ML20236K417
| ML20236K417 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 06/29/1998 |
| From: | Sepp H WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Collins S NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20036E514 | List: |
| References | |
| CAW-98-1261, NUDOCS 9807090312 | |
| Download: ML20236K417 (29) | |
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Westinghouse Electric Company, Energy Systems Nuclear Sennees Division a division of CBS Corporation Box 355 Pittsburgh, Pennsylvania 152304355 CAW-98-1261 June 29,1998 Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: MR SAMUEL J. COLLINS APPLICATION FOR WITHHOLDING PROPRIETARY
. INFORMATION FROM PUBLIC DISCLOSURE
SUBJECT:
-RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION ON SOUTH TEXAS PROJECT RCS FLOW REDUCTION EVALUATION (PROPRIETARY)
Dear Mr. Collins:
The Proprietary information for which withholding is being requested in the above referenced '
report is further identified in Affidavit CAW-98-1261 signed by the owner of the proprietary information, Westinghouse Electric Company.- The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.790 of the Commission's regulations.
Accordingly, this letter authorizes the utilization of the accompanying Affidavit by South Texas
' Project Nuclear Operating Company.
Correspondence with respect to the proprietary aspects of the application for withholding or.the Westinghouse affidavit should reference this letter, CAW-98-1261 and should be addressed to
_ the undersigned.
I Very truly yours, enry . p , Manager Regulatory and Licensing Engineering Enclosures cc: T. Cader/NRC (5E7) l 9807090312 990630 PDR ADOCK 05000498 P PDR
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l CAW-98 1261 l
l AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:
l ss COUNTY OF ALLEGHENY:
Before me, the undersigned authority, personally appeared Henry A. Sepp, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company, a division of CBS Corporation (" Westinghouse"), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:
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-f v ~
AA l l / /\/
Henry A. Sepp, Manager Regulatory and Licenseing Engineering Swom to and subscribed before me this 80% day of od- ,1998 i
%g Janet A. Snhwab. Nota Monftevi% B6ro. #W9he
% C0mMitsbn Empires h4 g Notary Public
2- CAW-98-1261 (1) I am Manager, Regulatory and Licensing Engineering, in the Nuclear Services Division, of the Westinghouse Electric Company, a division of CBS Corporation (" Westinghouse"), and as such, I have been specifically delegated the function of reviewing the proprietary information scught to be withheld from public disclosure in connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Energy Systems Business Unit.
(2) I am making this Affidavit in conformance with the provisions of 10CFR Section 2.790 of the Commission's regulations and in conjunction with the Westhighouse application for withholding accompanying this Affidavit.
(3) I have personal knowledge of the criteria and procedures utilized by the Westinghouse Energy Systems Business Unit in designating information as a trade secret, privileged or as confidential commercial or financial information.
(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.
(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.
(ii) The information is of a type customarily held in confidence by Westinghouse and not customanly disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.
Under that system, information 12 held in confidence if it falls in one or inore of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:
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CAW-98-1261 (a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.
(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.
(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.
(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.
(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.
(f) It contairs patentable ideas, for which patent protection may be desirable.
There are sound policy reasons behind the Westinghouse system which include the following:
(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position. 1 l
l (b) It is information which is mart:e.,:ble in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to 1
sell products and services involving the use of the information. i l
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CAW-93-1261 (c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.
(d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.
(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.
(f) The Westinghouse capacity to invest corporate assets in reseasch and development depends upon the success in obtaining and maintaining a competitive advantage.
l l t i (iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10CFR Section 2.790, it is to be received in confidence by the Commission.
- (iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.
(v) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in " Responses to NRC Request for Additional Information on South Texas Project RCS Flow Reduction Evalution," (Proprietary), June 1998 for South Texas Project, being transmitted South Texas Project Nuclear Operating Company letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. Attention Mr. Samuel J. Collins. The proprietary information as submitted for use by South Texas Project Nuclear Operating Company for South Texas Project Nuclear Power Plants is expected to be applicable in other licensee submittals in response to certain NRC requirements for justification of use of RCS flow reduction.
CAW-98-1261 This information is part of that which will enable Westinghouse to:
(a) Provide documentation to support related license amendments for RCS flow l reduction.
(b) Establish applicable codes and standards which are used in the evaluation.
