ML20211B261
| ML20211B261 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 09/16/1997 |
| From: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Collins S NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20046D804 | List: |
| References | |
| CAW-97-1170, NUDOCS 9709250135 | |
| Download: ML20211B261 (35) | |
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September 16,1997 CAW-97-1170
- Document Control Desk
- U.S. Nuclear Regulatory Commission Washington, DC 20555
" Attention: Mr. Samuel J. Collins _
APPLICATION FOR WITHHOLDING PROPRIETARY ,
INFORM ATION FROM PUBLIC DISCLOSURE
Subject:
"Farley Uprate RAI," SAE-SSO-97 Il6 (Proprietary)
Dear Mr. Collins:
The proprietary information for which withholding is being requested in the above-referenced report is further identified in Aftidavit CAW-97-Il70, signed by the owner of the proprietary information, Westinghouse Electric Corporation, ne aftidavit, 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.
Acccrdingly, this letter authorizes the utilization of the accompanying Affidavit by Southern Nuclear Operating Company.
Correspondence with respect to the proprietary aspects of the application for withholding or the L Westinghouse affidavit should reference this letter, CAW-971170, and should be addressed to the undersigned.
i Very truly yours, 1 [-
9709250135 970922 E ** '"' an ger PDR ADOCK 05000348 P pon Equipment Design and Regulatory Engm, eerm, g Enclosures
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cc: Kevin Bohrer/NRC (12H5) __
"The mission ofNSD is to provide our customers with people, equipment and urtices that ut the standards ofexcellence in the nuclear industry."
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CAW-97-1170 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:
ss COUNTY OF ALLEGHENY:
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~ Before me, the undersigned authority, personally appeared Nicholas J. Liparuto, 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 Corporation (" Westinghouse") and that the averments of fact set forth :
! -- -in this Affidavit are true and correct to the best of his luewledge, information, and belief:
Nicholas J. Lihulo, Manager Equipment Design and Regulatory Engineering Sworn to and subscribed before m'this /4 Mday of, /e N_e/M.<- ,1997-( -
NotNial Seal 1 i n12 '
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Notary Public
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d 2 CAW-97-1170 (1) - l'am Manager, Equipment Design and Regulatory Engmeering, in the Nuclear Services Division, of the Westinghouse Electric Corporation and as such, I have been specifically
- delegated the function of reviewing the proprietary information sought 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 Westinghouse 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 cJt he Commission's regulations, the following is furnished for consideration by th, ammission in determining whetter 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 customarily 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, L utilizes a system to determine when and whether to hold certain types ofinformation in confidence. The application of that system and the substance of that system constitutes lL L Westinghouse policy and provides the rational basis required.
l Under that system, information'is held in confidence if it falls in one or more of
! several types, the release of which might result in the loss of an existing or potential f
L competitive advantage, as follows:
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(a)= The information reveals the distinguishing aspec ; af a process (or component,
- i. structure, tool, method, etc.) where prevention c'its use by any of 2nenuwwn -
-3 CAW-971170 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.
(b) It consists of supporting data, including test dr.ta, 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 contains 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.
(b) It is information which is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.
(c) Use by our cor petitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.
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-4 CAW 971170 (d) Each component of proprieiary 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 puule, tbmhy depriving Westinghouse of a competitive advar.tage.
(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby givw a market advantage to the competition of those countries.
(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.
(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 Yarley Uprate RAI," SAE SSO-97 Il6 (Proprietary),
September,1997 for Farley Units I and 2, being transmitted by Southern Nuclear Operating Company letter and Application for Withholding Proprietary information from Public Disclosure, to the Document Control Desk, Attention Samuel J. Collins.
The proprietary information as submitted for use by Southern Nuclear Operating Company for Farley Units 1 and 2 is expected to be applicable in other licensee submittals in response to certain NRC requirements for justification of the use of the instrument uncertainty methodology.
This .information is part of that which will enable Westinghouse to:
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(a) Provide application of the uncertainty methodology.
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(b) Provide Farley specific uncertainties used in the application of uncertainty methodology. .
i (c) Assist the customer in obtaining NRC approval.
Further this information has substantial commercial value as follows:
(a) Westinghouse plans to sell the use of similar infonaation * ) its customers for purposes of meeting requirements for licensing documentation.
(b) Westinghouse can sell support and defense of the technology to its customers in the licensir;! 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 methodologies 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.
The development of the technology described in part by the information is the result of applying the results of many years of experience in aa 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 testing.
Further the deponent sayeth not.
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s, i Proprietary information Notice Transmitted herewith are proprietary and/or non proprietary vwslons of documents fornished to the NRC in connection with requats for generic and/or plant 4pecific review and approval.
In order to conform to the requitanents of 10 CFR 2.790 of the Conuniulon's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary vwsions la contained within brackets, and whwe the proprietary information has been deleted in the non proprietary vwslons, only the brackets remain (the information that was contained within the brackets in the proprietary ywslons having been deleted).
The justification for claiming the information so designated as proprietary is indicated in both versions by means of loww case letters (a) through (f) contained within parentheses located as a superscript immediately following the brackets enclosing each item of informadon belng identified as proprietary or in the margin opposite such information. These lower une lettes refw to the typw of information Watinghouse customarily Solds in confidence identified in Sections (4)(li)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.790(b)(1).
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Copyright Notice De reports transehead herewith each bear a Wutinghouse copyright notice, no NRC la pwmitted I to make the number of copies of the information contained la these reports which are necessary for its internal use in connection with generic and plant speel6c reviews and approvals as well as the lasuance, dealal, amendment, transfw, renewal, moJi6 cation, suspension, revocation, or violation of a license, permh, ordw, or regulation subject to the requirmnents of 10 CFR 2.790 regarding restriction on pabik disclosure to the entent such information has been identined as proprietary by Watingh:gse,'ccpyright protection notwithstanding. With respect to the non-proprietary vwsions of these repovis, the NRC la permitted to make the number of copies beyond those necessary for its -
laternal use which are necessary in ordw to have one copy available Ibr public viewing in the appropriate docket Sles la the public document room in Washington, DC and la local public -
document rooms as may be required by NRC regulations if the number of copies submined is insufficient for this purpose, Copies made by the NRC must include the copyright notice in all instances and the proprietary notics if the original was identined as proprietary.
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ATTACHMENT I SNC Response to NRC Request For Additional Information
- Related To Power Uprate Submittal . Joseph M. Farley Nuclear Plant, Units I & 2 SNC RESPONSES TO NRC QUESTION NOS.1 - 6 & 9 11
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SNC Response to NRC Request For AdditionalInformation Related To Power Uprate Submittal . Joseph M. Farley Nuclear Plant, Units 1 & 2 NRCRussiienEtl i
r%scuss whether the power uprate will change the type and scope of plant emergency and abnormal operating procedures. Will the power uprate change the type, scope, and nature of operator actions needed for accident mitigation and will it require any new operator actions?
SHC.Respenscho L ne type and scope of the Farley emergency response procedures and abnormal operating procedures are not impacted by power uprate. %c type and nature of operator actions needed for nccident mitigation will not change, and new operator actie is will not be needed for power uprate with the possible exception of the emergency response ATWS procedure and the emergency boration abnornal operating procedure. Due to an assumed increase in allowable charging /SI pump head degradation allowance ftom 8% to 10%, opening a pressurizer PORV may be required to support emergency boration in certain scenarios such as ARVS. His action is consistent with the generic guidance pro ided in the Westinghouse Owner's Group (WOG) Emergency Response Guidelines (ERGS). %c increase in allowable charging pump degradation is not directly due to power uprate but is a change that has been factored into the power uprate rumlysis, and therefore, appropriate procedure changes will be made in accordance with ERG guidelines.
Power uprate does impact several of the emergency response procedure setpoints for operator actions. As part of power uprate, evaluations and analyses were perfomied to identify the impacted cmcrgency response procedure setpoints and to develop the setpoint values applicable for power uprate. He emergency response procedures will be revised to incorporate these revised setpoint values prior to implementation of power uprate. Rese emergency response procedure setpoint revisions will not impact the type and scope of the emergency response procedures or the type and nature of the operator actions contained therein, Whn 9/16/97 A INP/rwu & cdc 9/11/97 NRCAlteslier No. 2 Provide examples of operator actions potentially sensitive to power uprate and address whether the power uprate will have any effect on operator reliability or performance. Identify operator actions that would necessitate reduced response times associated with a power uprate. Please specify the expected response times before the power uprate and the reduced response times. What have simulator observations shown relative to operator response times for operator actions that arc potentially sensitive to power uprate? Please state why reduced operator response times are needed. Please state whether reduced time available to the operator due to the power uprate will significantly aflect the operators ability to complete nuumal actions in the times required.
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SNC Response No. 2 i
ne power uprate analyses and evaluations performed for the Chapter 15 accidats and transients
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are described In Section 6.0 of the NSSS Licensing Report.- There were no changes made to l operator action assumptions in the Chapter 15 accidents and trar. Ants which resulted in reduced operator response times, nerefore, the operator's ability to co.nplete manual actions in the time required will not be significantly a%cted.
Although not a Chapter 15 revision, a change made to operator action assumptions concerning accident analyses was for the small break loss of coolant accident (SBLOCA). De presious .
. SBLOCA analysis modeled ECCS flow rates that included the fluid systems assumpdon of the i
- charging /SI pump miniflow isolation valves being closed. Consistent with this fluid systems assumptior, the previous SBLOCA analysis implicitly assumed operator action (with an operator action delay time allowance of 10 minutes) to close the charging /SI pump miniflow isolation -
valves. His action is in concert with the current Farley emergency response procedures. The power uprate SBLOCA analyses modeled ECCS flow rates that include the conservative fluid a systems assumption that the charging /SI pump miniflow isolation valves are open. Consequently, the SBLOCA analysis for power uprate no longer implicitly assumes operator action to close the charging /SI pump miniflow isolation valves. The Faricy procedures will continue to require ,
manual miniflow isolation. %c SBLOCA analysis operator action assumption change is not a ;
i direct result of power uprate but is related to a conservat:ve input for the analysis performed to develop ECCS flow rates for use in power uprate analysis.
