ML20091F484
| ML20091F484 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 11/22/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20091F483 | List: |
| References | |
| NUDOCS 9112060378 | |
| Download: ML20091F484 (5) | |
Text
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i NUCLEAR REGULATORY COMMISSION 3
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SAFETY EVALUATION BY THE OFFICE Or HUCLEAR REACTOR REGULATION PELATED TO AMENDMfNT N05, 149 AND 133 TO FAClllTY OPERATING LICENSE N05. NPF-4 AND NPF-7 VIRGIN 1A ELECTRIC AND POWER COMPANY 30DOMINIONELECTRICCOOPERATIVE NORTH ANNA POWER STATION, UNITS NO. 1 AND No. ?
00CKET N05. 50-338 AtJD 56 339
1.0 INTRODUCTION
By letter dated August 11, 1989, the Virginia Electric and Power Company (the licensee) proposed changes to the Technical Specifications (TS)' for the North Anna Power Station, Units No I and No. 2 (NA-1&2). The proposed changes would revise the NA-1&E TS which govern the control rod insertion limits. The changes would allow greater operational flexibility with respect to control rod bank positioning as a means of minimizing localized rod control cluster assembly (RCCA) wear. Similar changes for other nuclear facilities have already been approved by the NRC.
Westinghouse, under contract to the Electrical Power Research Institute (EPRI),
investigated the impact of fretting wear on control rod lifeti'ne (Report No.
NP-4512 dated March 1986). Two methods of control rod wear were identified:
(1) sliding wear along the entire length of rodlets, where the rodlets contacted the upper internal guide cards during rod insertion and withdrawal, and (?) fretting wear at each of the eight guide card locations due to flow induced rodlet vibration.
Based on hot cell data, it has been determined that the RCCA wear depth can reach the Westinghouse-recommended maximum as early as 74,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of critical operation time.
Repositioning the RCCA bants so that a fresh surface is contacting each guide card is estimated to extend the time at which the maximum recommended wear depth is reached by about 50 percent.
The NA-1&2 Core Operating Limits Report (COLR) currently defines the RCCA fully withdrawn position as P28 steps.
The licensee is proposing repositioning the RCCA ba-h on a routine basis to minimize fretting wear and extend the RCCA operating ii._ time.
TS changes are therefore being requested to allow the NA-1&P fully withdrawn position to vary between P25 and ??9 steps, inclusive.
The value applicable to a specific cycle would then be defined by the licensee on a case-by-case basis, 9112060370 911122 PDR ADDCK 05000338 P
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2.0 DISCUS $10N To implement the proposed change in the fully withdrawn position, the control bank insertion limits must be modified.
Currently, the NA-1&P COLRs define the fully withdrawn position as 278 steps.
These figures would be changed to show rod positions up to a maximum fully withdrawn position of ??9 steps. The fully withdrawn position, which is to be used for a giv*n cycle would be defined prior to initiation of the nuclear design calculations for that cycle, and documented in the nuclear design report and reload Safety Evaluation.
The cycle-specific fully withdrawn position would be provided to the operators in the rod insertion limit operator curve.
That curve would also define the C-bank insertion limit endpoint corresponding to the fully withdrawn position, Aside from changing the fully withdrawn position, the insertion limits would not change. Therefore, these same curves would be applicable for fully withdrawn positions f rom 725 to P29 steps, with the only difference being the definition of the fully withdrawn position (i.e., the applicable upper bound on the y-axis).
Inserting the RCCAs a small number of additional steps from their current withdrawn position would result in a negligible change to core power distributions and peaking factors. This conclusion is supported by two facts:
(1) there is very little power generated in the top portion of the core, and (2) when the rods are positioned at ??S steps, they are inserted to the top of, but not into, the active fuel region.
(in the current fully withdrawn position, the RCCAs are positioned above the top of the active f uel stack).
The licensee has evaluated the impact of changing the RCCA fully withdrawn position. This assessment consisted of two portions:
(1) the potential impact on the models and methods used for neutronic calculations, and (2) the impact on the core physics-related key analysis parameters for reload safety analysis.