(c) Assist the customer to obtain NRC approval.
Further this information has substantial commercial value as follows:
(a) Westinghouse plans to sell the use of similar information to its customers for purposes of meeting NRC requirements for licensing documentation.
(b) Westinghouse can sell support and defense of RCS flow reduction methodology to its customers in the licensing process.
Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar advanced nuclear power designs and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.
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CAW-98-1261 The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.
In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended for developing testing and analytical methods and performing tests.
Funher the deponent sayeth not.
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f ATTACHMENT 4 Response to Request for Additional Information on Alternate Operation with Reduced Measured Reactor Coolant System Flow NON PROPRIETARY l
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1 EAWN4L\NRC WK\TSC 9 mull 03N DOC u__________ _ _
Westinghouse Proprietary Class 3 Page1 Response to Request for Additional Information on Alternate Operation with Reduced Measured Reactor Coolant System Flow 1.0 Introduction The following information is being provided in response to an NRC Request for Additional Information (RAl) on the 3% Flow Reduction evaluation for South Texas Units 1 and 2.
The responses have been organized in the following manner: First, background information is
- provided on the basis of the plant operating parameters used in the flow reduction evaluation, l followed by responses to the individual questions. The RAI response may refer back to the
' Background Information'section as needed.
2.0 Background Information on Program ODeratine Parameters 2.1 In 1991, the South Texas Project evaluated a range of NSSS operating temperatures at the 3800 MWt core thermal power rating with up to 10% steam generator tube plugging. The South Texas Project desired the ability to decrease the South Texas Units 1 and 2 operating temperatures by as much as 10 F in order to reduce steam generator corrosion propensity. Additional accident analyses were performed in 1993 to support the use of Vantage 5H fuel. The Reactor Coolant System (RCS) and steam parameters associated i with the current Analyses of Record are provided in Table 1. f 2.2 Column 1 of Table I represents the original design basis values licensed for South Texas Units 1 and 2. Key parameters include 3800 MWt core power,95,400 gpm/ loop thermal design Pow, and 593 F vessel average temperature. Columns 2 and 3 of Table 1 represent the Reactor Coolant System (RCS) design basis parameters for operation at the upper end of the analyzed temperature range (i.e., vessel outlet temperature of 625.6 F, j
vessel average temperature of 593 F) for 0% and 10% steam generator tube plugging, l
respectively. 1 2.3 Based on calculations performed by Westinghouse, the licensed thermal design flow of 95.400 gpm/ loop can be supported for up to 10% steam generator tube plugging in each steam generator. Therefore, the original RCS thermal design flow rate is maintained.
The South Texas Project also desired to maintain the licensed 3800 MWt core power level and RCS vessel average temperature of 593 F. Westinghouse standard practice is to calculate the thermal design flow based on, or bounded by, the upper limit of steam generator tube plugging to be analyzed and to provide steam parameters for conditions with no steam generator tube plugging as well as with the upper bound value (10% in this case). As indicated in Table 1, all column 2 and 3 primary side RCS parameters are
! identical. The steam pressure, steam temperature and steam flow rate for the 0% and l 10% plugging cases are different.
l The original design basis parameters (column 1) incorporated no specific plugging level, which corresponds exactly with the column 2 parameters.
l Westinghouse Proprietary Class 3 '
Page 2 2.4 As steam generator tubes are plugged, the heat transfer area is reduced and, as a result, the l
steam temperature and pressure decrease (e.g., i100 psia -+ 1081 psia). The decreased I steam pressure results in an increase in the latent heat of vaporization and yields a slight l 6
l reduction in the steam flow (i.e.,16.96 x 10 lbm/hr -+ 16.94 x 106 lbm/hr). If the NSSS power and RCS How are maintained, and the unit is operated with a 10 F reduction in vessel hot leg temperature (i.e., 625.6'F -+ 615.6 F), the corresponding vessel average !
and vessel inlet temperatures are 582.3 F and 549.0 F, respectively.
2.5 Columns 4 and 5 of Table i provide the design basis NSSS and steam parameters for operation with a 615.6 F vessel outlet temperature with 0% and 10% steam generator plugging, respectively. As a result of the reduced RCS temperature, the steam pressure, temperature, and flow are reduced.