With respect to the emergency response procedures, operator actions potentially sensitive to power j uprate are those that are performed based on setpoint values that are calculated using the design ,
parameters for power uprate as shown in Tabic 2.1 2 of the NSSS Licensing Report, including reactor thermal power, RCS hot leg temperature, and steam generator secondary sido pressure. ,
Changes to these design parameters can impact the setpoints calculated for operator actions but should not impact the type and r, cope of operator actions. For exampic, the change to core power potentially impacts operator actions that control ECCS flow to remove co*e decay heat and operator actions to prepare for and initiate switchover from cold leg to hot leg recirculation.
Changes to RCS hot leg temperature potentially impacts operator actions that are based on RCS saturation pressure corresponding to full power RCS hot leg temperature. As stated in the SNC response to NRC question No I above, applicable emergency response procedures resisions will be incorporated prior to implementation of power uprate.
Operator actions in nomal and abnormal operating procedures that are based on core power related setpoints may also be impacted. Examples include operator actions such as initiation of mid loop operation and vacuum fill operations which are dependent on decay heat. Additionally, in response to failure of certain components such as a main feedwater pump or a circulating water pump, operator action may require power to be reduced to a specified power level. The power level specified in the Farley operating procedures will be evaluated, and where appropriate,
. changes will be madec Also, current operating restrictions for removal of feedwater heaters during -
power operation are intended to prevent a reactor overpower condition due to addition of cold l feedwater and to ensue that the turbine blading groups do not exceed the design nutximum flow.
pressure drop or work output. Power uprate will necessitate appropriate revisions to these operating restrictions. ,
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NRC Ountion No.1 i i
Discuss any changes the power uprate will have oc control room instruments, alamis, and displays, j Are zone markings on meters changed (c g., normal range, marginal range, and out of tolerance- l range)7 SNC Response No. 3 Engineering is presently dcrming the scope of control board indicators and alarms and computer points that will be impacted by the Farley power uprate. Required changes will be implemented.
Preliminary engineering reviews indicate that power uprate will have minimum impact on the control instruments, alanns, and displays. Examples of potential changes are discussed below.
One example of a potential indicator color coding change is steam line pressure. Presently these indicators are color coded from about 780 to 1000 psig in green to indicate the expected normal operating steam pressure range associated with full power and hot standby no-load operations. In that the uprated best estimate steam pressure used in the turbine modification design is 787 psia, it is anticipated that the low cnd of the green band will be adjusted to approximately 770 psig.
One exampic of a potential main control board alami change is the setpoint for the RCS liigh Tavg annunciator. He setpoint for this alarm is 4 'F above the nominal operating Tavg at full power.
Ilecause Farley has allowed for a Tavg range (567.2 to 577.2 'F) at full power in the uprate analyses, this alann setpoint will have to be evaluated for a possible serpoint change aner unit steady-state full power operation is achieved and the full load reference temperature (Tref) is established. Dased on current SO plugging, the projected full power Tavg for the first uprated cycle should be within 3 'F of the current operating Tavg at full power.
An example of a potential plant computer clarm/ display change is the steam generator feed pump (SGFP) low suction pressure setpoints. Presently, the computer alarm setpoints are as follows: low alert at 350 psig (ycilow); and low alarm at 340 psig (red). These computer alarms allow advance warning of decreasing suction pressure prior to initiation of automatic control system actuations (i.e., start of the standby condensate pump and time delay trip of the SGFP). In conjunction with the condensate pump modifications, engineering established a new automatic actuation setpoint of 275 psig for the condensate pump start and feed pump trip. As such, it is anticipated that the associated computer-generated low suction pressure alann setpoints will also change.
SCS/jth & SNC/mse 9/17/97 NRC Ouestion No. 4 Discuss any changes the power uprate will have on the Safety Parameter Display System.
SNC Response No. 4
%c SPDS computer point list has been reviewed by the plant staff. The conclusion of the resiew is that no SPDS setpoint changes are u ticipated at this time. There is, however, a potential for changes in the plant process computer and/or SPDS scaling / calibration curves. The high/ low A1Tl 3
1 alarm limits for some uprate afTected instrumentation inputs that are non SPDS computer points may change as well. Any required changes to scaling and alann limits will be implemented.
lW1'/rdt A SCS/jth 09/17/97 NRC Ountion No. 5 Describe any changes the power uprate will have on the operator training program and the plant simulator. Provide a copy of the post modification test report (or test abstracts) to document and support the effectiveness of simulator changes as required by ANSI /ANS 3.5-1985, Section 5.4.1.
Specincally, please propose a license condition and/or commitments that address the following:
(a) Provide classroom and simulator training on the power uprate modification.
(b) Complete simulator changes that are consistent with ANSI /ANS 3.51985. Simulator fidelity will be revalidated in accordance with ANSI /ANS 3.5 1985, Section 5.4.1,
" Simulator perfonnance Testing." Simulator revalidation will include comparison of individual simulated systems and components and simulated integrated plant steady state and transient performance with reference plant responses using similar startup test procedures.
(c) Complete control room and plant process computer system changes as a result of the power uprate.
(d) Modify training and plant simulator relative to issues and discrepancies identified during the star:up testing program.
SEC_ Response No. 5 The Failey simulator is Unit I referenced; therefore, the final simulator modifications will be i implemented following the Unit I uprating. To facilitate training, the simulator will be temporarily modified to implement only those uprate changes made to systems directly affected by the Unit 2 power uprate in the spring of 1998. %csc simulator r.hanges may include hardnure modifications to control board meters as well as software changes associated with instrumentation scaling, alarms, setpoints, pump performance characteristics and system operating parameters, as well as modi 0 cation to the reactor core physics data. Additional changes may be required due to the new accident analyses (e g., containment temperature & pressure response) which were perfonned in support of the uprating De need for such simulator changes will be evaluated and implemented at a later date consistent with ANSI /ANS 3.51985. Following ANSI /ANS 3.51985 required testing, the post modification tests results will be provided for your review, if determined to be necessary. The initial post modification testing will be limited in scope since the Farley simulator is Unit I referenced and the changes will be removed following Unit 2 startup training.
(a) %c Farky training staff will provide classroom and simulator training on the power uprate changes for the Unit I and 2 operations crews prior to the Unit 2 startup in the spring of 1998. As in the past, Farley will pro ide positive moderator temperaturc l
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coefficient startup training on the simulator for those operations crews involved in BOL Unit startup following refueling.
(b) Using existing Simulator Certification Tests, consistent with the requirements of ANSI /ANS 3.5 1985, changes to the simulator will be validated against best estinute data provided in the applicable design change package (s) and/or the supporting NSSS and BOP cngineering reports for power uprate before conducting startup training.
(c) Those changes made to the main control board and plant process computer system as a result of the power uprate will be temporarily implemented, where applicable, on the simulator prior to the Unit 2 startup training in the spring of 1998. The final modification will be implemented following the Unit i 1998 fall outage in accordance with ANSI /ANS 3.5 1985.
(d) After final mcxlification, tha simulator response will be further evaluated against actual unit operating data collected after the Unit I uprate when such data becomes available for comparison. In addition, new accident analyses conducted for the uprate will be evaluated to determine if any simulator malfunctions are affected. This data will then become part of the normal process for all future certification testing requirements, lHP/gpc & Idm - 09/I7/97 NRC Ouestian No. 6 Provide a proprietary and non-proprietary version of WCAP-10263, "A Resiew Plan for Uprating the Licensed Power of a Pressurized Water Reactor Power Plant,"(1983).
SNC Response No. 6 A copy of Westinghouse WCAP-10263, "A Review Plan for Uprating the Licensed Power of a Pressurired Water Reactor Power Plant,"is provided for your information in Attachment 111 of this letter. This is a non proprietary Westinghouse report; there is no < roprietary version.
SNC/mre 09/IM97 NRCRusstion No 9 Provide studies or calculations performed to support that the electrical load increases and voltage changes are minimum throughout the onsite and offsite (switchyard) power system at Farley for power uprate (i.e., one line diagram that plotted load flows and voltages for before and after powrr uprate).
SNC Regengh_9 Attached, Figures 1 - 4 are provided to illustrate that the results of the electrical load and voltage changes throughout the onsite and offsite (switchyard) power system as a result of power uprate are minimal. These figures, i.e., one line diagrams, illustrate representative cases studied for Unit
- 1. These diagrams show key load flows and voltages for pre-uprate conditions and the load flows and voltages [due to the increase in load on the Reactor Coolant Pump (RCP) and Condensate A1TI5
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Pump motorsj as a result of power uprate. Because the distribution system configuration for both j Units are similar, the Unit 2 case results are comparable. Therefore, for simplicity, only the Unit ! l results are shown.
The 4160V clectrical distribution system is c: nprised of two sections, normal and emergency. )
%c normal buses (A, B, C, D, and E) supply nonsafety-related loads and have two soure:s of ;
offsite power available - cither the Unit Auxiliary Transformers A and B or the Startup (S/U)
Transformers A and B. %e emergency buses (F. O,11, J. K, and L) receive power from the S/U Transformers A and B and supply equipment essential for the safe shutdown of the plant. He emergency buses are further divided into two independent, redundant sets of buses referred to as train A (F,11, and K) and train B (G, J, and L). The train A buses roccive power through S/U Transformer A with an alternate supply through S/U Transformer B. Bus F supplies power to buses 11 and K. %e train B buses receive power through S/U Transformer B with an alternate supply through S/U Transformer A. Bus G supplies power to buses J and L.