Current design predictions and reload evaluations were performed using the NRC-approved pD007 FLAME < and N0ftAD models (VEP-FRD-19A. "The PDQ 07 Discrete Model," Virginia Electric and power Company, July 1981; VEP FRD-2aA, "The VEPC0 FLAME Model," Virginia Electrk. and power Company, July 1901; and VEP-NFL.1-A, " NOMAD Code and Model," Virginia Electric and Power Company, May 1985).
It was determined that these models are suitable for modeling an all rods out (ARO) position between 225 and ?29 steps, inclusive. The effects of the AR0 chan9es in this range are easily compensated for, because the RCCAs would continue to be positioned above the top of the active core when they are in the fully withdrawn position.
A generic assessment of the impacts of changing the RCCA fully withdrawn position on safety analysis inputs was performed by the licensee, as discussed below. A cycle-specific analysis would also be performed as part of the reload safety evaluation process prior to implementation of a new ARO position. This analysis would confirm that, for the cycle in question, the amount of margin remaining for any affected key analysis parameter is sufficient to accommodate the effects of the ARO position change.
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- The generic evaluation indicated that a core reload would be irnpacted in the following areas:
(1) integral rod worths (?) power distributions, (3) reactivity redistribution (4) boron concentrations, and (5) burnup distributions.
When the RCCAs are inserted further than the current ??8 step " parked" position, the rod worths associated with that insertion are not available during a trip, which results in a decrease in excess shutdown margin and trip reactivity magnitude. For these parameters, the differences would be negligible in comparison with the excess margin which has been available in recent reload evaluations.
The reduction in trip reactivity at hot full power would produce more limiting results for trip reactivity vs. position.
However, for those transients affected by trip reactivity vs. position (e.g.,
the loss of flow accident), conservatisms exist in the key analysis parameters to acconinodate the impact of the trip reactivity change with no change to the l'pdated Final Safety Analysis Report (UFSAR) results.
A three step rod insertion (to step 225) would result in slight changes to the axial and radial power distributions. These changes may result in slight increases in FdH and Fq. These differences would be taken into consideration during the reload design phase, using physics models capable of determining the impact on power distributions.
Tti Fq and FdH values would rertain within their respective limits.
Increased rod shadowing caused by the slight insertion of the rods would result in a slight increase ir the end-of-cycle reactivity redistribution factor when changing from hot full power to hot zero power.
This results in a decrease in shutdown margin and available trip reactivity.
Again, sufficient margin is available to acconinodate these slight changes.
Boron concentrations would be reduced sliqhtly when the RCCAs are inserted further. This would actually provide a small amount of additional margin for inadvertent boron dilution and beginning-of-cycle temperature coefficients.
The rod shadowing effects would also result in slight changes in core isotopics, resulting in an impact on temperature coefficients. These changes are expected to be minor, and margin exists in the key analysis parameters to acconrnodate the impact.
Changes in burnup distribution due to rod insertion would impact differential rod worth (rod withdtawal), trip reactivity vs. position, and power distributions. These differences would be small and can be readily absorbed in available margin, or can be accomodated by conservatisms in other key analysis parameters, with no adverse impact on the UFSAR results.
It was determined that the differences in these five areas would be minor.
For key analysis parameters which becane more limiting, either available margin exists or the difference would be properly accounted for with the physics models.
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Withdraw?ng the RCCAs an additional step from their current withdrawn position would have a negligibic impact on the core power distributions and peaking factors because in both cases, the pCCAs are positioned above the top of the active fuel stack in the fully withdrawn position, increasing the fully withdrawn position would also increase the red drop tirre.
However, it was determined that increasing the fully withdrawn position by one step would increase the control red drop time by only about 0.01 sec. This time change is minimal, and of insufficient magnitude to affect the overall rod drop time used in any safety calculations.
In addition, the NA-1&P TS limit on maximum rod drop tine would be unchanged.
Control rod drop time measurements performed prior to the startup of each operating cycle would verify that the rod drop time from the maximum fully withdrawn position of ??9 steps is less than the maximum 2.7 seconds allowed by the NA-182 TS.