2.6 The South Texas Project and Westinghouse performed bounding analyses to support operation over the range of parameters presented in Table 1. The analyses associated with postulated accidents and fuel are documented in Reference 1 of this attachment. The Updated Final Safety Analysis Report documents analyses associated with the NSSS systems and components including the development of bounding NSSS system transients which were utilized to perform analyses (e.g., stress and fatigue) to demonstrate continued compliance with regulatory and industry requirements (e.g., ASME code) applicable to the South Texas Project. The transients developed bound operation over the following ranges:
Parameter Range Bounded Full load vessel outlet temperature 615.6 F to 625.6 F Full load vessel average temperature 582.3 F to 593.0 F Fullload vesselis temperature 549.0 F to 560.4 F Full load steam pre 3sure 975 psia to 1100 psia The above accident, system, and component analyses currently serve as the design bases for operation of South Texas Units I and 2. Appropriate revisions were incorporated into Revision 5 of the South Texas Project UFSAR.
2.7 In Reference 2, the South Texas Project proposed an alternate mode of operation in which the assumed RCS Thermal Design Flow would be reduced by 3% in order to address i concerns that anticipated steam generator tube plugging, combined with hot leg streaming effects, would reduce the flow to the point where the Technical Specification Departure
. from Nucleate Boiling flow requirement for that parameter could no longer be satisfied.
The 3% flow reduction bounds the effects of expected steam generator tube plugging in the upcoming Unit 2 refueling outage. When the allowed 3% reduction is combined with j the elbow tap measurement change proposed in Reference 3 of this attachment, the South j Texas units 'will be able to plug up to 10% of the tubes in each steam generator and still i meet Technical Specification RCS flow requirements.
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l 2.8 Relative to the NSSS parameters, a 3% reduction in thermal design How results in a 1.8 F i
increase in the vessel temperature differential at 3800 MWt core power. NSSS and steam parameters were developed for conditions such that the existing system and component analyses remained bounding and that DNB margins were maintained. The resultant ,
parameters are shown in Table 1 of attachment 2 to Reference 2. In order to maintain the
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current Departure from Nucleate Boiling margin, it is necessary to reduce the upper bound vessel average temperature from 593.0 F to 590.0 F. Relative to the systems and components, by maintaining the vessel inlet temperature at a value greater than or equal to 549 F and the vessel outlet temperature at a value less than 625.6 F, the existing NSSS q system and component analyses remain bounding. In order to achieve this, it is necessary {
to: I e
increase the lower bound value of vessel average temperature by 0.9'F to 583.2 F e
decrease the upper bound value of vessel average temperature by 0.9 F to 592.1 F.
The upper bound vessel average temperature is further constrained by the 590.0 F value associated with the 3 F reduction in vessel average temperature to maintain DNB margins discussed above.
2.9 The resulting pararneter values associated with the 3% thermal design How reduction program bound the following conditions:
Parameter Range Bounded
- Full load vessel outlet temperature 617.4 F to 623.8 F Full load vessel average temperature 583.2 F to 590.0 F Full load vessel inlet temperature 549.0 F to 556.2 F Full load steam pressure 979 psia to 1067 psia These parameters are contained within the range of parameters provided in Table I prior to the 3% thermal design How reduction. 1 3.0 Ressmses to Reauest for Additional Information (Reference 4)
RAI 1. In the six items (addressed by safety evaluations] you indicate that the evaluation is done by several means, such as:
i e sensitivity studies.
- recalculation and e " evaluation."
When it is stated that the method used is by " evaluation," is it by 1) sensitivity studies,2) recalculation or 3) by engineering judgment?
l For example, on page 6 of Attachment 2, for Category 1, you state that " Evaluation of the current licensing-basis analysis results with the new Core Thermal Safety Limits (based on 3% Dow reduction) confirms that the Departure from Nucleate Boiling Ratio limit remains satisfied for each case." For Category 2 you state "Therefore, it is sufficient to
Westinghouse Proprietary Class 3 ;
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evaluate, or re-calculate, the minimum Departure from Nucleate Boiling Ratio value for ,
each of these events based on the transient statepoints from the current licensing-basis l analyses with revised reference conditions based on a 37c flow reduction." '
Please make clear which items are recalculated and which items are not based on a calculation or sensitivity analysis. Please explain what is meant in cases when the term -
" evaluation" is used for an item in your letter of 12/31/97. i l
Response to RAI #1 Tables 2,3 and 4 (attached) specifically identify each transient evaluated for the flow reduction program and describe what type of evaluation was made. The DNB event categories (i.e., third column of Table 2), as defined in the original submittal, continue to be used.
