ML20070D954
| ML20070D954 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 07/06/1994 |
| From: | Powers K TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9407130223 | |
| Download: ML20070D954 (10) | |
Text
>
4 A
we Valley Authonty, Post Othce Box 2000. Ekxjd Dasy Tennessee 37379 2000 f
Ken Powers vre Pmsdont. Eequoyan Nudear Plant July 6, 1994 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C. 20555 Gentlemen:
In the Matter of
)
Docket Nos. 50-327 Tennessee Valley Authority
)
50-328 SEQUOYAH NUCLEAR PLANT (SQN) - 10 CFR 50.46 ANNUAL REPORT
References:
1.
TVA letter to NRC dated July 28, 1993 2.
TVA letter to NRC dated November 12, 1993, "Sequoyah Nuclear Plant (SQN) - 10 CFR 50.46 Updated Annual Report" 10 CFR 50.46 requires reporting, at least on an annual basis, each change to or error discovered in an acceptable loss-of-coolant accident (LOCA) evaluation model or in the application of such a model that affects the peak clad temperature (PCT) calculation. Additionally, 10 CFR 50.46 requires that a 30-day report be furnished if a significant change or error is discovered.
The purpose of this letter is to provide both the annual report and the 30-day notification.
In 1992, Westinghouse Electric Corporation identified an emergency core cooling system (ECCS) model issue dealing with the swelling and bursting of fuel rods.
For conservatism, while the issue was analyzed, a temporary 103-degree Fahrenheit (F) 1 penalty was assigned to the small break LOCA PCT.
The issue has been resolved and an 86-degree F penalty is permanently assigned. This identified change did not result in exceeding the limits of 10 CFR 50.46, and PCT margin allocations will ensure these limitations are not exceeded. Therefore, further reanalysis or actions are not planned at this time.
i The enclosed documentation contains the recent changes to SQN's ECCS evaluation model and the effect on the PCT during the reporting period beginning July 29, 1993. Reference 1 supplied SQN's last 10 CFR 50.46 annual report and Reference 2 supplied an update to that report, j
oe<.~
s 9407130223 94o706 h90 DR ADDCK 05000327 p
.PDR
s i
U.S. Nuclear Regulatory. Commission Page 2 July 6, 1994 Additionally. potential issues are under investigation by Westinghouse that may impact the PCT for both large and small break LOCA. The i
potential issues have had PCT margin temporarily allocated to ensure that the cumulative efforts are tracked such'that the 10 CFR 50.46 PCT limit of 2200-degrees F is not exceeded. Upon their resolution, these issues will continue to be reported as appropriate.
Please direct questions concerning this issue to W. C. Ludwig at (615) 843-7460.
Sincerely Ken Powers Enclosure cc (Enclosure):
Mr. D. E. LaBarge, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 NRC Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy-Daisy, Tennessee 37379-3624 Regional Administrator U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323-2711 l
~ _.
F ENCLOSURE l
10 CFR 50.46 REPORT DOCUMENTATION
-l i
i 1
i
- l
)
}
I
'i i
i i
l
}
Largt, Break Lean _nLCnolanLAccident_(LOCAL JCI__
Attacinent Previous Licensing Basis Peak 2164'F j
Fuel Cladding (PCT) (June 10, 1993)
(Reference 1) 1.
Vessel and Steam Generator
-6*F 1
Calculation Errors in the LUCIFER Computer Code 2.
Cold Leg Temperature
+10*F 2
Gradient 3.
Hot Leg Injection Valve
+6*F 3
Leakage Updated Licensing Basis PCT 2174*F Net Change
+10*F Smaillreak_LQCA PCT A11achment Previous Licensing Basis PCT 2116*F (June 10, 1993) (Reference 1)
Revised Licensing Basis PCT
+1932*F (Reference 2) 1.
Hot Assembly Average Rods
+9'F 4
Burst Effects 2.
Revised Burst Strain Limit Model
-9'F 5
3.
Vessel and Steam Generator
-16*F 1
Calculation Errors in the LUCIFER Computer Code 4.
Cold Leg Temperature Gradient
+2*F 2
5.
Burst / Blockage Modeling
+86'F 6
Updated Licensing Basis PCT 2004*F Net Change
+72*F A detailed discussion of each of the changes outlined above is included in the indicated attachment.
~..
VESSEL AND STEAM GENERATOR CALCULATION ERRORS IN LUCIFER Backoround The LUCIPER code is used to generate the component databases, from raw input data, to be used in the small and large break LOCA analyses.
