ML20148E762
| ML20148E762 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 01/06/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20148E744 | List: |
| References | |
| NUDOCS 8801260107 | |
| Download: ML20148E762 (5) | |
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SAFETYEVALUATIONBYTHEOFFICEOFNUCLEARREACTORREGULATIE RELATED TO AMENDMENT NO. 116 TO FACILITY OPERATING LICENSE NO. DPR-22 AND AMENDMENT NO.116 TO FACILITY OPERATING LICENSE NO. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION, UNIT N05. 1 AND 2 DOCKET NOS. 50-280 AND 50-281 INTRODUCTION By letters dated October 7, 1986, as supplemented June 8, 1987; April 1, 1987; and May 26, 1987, Virginia Electric and Power Company (the licensee) requested amendments to Facility Operating License Nos. DPR-32 and DPR-37, issued to the licensee for operation of the Surry Nuclear Power Station, Units 1 and 2, located in Surry County, Virginia.
By letter dated October 7, 1986, as supplemented June 8, 1987, the licensee proposed to revise Section 3.12 of the Surry Technical Specifications (TS) by revising the actions to be taken by the licensee while operating with an inoperable, misaligned or dropped control rod.
By letter dated April 1, 1987, the licensee requested to revise Figures 3.12-1A
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and 3.12-1B of the Surry TS, which govern the control rod insertion limit.
The licensee proposed redefining the fully withdrawn position of all rod cluster control assembly (RCCA) banks to 225 steps, instead of the current 228 steps.
The change will allow greater operational flexibility with regard to control rod bank positioning as a means of minimizing localized RCCA wear.
By letter dated May 26, 1987, the licensee requested changes in the TS to support the planned fuel design change from the Westinghouse Low Parasitic (LOPAR) 15 x 15 Fuel Assembly to the 15 x 15 Surry Improved Fuel (SIF) Assembly during Cycle 10 for both Surry units.
DISCUSSION AND EVALUATION Control Rod Insertion Limits j
By letter dated October 7, 1986, as supplemented June 8, 1987, the licensee proposed to revise Section 3.12 of the Surry TS.
The proposed revision would alter the manner i shich inoperable rods are treated.
Currently, when an inoperable rod is aiscovered, alternate insertion limits are invoked depending on the position of the inoperable rod and whether or not it is stuck.
Opera-tion may then continue indefinitely with the new limits.
The limits are pre-calculated so that the limiting inoperable rod condition is covered.
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1 The licensee has proposed to revise the actions required upon discovery of an inoperable rod so that they more nearly match those of the Westinghouse Standard Technical Specifications.
The proposed TS would require that the rod be repaired within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or that either the power be reduced to 75% of full power within the next hour or the bank be aligned to the position of the inoperable rod and the permitted power be determined by the bank position.
Reduction in power for the misaligned rod is not required for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in the current TS.
The proposed TS would also require that the shutdown margin be determined within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.
This is an additional action not required in the current TS.
The proposed TS are thus conservative with respect to the current ones, and are therefore acceptable.
Thg proposed TS would also require that the hot channel factors (Fn(Z) and (F
be shown to be within the limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from the time of diNov)eryoftheinoperablerod.
This action is required within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in the present TS for the purpose of determining the necessity for power reduction be-low 75% of full power.
Considering other immediate actions required by the licensee, the staff finds the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit to be acceptable.
- However, the staff requested that confirmation (by measurement or calcylation) be made within the same time frame, which is required for F (Z) and F that the value oftheaxiallydependentradialorplanarpeakingf$ctor(F 3fZ))usedinthe constant axial offset control analysis is still valid.
By N tter dated June 8, 1987, the licensee submitted a revised proposal to change TS 3.12.C.5.b.3, which satisfied the staff's request.
The proposed TS would also require a reanalysis of the transients and accidents that are affected by the ino'perable rod within 5 days to confirm that the pre-vious analyses are valid.