RAI2. When you state that the evaluation for accident or transient analysis is by sensitivity studies, please provide the sensitivity. For example, provide the change in departure from j nucleate boiling ratio (DNBR) or pressure for the parameter compared to. Please provide j the new value for DNBR and pressure expected from the change and also the limit value. I Also, provide the same for the results of any other sensitivity parameters . l 1
Response to RAI #2 The attached Table 2 specifies for which transients a sensitivity study was performed to support the DNB evaluation. For each of these events, a sensitivity run was made to confirm that the current limiting transient statepoint satisfies the minimum DNBR acceptance criterion with respect to the revised core limits. (The revised core limits are based upon the reduced RCS flow and reduced vessel average temperature). The minimum DNBR acceptance criterion was verified by comparing the calculated core inlet temperature for the limiting point in the transient with the limit value associated with the I revised core limits. Table 3 specifies which type of evaluation method (i.e., sensitivity study or engineering judgment) was used for the non-DNB criteria evaluation, and provides information related to the sensitivities. I RAI 3. Where you state that there are minor changes in parameters (example, operating temperatures for L.arge Break LOCA on page 3 of Attachment 2), please explain by providing comparisons of the changes both by numerical values and by percent changes.
Response to RAI #3 Please see response to RAI #6, Part b. j i
RA14. Please provide the UFSAR sections for each of the accident and transient evaluations presented so that a cross reference can be more easily made for each of the evaluations and also to aid in checking for completeness.
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l Westinghouse Proprietary Class 3 Page 5 Response to RAI #4 The UFSAR section identifiers have been specified in Table 2 for the Non-LOCA transient events and in Table 4 for the LOCA events.
RAI 5. On page 1 of Attachment 2:
- a. You state that "To offset the decrease in Departure from Nucleate Boiling (DNB]
margin,
- l. the upper end of the nominal Tavg range is reduced from 593 to 590 F, and
- 2. the K1 and K4 terms in the Overtemperature Delta-T (OTAT) and Overpower Delta-T (OPAT) reactor trip setpoints are reduced respectively.
- 3. The Limiting Condition for Operation maximum Tavg for the Reactor Coolant System is reduced from 598 F to 595 #F.
Please state the total offset in DNB margin and the contribution to margin that is obtained from each of the above three modifications.
- b. You state that "To ensure that the NSSS component analyses remain bounding, the lower end of the nominal Reactor Coolant System [RCS) average temperature (T ,g) range is raised from 582.3 to 583.3 F."
Please explain how this 1 F of increase in Tavg temperature ensures that the NSSS component analyses remains bounding.
Response to RAI #5. Part a As discussed during the telecon on May 20,1998 with the NRC, South Texas Project, and Westinghouse, our evaluations do not enable us to break down the individual contributions to DNB margin as requested above. Table 2 specifies how each event has been evaluated with respect to available DNB margin and the revised core limits associated with the reduced RCS flow and reduced vessel average temperature (Tavg). As mentioned in the " Basis for Conclusion" entry in Table 2 for several cases, the change to the OTAT and OPAT setpoints will result in an earlier reactor trip. Subsequently, less limiting transient conditions will occur for those events that rely on protection provided by the OTAT and OPAT reactor trip fur.ctions. Refer to Table 2 for additional information. For these evaluations, the DNB benefit associated with a cooler T-inlet due l to the lower RCS flow has been ignored. Furthermore, the DNB benefit due to the l revised OTAT setpoint equation which would result in an earlier trip and less severe transient conditions has also not been credited.
I Response to RAI#5. Part b The response to this RAI is contained within the " Background Information" section above. (para. 2.8)
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I l RAI 6. On page 3 and 4 of Attachment 2:
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- a. For " Power Capability Parameters," you state that you have evaluated for two l RCS Tavg levels of 583.2 F for a lower bound and 590 F for a upper bound. The l lower bound has been increased by 0.9 F and the upper bound has been decreased by 3 F.
l Please explain the backg . und on how the lower and upper bound average temperatures are arrived at and also what instrumentation is used to measure T vg .
