ML20059G735
| ML20059G735 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 09/06/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059G733 | List: |
| References | |
| NUDOCS 9009130121 | |
| Download: ML20059G735 (4) | |
Text
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9 NUCLEAR REGUL.ATORY COMMISSION t
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l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION j
r RELATED TO AMENDMENT N05.139 AND 122 TO FACILITY OPERATING LICENSE NOS. NPF-4 AND NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE NORTN ANNA POWER STATION. UNITS NO. 1 AND NO. 2 D_0,CKET NOS. 50-338 AND 50-339 I
1 INTRODUCTION By letter dated January 15, 1990 and as supplemented by letter dated August 15,
'1990 Virginia Electric and Power Company (the licensee) requested changes to the Technical Specifications (TS) for North Anna Power Station, Units 1 and 2 (NA-1&2).
The changes would increase the maximum allowable control rod. drop time.
These changes are needed to support a planned fuel design change from the Westinghouse Low Parasitic 17X17 (LOPAR) fuel assembly to a new 17X17 assembly with Westinghouse VANTAGE SH fuel assembly design features.
The new
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. fuel design, designated North Anna Improved Fuel (NAIF), utilizes Zircaloy grids and smaller diameter thimble tubes. Because of the smaller diameter (thimble tubes, a change to the TS governing the maximum allowable control rod
- drop time is needed.
The NAIF design includes the following VANTAGE SH design features:
i (1) Low pressure drop intermediate Zircaloy grids (2) Reconstitutable top nozzle (RTN) i (3) Extended burnup capability (4) Reconstitutable debris filter bottom nozzle (DFBN)
The RTN, DFBN and extended burnup capability w ro previously evaluated as part of the Reload Safety Evaluation and implemented in NA-1 Cycle 8 and NA-2 Cycle 7.
The Zircaloy grids will decrease the neutron parasitic capture and thereby permit more efficient fuel usage.
In addition to the fuel assembly modifications.
i thimble plug assemblies will be removed from the NA-1 Cycle 9 and NA-2 Cycle 8 and subsequent cycles.
The August 15, 1990 16tter provided additional information requested by the staff regarding tables for retained departure from nucleate boiling ratio l"L (DNDR) for each cycle-specific Reload Safety Evaluation.
The additional information did not alter the proposed action or affect the staff's initial determination of no significant hazards consideration as noticed in the Federal Register on February 21, 1990 (55 FR 6122).
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l-9009130121 900906 PDR ADOCK 05000338 P
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a ll 2-DISCUSSION The mechanical *, thermal / hydraulic, nuclear and accident evaluations performed by the licensee considered the impact of the Zircaloy grids and the thimble plug removal.
The VANTAGE SH hydraulic test program as stated in the Westinghouse Reference, Core Report-VANTAGE SH Fuel Assembly, Addendum 2-A to WCAP-104444-P-A, February 1989, showed that the VANTAGE SH grid and LOPAR fuel Inconel mixing vane grid had the same hydraulic characteristics.
Therefore, no transition core DNBR penalty or loss of coolant accident (LOCA) PCT transition core penalty was needed for the mixed assembly cores.
The NAIF assembly guide thimbles have a 8 mil inner diameter (ID) and outer diameter (00) reduction abova the dashpot.
The thimble tube ID provides adequate diametral clearance for the control rods.
The time to the dashpot for accident analyses has been increased from 2.2 seconds to 2.7 seconds.
The physics characteristics for an NAIF core as compared to the LOPAR core are slightly different, but well within the normal range of variations seen from cycle to cycle.
The main impact of thimble plug removal is the increase in core bypass flow.
Hydraulic calculations show that the design value of core bypass flow assumed in the deterministic DNB evaluations increases from 4.5% to 6.5%.
The best estimate bypass'flowrate used in the statistical DNB evalua'.lons increases by 1.5%.
The accidents analyzed in the Updated Final Safety Analysis Report (UFSAR) which could potentially be affected by the NAIF reload were reviewed by the licensee.
Accident transients for which the protection system trips the reactor within a few seconds after the transient begins were reanalyzed with the rod drop time of 2.7 seconds..These accidents were the loss of flow, locked cotor and rod ejec-tion.
The results of these accident analyses showed that the safety limits and criteria are satisfied for the increased rod drop time.
Other non-LOCA accidents analyzed in the UFSAR were individually reviewed and were found to be minimally affected by the increased rod drop time.
The increase in bypass flow was directly accounted for in the DNB transients which were reanalyzed, locked rotor and loss of flow.
For those' transients which were not reanalyzed, the loss of margin resulting from the increased bypass flow and the associated reduction of core flow is accommodated by retained DNBR margin.
Tables for retained margin are included in the cycle-specific thermal-hydraulic evaluation in the technical report for each cycle-specific Reload Safety Evaluation.
The existing analysis for a full core of 17X17 LOPAR fuel bounds a full core of 17X17 NAIF fuel, and since there is no transition core penalty, the current LOCA analysis remains bounding.
The TS changes to support the transition to the NAIF fuel are:
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Modifications to IS 3.1.3.4 would permit an increase in the control drop time from 2.2 to 2.7 seconds.
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Modification to TS Bases 3/4.4.1 would reflect the use of the Wk8-1 correlation which would make the section consistent with Section 8 3/4.2.4 and the Aases for Section 2.0.
This change was overlooked in a previous modification.
EVALUATION j
t Based on the review of ths material submitted by the licensee, we find that i
operation in the proposed manner is within the acceptance criteria and, thus, the proposed TS changes to be acceptable.
ENVIRONMENTAL CONSIDERATION These amendments involve a change to a requirement with respect to installation or use of a facility component located within the restricted area as defined in 7
We have determined that the amendments involve no significant t
increase in the amounts, and no significant change in the types, of any effluents i
that may be released offsite, and that there is no significant increase in i
individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that these amendments involve no significant hazards consideration and there has been no public comment on such finding.
Accordingly, these amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to-10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of these amendments.-
CONCLUSION We have concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
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.3te: September 6, 1990 Principal Contributor:
M. Chatterton
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0' DATED:
September 6, 1990
. AMENDMENT N0 139 TO FACILITY OPERATING LICENSE NO. NPF-4-NORTH ANN?. UNIT 1
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. AMENDMENT NO.: 122 TO FACILITY OPERATING LICENSE NO. NPF-7-NORTH ANNA UNIT 2 gg
'NRC & Local PDRs PDII-2 Reading o
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S. Yarga, 14/E/4 G.' Lainas -14/H/3 i-H. Berkow D.-Miller L. Engle OGC-WF D. Hagan, 3302 MNBB.
E. Jordan, 3302 MNBB B. Grimes, 9/A/2 G. Hill (8),P-137 Wanda' Jones, P-130A J. Calvo, 11/F/23 M. Chatterton ACRS(10)
GPA/PA OC/LFMB M. Sinkule. R-II cc: Plant Service list i
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