ML20137H575
| ML20137H575 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/27/1997 |
| From: | Shell R TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-M95958, TAC-M96599, NUDOCS 9704020283 | |
| Download: ML20137H575 (6) | |
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Tennessee Vaney Authority, Post omce Box 2000, soddf-Daisy, Tennessee 37379-2000 March 27,1997 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Gentlemen:
In the Matter of
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Docket No. 50-327 Tennessee Valley Authority
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50-328 SEQUOYAH NUCLEAR PLANT (SON) - RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING TECHNICAL SPECIFICATION CHANGE 96-07.
Reference:
NRC letter to TVA dated March 18,1997, " Request for Additional Information - Technical Specification Change Request TS 96-07 for Sequoyah Nuclear Plant Units land 2 (TAC NOS. M95958 and M96599).
In response to NRC questions contained in the referenced letter, enclosed are TVA's answers to questions 1,4, and 5. These three questions are being addressed at this time to support concurrent NRC review of TS change 96-01, SON Mark-BW Fuel Transition. TVA will provide responses to the balance of the questions (questions 2 and 3) in a subsequent letter.
Please direct questions concerning this issue to Don Goodin at (423) 843-7734.
af Sincerely, h
bl. Oleff R. H. Shell Site Licensing and Industry Affairs Manager 9704020283 970327
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4 U.S. Nuclear Regulatory Commission
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March-27,1997 '
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R. W. Hernan, Senior Project Manager j
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 11 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323-2711 4
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i ENCLOSURE i
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i Sequoyah Nuclear Plant
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Responses to Questions 1,.4, and 5 provided in
- I NRC letter to TVA letter dated March 18,1997.
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Sequoyah Nuclear Plant Request for Additional Information Regarding Pressurizer Safety Valve and Main Steam l
Safety Valve Setpoint Tolerance Increase Q1)
Justify the following changes in assumptions from the current FSAR analysis:
a) Current -
Initial RCS Temperature is assumed at its maximum value consistent with steady-state full power.
New -
Initial RCS Temperature was set to its nominal value minus four degrees for control band and measurement uncertainty.
b) Current -
Initial RCS Pressure is assumed at a minimum value consistent with the steady state full power operation.
New -
Initial RCS Pressure is assumed at its nominal value consistent with steady-state full power operation.
i Initial RCS Power is assumed at its maximum c) Current value consistent with steady-state full power.
New -
Initial RCS Power is assumed at its nominal value consistent with steady-state full power operation.
Response
All of the portions of this question are assumed to address the initial conditions utilized in the loss of electric load analysis, performed by FCF in the support of the Mark-BW fuel reload.
a) The RCS temperature is initialized at nominal conditions minus 4 F in the RELAPS model to conservatively account for control band and measurement uncertainty.
This choice of initialization requires the reduction of secondary heat sink temperature and results in a minimum initial secondary 4
pressure.
Upon transient initiation, a loss of load or i
turbine trip, the secondary pressure increases to the main steam safety valve setpoint.
By minimizing the secondary pressure initially, the heat sink temperature transient and the reduction in primary heat sink is maximized.
The result is a conservative primary pressurization.
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I Initi'alizing'the RCS-temperature at-nominal conditions minus 4 F is also conservative for secondary pressurization.
This result arises from the correlation of secondary l
. pressurization to the combined core power level and MSSV setpoints.
For loss of electric load an extremely conservative moderator temperature coefficient of +7 pcm/ F is assumed.
Core power increases.with increased coolant temperature.
By maximizing the' reduction in primary to secondary heat transfer (via the initially minimum secondary
. pressure and subsequent-increase to the MSSV setpoint), the f
rise in core power is also maximized.
A maximum core power assures a conservative simulation of secondary steam production and challenge of the MSSVs.
A conservative estimate of peak secondary pressure results.
The secondary i
pressure also includes the effects of pressurizer pressure control in delaying reactor trip.
i b)fThe RCS pressure'is initialized at nominal conditions in the RELAP5 model.
This assumption is consistent with the approved methods of the non-LOCA topical BAW-10169 (please
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see Table-5.1.1).
I c) The discussion of loss of electric load initial conditions, Section 6.2.7.1 of the Mark-BW fuel reload l
topical is incorrect in the specification of initial core power for this event.
The initial RCS power assumed in the i
loss of electric load RELAPS analysis for Sequoyah is the l
maximum value, 102% of rated thermal power.
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T Q4)
What initial pressurizer level was assumed in the analysis?
Justify your' answer with respect to peak pressure considerations.
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Response
This question is assumed to address the initial conditions utilized in the loss of electric load analysis, performed by FCF in support of the Mark-BW fuel reload at Sequoyah.
In 3
this analysis, the pressurizer is initialized with 1242 ft of liquid.
This is equivalent to a pressurizer level of about 69% of span and is higher than the nominal plus uncertainty pressurizer level of 65%.
A smaller initial i
steam volume in the pressurizer yields a more rapid pressurization rate and a maximized pressure result.
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D 05) hhat were the peak primary and secondary pressures achieved in the analysis?
This question is assumed to address the Sequoyah loss of electric load analysis, performed in support of the Mark-BW fuel reload.
A peak primary pressure of 2740 psia, below the limit of 2748 psia resulted from the analysis.
A peak secondary pressure of 1201 psia, below the limit of 1208 psia was achieved in the analysis.
Note that these results are characterized by a cctrum of cases wherein the primary and secondary safety val tolerances are increased to the maximum extent possible within the design limits.
Significant added conservatism is characterized by the assumption of a positive moderator temperature coefficient.
The peak secondary pressure is maximized by the added assumption of operable pressurizer pressure control.
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