The cases illustrated in Figures 1 - 4 assume that the 230kV switchyard voltage is at its minimum expected value of 230kV (100% of 230kV). %c 230kV voltage is nonnally controlled between 235kV (102.2% of 230kV) and 239kV (103.9% of 230kV). De mininmm expected value of 100% is based on system studies performed by Southern Company Senices (SCS) Transmission Planning, which assume both Unit I and Unit 2 are tripped with maximum station senice loading on the Farley buses.
Note that all steady state voltages are expressed in % of bus rated voltage (4.16kV or 600V) on Figures I and 4. Motor starting voltages are expressed in % of motor rated voltage (4000V or 575V) on Figures 2 and 3.
Figure 1," Unit i Voltage & Load During LOCA Steady State Conditions," illustrates LOCA steady state conditions with all the 4160V buses aligned to the S/U Transformers. His case conservatively assumes the RCP and Condensate Pump motors are nmning and aligned to the S/U Transformers.
Figure 2, " Unit i Typical Starting Voltages for LOCA Motor Loads," shows typical LOCA starting voltages a safety-related loads with all the 4160V buses aligned to the S/U Transformers.
He values shown are representative of the change in voltages expected at the other safety-related loads not shown for simplicity. These voltages are calculated assuming a simultaneous start of the loads that receive a safety injection (SI) signal with no intentional time delay. His case conservatively assumes the RCP and Condensate Pump motors are running and aligned to the S/U Transformers.
Figure 3," Unit 1 Motor Starting of RCP & Cond. Pumps with 4kV Buses Fed from Start Up Auxiliary Transfonners," shows the change in the motor starting voltages for the RCPs and the Condensate Pumps with all the 4160V buses aligned to the S/U Transformers.
Figure 4," Unit i Voltages at RCPs & Cond. Pump with Buses A, B, C on Unit Auxiliary l Transfonners (Normal Operation)," shows the r.hange in the minimum steady state voltages for the l RCPs and the Condensate Pumps with 4160V buses A, B, and C aligned to the Unit Auxiliary
! Transformers. Rese voltages are calculated assuming the generator is at its minimum rated voltage (95%). .-
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Since no new loads or alignments are involved, short circuit levels are not impacted and are, therefore, not included.
Consequently, as illustrated by the attachal figures, the electrical load and voltage changes throughout the onsite and offsite (switchyard) power system as a result of power uprate are minimally impacted when compared to the as-built or pre-uprate values. Due to the similaritics in the Unit I and Unit 2 Station Auxiliary Electiical Distribution Systems, the results shown for Unit I apply to both Units. The Unit I and Unit 2 Station Auxiliary Electrical Distribution System will not be adversely impacted by the proposed power uprate of Plant Farley, and plant equipment raluired Ibr safe shutdown will continue to be ahic to perform their safety related functions.
SCS/ujb & tic .09/09/97 NRC_Quntion No. 9 Supplement (Refergnec August 26.1997 NRC/SNC Conferenge Call)
For power uprate, SNC gave a peak expected uprated turbine generator MWe output value of 920 h1We. SNC also stated that the delta between the existing output and the uprated output is approxinutely 25 MWe. Ilowever, upon reviewing Farley design information, the original rating was listed as 861 MWe. Please explain the apparent differences.
SMC.ResponssjQurstinn No. 9 SuppJrmcal The original gross electrical output for Farley was 861 MWe (ref. FSAR Section 1.1.5.2), his is a design number based on the original turbine rating @ 750 psia throttle steam pressure and 3.5" lig absolute exhaust backpressure. Typical operating values were higher due to higher operating steam pressures and lower operating backpressures. In addition, modifications to the llP turbine and main steam reheaters in the 1980's improved the turbine efficiency and optimized the MWe output (ref. Il0P Uprate Licensing Report Section 2.4). Calculations today using measured data yield a predicted output of approximately 885 MWe @ 2.8"lig absolute (average annual ambient backpressure); note that actual operating values vary. After power uprate, the predicted output value is approximately 910 MWe @ 2.97" lig absolute (average annual ambient backpressure).
The difference beturen the uprate value of 910 MWe c * *he existing value of 885 MWe is 25 MWe.
A peak value of 920 MWe was used for conservatism in the uprate electrical analyses described in the 110P Licensing Report.
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NRC Ourgien No.10 Provide summary (case descriptions and fmdings) of system stability cases that were run for the power uprate.
SNC Response No.10 The Farley off-site power system is designed to prevent a complete loss of preferred power (LOSP) due to a single event such as electrical fault, loss of a generator, loss of load, or loss of a transmission line. Stability studies simulating these events are performed to verify this, in additica, in order to optimize the stable operation and design of the FNP units and the system grid, severe but unexpected transients are studied such as those due to three-phase faults with the
. additional contingency of breaker failure. Grid stability is maintained if, after a disturbance such as a fault, the power system returns to equilibrium without experiencing cascading trips oflines that could result in the system or voltage collapse.
He attached tables provide a summary of the Farley Nuclear Plant stability cases that were run specifically for comparison before and after power uprate. He attached figure (FSAR Figure 8.2-1, Switchyard Arrangement One-Line Diagram)is a one line diagram of the switchyard arrangement.
Cases R0 and R0A in Table 1, " Single Event Cases Used to Validate FSAR Commitments,"
represent the limiting cases for demonstrating that the grid will remain stable and safety-related buses will continue to be supplied by the off site preferred power source for single contingency events and faults. Case R0 (close-in normally cleared 3 phase fault on Webb 230 KV line) represents the worst single contingency case for a 230 KV line, and case R0A (close-in normally cleared 3 phase fault cn North Tifton 500 KV hne) represents the worst single contingency case for a 500 KV line. Both cases assume all Farley transmission lines are in senice, which is the normal configuration. The results of these cases demonstrate transient stability is maintained for the expected power uprate load level and system load levels above 12 GW. The lowest system load during 1996 was 12.6 GW, Both of these cases are worse than other single events such as loss of a generator, loss ofload, or loss of other transmission lines.
Cases RI through R5 in Table 2," Operational Cases Studied to Assess and Optimize Unit and Grid Stability," represent the cases studied to assess and optimize the stable operation and design of the FNP units and the system grid For double contingency events such as fault plus breaker failure or a line out with a close-in 3 phase fault, sensitivity cases for various system load levels show that FNP may experience unit trips at higher system load levels afler uprate than today.
Plant procedures FNP 1/2 UOP 3.1, " Power Operation," provide guidance for operation with a transmission line out of senice and both Units on line. Where applicable, limits are given for hasing both Units' power system stabilizers (PSSs) in senice and for having either PSS out of senicc. Rese procedures will be revised to reflect the new limitati.ms attributed to the Farley uprating.
I SCS/jms & tle - 09/09/97 ATft 8
Stability Comeparisons of Farley Before and After Uprate Table 1 - Single Event Cases Used to Validate FSAR Comesnitanests Csee Castiegency Descripties Femk Femk Type SES Imed Failed Bresher Today A9erUprete 'i No. Imentsee Inwlle CW RO Webb 230 kV1me fault Close4n 3P11 12 N/A StaNe Stable ROA N. Tifkm 500 kV 1me fault Close-In 3PH 12 N/A StsNe Stable M.
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n-Stability Cotuparisons of Farley Before and After Uprate Table 2 - Operational Cases Stadied to Assess and Optimize Unit and Grid Stability Case Centsagency Descriptics Fas!t Faek Tne SES Imd Failed Breaker Tedey After Upeace No. tecation leelle GW RI N. Tifton 500 kV 1me fault Close-In 3Pil 12 N/A- Umt l&2 tnp Urut 1&2 tnp Snowdoun 500 kV Ime out N. Tifton 500 kV 1me fault Close-In 3Pil 13 N/A Stable Umt IA2 tnp ,
Snomdoun 500 kV 1me out N. Tafton f00 kVIme fault Close4n 3Pli 19 N/A- StaNe Umt 1&2 tnp Snowdoun 500 kV 1me out N. Tifton 500 kV Ime fault Close4n 3PII J) N/A Stable StaNe Snowdoun 500 kV line out R2 Snowdoun 500 kV Ime fault Close4n 3PII 12 N/A StaNe Umt I A2 tnp N Tifton 500kVimeout Snowdoun 500 kV 1me fault Close4n 3PIi 16 N/A StaNe Umt I A2 tnp l N. T ifton 500 kV Ime out L
> Sne 6doun 500 kV 1me fault Close4n 3PII 17 N/A StaNe Stable ,
4 N Tifton 500 kV1me out i 7 R3 S. Ik inbridge 230 kV Ime fault Close4n 3P1I w1BF 12 826 Umt 1&2 tnp Umt 112 tnp '
- 5 S. B sinhndge 230 kV 1me fault Close4n 3P1I m1BF 13 826 StaNe Umt 1&2 tnp S. B unbridge 230 kV 1me fault Close4n 3Pflw/BF 17 826 StaNe Umt I tnps Unit 2 stable i S. BU W 230 kV 1me fault Close4a 3P1Iw/BF 24 826 StaNe Umt I tnps i Unit 2 sesNe j S. Bai.tbndre 230 kV line fault Close4n 3PH w/BF 25 826 StaNe StaNe '
R4 Fault sn 230 kV Sxie of Auto Close4n 3Pli w/BF 12 8% StaNe Unit 1&2 tnp l Fault on 230 kV Side of Auto Close4n 3P11w/BF 13 8% StaNe Umt I tnps [
- Umt 2 staNe l Fault on 230 kV Side ofAuto Close4n 3Pliw/BF 24 8% Stable Umt I trips
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i NEC_Qncition No. I1 in your latest response No. 3 on page 23, you stated that "[ajlthough the power uprate composite temperature profile in Section 3 exceeds the existing composite temperature profile by approximately 5 'F, a review of the test profiles for EQ [ equipment qualification] equipment inside contairmient indicates that there is sufficient margin in the test profiles to envelop the power uprate composite temperature profile." Provide a representative EQ test profile curve that demonstrates that the EQ test profile is still bounded by the power uprate composite temperature profile by superimposing it with the latest "FNP Composite LOCA/MSLB Containment Temperature Profile" submittal.
