View source for TXX-9826, Application for Amends to Licenses NPF-87 & NPF-89,proposing Increase in Licensed Power for Operation of Cpses,Unit 2 to 3445 Mwt (Approx 1%) & Changes Administratively for Unit 1.Supporting TSs & TR WCAP-10263,encl

{{Adams
| number = ML20198J307
| issue date = 12/21/1998
| title = Application for Amends to Licenses NPF-87 & NPF-89,proposing Increase in Licensed Power for Operation of Cpses,Unit 2 to 3445 Mwt (Approx 1%) & Changes Administratively for Unit 1.Supporting TSs & TR WCAP-10263,encl
| author name = Terry C, Walker R
| author affiliation = TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000445, 05000446
| license number = 
| contact person = 
| document report number = TXX-98265, NUDOCS 9812300138
| package number = ML20198J312
| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS &amp; PERMITS
| page count = 37
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:N                                                     l FE                      Log # TXX-98265               I
                                                              .:-            File # 916 (1.0) r        r                      10010 Ref. # 10CFR50.90 7UELECTRIC                         10CFR50.36
: c. t ce 'n,r,y                                                           December 21,1998 Senior Vice President
    & PrincipalNuclear Officer U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 2

==SUBJECT:==
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 SUBMITTAL OF LICENSE AMENDMENT REQUEST 98-010 INCREASE IN UNIT 2 REACTOR POWER TO 3445 MWt REF:       1)     TU Electric Letter logged TXX-98180, from C. L. Terry to the NRC dated July 17,1998
: 2)    TU Electric Letter logged TXX-98183, from C. L. Terry to the NRC dated August 13,1998
: 3)    TU Electric Letter logged TXX-98240, from C. L. Terry to the NRC dated November 13,1997
: 4)     TU Electric Letter logged TXX-97105, from C. L. Terry to the NRC dated May 15,1997                                                    i Gentlemen:
Pursuant to 10CFR50.90, TU Electric hereby requests an amendment to the CPSES Unit 1 Operating License (NPF-87) and CPSES Unit 2 Operating License (NPF-89) by incorporating the attached changes into the CPSES Unit 2 Operating License and                 i the CPSES Units 1 and 2 Technical Specifications. These changes apply to CPSES Unit 2 and administratively to CPSES Unit 1.
Based on the Caldon topical report submitted per Reference 1 and the exemption request submitted per Reference 2 above, TU Electric requests an increase in the licensed power for operation of CPSES Unit 2 to 3445 MWt (approximately 1%).
The revised core safety limit curves and revised Overtemperature N-16 reactor trip setpoints for CPSES Unit 2, Cycle 5 (submitted in Reference 3) anticipated submittal of this change and, therefore, remain valid at the uprated power condition,             j l
Attachment 1 is the required affidavit. Attachment 2 provides a detailed description of the proposed changes, a safety analysis of the proposed changes and TU Electric's determination that the proposed changes do not involve a significant p'
hazard consideration. Attachment 3 provides a markup of the affected page of             \)p' Operating License NPF-89. Attachment 4 provides the Technical Specification pages marked-up to reflect the proposed changes. Attachment 5 provides a markup of the improved Tecb_n,ical Specifications pages submitted per Reference 4.
9812300138 981221             1' PDR     ADOCK 05000445 P                         PDR L w,,,,,,,,,,,,, ,,EAK STEAM ELECTRIC STATION P.O. Box 1002 Glen Rose. Texas 76043-1002 J