- b. For "Large Break LOCA and Small Break LOCA," you state that "The reduced i flow results in very minor changes in operating temperatures throughout the Reactor Coolant System. The changes in these operating temperatures were evaluated based on plant-specific sensitivity studies performed to support the current operating temperature range for South Texas Project Units I and 2."
Please give an example of a minor change in operating temperature and a corresponding sensitivity study application.
Response to RAI #6. Part a The response to this RAI for determining the lower and upper bound average temperatures is contained within the " Background Information" section above. (para. 2.8)
T vg is measured using fast-response, narrow range resistance temperature detectors. The average hot leg temperature is obtained by averaging the temperature reading from three RTDs located in thermowells in each of the four hot legs. The cold leg temperature is measured using one RTD in a thermowell at the discharge of each reactor coolant pump.
Response to RAI #6. Part b -
Section 15.6.5.4.3 of the South Texas Project UFSAR (Rev. 5) contains the following statements under both the Laree Break Results and Small Break Results headings:
" Analysis results for conditions with the reduced loop operating temperatures confirmed that the predicted peak clad temperature was lower for lower loop temperatures."
Although not detailed in the UFSAR, these conclusions are supported by plant-specific sensitivity studies for South Texas which support a range of operating temperatures l 582.3 F s T. , s 593.0 F performed at the time of the UFSAR analyses.
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! At an assumed operating pressure of 2296 psia (2250 psia + 46 psi uncertainty),
consistent with the LOCA analysis assumptions, a 3% decrease in the Thermal Design Flow translated into a 3% decrease in core flow will result in a change in core enthalpy rise of - 3 BTU /lbm. With operating temperatures on the order of 600 F, this corresponds to an increase in temperatures across the core of ~ 2 F, or an - 1 F decrease in Tcold and an ~ l F in Thot. This is the "very minor" change in operating temperatures
Westinghouse Proprietary Class 3 Page 7 referred to in the LOCA evaluations (in terms of absolute temperatures, *R, this change is
< 0.1%). However, for the current UFSAR analyses, low operating temperatures in this range yield lower calculated Peak Cladding Temperatures, so in these circumstances, quanti 6 cation of the effects would show a very small benefit in Peak Cladding Temperature.
Subsequent to the completed LOCA evaluation, considerations other than LOCA resulted in limiting the operating temperature range so that the minimum Tcold would be no lower than that previously analyzed. Under these assumptions, operating temperatures in the RCS fall entirely within the previously analyzed range.
Similarly, for the upper end of the operating range, considerations other than the LOCA resulted in limiting the operating temperature range so that the maximum Thot would be no . higher than that previously analyzed. Under then assumptions, operating temperatures in the RCS fall entirely within the previouse cauh .ed range.
RAI 7. On page 6 of Attachment 2:
- a. For Category 2, please explain the statement " based on the transient statepoints from the current licensing-basis analyses."
- b. You state that sufficient margin has been identified for the 5 events listed. Please provide these values and state from what variables they are obtained from.
Response to RAI #7. Part a Transient statepoints are a set of parameters which include RCS pressure, core heat flux, core inlet temperature and/or core average temperature, and RCS flow for discrete transient time steps. These parameters are used to determine the minimum DNBR and the corresponding time that it occurs.
Response to RAI #7. Part b Table 2 specifies what form of evaluations were performed for these five events with respect to DNB.
The following terms have been used in Table 2 to define the amount of DNBR margin and the size of DNBR penalty resulting from the flow reduction. Please refer to these values as appropriate.
l l Negligible = about 1%
Small = about 5%
Sufficient = about 10%
Large = more than 20%
Page 8 RAI 8. On page 7 of Attachment 2:
You state that several current non-LOCA analyses for South Texas Project Units 1 and 2 have very little margin available and refer to Table 3. Table 3 is entitled " Sensitivity Analyses Performed With a 3 % Reduction in Reactor Coolant System Flow." The events listed in Table 3 are loss ofload, loss of feedwater, station blackout, feedline break, and rod ejection. The corresponding limiting parameters for each event are listed in Table 3.
Please provide the results from the analyses for these limiting parameters.
Resnonse to RAI#8 The attached Table 3 describes the results of evaluations performed to verify that the listed acceptance criteria were met for the 3% flow reduction for each of these events.
