ML20011E736
| ML20011E736 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 02/07/1990 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20011E734 | List: |
| References | |
| NUDOCS 9002220335 | |
| Download: ML20011E736 (5) | |
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ATTACllMENT I PROPOSED CllANGES TO THE TECHNICAL SPECIFICATIONS
.i 9002220335 900207 N
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DESIGN FEATURES I"
OESIGN PRES $WRE AND TEMPERATURE
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5.2.2 The reactor containment vessel is designed and shall te maintained for a maximum internal pressure of 15 psig and a temperature of 32B'F.
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5.3 REACTOR CORE b
FUEL. ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly 1,
nominally containing 264 fuel rods clad with Zircaloy 4, except that substity-tions of fuel rods by filler rods consisting of Zircaloy-4 or stainless steel, y
or by vacancies, may be made in fuel assemblies if justified by cycle-specific
-'j reload analyses using NRC-approved methocology.
Should more than 30 rods in the core, or 10 rods in any assembly, be replaced per refueling, a special
}y report describing the number of rods replaced will be submitted to the commis-g sion pursuant to Specification 6.9.2 within 30 days after cycle startup.
Each W8 f fuel rod shall have a nominal active fuel length of 144 inches.
Reload fuel V
shall be similar in physical design to the initial core loading and shall have 4
a maximum enrichment of 4.0 weight percent U-235 with a maximum enrichment g
tolerance of 2 0.05 weight percent U-235.
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CONTROL R00 ASSEMBLIES 4 NW q
5.3.2 The core shall contain 53 full-length control roa assemblies.
The full-9j length control rod assemblies shall contain a nominal 142 inches of absorber tA ' 7 material of which 102 inches shall be 100% boron carbide and remaining 40 inch
$gg tip shall be 80% silver, 15% indium, and 5% cadm b E
TS For U 1, al ntrol r shall b ad wit
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Unit all con rocs, e pt for contre' ocs in o nod C1 er ontro ssembly CA), sha De clad h sta iss steel
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rema' ina Rod ster Conte Assembi control recs shal e clac th Incone' Ea 5.4 REACTOR COOLANT SYSTEM OESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and snall De maintained:
a.
In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal :egrscation oursuant to the applicable Surveillance Requirements, b.
For a pressure a 2485 psig, and c.
For a temperature of 650*F, except for tne pressuri:er which is 680'F.
VOLUME
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5.4.2 The total water and str i volume of th. Reactor Coolant System is 12,040 2 100 cubic feet at a nominal T o' 32i'F.
avg 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown in Figure 5.1-1.
CATAWBA - UNITS 1 & 2 5-6 Amencment No. 64 (Unit 1)
Amencment No. 58 (Unit 2)
i ATTACIMENT 11 DISCUSSION, NO SIGN 171 CANT llAZARDS ANALYSIS AND ENVIRONMENTAL IMPACT STATEMENT l
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h DISCUSSION. NO SIGNIFICANT HAZARDS ANALYSIS AND ENVIRONMENTAL 1MPACT STATEMENT On May 23, 1989 the NRC issued License Amendment Nos. 64 and 58 to Catawba Units 1 and 2 Facility Operating Licenses NpF-35 and NpF-52, respectively. These amendments require that one Catawba Unit 2 RCCA be clad with inconel. The proposed changes to Technical Specifications Design Feature 5.3.2 provide the option to withdraw the inconel clad RCCA from the Unit 2 core and replace it L
with a Westinghouse 17x17 kCCA should unexpected wear be discovered during upcoming inspections.
Duke is currently conducting the RCCA demonstration program at Catawba Unit 2.
Three Babcock and Wilcox Fuel Company (BWFC) supplied 17x17 RCCAs were inserted into the Unit 2 core at the beginning-of-cycle (BOC) 3.
These RCCAs incorporate clad coating and plating materials that are resistant to wear. Two of the demonstration assemblics consist of control rods fabricated with Armoloy plated 304 stainless steel cladding. Tne third assembly is comprised of control rods fabricated with chromium carbido coated Inconel 625 cladding. The chromium carbide coating is supplied by Union Carbido and is applied by their patented D-Gun process. The objectives of the demonstration program are as follows:
- 1) Demonstrate the compatibility of the BWFC RCCAs with Westinghouse internalst
- 2) Demonstrate that BWFC RCCAs function as required during RCCA SCRAMS and stepping axercisest
- 3) Determine the wear characteristics of various RCCA clad coatings versus typical clad materials.
Wear measurements will be performed on the BWFC RCCAs and the upper internals guide structures to (a) quantify the performance of RCCAs relative to mitigating clad wear and (b) determine the impact of the wear resistant coatings on the mating surfaces of the upper internals.
Upper Internals guido card inspections are planned at Catawba Unit 2 during the end-of-cycle (EOC) 3. EOC-4, E00-5, and EOC-6 refueling outages to evaluate the impact of the BWFC demonstration RCCAs on guide cards. A B&W inspection system is currently being developed to completely profile 9 guide cards on each upper Internals guide tube. Duke will evaluate the guido card and RCCA wear data obtained as a result of BWFC demonstration RCCA operation to determine the acceptability of wear resistant coatings. The proposed amendment will provide the option to withdraw the inconel clad RCCA and replace it with the 17x17 Westinghouse RCCA currently used at Catawba should excessive wear be detected during the upcoming inspections.
I Further details regarding the RCCA demonstration program are provided por my April 6, 1989 and April 21, 1989 submittals to the NRC.
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Modifying the Catawba Technical Specifications as identified in Attachment 1 to allow the withdrawal of the demonstration inconel clad RCCA from the Catawba Unit 2 core does not increase the probability or consequences of the accidents or safety related equipment malfunctions that are evaluated in the FSAR. The possibility of an accident or equipment malfunction which is different than any previously evaluated in the FSAR is not created. Also, the margins of safety which are defined in the bases of the Technical Specifications are not reduced.
The Chromium carbide coated /Inconel 625 RCCA would be replaced with a-Westinghouse 17x17 RCCA. Westinghouse 17x17 RCCAs are normally used at Catawba Nuclear Station.
10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations if operation 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 accident previously evaluated: or
- 3) Involve a significant reduction in a margin of safety.
Withdrawal of the inconel clad RCCA and its replacement with a Westinghouse 17x17 RCCA does not involve any increase in the probability or consequences of an accident previously evaluated.
FSAR Chapter 15 accidents were evaluated assuming all RCCAs are Westinghouse 17x17 assemblies. The proposed Technical Specification 5.3.2 only provides the ficxibility to withdraw the demonstration inconel clad assembly should unexpected wear be discovered during upcoming inspections.
The proposed amendment cannot create the possibility of a new or different kind of accident irom any accident previously evaluated.
All accidents previously evaluated assumed RCCAs are Westinghouse 17x17 assemblies. No new mode of.
i operation is introduced by this change.
The proposed araendment does not reduce any margin of safety. Drop times for the control rods will be verified within the Technical Specification limits before unit operation following a refueling outage. No other margin of safety is affected by this change.
Based on the above discussion. Duke Power concludes that this proposed amendment does not involve any significant hazards considerations.
f-Environmental Impact The proposed Technical Specification change has been reviewed against the criteria of 10 CFR 51.22 for environmental considerations. As shown above, the proposed change does not involve a significant hazards consideration, nor increase individual or cumulative occupational radiation exposures. Based on this, the proposed Technical Specification change meets the criteria given in 10 CFR 51.22(c)(9) for a categorical exclusion from the requirement for an l
environmental Impact Statement.
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