ML20092A593
| ML20092A593 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 08/31/1995 |
| From: | Berkow H NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20092A596 | List: |
| References | |
| NUDOCS 9509110102 | |
| Download: ML20092A593 (16) | |
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UNITED STATES ej j
NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 201 6 0001
.....,o DUKE POWER COMPANY NORTH CAROLINA ELECTRIC MEMBERSHIP CORPORATION SALUDA RIVER ELECTRIC COOPERATIVE. INC.
DOCKET NO. 50-413 i
CATAWBA NUCLEAR STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE i
Amendment No. 134 License No. NPF-35 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility) Facility Operating License No. NPF-35 filed by the Duke Power Company, acting for itself, North Carolina Electric Membership Corporation, and Saluda River Electric Cooperative, Inc. (licensees), dated September 19, 1994, as supplemented April 26 and June 19, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations as set forth in 10 CFR Chapter I B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; d
I C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
9509110102 950831 PDR ADOCK 05000413 P
. 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No.
NPF-35 is hereby amended to read as follows:
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 134
, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Duke Power Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION i
erbertN.Berkow,Diehtr Project Directorate II-2 l
Division of Reactor Projects - I/II i
Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes l
Date of Issuance:
August 31, 1995 1
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uagk UNITED STATES y-g g
NUCLEAR REGULATORY COMMISSION j
WASHINGTON. D.C. 2006H001
\\...../
DUKE POWER COMPANY NORTH CAROLINA MUNICIPAL POWER AGENCY NO. 1 PIEDMONT MUNICIPAL POMER AGENCY DOCKET NO. 50-414 CATAWBA NUCLEAR STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 128 License No. NPF-52 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility) Facility Operating License No. NPF-52 filed i
by the Duke Power Company, acting for itself, North Carolina Municipal Power Agency No. I and Piedmont Municipal Power Agency (licensees), dated September 19, 1994, as supplemented April 26 and June 19, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations as set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
. 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No.
NPF-52 is hereby amended to read as follows:
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.
128, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license.
Duke Power Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.
FOR HE NUCLEAR REGULATORY COPMISSION
/dh 47
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erbert N. Berkow, Director Project Directorate 11-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
1 Technical Specification Changes Date of Issuance:
August 31, 1995
t ATTACHMENT TO LICENSE AMENDMENT NO.134 FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 8HQ TO LICENSE AMENDMENT NO.128 FACILITY OPERATING LICENSE NO. NPF-52 DOCKET NO. 50-414 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.
Remove Paaes Insert Paaes XI XI 3/4 9-17 3/4 9-18 3/4 9-19 3/4 9-20 3/4 9-21 B 3/4 9-4 B 3/4 9-5 5-7 5-7 6-19 6-19 6-19a 6-19a
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i LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION EA(if 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level...................................
3/4 9-10 Low Water Level....................................
3/4 9-11 3/4.9.9 WATER LEVEL - REACTOR VESSEL.......................
3/4 9-12 3/4.9.10 WATER LEVEL - STORAGE POOL........................ 3/4 9-13 3/4.9.11 FUEL HANDLING VENTILATION EXHAUST SYSTEM..........
3/4 9-14 3/4.9.12 SPENT FUEL POOL BORON CONCENTRATION................ 3/4 9-17 3/4.9.13 SPENT FUEL ASSEMBLY ST0 RAGE........................
3/4 9-18 23 Table 3.9-1 MINIMUM QUALIFYING BURNUP Vs. INITIAL ENRICHMENT FOR UNRESTRICTED ST0 RAGE.........................
3/4 9-19 Table 3.9-2 MINIMUM QUALIFYING BURNUP Vs. INITIAL ENRICHMENT FOR FILLER ASSEMBLIES............................
3/4 9-20 Figure 3.9-1 REQUIRED 3 OUT OF 4 LOADING PATTERN FOR RESTRICTED ST0 RAGE............................... 3/4 9-21 4
3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN....................................
3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION l
LIMITS...........................................
3/4 10-2 3/4.10.3 PHYSICS 1ESTS......................................
3/4 10-3 3/4.10.4 REACTOR COOLANT L00PS.............................. 3/4 10-4 a
3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN..............
3/4 10-5 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Liquid Holdup Tanks................................
3/4 11-1 i
3/4.11.2 GASE0US EFFLUENTS Explosive Gas Mixture..............................
3/4 11-2 Gas Storage Tanks..................................
3/4 11-3 CATAWBA - UNITS 1 & 2 XI Amendment No. jjg (Unit 1)
Amendment No.
(Unit 2)
REFUELING OPERATIONS 3/4.9.12 SPENT FUEL P0OL BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.12 The boron concentration in the spent fuel pool shall be within the limit specified in the COLR.