RAI 9. On page 8 of Attachment 2:
Under the heading " DESIGN TRANSIENT EVALUATION" you state "Using the Thermal Hydraulic Reactor Coolant System Parameters listed in Table 1, an evaluation of -
the Design Transients used for the Reactor Coolant System Component Fatigue Analysis was conducted for the South Texas Project. The alternate full power nominal Tavg range coupled with the alternate Reactor Coolant System flow continue to maintain the plant T-hot and T-cold values within the operating range defined for the current plant design basis used in the development of the NSSS design transients. Therefore, the transient response of the NSSS parameters used in the component fatigue analysis (T-hot, T-cold, Reactor Coolant System flow, Reactor Coolant System and pressurizer pressure, pressurizer spray and surge flow, and steam and feedwater flows and temperatures) do not need to be revised from those included in the present design transients. There is also no impact on the installed capacity of the major Reactor Coolant System pressure relieving devices (i.e., pressurizer spray, pressurizer power-operated relief valves and safety valves, and the steam generator safety valves); the Reactor Coolant System flow changes do not require changes in the required relieving capacities."
In Table 1, which is labeled "NSSS Power Capability Parameters With a 3% Reduction In Thermal Design Flow,' the parameters are provided for two temperatures,590 F Tavg and 583.2 F Tavg. For each of these two temperatures there are parameter values for two steam generator tube plugging (SGTP) conditions of 0% and 10%.
- a. Discuss the basis for the two temperatures (590 F and 583.2 F).
- b. For the two changes in Tavg (590 F and 583.2 F) in Table 1 [of Reference 2]:
- 1) Please discuss the methodology for determining the change, and the reasons for the change, in reactor coolant system temperatures for the 6 affected parameters listed in Table 1 (of Reference 2]. Also, it is indicated that SGTP does not affect the parameters. Please explain.
Westinghouse Proprietary Class 3 Page 9
- 2) Please discuss the methodology for determining the change, and the reasons for the change, in steam pressure, steam temperature, and steam flow. Also, it is indicated that SGTP does affect the parameters. Please explain.
Response to RAI #9. Part a The response to this RAI is contained within the " Background Information" section above. (para. 2.8)
Response to RAI #9. Part blI)
The response to this RAI is contained within the " Background Information" section above. (para. 2.3 and 2.8)
Response to'RAI #9. Part bl2)
The response to this RAI is contained within the " Background Information" section above. (para. 2.4)
RAI 10. On page 1 of Attachment 3, it is stated that the original Thermal Design Flow (TDF) rate was 381,600 gpm and the proposed new TDF rate is reduced by 3% to 370,000 gpm, a reduction of 11,600 gpm. You state that the new TDF rate is for alternate operations
" based on the 10% Steam Generator Tube Plugging limit of the current Licensing Basis."
Please explain how the i1,600 gpm reduction in TDF is obtained. If from reduced margins please provide the tabulation of items from where the 11,600 gpm was gained.
Has this reduction of i1,600 gpm been arrived at by a recalculation of analysis codes? If so, please provide the name(s) of the approved code (s) used.
Response to RAI #10 The response to this RAI is contained within the " Background Information" section above. (para. 2.7) 4.0 Response to Request for AdditionalInformation (Reference 5) i RAI 1. Discuss the effects of the reduced reactor coolant system (RCS) flow on the design l transients that were used in the fatigue analysis evaluation of affected components with j respect to the number of occurrences for each transient.
l Resoonse to RAI #1 l
The number of postulated occurrences of the design transients was not impacted as a j result of the 3% reduction in flow. Typically, the number of occurrences are based on the l category of the transient (i.e. normal, upset, emergency, faulted or test). Therefore, the l l number of occurrences of the design transients has not changed for the flow reduction i evaluation.
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RAI 2. Provide a summary of the evaluations (including analytical methodology, assumptions, and maximum stress and fatigue usage factors) for the structural integrity of the reactor vessel and internals, RCS piping, control rod drive mechanisms (CRDMS) and housing, pressurizer, surge line (stratification), pressurizer spray nozzles, steam generators (SGs),
reactor coolant pumps (RCPs), and pressurizer power-operated valves and safety valves.