This is an expanded requirement from that in the present TS and is, therefore, acceptable.
In the present TS, the power may be increased to greater than 75% of full power after the determination of the acceptability of the hot channel factors.
The proposed TS include an additional requirement that the effect of increased power operation on the accident analyses be determined prior to increasing the power.
Thus, this change is acceptable.
Based on the above evaluation, the staff concludes that the proposed revisions to the Surry TS are acceptable.
Fully Withdrawn Control Rods By letter dated April 1, 1987, the licensee proposed to redefine the fully with-drawn position of all RCCA banks to 225 steps, which would eliminate localized RCCA wear at the top of the control rods.
The current fully withdrawn position is 228 steps.
At 225 steps withdrawn, the RCCAs are only 0.31 inches into the active fuel region (228 steps withdrawn is above the active fuel region).
Since the top region of the core has such low worth, the effect of the proposed change was expected to be minimal.
To confirm this, neutron calculations were performed i
by the licensee.
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The shutdown margin calculations showed a 0.003% ap decrease at beginning-of-life (BOL) and a 0.006% Ap decrease at end-of-life (EOL) (recent cycles have excess shutdown margin of approximately 1.75% ap at BOL and 1.6% Ap at E0L).
The changes in the other parameters which would be affected are similarly minimal with respect to available margins.
Because the proposed change will result in slight insertion of the RCCAs into the active fuel region, the staff expects essentially negligible effects due to the proposed change.
Therefore, the staff concludes that the proposed change is acceptable.
Surry Improved Fuel By letter dated May 26, 1987, the licensee proposed amendments which support the planned fuel design change from the Westinghouse Low Parasitic (LOPAR) 15 x 15 Fuel Assembly to the 15 x 15 Surry Improved Fuel (SIF) Assembly.
The Surry units currently use the Westinghouse LOPAR fuel assemblies.
In recent years, Westinghouse offered two advanced fuel designs, known as the Optimized Fuel Assemblies (0FA) and the VANTAGE 5 assemblies.
The proposed SIF fuel is similar to the 0FA fuel but includes some features of the VANTAGE 5 fuel.
The most significant differences between the LOPAR and SIF fuel assemblies are common to both the OFA and VANTAGE 5 assemblies.
These include the use of Zircaloy grids instead of Inconel grids, smaller diameter thimble tubes, and three-leaf holddown springs instead of two-leaf springs.
VANTAGE 5 features in the SIF design include slightly shorter nozzles, which result in slightly longer thimble tubes and fuel rods and a removable top nozzle.
In addition to use of the SIF fuel assemblies, the licensee has proposed to eliminate the use of thimble plugs in reload cores.
These devices are used to inhibit flow in those assemblies which do not have either control rods or burn-able poison assemblies.
Their elimination results in a significant increase in core bypass flow and a slight increase in overall core flow.
Both the OFA and VANTAGE 5 fuel designs have been approved for use in Westing-house reactors.
The major effect on the neutronic behavior of the core from the use of these fuels is due to the increase in the drop time for the control rods.
This increase is 0.6 seconds (from 1.8 to 2.4 seconds), due to the reduc-tion in diameter of the thimble tubes.
There is no change in the fuel rod design (over its fueled length), and core neutronics parameters are not affected by the change in assembly design.
There are small changes in the thermal-hydraulic parameters of the core, due primarily to the bulkier grid straps in the advanced designs.
The thermal-hydraulic comparability of the LOPAR and the advanced designs has been confirmed by hydraulic tests.
Since the SIF fuel is essentially the same as the OFA fuel, these tests also apply to that fuel.
A more significant difference in the thermal-hydraulic performance results from the removal of the thimble plugs from the core.
Westinghouse calculates that the increased core bypass flow results in a 2% loss of DNBR margin due to the 1.5% flow decrease in the fuel rod channels.