APPLICABILITY:
During storage of fuel in the spent fuel pool.
ACTION:
a.
Immediately suspend movement of fuel assemblies in the spent fuel pool and initiate action to restore the spent fuel pool boron concentration to within its limit.
b.
The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.9.12 Verify at least once per 7 days that the spent fuel pool boron l
concentration is within its limit.
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CATAWBA - UNITS 1 & 2 3/4 9-17 Amendment No.134 (Unit 1) l Amendment No.128 (Unit 2) l l.
REFUELING OPERATIONS 3/4.9.I3 SPENT FUEL ASSEMBLY STORAGE LIMITING CONDITION FOR OPERATION 3.9.13 New or irradiated fuel may be stored in the Spent Fuel Pool in accordance with these limits:
a.
Unrestricted storage of fuel meeting the criteria of Table 3.9-1; or b.
Restricted storage in accordance with Figure 3.9-1, of fuel which does n21 meet the criteria of Table 3.9-1.
APPLICABILITY:
During storage of fuel in the spent fuel pool.
ACTION:
a.
Immediately initiate action to move the noncomplying fuel assembly to the correct location.
b.
The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.9.13 Prior to storing a fuel assembly in the spent fuel storage pool, verify by administrative means the initial enrichment and burnup of the fuel assembly are in accordance with Specification 3.9.13.
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i CATAWBA - UNITS 1 & 2 3/4 9-18 Amendment No.134 (Unit 1)
Amendment No.128 (Unit 2)
l Table 3.9-1 Minimum Oualifyina Burnup Versus Initial Enrichment for Unrestricted Storace Initial Nominal Enrichment Assembly Burnup (Weight % U-235)
(GWD/MTU) 4.05 (or less) 0 4.50 2.73 5.00 5.67 10 8
s@
o 8
X UNRESTRICTED l4 STORAGE cn 2
$2 RESTRICTED 4
STORAGE 0
4.00 4.25 4.50 4.75 5.00 1
initial Nominal Enrichment (Weight % U-235)
Fuel which differs from those. designs used to determine the requirements of Table 3.9-1 may be qualified for Unrestricted storage by means of an analysis using NRC approved methodology to assure that k,,, is less than or equal to 0.95.
Likewise, previously unanalyzed fuel up to 5.0 weight % U-235 may be qualified for Restricted storage by means of an analysis using NRC approved methodology to assure that k,,, is less than or equal to 0.95.
CATAWBA - UNITS 1 AND 2 3/4 9-19 Amendment No.
134 Amendment No.
128
Table 3.9-2 Minimum Oualifyina Burnuo Versus Initial Enrichment for Filler Assemblies Initial Nominal Enrichment Assembly Burnup 4
(Weicht% U-235)
(GWD/MTV) 1.90 (or less) 0 1
2.00 16.83 I
2.50 26.05 3.00 35.11 3.50 43.48 4.00 51.99 4.48 60.00 I
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70 UNACCEPTABLE For Use As Filler Assembly g 60 a
a N 50 ACCEPTABLE 9
40 For Use As FiHer Assembly b 30
( 20 -
UNACCEPTABLE 8
For Use As Filler Assembly 4 10 0
i 1.50
'2.00 2.50 3.00 3.50
.4.00 4.50 j
Initial Nominal Enrichment (Weight % U-235) l Fuel which differs from those designs used to determine the requirements of Table 3.9-2 may be qualified for use as a Filler Assembly by means of an analysis using NRC approved methodology to assure that k,,, is less than or equal to 0.95.
CATAWBA - UNITS 1 AND 2 3/4 9-20 Amendment No. 134 Amendment No. 128
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Fiqure 3.9-1 I
Reauired 3 out of 4 Loadina Pattern for Restricted Storace l
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RESTRICTED RESTRICTED RESTRICTED RESTRICTED FUEL FUEL FUEL FUEL I
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1 RESTRICTED FLLER
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-RESTRICTED FK.LER FUEL LOCATION FUEL LOCATION nu;;uu:nua; tu;anu: anus MM"FH994Mi
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Restricted fuel:
Fuel defined for Restricted Storage in Table 3.9-1.
I (Fuel defined for Unrestricted Storage in Table 3.9-1, i
or non-fuel components, or an empty location may be placed in restricted fuel locations as needed)
Filler Location:
Either fuel which meets the minimum burnup i
requirements of Table 3.9-2, or an empty cell.
4 Boundary Condition:
Any row bounded by an Unrestricted Storage Area shall contain a combination of restricted fuel assemblies and filler locations arranged such that no restricted i
fuel assemblies are adjacent to each other.