Sd.C.Effp9R$eNo.II In general, EQ test programs are not performed for plant specific parameters. Moreover, there is no single representative EQ test profile. For example, in some equipment suppliers test programs, thnceident test durations are shorter than the required plant specific post accident equipment cperating time while others exceed the plant specific requirements. Figure 1," Comparison of Typical < 304ay Test Profile vs. Power Uprate Composite Profile," is typical of a shorter duration test, whereas, Figure 2 " Comparison of Typical 30-day Test Profile vs. Power Uprate Composite l'rofile,"is typical of a 30-day test. In both figures, the power uprate composite profile has been superimposed. For the shorter duration test, an equivalency evaluation using Arrhenius metinlology is perfonned (in accordance with the guidelines provided by EPRI and Regulatory Guide 1.89, Rev.1, Section C.5.b) to extend the test duration to the Farley specific accident requirements. As can be observed from these typical profiles, there is adequate margin between the test profiles and the power uprate composite temperature profile.
II/d AJcl 09/ll/97 J
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l 3 WES11NGllOUSE NON-PROIARIETARY CLASS 3 '
?TAK!2Y UFRATE RAl* : SAE-SSO-97117 ~ l SNC Response to NRC Request For Additional Information -
Related To Power Uprate Submittal Joseph M. Fadey Nuclear Plant, Units 1 & 2
' NRC Ouestion NoL7 ;
How have you ensured that the revised instrument setpoints are within the safety limits? .
SNC Response No. 7
, _ As part of the uprate etTort for Farley, safety analyses and/or evaluations were performed as described in Section 6 of WCAP-14723, " Power Uprate Project NSSS Licensing Report, Farley Nuclear Plant Units 1 & 2.". These analyses /cyaluations demonstrate that the reactor safety limits
- as defined in Section 2.1 of the Faricy Technical Specifications are not challenged for any FSAR i- accident. Some of these analyses / evaluations assume that one or more Reactor Trip System (RTS) i functions or Engineered Safety Features Actuation System (ESFAS) functions actuate to mitigate '
- the transient; For those analyses which explicitly model an RTS or ESFAS actuation setpoint, that actuation setpoint or Safety Analyses Limit (SAL) is given in Section 6 of WCAP-14723.
Each RTS and ESFAS function that was potentially impacted by the uprate was evaluated to <
ensure that the associated nominal trip setpoint provides positive margin to the S AL (where an -
- explicit SAL exists) after accounting for all known uncertaintics. Based on the evaluations, and in -
order to assure that the RTS and ESFAS functions will provide the appropriate response before the :
assumed S AL values are reached, setpoint uncertainty calculations were performed for selected RTS and ESFAS setpoints which were potentially impacted by the uprate. He uncertaintics were calculated using the Westinghouse statistical actpoint methodology as presiously applied in the analyscs and evaluations supporting the Farley Technical Specifications change to revise the OTAT and OPAT reactor trip setpoints. (Reference SNC letwr to NRC dated June 12,1996,
" Joseph M. Farley Nuclear Phnt Tecimical Specifications Change Request Resision to Core Limits and OTAT & OPAT Setpoints and implementation of RAOC," which was approved by
- NRC letter to SNC dated September 3,1996.) All known uncertainty terms associated with the sensors, signal processing equipment, and calibration methods were accounted for in the calculations. In addition, process measurement accuracy allowances and adverse emironmental allowances were included for setpoints which are subject to these effects. Only three setpoints were
. revised for the uprate: Steam Generator Level High-High; P-8 Interlock; and P-12 Interlock. These are discussed further in response to Quecion No. 8 below. All other nominal trip setpoints were
' found to be acceptable as currently specitled in the Faricy Technical Specifications. As a result of
- the calculations performed for uprate, the Allowable Values were revised to reflect equipment, calibration procedures, and calculation methodology refinements at Farlej.
W/wm A sva 9/)6/97 ERC. _Qaestion N18 Provide the following information for each changed mstrument setpoint: ,
(a) : The change in the setpoint.
A'!Til 1
.L,_ - . . - .= -. . - . . . ._ ,- - . .~ . . - , .- - ,
s Wl'STINGlIOtJSE NON-I'ROPRIETARY Cf. ASS 3
- TARIEY UI' RATE RAl" SAE-SSO 97117 (b) The calculated worst case drin, with the page reference to the calculation that covers the drift, including uncertainty calculation where applicable.
(c) h allowable value, (d) *Re safety limit, with the page reference to the document that specified the safety limit.
SNC Reponse No. 8 (a) As discussed in the response to question No. 7 above, the RTS and ESFAS functions which were potentially impacted by the uprate were evaluated and setpoint uncertainty calculations were performed for selected RTS and ESFAS functions. Sctpoint and Allowable Value changes resulting from the uncertainty calculations are shown in Tables 6.7-1 and 6.7 2 of WCAP 14723. Three nominal trip setpoints were revised for the uprate; the changes to the nominal trip setpoints and allowable values for these three functions are shown in Table i "Faricy Uprate Reactor Trip System Setpoint Changes,"
and Table 2 "Faricy Uprate Engineered Safety Features Actuation Systems Setpoint Changes," below.
(b) ne drin values along with the other terms accounted for in the uncertainty calculations for these three functions aic listed in Table 3 (Power Range, P 8 Interlock), Table 4 (Steam Generator Water level - liigh-liigh), and Table 5 (Tavg - Low-Low, P 12 Interlock) which follow. %c drin values for all the setpoints are allowances. The driR allowances have been determined to be conservative based on a qualitative assessment of Faricy as letVas found data.
(c) The allowable values associated with the revised trip setpoints were given in Tables 6.7-1 and 6.7 2 of WCAP-14723, and are repeated in Tables I and 2 below.
(d) P 8 Interlock: The SAL is 39.9% of Rated Thermal Power (RTP), and is an implicit assumption in the Loss of Flow analysis (Section 6.2.5 of WCAP-14723). Based on the revised nominal setpoint of 30% RTP and the total channel statistical allowance of
[ l'", as shown on Table 3, the margin to the S AL is [ ] *"
Steam Generator Water Level liigh-liigh: The SAL is 100% narrow range (NR) span as shown on page 6-63 of WCAP-14723. i;owever, a more restrictive (lower) limit has been placed on this channel to account for the impact on the level measurement system of void fractions at high water levels. Therefore, the limit used in evaluating the setpoint for this channel is 90.4% NR span, which is midway between the steam generator upper tap and the mid deck plate. Based on the revised nominal setpoint of 78.5% NR span and the total channel statistical allowance of[ l'", as shown on Table 4, the margin to the limit is 1. j '"
P-12 Interlock: This interlock (Low-Low Tavg) is not explicitly modeled with a S AL in any safety analyses.
AIT 11 2
WESTINOllOUSE NON PROPRILTARY CLASS 3 "FARI.EY UPRATE RAl" SAE-SSO 97117 TABLEI FARLEY UPRATE REACTOR TRIP SYSTEM SETPOINT Cl{ANGES Trip Setpoint Allowable Value Functional Power Uprate Power Uprate Unit 20.C Current Value Value Current Value Value Power Range 535% RTP $30% RTP $36% RTP 530.4% RTP Neutron Flux, P-8 Interlock TABLE 2 FARLEY UPRATE ENGINEERED SAFETY FEATURES ACTUATION SYSTEMS SETPOINT CIIANGES_
Trip Setpoint Allowable Value Functional Power Uprate Power Uprate Unit 5.a Current Value Value Current Value Value Steam Generator $79.2% of span $78.5% of span 580.5% of span s78.9% of span Water Level liigh-liigh, Turbine Trip &
Feedwater Isolation Functional Unit 8.h Low Low Tavg, s544'F $545'F $547'F $545.4 F P-12 (Increasing)
ESFAS Interlock W/wm & ava 9/16/97 ATT II- 3
WESTINGilOUSE NON-PROPRIETARY CIASS 3 "FARLEY UPRATE RAP' 5AE-SSO-97-l17 TABLE 3 POWER RANGE, P-8 INTERLOCK Parameter Allowance
- Process Measurement Accuracy
+u Primary Element Accuracy Sensor Calibration
.{ l'"
Sensor Pressure Effects Sensor Temperature Effects
[ ).u Sensor Drift
[ j+u Environmental Allowance Rack Calibration Rack Accuracy Measurement & Test Equipment Accuracy Comparator Rack Temperature Effects Rack Drift
- In % span (120% Rated Thermal Pour)
Channel Statistical Allowance =
--* = 4y
==
ATT II- 4
. . ,. . . _ _ . _. .~ ~ . . . - . .-
WESTINO!!OUSE NON-PROPRIETARY CLASS 3 "FARLEY UPRATE RAI"- SAE-SSO-97117 TABLE 4 ,
STEAM GENERATOR WATER LEVEL - HIGH-HIGH Parameter.- Allowance'.-
Process Measurement Accuracy
_ _ +s. - -+.
Primary Element Accuracy Sensor Reference Accuracy
- Sensor Calibration Measurement & Test Equipment Accuracy .
Sensor Pressure Effects Sensor Temperature EfTects Sensor Drift
- Environmental Allowance Rack Calibration Measurement & Test Equipment Accuracy Comparator Rack Temperature Eficcts
' Rack Drift _
~ ' In % span (100 percent span)
Channel Statistical Allowance =
_. _ +=.
A1T II- 5
WESTINGllOUSE NON-PROPRIETARY CIASS 3
?FARLEY UPRATE RAr* SAE SSO 97-117 TABLES.