Thimble plug removal also results in a reduction to the fuel
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assembly hydraulic loss coefficient; however, Westinghouse hydraulic tests show that the reduced fuel assembly loss coefficient results in a net reduction in the hydraulic lift force which more than compensate for the slight increase in core flow rate.
The licensee performed an evaluation to confirm that all safety criteria will be met when the SIF fuel is substituted for the LOPAR fuel.
Care was taken in the analysis to choose parameters which would bound transition cores as well as a core fully loaded with SIF fuel.
The SIF design is mechanically similar to the approved 0FA and VANTAGE 5 fuel, and is expected to have the same mechanical response.
Also, when compared to LOPAR fuel, the SIF design is expected to have the same mechanical response except that it is expected to experience less bowing.
The Zircaloy grids result in larger lifting forces, but these are offset by use of the three-leaf holddown springs.
These springs have been used on 0FA and VANTAGE 5 fuel and their use on SIF fuel is acceptable.
The shorter bottom and top nozzle designs result in longer fuel rods, since the overall envelope of the assemblies is not changed.
The fuel pellet stack length is not changed and as a result, the upper plenum length is increased.
This will permit increased fuel burnup.
Analysis has been performed to show that fuel design criteria for fission gas pressure is not violated for batch
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average discharge burnups as high as 45,000 MWD /MTV.
The effects of mixed core loading and thimble plug removal on fuel and control rod wear have been evaluated.
It was concluded that the resultant cross flow has a negligible effect on fuel rod wear.
In fact, the removal of the thimble plugs has a beneficial effect on control rod wear.
The similarity between the LOPAR and SIF fuel with respect to fuel rod diameter, rod-to-rod spacing and cladding results in negligible differences in neutronics parameters between the two fuels.
Cycle-to-cycle variations in these para-meters are dictated by fuel management policy, rather than by the differences in fuel design.
These differences are evaluated for each cycle as part of the cycle design using approved reload methodology.
The hydraulic compatibility of LOPAR and 0FA fuel has been confirmed by tests conducted at the Westinghouse Fuel Test Systems facility.
These tests are also valid for the SIF assembly.
The W-3 correlation is currently used for the DNB analysis of the Surry plants and will continue to be used for LOPAR fuel in the mixed core cycles.
The SIF assemblies will be analyzed with the THINC-I code using the WRB-1 CHF correlation. This procedure has been approved for the OFA and VANTAGE 5 fuel and is also acceptable for SIF fuel.
The 95/95 limit for the WRB-1 correlation is 1.17.
However, a plant-specific margin of 20% has been added to arrive at a design timit of 1.46 for the SIF fuel.
The 1.3 limit value for the W-3 correlation contains an 18 percent margin for the LOPAR fuel.
These margins are sufficient to account for the rod bow penalty, the transition core penalty and the impact of the removal of the thimble plugs.
The only significant changes in core characteristics that affect transient and accident evaluations are the increased scram time and thimble plug removal.
The first of these changes affects the "fast" accidents such as rod ejection, loss of flow, and locked rotor.
The second affects the large-break LOCA analysis.
The first three events were reanalyzed using approved methods and all safety criteria were shown to be met.
A reevaluation of the large-break LOCA resulted in a peak clad temperature of 1979 F, including a 10*F transition core penalty.
This meets the acceptance criterion of 2200 F and is acceptable.
Based on the above review, the staff concludes that the use of SIF fuel in Surry Units 1 and 2 is acceptable.
Further, the staff has reviewed the proposed TS changes and concludes that they are consistent with analyses provided and are acceptable.
ENVIRONMENTAL CONSIDERATION These amendments involve a change in the installation or use of the facilities components located within the restricted areas as defined in 10 CFR 20.
The staff has determined that these 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 these amendments involve no significant hazards consideration and there has been no public comment on such finding.
Accord-ingly, 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 these amendments will not be inimical to the common defense aid security or to the health and safety of the public.
Dated: January 6, 1988 Principal Contributors:
W. Brooks M. Chatterton
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