse impact upon the probability or consequences of any previously analyzed accident.

Likewise, the proposed change will not create the possibility of a new or different kind of  ;

accident, since no new failure modes are identified. l Finally, no negative impact upon any safety margin is created since the proposed change is conservative.

INCREASE IN BORON CONCENTRATION LIMITS FOR TIIE RCS AND REFUELING CANAL IN AIODE 6 The increase in the required RCS and refueling canal minimum boron concentration was added only to maintain consistency between the boron concentration of the RCS and refueling canal and the RWST in Mode 6.

The change in boron concentration limits for the RCS and refueling canal will not increase the probability of an accident since no accident initiators are involved with this change. Since the

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change is conservative, the consequences of an accident previously evaluated will not be increased. The increase in the boron concentration limit for the RCS and refueling canal in Mode 6 adds further margin to the initial conditions assumed for the boron dilution accident in the safety analysis. Therefore, the consequences of the boron dilution accident previously evaluated will not be increased.

The Imssibility of a new or different kind of accident from any previously evaluated will not be created since this change is bounded by previously evaluated accidents and does not introduce any new failure modes.

This change does not involve a significant reduction in the margin of safety since the analyses perfonned demonstrate that the limits imposed meet all accident analysis and design basis requirements.

ADDITION OF TWO RELOAD RELATED TOPICAL REPORTS This change is administrative in nature and adds two previously approved topical reports to the list of methodologies used to detennine core operating limits. The change will have no impact upon either the probability or consequences of a previously analyzed accident. The methodologies described in the topical reports have been previously reviewed and approved by the NRC. Also, no new accident possibilities are created, .since this is an administrative change.

Finally, no impact upon any safety margin is created, since the change is administmtive in nature and the described topical repons have received full NRC approval.

CORRECTION OF ERRORS IN NOAIENCLATURE AND REAIOVAL OF OBSOLETE FOOTNOTES These changes are also administrative in nature and are intended to correct miscellaneous errors and obsolete references. As such, the changes will have no impact upon either the probability or consequences of any previously analyzed accidents, will not create the possibility of any new accident scenarios, and will not impact any safety margins.

Environmental Impact Statement The proposed amendment has been reviewed against the criteria of 10 CFR 51.22 for environmental considerations. As described above, the proposed amendment does not involve l any significant hazards consideration, nor a significant increase or change in the types or  !

amounts of effluents that may be released offsite, nor a significant increase in the individual or cumulative occupational radiation exposures. Therefore, the proposed amendment meets the criteria given in 10 CFR 51.22(c)(9) for categorical exclusion from the requirement for an Environmental Impact Statement.

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