. TAVG . LOW LOW, P-12 INTERLOCK Parameter Allowance
- Process' Measurement Accuracy
<s. +s.
t Primary Element Accuracy Sensor Calibration Accuracy
-l Sensor Reference Accuracy Sensor Pressure Effects ,
~ Senso Temperature Effects Sensor Drift Emironmental Allowance Rack Calibration
+4,C
'I Measurement & Test Equipment Accuracy
_- _ +v l
A1T11 6
WESTINGilOUSE NON-PROPRIETARY CLASS 3 -
"FARLEY UPRAll! RAl", - SAE.SSO 97-117 TABLE 5 (Continued) .
TAVG - LOW-LOW, P-12 INTERLOCK -
Parameter Allowance"
+"
i Comparator .
. Rack Temperature Effects
~ Rack Drift _
- In % span (100,0 'F)
S Number of Hot Lcg RTDs used
$$ Number of Cold Leg RTDs used Channel Statistical Allowance =
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a o
- A1T II- 7
ATTACHMENT Ill SNC Response to NRC Request For Additional Information Related To Power Uprate Submittal- Joseph M. Farley Nuclear Plant, Units 1 & 2 WCAP-10263,"A REVIEW PLAN FOR UPRATING Tile LICENSED POWER OF A PRESSURIZED WATER REACTOR POWER PLANT" u
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- W.STINGHOUSE CLASS -.3 '
i A REVIEW PLAN FOR UPRATING THE LICENSED POWER '
OF A PRESSURIZED WATER REACTOR POWER PLANT R.' H. McFetridge '
R. T.11archase R. H. - Faas JANUARY,1983 i Approved: .br uschi, Manager Approved:
E. P. Rahe, FQ4ger K. s M Systems Engineering Nuclear Safety \/
1 Westinghouse Elt ctric Corporation Nuclear Energy Systems-P.O. Box 355 -
Pittsburgh, PA 15230~
ABSTRACT This report defines a review plan for increasing the licensed power rating of a nuclear plant. It describes the evaluations required to support an uprating application for a typical plant, and proposes a basis for setting the ground rules and criteria for perfonning those -
evaluations. Its purpose is to develop guidelines for licensees to use when applying for increases in their licensed power ratings.
The review plan is based on three propositions fundamental to the feasibility of uprating an operating nuclear power plant:
- 1. Power related aspects of the plant design will be reviewed.
- 2. The licensing criteria and acceptance standards applicable to current plant operation will apply to uprated plant operation.
- 3. Analyses required to support an uprating application will be perfonned using current analytical techniques.
The NRC must establish its position regarding these issues in order for the applicant to provide sufficient and appropriate information in support of an uprating application.
3353Q:1/012783
TABLE 0F CONTENTS 1.0! INTRODUCTION
- 1.1 - Background 1 1,2 History .
1 1.3 Objectives- 2 2.0 UPRATING REVIEW PROCESS 2,1 Ground Rules and Criteria 4
'2.1.1- Impact on Current Operating License 2.1. 2 Scope of 'deview 2.1. 3 CodG, Standards, and Criteria 2 .1.4 Analytical Techniques 2.2_- Uprating Review Process 6 2.2.1 Uprating Parameters 2.2.2 Pre-tendering Discussions
.2.2.3 Docketing and Approval 2.2.4 Uprating Implementation 3.0 SCOPE.0F REVIEW FOR A PLANT UPRATING 3.1 General 8 3.1.1 Design Limiting Uprating Evaluations ,
3.2 Detailed NSSS Evaluation 10 3.2.1 General 3.2.2 Reactor _ Coolant System (RCS) 3.2.3 Chemical and Volume Control System (CVCS) 3.2.4 Residual Heat Removal-System (RHRS)-
3.2.5 - Safety' Injection System (SIS)
-3.2.6 - Boron Thennal Regeneration System (BTRS) 3.3 Balance of Plant Systems and Equipment Evaluations 13 3353Q:1/123182 -
I TABLE.0F CONTENTS !
(Continued) 3.3.1 Typical BOP /NSSS Interfaces 3.4 Accident Analyses 17-
4.0 REFERENCES
- 21 TABLES 1 _ Comparison of Typical 2 Loop Plant Parameters 22 2 Comparison of Typical 3 Loop Plant-Parameters 23 3 Comparison of Typical 4 Loop Plant Parameters 24 ;
- 4. Typical _ Uprating Milestones 25 5 NSSS. Components 26' 6 Summary of Typical Reactor Coolant System Design 27 Transients
-7 List of Typical Accident Analyses _ 28
~ 3353Q:1/123182 1 -
1.0 .-INTRODUCTION
1.1 BACKGROUND
Due to the increasing lead time and the rising capital cost of new power plant construction, there has been a major trend by electric utilities to upgrade and uprate their existing generating plants. Increasing the number of kilowatt hours generated by an existing unit is a cost effec-tive way to' add generating capacity that benefits both tSe utility and its customers. Most of the uprating effort to date has been concen-trated on fossil fueled plants. Although there is a growing interest in uprating nuclear plants as well, many utilities have hesitated to pursue that option because the regulatory review and approval process is not clear at present. The impact of an uprating application on the current plant operating license, the criteria that will be applied by regulatory authorities in their review of an uprating application, and the time required to complete the review process are all critical factors in detemining if it is feasible to uprate a nuclear plant.
i 1.2 HISTORY Thermal- power uprating of. nuclear facilities is not a new concept.
During the 1960's and early 1970's a number of utilities and NSSS sup-pliers recognized the potential for uprating the thermal output of the nuclear unit to increase electrical generation. Conservatism was designed into the original plant systems and equipment with the under-standing that increased thennal power ratings would be requested at a later date based on the levels of safety and ope-ability demonstrated by the plant at the originally licensed power. The Robert E. Ginna and H.
B. Robinson II nuclear units are examples of Westinghouse plants uprated after the initial operating license was granted. Ginna was operated at
. a rating of 1320 MWt until an amenchnent to increase the licensed rating to 1520 MWt was approved. H. B. Robinson II was originally operated at 2200 MWt until a thermal power uprating to 2300 MWt was approved. Later plants have been uprated before initial power generation. The D. C.
'3353Q:1/123182 1
Cook Unit II, for example, was uprated from 3250 MWt to 3403 MWt during licensing of the plant. Several non-Westinghouse nuclear facilities have also been uprated, including Fort Calhoun, St. Lucie I, Crystal River, and Millstone II. Today there is a broad base of experience to support the operation of plant components at upratet levels. In an effort to streamline and standardize the licensinc review process, nuclear suppliers have standardized plant, component and system designs to envelope a spectrum of operating conditions over a broad range of themal power ratings. Tables 1 through 3 shew the progression of Westinghouse NSSS ratings with time for 2, 3 and 4 loop plants with an active fuel length of 12 feet. From the tables it can be seen that over the years thermal power has increased by approximately 30 percent.
During this period, many of the standard NSSS components have been licensed and operated at power levels beyond those of their initial application.
It is also significant that the safety related features of a Westinghouse PWR are typically designed for a thermal power rating about five percent greater than the licensed rating. This power rating is refen ad to as the Engineered Safeguards Design Rating (ESDR), and it is usually detemined by the turbine limiting flow capability. As a result of this practice, many of the Westinghouse pressurized water reactors operating today could be uprated to the ESDR with only minor software
! and hardware modifications. With appropriate modifications to the NSSS and to the BOP, some of these units could be uprated beyond the ESDR.
I 1.3 OBJECTIVES l The primary objective of this report is to develop guidelines for
! licensees to use when preparing applications for increases in their licensed power levels. It consists of two principal elements. One
! describes the safety evaluations and component design reviews that will be performed to demonstrate that a plant can continue to be operated without undue risk to the health and safety of the public if the licensed power level is increased as requested. The other proposes a 3353Q:1/012783 2
I j
-1
- set of ground rules and criteria that provide a uniform and well-defined
' base from which to evaluate changes in power rating, It is hoped that - .
through review of this report:and-discussions that_ follow, the NRC will
- establisht
' l) A por,ition regarding the~information required to permit the staff to l conclude its_ review of an upratingLapplication; and
- 2) ' A basis for defining the ground rules and criteria that will be-used ,
in evaluating that application.
1
- 3353Q
- 1/123182 3 p
i . --
2.0 UPRATING REVIEW PROCESS f
2.1 GROUND RULES AND CRITERIA In order to prepare an uprating application for submittal to the NRC, a licensee must be able to establish: '
- 1. The potential impact of the uprating application on the current design basis
- 2. Scope of regulatory authority review
- 3. Applicable regulatory codes, standards and criteria
- 4. Analytical techniques to be utilized The NRC position regarding these issues will have a major impact on the feasibility of uprating nuclear facilities. It will also facilitate the review process if the applicant is able to provide sufficient and appro-priate information to support the initial uprating application. Follow-ing is a discussion of the Westinghouse position on these issues.
2.1.1 IMPACT ON CURRENT DESIGN BASIS The proposed uprating will be analyzed in accordance with the codes and standards applicable to the plant at the time of submittal and, as such, will have no impact on the plant design basis.
2.1. 2 SCOPE OF REVIEW The scope of regulatory review should encompass all aspects of the faci-lity design and operation which are pacted by the proposed uprating.
Any aspect of the design that is impacted will be evaluated against the current codes anr1 regulations applicable to the plant. However, a review will be t., ode as defined in 10CFR50.59 to identify any potential 3353Q:1/123182 4
unreviewed safety questions that might result from the uprating.
~
Section 3 of this report provides a discussion of the scope of a typical uprating review.