Errors were found in the VESCAL subroutine of the LUCIFER code.
These errors were in the geometric and mass calculation of the vessel and steam generator portions of the needed data.
All LOCA analyses using the LUCIFER code outputs are affected by these error corrections.
The errors were corrected in a manner to maintain the consistency of the LUCIFER code.
The errors were determined to be a Non-Discretionary Change as described in Section 4.1.2 of WCAP-13451 and were corrected in accordance with Section 4.1.3 of WCAP-13451.
Estimated Effect Representative plant calculations indicate a net PCT effect of
-16oF for small break LOCA and a -6eF for large break LOCA.
These calculations are applicable to the Sequoyah large break /small break LOCA analyses.
l
COLD LEG TEMPERATURE GRADIENT Backaround In the late 1960s, Westinghouse identified temperature gradients in the RCS crossover legs between the steam generators and reactor coolant pumps.
The gradient resulted from differences in outlet temperatures between long and short steam generator U-tubes.
To obtain improved temperature measurements, the Westinghouse standard design was modified to measure T-Cold downstream of the reactor coolant pumps.
Because of the flow mixing provided by the reactor coolant pumps, this modification was thought to provide an accurate measurement of bulk T-Cold with only one location for cold leg temperature measurement.
In 1990, Sequoyah installed a second T-Cold measurement point in each loop in conjunction with the RCS resistance temperature detector bypass manifold deletion / EAGLE-21 process protection system upgrade program.
Data obtained from the two T-Cold measurement points indicated that there was a measurable temperature gradient in the RCS cold leg at Sequoyah.
Data from similar plants with two cold leg temperature measurements per loop confirmed the presence of a temperature gradient.
Both the Large Break LOCA and Small Break LOCA Evaluation Models use RCS T-Ave as an input parameter.
In late 1990, a Sequoyah plant-specific evaluation was performed which conservatively evaluated the effect of a 20F reduction in T-Ave.
This T-Ave l
reduction was intended to reflect T-Cold measurement errors due to cold leg thermal gradients.
This plant-specific evaluation resulted in a loof increase in calculated PCT for large break LOCA and a 2oF increace in calculated PCT for small break LOCA.
These results were considered to be temporary assessments pending review of the root cause of the cold leg thermal gradient.
A review of the root cause was conducted by the Westinghouse Owners Group Upper Plenum Flow Anomaly Subgroup.
(In addition to the upper plenum flow anomaly, this group conducted a root cause investigation of RCS hot leg / cold leg temperature streaming.
RCS temperature streaming was considered to be a contributtng factor to the flow anomaly.)
Estimated Effect The upper Plenum Flow Anomaly Subgroup completed the root cause evaluation in November 1993.
The review concluded that a singular root cause of RCS flow streaming (and a corresponding effective corrective action) could not be identified.
As such, it was decided to revise the classification of the Sequoyah plant-specific assessments from " temporary" to " permanent."
The increases in calculated PCT. discussed above are now considered to be permanent assessments.
l
r DOUBLE-DISK GATE VALVE PRESSURE EQUALIZATION Backarggnd Westinghouse identified a potential issue concerning the use of double-disk gate valves in the emergency core cooling system (ECCS) as hot leg isolation valves.
Use of double-disk gate valves may involve an inner disc pressure equalization line that could set up a leak path into the hot leg during cold leg injection following a loss of coolant accident (LOCA).
This condition could lead to inadequate cold leg injection resulting in an increase in PCT.
The design characteristic of a double-disk gate valve provides isolation by the downstream disk sealing against the valve seat.
The mechanical seating force and the hydraulic force from the upstream pressure (SI pump) act to provide force to the valve seal surfaces..The double-disk gate valve design results in a volume of fluid which is enclosed between the discs when the valve is closed.
As the fluid volume heats up, pressure greater than system pressure may develop and may cause the disks to bind against the seats to the extent that the valves cannot be opened.
To avoid this, many double-disk gate valves have been modified to include a pressure equalization line or a small hole in one of the disks to relieve the pressure between the disks.
Based on generic leakage calculatic 2, it was determined that the double-disk gate valves modified to eliminate concerns for thermal binding could leak as much as 30 gpm per valve.
This leakage into the RCS hot legs will increase steam binding during reflood and result in an increase in the calculated peak cladding temperature.
In response to the Westinghouse potential issue, it was noted that TVA Design Change Notices M08573A (Unit 1) and M08574A (Unit
- 2) require installation of a pressure equalization line on the Sequoyah residual heat removal hot leg recirculation isolation valve (1, 2-63-172) to address thermal binding.