Example:
In the figure above, row 1 or column 1 can not be adjacent to an Unrestricted Storage Area, but row 4 or column 4 can be.
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CATAWBA - UNITS 1 AND 2 3/4 9-21 Amendment No. 134 Amendment No. 128
l REFUELING OPERATIONS j
BASES 1
3/4.9.12 AND 3/4.9.13 SPENT FUEL POOL BORON CONCENTRATION AND SPENT FUEL J
ASSEMBLY STORAGE i
The requirements for spent fuel pool boron concentration specified in Specification 3.9.12 ensure that a minimum boron concentration is maintained
]
in the pool. The requirements for spent fuel assembly storage specified in Specification 3.9.13 ensure that the pool remains subcritical.
The water in i
the spent fuel storage pool normally contains soluble boron, which results in large subcriticality margins under actual operating conditions. However, the l
NRC guidelines based upon the accident condition in which all soluble poison l
l is assumed to have been lost, specify that the limiting k,, of 0.95 be 1
evaluated in the absence of soluble boron. Hencethedeslgnofthespentfuel i
storage racks is based on the use of unborated water, which maintains the spent fuel pool in a subcritical condition during normal operation with the pool fully loaded.
The double contingency principle discussed in ANSI l
N-16.1-1975 and the April 1978 NRC letter (Ref. 4) allows credit for soluble boron under other abnormal or accident conditions, since only a single
'1 accident need be considered at one time.
For example, the most severe j
accident scenario is associated with the accidental misloading of a fuel l
assembly. This could increase the reactivity of the spent fuel pool. To mitigate this postulated criticality related accident, boron is dissolved in the pool water.
Tables 3.9-1 and 3.9-2 allow for specific criticality analyses for fuel which does not meet the requirements for storage defined in these tables.
l These analyses would require using NRC approved methodology to ensure that i
k s0.95 with a 95 percent probability at a 95 percent confidence level as d,,s,cribed in Section 9.1 of the FSAR. This option is intended to be used for e
fuel not included in previous criticality analyses.
Fuel storage is still 4
limited to the configurations defined in TS 3.9.13.
The use of specific analyses for qualification of previously unanalyzed fuel includes, ut is not l
limited to, fuel assembly designs not previously analyzed which may ue as a l
result of new fuel designs or fuel shipments from another facility.
Currently j
analyzed fuel designs include the Babcock and Wilcox MkBW design, and the j
Westinghouse Standard and Optimized fuel designs. Another more likely, and expected use of this specific analysis provision would be to analyze movement j
and storage of individual fuel pins as a result of reconstitution activities.
assumed the most conservative conditions,,l,.e.0.95, the criticality analysis In verifying the design criteria of k 5
fuel of the maximum permissible reactivity for a given configuration.
Since the data presented in Specification 3.9.13.a and 3.9,13.b represents the maximum reactivity i
requirements for acceptable storage, substitutions of less reactive components 1
would also meet the k s0.95 criteria. Hence an empty cell, or a non-fuel component may be subsITtuted for any designated fuel assembly location.
These, or other substitutions which will decrease the reactivity of a particular storage cell will only decrease the overall reactivity of the spent l
fuel storage pool.
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CATAWBA - UNITS 1 & 2 B 3/4 9-4 Amendment No.134 (Unit 1)
Amendment No.128(Unit 2) i
l REFUELING OPERATIONS BASES 3/4.9.12 AND 3/4.9.13 SPENT FUEL POOL BORON CONCENTRATION AND SPENT FUEL l
ASSEMBLY STORAGE (Continued)
If both restricted and unrestricted storage is used, an additional criteria has been imposed to ensure that the boundary row between these two configurations would not locally increase the reactivity above the required limit.
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The action statement applicable to fuel storage in the spent fuel pool requires that action must be taken to preclude the occurrence of an accident t
or to mitigate the consequences of an accident in progress. This is most 1
efficiently achieved by immediately suspending the movement of fuel assemblies.
Prior to the resumption of fuel movement, the requirements of the j
LCOs must be met. This requires restoring the soluble boron concentration and the correct fuel storage configuration to within the corresponding limits.
This does not preclude movement of a fuel assembly to a safe position.
The surveillance requirements ensure that the requirements of the two LC0's are satisfied, namely boron concentration and fuel placement.
The boron concentration in the spent fuel pool is verified to be greater than or equal j
to the minimum limit. The fuel assemblies are verified to met the subcriticality requirement by meeting either the initial enrichment and burnup 3
requirements of Table 3.9-1 and 3.9-2, or by using NRC approved methodology to s0.95.
By meeting either of these requirements, the analyzed ensurethatk,7ullyaddressed.