2.1. 3 CODES, STANDARDS, CRITERIA The proposed uprating will be performed in accordance with the established licensing criteria and standards which apply to the current operating license of the specific plant being uprated. If the uprating involves a potentially unreviewed safety question, it will be identified and resolved during the uprating review process. This process will assure.that protection of the public health and safety can be maintained within the current licensing basis.
The need for plant modifications associated with the uprating will be established by .the results of component design reviews and analytical evaluations. based on operating conditions at the uprated power. These reviews and evaluations will be used to identify any areas where exist-ing plant components and designs fail to meet applicable licensing criterie and standards at the uprated power, as well as to determine -
appropriate modifications to re-establish compliance. The types of modifications which might be required to support a plant uprating are Judged not to be " material alterations" under 10CFR50.91 because they would not change the plant operations or purpose as originally licensed.
2.1. 4 ANALYTICAL TECHNIQUES The technology and data base of the nuclear industry have progressed r - significantly in many areas. To take advantage of that progress, current analytical techniques will be used for any analyses required to support an uprating. This will also facilitate performance of the l
analyses and the regulatory review of the results. Existing analyses will not be redone if they are.not affected by the uprating, or if they have already been analyzed at the uprated power for the FSAR.
f 3353Q:1/012783 5
2.2 UPRATING REVIEW PROCESS Table 4 summarizes the major milestones that must be accomplished during review and approval of a typical uprating. These milestones are applicable to upratings in general, and can be modified easily to suit the specific requirements of a particular uprating application. A discussion of the more signiilcant interface activities in the uprating process follows.
2.2.1 UPRATING PARAMETERS The initial step in an uprating program is for the utility to establish uprating parameters and to define an associated plant configuration for the evaluation of limiting plant transients and accidents. This evalua-4 tion is performed to confirm that compliance with the established plant licensing basis will be maintained with the proposed uprated parameters and plant configuration. Based on the results of this evaluation, the utility determines the feasibility of proceeding with the uprating program.
2.2.2 PRE-TENDERING DISCUSSIONS I
The utility will initiate pre-tendering discussions to inform the NRC of the impending uprating application, end to describe the proposed uprat-l ing program. This will pennit the conunission to plan and schedule the
- uprating review, and to provide comment on the utility uprating pro-gram. It is assumed that the NRC will have previously p ,vided guidance on the program content through its comment on this rep..et. Based on l these pre-tendering discussions, the utility decides wnether or not to make a final commitment to the uprating program.
The utility, NSSS supplier, and architect engineer will then meet with the NRC in a technical review of the evaluation and analysis of limiting
-transients and accidents. Results of this discussion are documented to 3353Q:1/012783 6
the NRC for_ infomation,;infomal review and, schedule planning. Follow - l ing this meeting =, the NRC responds with 'a schedular commitment, _ and .
identifies any technical constraints that could inhibit licensing of the
- uprated conditions.-
12.2.3 DOCKETING __AND APPROVAL Based on connents from the NRC, the remainder of the _uprating program is
- executed (e.g., evaluations, analyses and hardware modifications). A ~
final . licensing document is submitted containing all required analyses and evaluations ard describing any required plant modifications to demonstrate that compliance with the established licensing criteria is maintained. This document is docketed, and forms the basis for final NRC review and approval of the uprating.
2.2.4 ' UPRATING _IWLEENTATION-Af ter the' NRC has issued a license amendment for the -uprated condi-tions, the utility implements the uprating._ Plant design and operating documents are revised consistent with parameters for the uprated power.
Hardware modifications are completed ~and verified fur.ctional. When 4
these actions are complete, ^he plant can be operated at the uprated
. power. The next periodic updating of the plant Final Safety Analysis Report required by 10CFR50.71 will incorporate changes resulting from the plant uprating.-
l 3353Q:1/123182 7
= - . . . = - . . - _.
3.0 SCOPE OF REVIEW FOR A PLANT UPRATING 3.1 GENERAL The licensing review for a plant uprating is typically perfomed in two parts. In the first part, design limiting conditions and events are reviewed to demonstrate the feasibility of uprating to the desired power. This infomation is also used as a basis for pre-tendering discussions in whi@
feedback from the NRC is obtained to identify any licensing constraints. The review is then completed by performing all of the remaining evaluations and analyses required to license the uprating.
.3.1.1 DESIGN LIMITING UPRATING EVALUATIONS Initially, a set of plant parameters will be established as a basis for the uprating evaluations. These parameters will be established by the utility in conjunction with the NSSS supplier and architect engineer based on a knowledge of replicate plants / systems operating at higher power levels, available system / component margin, potential hardware / system improvements available and limitations of components and systems which would not be practical to replace or modify (e.g., containment or reactor vessel structures). Key parameters Include:
NSSS Power Feedwater Flow Rate Reactor Power Steam Generator Outlet Pressure Core Flow Rate Reactor Vessel Inlet Temperature Reactor Coolant Pump Flow Rate Reactor Yessel Outlet Temperature Steam Flow Rate Steam Generator Feedwater Temperature As the program progresses, these parameters will be used to determine more detailed plant parameters, such as heat rejection rates to the component cooling water systems, mass and energy release rates, radiation source terms and emergency core cooling system parameters.
Evaluation of the design limiting accidents and transients are perfomed next to detemine the adequacy of the existing plant for operation at 3353Q:1/123182 8
uprated conditions'.- These evaluations will also provide input to. define -
any plant modifications' that might be required to satisfy the acceptance criteria.7 All analyses will be made_to FSAR quality standards using NRC ,
approved calculational techniques so that- theylneed not be re-done- ,
during the balance of the uprating evaluations.
Accidents and. transients that would be analyzed during this_ part of a typical plant uprating review include design limiting events for o DNB Menin 0- Reactivity _ Excursions o ECCS Capability o Peak RCS Pressure o Heatup o= Auxiliary Feedwater System o Containment Design In parallel with the review of the design limiting accidents and
- transients, an analysis of the NSSS systems and components will be perfomed to determine their capability for operation at the uprated power. These analyses and evaluations will either 1) verify compliance of existing systems and operating procedures with applicable plant design bases.and regulatory requirements, or 2) identify those areas-where revisions and/or modifications are required. This review will include all of the classical NSSS fluid systems components listed in L Table 5, as well as any components provided by the NSSS supplier in optional systems. The impact of the uprated parameters on functional design requirements and structural integrity of these components will-be
! reviewed. Typical NSSS-operating transients to be considered during this review are listed in Table 6. < Where the uprating requirements are not bounded by current component design, revisions and modifications-will be made as necessary to demonstrate compliance with applicable codes and standards.
l L3353Q:1/123182__ d'
The' plant technical specificatiens will be reviewed to identify- required revisions to protection setpoints and/or limiting conditions ~ for
. operati on. .
3.2: DETAILED NSSS EVALUATION'-
3.2.1 GENERAL The detailed evaluation will differ from the design limiting evaluation
-in that it is focused on those specific areas in which the need for further evaluation and possible plant changes has been identified.
When the' design limiting evaluation has indicated that the uprating has an impact on a particular system and/or component, the designer will receive revisions to the design bases and/or functional requirements for the specific system / component and will detemine if the installed
~
system / component remains in compliance with the plant specific stan-dards, design criteria, and regulatory requirements for the uprated conditions.-
An uprating can increase the operating power level and temperatures of the RCS. This necessitates the verification that each installed system component and the associated analyses are in compliance with the design codes, standards and criteria for the revised nominal operating condi-l tions. In some instances it will be necessary to revise the documented analyses to account for the increased power level. Three levels of
- effort may be necessary to accomplish this review. Each of the three l 1evels'is discussed below
The first level of effort. is to identify for which NSS$ systems and associated components no change in the original design bases and l- functional: requirements is required.. For these components and/or systems, no additional effort is required with. respect to the
! .uprating.
l 3353Q:1/123182 10 L
The second level of effort is to identify for which NSSS comp nents l the uprated conditions are bounded by analyses performed for a generic design or for a plant with the identical systems component at power levels equal to or greater than those associated with the proposed change. For these cases, an evaluation is provided to document the acceptability of the installed system or component.
The third level of effort is to confirm compliance with the applica-ble design codes, standards and criteria for specific instances where the uprated conditions are not bounded by analyses perfonned for a generic design or for a unit with the identical components at duty ratings equal to or greater than those associated with the proposed change.
In summary, the majority of the NSSS components will be enveloped by either the original analyses for the specific unit or analyses for other plants with identical structures at a higher duty rating. For specific components where additional analyses are necessary, it must be deter-mined if the structures remain in compliance with the design codes, standards and criteria applied, to the current license for the specific unit. Should it be necessary, appropriate action will be taken to assure compliance with the unit's current licensing bases at the uprated condition.
3.2.2 REACTOR COOLANT SYSTEM (RCS)
As a minimum, the impact of the proposed uprating on the functional, operational, and safety related aspects of the RCS will be evaluated and/or analyzed in the following areas:
Analyses will be performed to detemine the pressurizer spray, power operated relief and safety valve relief capacity necessary to maintain the original design bases for the increased power level. The specific plant Safety Analysis Report discusses the design bases for that unit.
Evaluations will be perfonned to detennine the necessary operating range of the Reactor Coolant System control, protection and measurement 3353Q:1/123182 11
instrumentation (e.g., pressure, temperature, flow, level, flux mapping-and nnelear_ power) and-the associated systems (e.g. , nuclear instrumen-tation, flux mapping, bottom mounted instrumentation and incore thermo-couple ' systems) at the increased power level. Any necessary revisions to the current operating ranges or functional requirements will be identified.
3.2.3 CHEMICAL AND VOLUE CONTROL SYSTEM (CVCS)
All' functional requirements of the CVCS will be reviewed. The areas which are most likely to be impacted by the uprating are:
- 1. CYCS heat exchanger heat rejection rates - If the uprating results in an increased RCS cold leg temperature, the heat loads from the CVCS heat exchangers to the component cooling water system will increase.