The Unit 1 valve was modified during the Cycle 6 refueling outage.
The Unit 2 valve will be modified in the Cycle 6 refueling outage.
Estimated Effect The PCT effect on the Large Break LOCA Evaluation Model for this issue was evaluated for Sequoyah on a plant-specific basis.
The calculation conservatively established a 6oF PCT increase for the.
addition of the pressure equalization line.
An assessment of this issue on the Small Break LOCA Evaluation Model PCT results showed a nominal benefit (which is being considered a OoF impact for Sequoyah).
l
i attachment 4 HOT ASSEMBLY AVERAGE ROD BURST EFFECTS Backaround The rod heat up code used in Small Break LOCA calculations contains a model to calculate the amount of clad strain that accompanies rod burst.
However, the methodology which has historically been used is to not apply this burst strain model to the hot assembly average rod.
This was done so as to minimize the rod gap and therefore maximize the heat transferred to the fluid channel, which in turn would maximize the hot rod temperature.
However, due to mechanisms governing the zirc-water temperature excursion (which is the subject of the SBLOCA Limiting Time-in-Life penalty for the hot rod), modeling of clad burst strain for the hot assembly average rod can result in a penalty for the hot rod by increasing the channel enthalpy at.the time of PCT.
Therefore, the methodology has been revised such that burst strain will also be modeled on the hot assembly average rod.
r This was determined to be a Non-discretionary Change as described in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of WCAP-13451.
i Estimated Effect Representative plant calculations have shown that this change introduces an approximate 10 percent increase in the SBLOCA Limiting Time-in-Life penalty on the hot rod.
For Sequoyah, this i
results in a 90F PCT increase.
However, this penalty is offset by the Revised Burst Strain Limit Model described on the following page.
I i
4
I REVISED BURST STRAIN LIMIT MODEL dackoround A revised burst strain limit model which limits strains was implemented in the rod heat up codes used in both Large Break and Small Break LOCA.
This model is identical to that previously approved for use for Appendix K analyses of Upper Plenum Injection plants with WCOBRA/ TRAC.
It is described in WCAP-10924-P-A, Rev.
1, Vol.
1, Add.
4,
" Westinghouse Large Break LOCA Best Estimate Methodology: Volume 1:
Model Description and Validation, Addendum 4:
Model Revisions," 1991.
This has been determined to be a Non-Discretionary Change as discussed in Section 4.1.2 of WCAP-13451 and is being implemented in accordance with Section 4.1.3 of WCAP-13451.
Estimated Effect The estimated.effect on Large Break LOCA PCTs ranges from negligible to a moderate, unquantified benefit.
For Sequoyah, this benefit will be considered a OoF impact on calculated PCT.
In Small Break LOCA, representative plant calculations indicate that the magnitude of the benefit is conservatively estimated to be exactly offsetting to the penalty introduced by the Hot Assembly Average Rod Burst issue documented on the previous page.
For Sequoyah, this represents a 9eF PCT reduction.
f, o
SMALL BREAK LOCA LIMITING TIME IN LIFE - ZIRC/ WATER OXIDATION TEMPERATURE EXCURSION
Background
Westinghouse recently completed an evaluation of a potential issue with regard to burst / blockage modeling in the Westinghouse small break LOCA evaluation model.
This potential issue involved a number of synergistic effects, all related to the manner in which the small break model accounts for the swelling and burst of fuel rods, modeling of the rod burst strain, and resulting effects on clad temperature and oxidation from the metal / water reaction models and channel blockage.
Fuel rod burst during the course of a small break LOCA analysis was found to potentially result in a significant temperature excursion above the clad temperature transient for a non-burst case.
Since the methociology for SBLOCA analyses had been to l
perform the analyses ac a near beginning of life (BOL) condition, where rod internal pressures are relatively low, most analyses did not result in the occurrence of rod burst, and therefore may not have reflected the most limiting time in life PCT.
In order to evaluate the effects of this phenomenon, Westinghouse has developed an analytical model which allows the prediction of rod burst PCT effects based upon the existing analysis of record.
Estimated Effect Resolution of this issue resulted in a PCT increase of 860F for j
Sequoyah.
Since the initial evaluation of the issue in 1992, i
Sequoyah has conservatively carried a " temporary" PCT penalty of 1030F against the small break LOCA evaluation model results.
{
Since the evaluation of this issue has been finalized, the actual PCT increase is being reported as a " permanent" change.
_