The requested summary should discuss the evaluations based on the design parameters (i.e., RCS pressure, hot leg temperature (T-hot), cold leg temperature (T-cold),
differential pressure, differential temperature, and steam pressure and temperature) for the current operation, the proposed operation, and the design basis analysis.
Response to RAI #2 Table 5 contains a summary of the information requested on the NSSS system and component evaluations. These evaluations were performed using the plant operating parameters (RCS pressure, Thot, Tcold steam pressure, etc.) found in Table 1 above.
Please see " Background Information' section 2.6 and Table I for detailed information on the current, proposed and original design basis parameters. As is summarized in Table 5 below, there are no changes to the results (maximum stress and fatigue usage factors) for these areas as a result of the flow reduction, as each one is bounded by the existing UFSAR analysis or evaluation. Therefore, since there is no reanalysis required for the reduced flow conditions, our evaluations do not enable us to break down the individual contributions to stresses or fatigue usage as requested.
RAI3. The decrease in steam pressure results in the increase in the primary-to-secondary pressure difference and the break flow inside the SGs. Discuss the evaluation of the effect of decreased steam pressure on the design basis analysis of the SGs with respect to the stresses and fatigue usage.
Response to RAI #3 As is documented in the response to RAI #2 of Section 4.0 above, the steam generators have been evaluated for the reduced RCS flow, and it was concluded that the range of conditions used in the existing UFSAR evaluation (see Table 1) remain bounding over the revised operating conditions, with no changes in the stresses or fatigue usage. This conclusion, along with the applicable ASME code, is documented in Table 5 above.
Therefore, since there is no reanalysis required for the revised conditions, our evaluations do not enable us to break down the individual contributions to stresses or fatigue usage as requested.
RAI4. On page 7 of 12 of Attachment 2, there is discussion on steam generator tube rupture associated with the proposed 3% reduction in thermal design flow. Please provide information regarding any change in the mass flow of the steam / water release by indicating if there is any increase or decrease, and if so, how much. Please provide supporting background information.
Westinghouse Proprietary Class 3 Page11 Response to RAI #4 No steam generator tube rupture (SGTR) analyses were performed to determine changes in mass Dow rates resulting from the reactor coolant system (RCS) flow reduction program. The analysis assumptions that are impacted by the flow reduction were identified and the impact of the changes evaluated. The evaluations for the impacted areas are summarized below.
The Thermal Design Flow reduction program lowers the maximum allowable Tavg and increases the minimum allowable Tavg. This change in operating temperature associated with the Thermal Design Flow reduction provides a benefit for both the margin to steam generator overfill and offsite dose analyses. This conclusion is based on analyses performed for the South Texas Thot reduction (Reference 1 of this attachment), which concluded that the reduction in operating temperatures was limiting for the margin to steam generator overfill analysis, while the higher operating temperature remained limiting for offsite dose considerations.
The reduction in RCS How increases the initial mass in the steam generators by less than 0.5% and reduces the initial steam pressure by less than 10 psi. Based on engineering judgment, both of these effects would have an insignificant impact on the SGTR transients (both margm to SG overfill and offsite doses). The change in RCS flow itself has only a minimal impact on the transient. This is because the case where the RCPs are tripped due to an assumed loss of offsite power following reactor trip is bounding.
- The time of reactor trip could be impacted by the RCS flow reduction, because of the change in the Over-Temperature Delta-T (OTAT) setpoint equation. An analysis performed modeling the revised OTAT setpoints showed that the combined effect of the reduced RCS flow and revised setpoints was a delay in reactor trip relative to the licensing basis analysis. A later reactor trip provides a benefit for both the margin to SG overfill and offsite dose analyses. This is addressed in Reference 6.
Based on these evaluations, it was concluded that although the changes in steam / water releases had not been quantified, the proposed reduction in RCS flow rate would result in SGTR analysis results that are bounded by the current licensing basis analysis.
RAI 5. On page 5 of 12 of Attachment 2, there is discussion on the radiological consequences of a design-basis loss-of-coolant accident. Please verify that the 3% reduction in thermal design flow does not change any of the input parameters in the radiological consequences evaluation.
Westinghouse Proprietary Class 3 Page 12 Response to RAI #5 The 3% reduction in thermal design flow does not change any of the input parameters in the radiological consequences evaluation. The LOCA radiological source term is based upon Regulatory Guide 1.4 and TID-14844 methodology (" Calculation of Distance Factors for Power and Test Reactor Sites," TID-14844, J. J. DiNunno, R. E. Baker, F. D.