- 2. Components and systems located upstream of the letdown heat exchanger - Should the RCS cold leg temperature be increased at the uprated conditions, the uprated functional requirements may not be enveloped by the current component design bases. The capability of the components to perfom at the uprated conditions will be confimed and appropriate modifications made. Should the RCS cold leg . temperature be reduced, the existing design bases would bound the uprated condition.
3.2.4 RESIDUAL HEAT REMOVAL SYSTEM (RHRS) l A higher power level results in an increase in the amount of decay heat i being generated in the core during norinal cooldown,- refueling operations and accident conditions. This will result in a higher heat load on the residual heat exchangers during the cooldown and also during the
! refueling outage. The increased heat loads will be transferred to the Component Cooling Water System (CCWS) and ultimately to the Service l Water Cooling System (SWCS). It will be necessary to evaluate the
! performance of the RHRS, CCWS and SWCS with the increased heat loads.
3353Q:1/011003 12-1
.- - ~. - - . . ,
N i;
- 0n some^ plants the RHRS pumps and heat exchangers are an intogral- part
- of_. theLSafety Injection System (SIS). For these plants, the ability' off
~
.the RHRS to' meet theidesign and functional- requirements of the SIS at -
I
- the uprated conditions _will. be'confinned.
, The' uprating does not impact the ability of the RHR pumps to transfer 7
water to or from the refueling water storage tank.
~
- 3. 2.' 5 SAFETY INJECTION SYSTEM The required-volume, duration and heat rejection capability of the safety injection flow in the event of a break is detennined based on
- analytical and empirical models which simulate reactor conditions subsequent to the postulated RCS and steam system breaks. As a result
. of these-analyses the system-and component requirements necessary to demonstrate compliance with regulatory requirements at the uprated power level will-be established. Should the requirements fall outside the .,
bounds of the installed system,-it may be necessary to implement-
- software / hardware _ modifications, provide revised heat rejection rate
! data for the CCWS and mvise the electrical-loading of the SIS equipment l on the safeguards electrical systems. In the event the current SIS provides adequate safety margin, no additional effort would be required.
3.2.6 BORON TERMAL REGEERATION SYSTEM (BTRS)
_ Evaluations;at the uprated conditions will~ be perfonned to assure that'
~
the installed system / component design bases and functional requirements
^
= bound the proposed operating conditions.
3.3 BALANCE OF PLANT SYSTEMS AND EQUIPENT EVALUATIONS
" Uprating the. electrical generation capability of the unit will also have
}
- an impact on the.B0P systems;and equipment. - As part of-the evaluation i of- the- NSSS, the 'NSSS/B0P interfaces will be reviewed and changes to the interface infonnation will be provided to the' utility. - The review and E ,
3353Q:1/123182 13-
Lanalyses forlthe BOP _ will follow:a pattern sinilar to the NSSS procedure as discussed in'section 3.2.
Initially the' plant conditions and configuration associated with the
- target uprating and a delineation of the necessary interface data will be identified. A review and analysis of the limiting B0P accidents (e.g. containment pressure and temperature) will be perfomed to confim ,
that the proposed uprat_ing parameters and associated plant configuration are-in compliance with the plant license.
. Subsequent to the evaluation of the-limiting BOP accidents, detailed evaluations of.the BOP systems and equipment will be perfomed. If the uprated system conditions are bounded by existing docun.entation, no
-additional effort will be required for that system or the_ equipment in
.that system.. If the uprated conditions are not bounded by the current
' design bases'and functional requirements, necessary software /hartiware revisions will be identified. Where revisions are identified, further evaluation will be perfomed to detemine if the equipment remains in compliance with the plant's current licensing basis. If necessary modi-fications to the equipment will be identified to assure compliance with the licensing basis is maintained.
3.3.1 - TYPICAL BOP /kSSS INTERFACES The following B0P/NSSS interfaces may be impacted by the uprating.
< These interfaces would only be affected as a result of modifying the design bases and/or functional- requirements of another NSSS or BOP system serviced by these B0P areas.
- a. AC and DC Emergency Power Syrtems - The plant is equipped with both onsite (AC and DC)-and offsite (AC). emergency electrical i power systems to provide reliable power to the NSSS and BOP I safety systems. Increases in the electrical power requirements of the NSSS essential _ systems, which result from the uprating will be identified.
3
_3353Q:1/123182 14
- b. AC and DC Power. Systems - The plant is equipped with electrical power systems which supply the NSSS equipment. Any increased NSSS electrical loads which may be required as a result of the uprating will be identified,
- c. Demineralized Water Makeup System - The purity requirements of.
the makeup water could be affected by the uprating,
- d. Auxiliary Feedwater System - TAe Auxiliary Feedwater System supplies feedwater to the secondary side of the steam generators whenever the main feedwater system is not available, in order to maintain the steam generator as the principal reactor shutdown heat sink. This system may also function as an alternate to the Main Feedwater System during startup, hot standby and cooldown conditions. The Auxiliary Feedwater System provides core cooling during abnonnal transients.
- e. Mass and Energy Release to the Containment - The mass / energy release data will be employed to detennine the containment pres-sure and temperature environment during the postulated accidents and to detennine the associated loadings on the structures and components within the containment in accordance with the licensing basis of the specific unit. Mass and energy release data for the uprated conditions will be provided.
- f. Spent Fuel Pit Cooling System - The functions of this system are:
- 1. Maintain desired water temperature in the spent fuel pit.
- 2. Maintain chemistry and activity level requirements in spent fuel pit water.
- 3. Provide refueling water cleanup and purification capabili-i ties.
3353Q:1/012783 15
i l
The increased decay heat rates will be identified to allow evaluation of the ability of the ini,talled spent fuel pit cooling system to maintain acceptable temperatures within the spent fuel pit. There are no NSSS/ BOP interface changes with respect to the other two functions.
- g. Main Steam System - The primary purpose of the steam system is to contain and transport. steam from the NSSS steam generators to the main turbine. The steam system also fonns part of the boundary between the radioactive fluid systems and the environ-ment.
The uprating will result in increased steam flow snd/or pressure in the main steam system.
- h. Component Cooling Water System (CCWS) - The CCWS is an inter-mediate system between the Reactor Coolant System and the Service Water Cooling Systems (SWCS). It ensures that leakage
-of radioactivity from the components being cooled is contained within the plant. The system typically removes heat from the NSSS and some B0P components. Revised heat rejection rates and/or cooling water flow requirements will be identified.
1 Radiological Source Terms - Radiological Source Tenns are used in assessing the radiological consequences of accidents. Any changes identified as a result of uprated parameters will be identified.
- k. Plant Testing - Numerous qualification and performance tests were completed for the initial startup to assure that all systems / components of the BOP and NSSS ara in compliance with the design and licensing bases for the unit. These tests also establish the operating margins of the plant systems. It will be necessary to verify that the performance of any system /
component modifications are in compliance with the requirements 3353Q: 1/123182 16
- of the uprating and the licensing bases. The recommended test program for NSSS and interfacing BOP systems would be developed on a plant specific basis, depending upon the magnitude of hardware modifications and the magnitude of the uprating.
3.4 ACCJDENT ANALYSES A reference analysis is nomally established as part of the initial licensing effort as documented in the FSAR. This is supplemented by reanalyses required for reload fuel or plant equipment or system changes. For a plant uprating, a safety evaluation is perfomed to confirm the validity of applicable reference analyses. If the reference analyses do not bound the uprated conditions, reanalysis using curre ntly approved methods and approprie'.c input parameters will be perfomed.
The elestinghouse Reload Safety Evaluation Methodology (R$EM) report y (Ref.1) sumarizes the overall process to assess changes. This report
_I was written primarily for reloads, but the process described is also applicable to upratings.
The uprating evaluation process includes:
- 1. A systematic evaluation to determine a) what parameters utilized in the reference safety evaluation are impacted by a change in plant rating and b) if these new parameters are bounded by the current reference safety evaluation.
- 2. A detemination of the effects on the reference safety analysis when a parameter per 1.b above is not bounded. This determination may require a reanalysis as appropriate. -
The specific steps in this process are the design initialization, design process and safety evaluation.
The design initialization process involves the collection and review of design basis information to ensure that the uprating safety evaluation will be based on the actual fuel and core components in the plant, the 3353Q:1/123182 17
i I
actual plant operating history, and any plant system changes associated with the uprating. The review includes the utility requirements, core design parameters, safety criteria and related constraints, specific operating limitations and past operating history. The initialization review identifies the objectives, requirements and constraints for the uprated cycle being designed.
The design process ensures that the utility power and energy require-ments established in the de.agn initialization phase are achieved. The key safety parameters for the cycle (i.e. uprating and reload para-meters) are then determined based on the preliminary design. The safety bases to be met for the uprated core are:
Departure from Nucleate Boiling Design Basis - There will be at least a 95 percent probability that departure from nucleate boiling (DNB) will not occur on the limiting fuel rods during nomal opera-tion, operational transients, or during any transient conditions arising from faults of moderate frequency (Condition I and Il events), at a 95 percent confidence level. In order to meet this basis, the minimum allowable DNB ratio is detemined. This minimum allowable DNBR depends upon the DNB correlation employed in the analysis. For example, this minimum DNBR was conservatively set at 1.30 for the original W-3 DNB correlation and 1.17 for tile WRB-1 DNB correl ation.
Fuel Temperature Design Basis - During modes of operation associated with Condition I and Condition Il events, there is at least a 95 percent probability that the peak kw/ft fuel rods will not exceed the U02 melting temperature, at the 95 percent confidence level.
The melting temperature of UO 2 is taken as 5080'F, unirradiated and decreasing 58'F per 10,000 MWD /MTV. By precluding UO2 melting, the fuel geometry is preserved and possible adverse effects of molten UO2 on the cladding are eliminated. To preclude center melting end to provide a basis for overpower protection system set-points a calculated centerline fuel temperature of 4700*F has con-servatively been selected as the overpower limit.