A change in the rod drop distance might be expected to affect the trip rod position versus time curve. This curve is used to compare the calculated reloao reactivity worth vs. position with the reactivity worth vs. time curve used in the safety analyses. The rod position versus time cbrve currently used for safety calculations (which is based on the maximum allowable rod drop time in the FA-l&P TS) would be bounding for fully withdrawn positions less than or equal to the current 228-step ARO position.
For a fully withdrawn position of 229 steps, the increase in the rod drop distance might be expected to have some small impact on the rod position versus time curve. However, sufficient margin exists in the reload reactivity worth versus posl tion curves to offset the impact on the rod position versus time curve. Therefore, the reactivity worth versus time curve currently used in the UFSAR would remain bounding for fully withdrawn positions between ??S and 229 steps, inclusive.
The impact of the rod position change on future cycles would be eddressed in the design analysis by incorporating the actual limits applicable to any given cycle into the reload design and safety evaluation for that cycle.
Overall, the anticipated changes to the nuclear design ptedictions are minor.
The models are expected to continue to provide predictions which are within acceptable criteria constraints.
3.0 EVALUATION i
The extremely small impact on power distributions and core physics key analysis parameters re:ulting f rom the change in the fully withdrawn pcsition can be accomodated within the existing NA-l&P core design limits. None of the parameter changes exceed the available margin to the key parameter safety analysis limits. The current control rod drop times and other tripped rod characteristics assumed in the safety analyses would not be changed as a result of the RCCA fully withdrawn position change.
Therefore, the current safety analyses will remain bounding.
Also, a cycle-specific evaluation of compliance with UFSAR acceptarce criteria would also be performed prior to implementation of the ititial change to the RCCA fully withdrawn position.
Future changes to the ARD position within the proposed allowable band would similarly be incorporated into the nuclear design and safety calculations for
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r 5-the cycle, and so would be assesset for compliance with the llFS"'
part of the NRC-approved normal reload design process; VEP '"
" Reload Nuclear Design tiethodology," Virginia Electric September 1986. Therefore the flRC staff finds the NA-1&2 TS acceptable.
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4.0 STATE CONSULTATION
in accordance with the Comission' regulatiers was notified of the proposed issuance of the had no comment.
5.0 ENVIRONMENTAL CONSIDERATION
These amcrdments change 6 requirement with r, f acility component beated within the restriz Part 20. lhe NRC staff has determined that increase in the amounts, and no significant that may be released offsite, and that ther.
individual or cumulative occupational radii previously issued a proposed finding that hazards consid2 ration and there has been n (54 FR 37055). Accordingly, these amendw for categorical exclusion set forth in 10 51.?2(b) no environmental impact statement prepared in connection with the issuarce
6.0 CONCLUSION
T9 Commission ha concluded, based on ti (1) there is rea; ble assurance that th not be endangered operation in the pro; be conducted in er ? (ance with the Commit of the amendments tot be inimical to s
the health and safety of the public, s
i Principal Contributor:
Leon B. Engle Odte:
November 22, 1991 f
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o l the cycle, and so would be assessed for compliance with the UFSAR criteria as part of the NRC-approved normal reload design process; VEP-FRD-42 Revision 1A,
" Reload Nuclear Design liethodology," Virginia Electric and Power Company, September 1986. Therefore, the NRC staff finds the proposed change to the NA-1&2 TS acceptable.
4.0 STATE CONSULTAT1,0N In accordance with the Commission's regulations, the
' gin b State official was notified of the proposed issuance of the amendrents. The State official nad no comment.
5.0 ENVIRONMENTALCONSlDERATQN These amendments change a requirement with respect to installation or use o' a facility component located witnin the restricted area as defined in 10 CFP Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration and there has been no public coment on such finding (54 FR 37055). Accordingly, these amendrents meet the eligibility criteria for categorical exclusion set forth in 10 CFP 51 ??(c)(9).
Pursuant to 10 CFR 51.22(b) no environmmtal impact statement or environmental assessment need be prepared in connection with the issuance of these amendments.
6.0 CONCLliSION The Commission has 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, (2) such activities will be conducted in complianco with the Comission's regulations, snd (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
Leon B. Engle Date:
November 22, 1991