Anderson, and R. L. Waterfield (1962)). As such, the release is based upon core isotopic inventory. A reduction in reactor coolant flow would not impact isotopic generation in the core.
Also, a reduction in reactor coolant flow would not impact the isotopic concentration in the reactor coolant system. The isotopic concentration of the reactor coolant system is used to determine a component of the post-LOCA dose from the release path via the supplemental purge valve. The system isotopic concentrations are a function of various parameters (e.g. system volume, volumetric cleanup rate, core release rate). However, system flow rate has a negligible effect on the isotopic concentration of the reactor coolant system.
5.0 References
- 1. Correspondence from S. L. Rosen (HL&P) to NRC Document Control Desk, " Proposed Licensing Amendment Concerning Technical Specifications Based Upon Nuclear Fuel Upgrade, Plant Safety Evaluation and Revised Thermal Design Procedure," dated May 27, 1993 (ST-HL-AE-4364).
- 2. Correspondence from T. H. Cloninger (STP Nuclear Operating Company to NRC Document Control Desk, " Proposed Amendment to Technical Specifications 2.1,2.2, and 3/4.2.5 for Alternate Operation with Reduced Measured Reactor Coolant System Flow," dated December 31,1997 (NOC-AE-000050).
- 3. Correspondence from T. H. Cloninger (HL&P) to NRC Document Control Desk, " Proposed Amendment to Technical Specification Table 2.2-1 and 3/4.2.5 for Reactor Coolant System Flow Monitoring - Revised," dated August 6,1997 (ST-HL-AE-5707).
- 4. Correspondence from T. Alexion (NRC) to W. Cottle (STPNOC), " Request for Additional Information on Alternate Operation with Reduced Measured Reactor Coolant System Flow, South Texas Project, Units I and 2 (TAC Nos. MA0350 and MA0351)," dated April 8,1998.
l l 5. Correspondence from T. Alexion (NRC) to W. Cottle (STPNOC), " Request for Additional l Information on Alternate Operation with Reduced Measured Reactor Coolant System Flow, l South Texas Project, Units 1 and 2 (TAC Nos. MA0350 and MA0351)," dated May 15,1998.
- 6. WCAP-12369, "LOFTTR2 Analysis for a Steam Generator Tube Rupture for the South Texas Project Units 1 and 2," September 1989.
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Westinghouse Proprietary Class 3 Page 13 TABLE 1 SOUTII TEXAS PROJECT NSSS POWER CAPABILITY PARAMETERS FUEL UPGRADE /Tuor REDUCTION PROGRAM i 2 3 4 5 Parameter @ 1007e Original 593'F T.,, 593 F T. , 582.3 F T.,, 582.3 F T.,,
Power Design Basis Orfc SGTP 10% SGTP 0% SGTP 107c SGTP NSSS Power [MWt) 3817 3817 3817 3817 3817 Reactor Power IMWtl 3800 3800 3800 3800 3800 Thermal Design Flow 95,400 95,400 95,400 95,400 95,400
[gpm/loopl Core Bypass Fraction [%) 4.5 4.5 4.5 4.5 4.5 Reactor Coolant System 2250 2250 2250 2250 2250 Press. Ipsial Reactor Coolant System Temperatures [ F) 1 Core Outlet 628.3 628.3 628.3 618.4 618.4 l Vessel Outlet 625.6 625.6 625.6 615.6 615.6 Vessel / Core Inlet 560.4 560.4 560.4 549.0 549.0 Core Average 596.5 596.5 596.5 585.6 585.6 Vessel Average 593.0 593.0 593.0 582.3 582.3 Steam Generator Outlet 560.1 560.1 560.1 548.7 548.7 Zero Load Temp. [ F] 567.0 567.0 567.0 567.0 567.0 Steam Pressure [psial i100 1100 1081 995 975 Steam Temp. [ F) 556.3 556.3 554.1 543.9 541.5 Steam Flow,10'[lbm/hr] 16.96 16.96 16.94 16.87 16.86 Assumed Steam Generator 0 0 10 0 10 Plugging Level [%)
Feedwater Temp. [*F] 440 440 440 440 440
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