3353Q:1/123182 -18
Reactor Coolant Systen Pressure - Peak RCS pressure is not to excsed 110 percent of the design pressure during Condition I and Condition II events.
Loss of Coolant Design Bases (10CFR50.46) - The LGCA design bases incorporates a review of peak cladding temperature, maximum cladding oxidation, maximum hydrogen generation, coolable geometry and long-term cooling.
Compliance with these bases ensures that the margin of safety as defined in the basis of the technical specification has not been reduced (a 10CFR50.59 requirement). These design bases are interpreted as safety limits for the safety evaluation.
The objective of the uprating safety evaluation is to verify compliance with the currently established safety limits for the specific unit with the uprated core and plant system design. This is accomplished by examining each accident pmsented in the FSAR or subsequent submittals to the NRC to determine if the reference analysis remains valid for the uprating. A typical listing of postulated accidents is pmsented in Table 7. The specific transients for each plant can be found in the unit's Safety Analysis Report. For those accidents which are affected by the uprating, an evaluation is perfomed to verify compliance with the applicable safety limits.
In the performance of an uprating safety evaluation, each accident is examined and the bounding values of the key safety parameters which could be affected by the uprating are determined based on the reference analysis. These parameters form the basis for determining whether the reference safety analysis remains valid. For an uprating, values of these safety parameters are determined for the core during the nuclear, themal and hydraulic, and fuel rod design proccss. Each of these para-meters is compared with the reference analysis value to detemine if any parameter is not bounded. If all of the parameters are bounded, the reference analysis remains valid and no new analysis is needed to verify
-that the safety limits are not exceeded. Should one or more of the 33530:1/123182 19
safety parameters not be bounded, a re-evaluation of the accident is performed.
The re-evaluation may be of two types. If the parameter is only slightly out of bounds, or if the transient is relatively insensitive to that parameter, a simple quantitative evaluation may be made. Al terna-tively, should the deviation be large or be expected to have a more significant or not easily quantifihble effect on the accident, a re-analysis of the accident is performed. If the accident is re-analyzed, the analysis methods follow standard procedures and will typically employ analytical methods which have been used in previous submittals to the NRC. These methods are those which have been presented in the FSAR or subsequent submittals to the NRC for that plant, reference SARs such as RESAR, or reports submitted for NRC approval. The re-analyzed acci-dent must continue to meet the appropriate safety limit for that event in order to be considered to have acceptable results.
Accident re-analysis may also be necessary if there are any changes made to the reactor plant systems, either in configuration, performance or setpoints as detemined during the design initialization phase. Should any plant or system changes affecting safety be incorporated, their impact will be detemined during the evaluation.
Measurements of nuclear and safety related parameters during and after cycle startup serve two purposes. The first is to insure that the measured parameters fall within the limiting values included in the Technical Specifications of the plant. The second is to confim the validity of the corresponding design calculations. For an uprating, as for any other reload, startup physics program will be perfomed to confirm the key safety parameters such as rod worths and moderator temperature coefficients. The testing will also confirm that the core is properly loaded. The values of all measured parameters are compared to those calculated using the design codes.
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4.0 REFERENCES
- 1. Bordelon F. M. et. al., WCAP-9272 Westinghouse' Reload Safety Evalu-ation Methodology w
b j3353Q:1/123182- _
21 -
b TABLE 1 COMPAPISON OF TYPICAL 2 LOOP PLANT PARAETERS First Second Third Future Gener- Gener- Gener- Gener-ation ation ation ation NSSS Power, MWt 1520 1650 1882 1967 NSSS System Pressure Nominal, psia 2250 2250 2250 2250 179,400 178,000 189,000 189,000 Total Core Inlet Thermal Flow Rate, gpm Reactor Coolant System Temperature, 'F ,
Nominal Reactor Vessel / Core Inlet 552.5 535.5 549.9 553.0 Average Rise in Vessel 57.3 63.6 66.2 68.6 Average in Vessel 581.2 567.3 $83.0 587.5 No Load 547 547 557 557 Rated Steam Pressure, psia 8 21 750 920 920 Major Components Fuel Type 14 x 14 14 x 14 16 x 16 10 x 16 Steam Generator Model 44 51 F F Reactor Coolant Pump Model/ Horsepower 93/6000 93A/6000 93A/7000 93A/7000 3353Q:1/123182 22 l
TABLE 2 COMPARISON OF TYPICAL 3 LOOP PLANT PARAMETERS First Second Third Future Gener- Gener- Gener- Gener-ation ation ation ation NSSS Power, MWt 2208 2441 2785 2910 NSSS System Pressure Nominal, psia 2250 2250 2250 2250 Total Core Inlet Thernal Flow Rate, gpm 268,500 265,500 292,800 278,400 Reactor Coolant System Temperature, 'F Nominal Reactor Vessel / Core Inlet 546.2 543.0 557.0 552.3 Average Rise in Yessel 56.1 62.6 62.9 68.9 Average in Vessel 574.2 574.3 588.5 586.8 No Load 547 547 557 547 Rated Stesm Pressure, psia 785 785 964 850 Major Components fuel Type 15 x 15 15 x 15 17 x 17 17 x 17 Steam Generator Model 44 51 F F Reactor Coolant Pump Model/ Horsepower 93/6000 93A/6000 93A/7000 93A/7000 3353Q:1/123182 23
TABLE 3 COMPARISON OF TYPICAL 4 LOOP PLANT PARAMETERS First Second Third Future Gener- Gener- kner- Gener-ation ation ation ation NSSS Power, MWt 2758 3250 3423 3600 NSSS System Pressure Nominal, psia 2250 2250 2250 2250 Total Core Inlet Themal Flow Rate, gpm 358,800 354,000 354,000 360,000 Reactor Coolant System Temperature, 'F Nominal Reactor Vessel / Core Inlet 543.0 536.3 552.5 547.6 Average Rise in vessel 53.0 63.0 64.3 66.7 Average in Yessel S69.5 567.8 584.7 580.9 No Load 547 547 557 b47 Rated Steam Pressure, psia 776 758 81 0 835 Major Components Fuel Type 15 x 15 15 x 15 17 x 17 17 x 17 Steam Generator Model 44 51 F F Reactor Coolant Pump Model/ Horsepower 93/6000 93A/6000 93A/6000 93A/6000 3353Q:1/123182 24
TABLE 4 TYPICAL UPRATING MILESTONES Milestone Est. Time Action (months)
- 1. Select Target Parameters and Plant Configuration 1-2 Utility, A/E and i
- 2. Perform Limiting Accident Analyses 4-6 Utility. A/E and y
- 3. Inform NRC of Intent to Submit Uprating Application Utility
- 4. Prepare and Submit Document Summarizing Limiting Accident Analyses and Identifying Scope of Implemenation Program 1-2 Utility, A/E and y
- 5. Review and Coment on Uprating Program 3-6 NRC
- 6. Perfonn Hemainder of Uprating Evaluations and Implement Hardware Improvements: Utility, A/E and y Analyses 6-9 Hardware 6-24
- 7. Final Review and Approval of Upruting Program 3-6 NRC
- 8. Issue Operating License Amendment NRC 3353Q:1/123182 25
TABLE 5 HSSS COMP 0NENTS Reactor Vessel Reactor Internals Control Rod Drive Mechanisms Incore Instrumentation Tubing Reactor Coolant Loop Piping Reactor Coolant Loop Isolation Valves Pressurizer Steam Generator Reactor Coolant Pumps Component and Piping Supports Tanks Heat Exchangers Pumps Valves Filters Evaporators Instrumentation Refueling and Fuel Handling Equipment-Chillers 3353Q:1/123182 26
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TABLE 6 SlHMARY OF TYPICAL REACTOR COOLANT SYSTEM DESIGN ACCIDENTS AND TRANSIENTS Normal Conditions
- 1. Heatup and Cooldown at 100*F/hr (pressurizer cooldown 200*F/hr)
- 2. Unit Loading and Unioading at 5 percent of full power / min
- 3. Step Load Increase and Decrease of 10 Percent of Full Power
- 4. Large Step Load Decrease
- 5. Steady State Fluctuations Upset Conditions
- 1. Loss of Load, without immediate turbine or reactor trip
- 2. Loss of Power (blackout with natural circulation in the RCS)
- 3. Loss of Flow (partial loss of flow one pump only)
- 4. Reactor Trip from Full Power
- 5. Operational Basis Earthquake (20 earthquakes of 20 cycles each)
Faulted Conditions
- 1. Main Reactor Coolant Pipe Break
- 2. Steam Pipe Break
- 3. Steam Generator Tube Rupture
- 4. Design Basis Earthquake Test Conditions
- 1. Turbine Roll Test
- 2. Hydrostatic Test Conditions
- a. Primary Side
- b. Secondary Side
- c. Primary Side Leak Test
-3353Q:1/011083 27
TABLE 7 LIST OF TYPICAL ACCIDENT ANALYSES Uncontrolled RCC Assembly Withdrawal
- 1. From a subcritical condition
- 2. At power RCC Assembly Misalignment Chemical Volume and Control System Halfunction
- 1. Dilution during refueling
- 2. Dilution during startup
- 3. Dilution at power Loss of Reactor Coolant Flow
- 1. Flow coast-down
- 2. Locked rotor accident Start-up of an Inactive !Mactor Coolant Loop Loss of External Electrical Load loss of Normal Feedwater Excessive Heat Removal Due to Feedwater System Halfunction Excessive Load Increase Incident Loss of all A.C. Power to Station Auxiliaries Steam Generator Tube Ruptu e Rupture of a Steam Pipe Rupture of a Control Rod Drive Mechanism Housing Reactor Coolant System Pipe Rupture 3353Q:1/123182 28
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