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accidents are l
The fuel storage requirements and restrictions discussed here and applied
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in specification 3.9.13 are based on a maximum allowable fuel enrichment of i
5.0 weight % U-235. The enrichments listed in Tables 3.9-1 and 3.9-2 are nominal enrichments and include uncertainties to account for the tolerance on the as built enrichment. Hence the as built enrichments may exceed the i
enrichments listed in the tables by up to 0.05 weight % U-235. Qualifying burnups for enrichments not listed in the tables may be linearly interpolated between the enrichments provided. This is because the reactivity of an j
assembly varies linearly for small ranges of enrichment.
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REFERENCES j
1.
" Regulatory Guide 1.13: Spent Fuel Storage Facility Design Basis", U.S.
Nuclear Regulatory Commission, Office of Standards Development, Revision
)
1, December 1976.
2.
" Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations", American Nuclear Society, ANSI N210-1976/
i ANS-57.2, April 1976.
I 3.
FSAR, Section 9.1.
l 4.
Double contingency principle of ANSI N16.1-1975, as specified in the j
April 14, 1978 NRC letter (Section 1.2) and implied in the proposed j
revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).
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CATAWBA - UNITS 1 & 2 B 3/4 9-5 Amendment No.134 (Unit 1)
Amendment No.128 (Unit 2)
EEllsM.((ATURES 5.6 FUEL STORAGE CRITICALITY 5.6.1 a.
The spent fuel storage racks are designed and shall be maintained with:
1) k so.95 if fully flooded with unborated water as described in S,e,c, tion 9.1 of the FSAR; and 1
2)
A nominal 13.5" center to center distance between fuel assemblies placed in the spent fuel storage racks.
b.
The new fuel storage racks are designed and shall be maintained
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with:
1) k,, s0.95 if fully flooded with unborated water as described in S,ection 9.1 of the FSAR; and k, 50.98 if moderated by aqueous foam as described in Section 2) 9,.1,of the FSAR; and 3)
A nominal 21" center to center distance between fuel assemblies placed in the new fuel storage racks.
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DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 596 feet.
CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1418 fuel assemblies.
i 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.
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CATAWBA - UNITS 1 AND 2 5-7 Amendment No.134 (Unit 1)
Amendment No.128 (Unit 2) 1
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)
- 10. Accumulator and Refueling Water Storage Tank baron concentration limits for Specifications 3/4.5.1 and 3/4.5.4.
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- 11. Reactor Coolant System and refueling canal boron concentration limits for Specification 3/4.9.1.
- 12. Standby Makeup Pump water supply boron concentration limit of Specification 4.7.13.3.
- 13. Spent Fuel Pool boron concentration limit of Specification 3/4.9.12.
l The analytical methods used to determine the core operating limits shall be
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those previously reviewed and approved by NRC in:
1.
WCAP-9272-P-A,." WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY,"
July 1985 (W Proprietary).
I (Methodology for Specifications 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor.)
t 2.
WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL 0FFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION," June 1983 (W Proprietary).
(Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2 - Heat Flux Hot Channel Factor (W(Z) surveillance requirements for F Methodology.)
3.
WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL j
USING BASH CODE," March 1987, (W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
4.
BAW-10168PA, Rev.1, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," January 1991 (B&W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
U j
CATAWBA - UNITS 1 & 2 6-19 Amendment No.134 (Unit 1)
Amendment No.128 (Unit 2)
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o ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued) 5.
DPC-NE-2011P-A, " Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," March, 1990 (DPC Proprietary).
(Methodology for Specifications 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.1.3.5 - Shutdown Rod Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)
6.
DPC-NE-300lP-A, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," November 1991 (DPC Proprietary).
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coeffi-cient, 3.1.3.5 - Shutdown Rod Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)
7.
DPC-NF-2010P-A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June 1985 (DPC Proprietary).
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, Specification 4.7.13.3 - Standby Makeup Pump Water Supply Boron Concentration, and Specification 3.9.1 - RCS and Refueling Canal Boron Concentration, and Specification 3.9.12 - Spent Fuel Pool Boron Concentration.)
8.
DPC-NE-3002A, "FSAR Chapter 15 System Transient Analysis Methodology,"
November 1991.
(Methodology used in the system thermal-hydraulic analyses which determine the core operating limits) 9.
DPC-NE-3000P-A, Rev. 1, " Thermal-Hydraulic Transient Analysis Methodology," November 1991.
(Modeling used in the system thermal-hydraulic analyses) i i
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i CATAWBA - UNITS 1 & 2 6-19a Amendment No. 134 (Unit 1) j Amendment No.128 (Unit 2)
-