TXX-98265 Page 2 of 2 TU Electric requeats approval of this proposed license amendment be targeted for the end of the spring,1999 Unit 2 refueling outage (2RFO4), with implementation of the changes to occur within 30 days after NRC approval.
In accordance with 10CFR50.91(b), TU Electric is providing the State of Texas with a copy of this proposed amendment.
Should you have any questions, please contact Mr. J. D. Seawright at (254) 897-0140.
l Sincerely,                            )
: e.  % Qerry C. Lance  T l
By:      @           r Roger D.' Walker Manager Regulatory Affairs            i JDS/grp Attachments: 1. Affidavit
: 2. Description and Assessment
: 3. Affected pages to the Unit 2 Operating License (NPF-89)
: 4. Affected pages to the Technical Specifications
: 5. Affected pages to the ITS Technical Specification c-     Mr. 2. W. Merschoff, Region IV                                                   l Mr. T. J. Polich, NRR                                                            <
Mr. J. l. Tapia, Region IV Resident inspectors, CPSES Mr. Arthur C. Tate                                                                '
Bureau of Radiation Control Texas Department of Public Health 1100 West 49th Street Austin, Texas 78704 l
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Attachment 1 to TXX-98265 Page 1 of 1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of                                               )
                                                                 )
Texas Utilities Electric Company                               )       Docket Nos. 50-445
                                                                 )                       50-446 (Comanche Peak Steam Electric                                  )       License Nos. NPF-87 Station, Units 1 & 2)                                          )                       NPF-89 AFFIDAVIT Roger D. Walker being duly sworn, hereby deposes and says that he is Regulatory Affairs Manager of TU Electric, the licensee herein; that he is duly authorized to sign and file with the Nuclear Regulatory Commission this License Amendment Request 98-01G; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.
0C>91    b Roger D. Walker Regulatory Affairs Manager STATE OF TEXAS                              )
                                              )
COUNTY OF SmNd                              )
Subscribed and sworn to before me, on this 2ld                      day of b=cenh-                        ,1998.
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ATTACHMENT 2 to TXX-98265 l
DESCRIPTION AND ASSESSMENT l
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l Attachment 2 to TXX-98265 Page 1 of 32 1
l DESCRIPTION AND ASSESSMENT
: l.      BACKGROUND Comanche Peak Steam Electric Station (CPSES) Unit 2 was originally licensed for a 100% core thermal power rating of 3411 MWt. CPSES evaluated the impact of a
;                1% uprating to 3445 MWt for systems, structures, and components.
TU Electric submitted a letter (logged TXX-98183) to the NRC, dated August 13, 1998, requesting an exemption from certain requirements applicable to ECCS evaluation models performed in accordance wita 10CFR50, Appendix K for Comanche Peak Steam Electric Station (CPSES) Units 1 and 2. This regulation imposes a 2% licensed power margin on ECCS evaluation models of light water power reactors licensed in accordance with the requirements of Appendix K.
Based on the proposed use of the improved Caldon, Inc. (Caldon) instrumentation to determine core power level, TU Electric is seeking an exemption to reduce the licensed power uncertainty required by 10CFR50, Appendix K, for modest increases of up to 1% in the licensed power levels using the current NRC approved methodologies.
The basis for the axemption is that the Caldon instrumentation provides a more accurate indication of feedwater flow (and correspondingly reactor thermal power) than assumed in the development of Appendix K requirements. Complete technical support for this conclusion is discussed in detail in the Caldon Topical Report ER-80P, " Improving Thermal Power Accuracg        and Plant Safety While increasing Operating Power Level Using the LEFM/7 System", which was previously submitted to the NRC on July 17,1998 by TU Electric letter, logged TXX-98180, from C. L. Terry. The improved thermal power measurement accuracy obviates the need for the full 2% power margin assumed in Appendix K, thereby increasing the thermal power available for electrical generation, while improving the certainty that actual reactor thermal power remains at or below the value used to analyze ECCS performance during a LOCA.
Along with the proposal to increase the reactor thermal power to 3445 MWt, TU Electric also proposes continued use of topical reports identified in CPSES Technical Specification 6.9.1.6b. These topical reports describe the NRC approved methodologies which support the CPSES safety analysis, including the small break and large break loss of coolant accidents analyses. In many of these topical reports, reference is made to the use of a 2% uncertainty applied to the reactor power, consistent with Appendix K. TU Electric proposes that these topical reports be approved for use consistent with the exemption request and this license amendment request, and further, the acknowledgment that the change in the power uncertainty does not constitute a significant change as defined in 10CFR50.46 and Appendix K.
II. DESCRIPTION OF TECHNICAL SPECIFICATIONS CHANGE REQUEST The Operating License for Unit 2 (NPF-89), section 2.C(1), identifies the maximum core thermal power level for which CPSES Unit 2 is authorized to operate as 3411 MWt. With the use of the Caldon, Inc., Leading Edge Flow Meter (LEFM(check)),
TU Electric proposes changing the maximum core power level to 3445 MWt.

_ _                                    _  _  to TXX-98265 Page 2 of 32 TU Electric proposes changing the definition of RATED THERMAL POWER (definition 1.28) in the Technical Specifications to reaa:
1.28 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3411 MWt for Unit 1 and 3445 MWt for Unit 2.
TU Electric also proposes changing the Unit 2 overpower N-16 trip setpoint from "s112%" of rated thermal power to "$111%" of rated thermal power and the Unit 2 overpower N-16 allowable value from "s114.5%"of rated thermal power to "s113.9%"of rated thermal power.
TU Electric also proposes adding the revised Unit 2 secondary steam flow at 100%
RATED THERMAL POWER of 15,268,734 lbs/hr to the BASES section for specification 3/4.7. In addition, a typographical error was found for the Unit 1 secondary steam flow. The value in the BASES section for specification 3/4.7 was "15,140,106" and should be "15,140,016" as noted in FSAR Table 10.1-1 and in Figure 10.1-1.
TU Electric has performed calculations to demonstrate that with the use of the LEFM(check), the uncertainty associated with the core power measurement may be reduced from the current value of 2% rated thermal power (RTP) to less than 1% RTP. Many of the safety analyses supporting the design and operation of CPSES include a power uncertainty of 2% RTP. Where practical, TU Electric proposes to use the existing analyses to justify operation at the uprated power condition. That is, analyses performed at 3411 MWt with the application of a 2%
RTP uncertainty allowance are equivalent to analyses performed at 3445 MWt with the application of a 1% RTP uncertainty allowance. G:ven this philosophy, the safety analysis limits of those Reactor Protection System functions that are expressed as a percentage of RTP will change. The affected functions are the overtemperature N-16, overpower N-16, and neutron flux, power range reactor trip functions. The overtemperature reactor trip function was explicitly analyzed in anticipation of the change to the RTP definition. Expressed as a percentage of RTP, the safety analysis limit used for the power range neutron flux and overpower reactor trip functions was decreased from 118% of RTP to 116.9% RTP. Sufficient margin to accommodate this change existed within the setpoint analysis for the high neutron flux function; however, it was necessary to revise the nominal trip setpoint and the allowable value associated with the overpower N-16 reactor trip function. The nominal trip setpoint was reduced from s112% RTP to s111% RTP. The allowable value was similarly reduced from s114.5% RTP to s113.9% RTP.
To summarize, TU Electric proposes changing the reactor core licensed power level in CPSES Unit 2 from 3411 MWt to 3445 MWt. Associated with this change are revisions to the nominal trip setpoint and allowable value of the overpower N-16 reactor trip function. The analyses for Unit 2, Cycle 5 to determine the core safety limits was performed using methodologies which are NRC approved and satisfy the applicable safety analyses limits. The core safety limits were developed in anticipation of a core thermal power of 3445 MWt and, therefore, remain valid for an increase in core thermal power from 3411 MWt to 3445 MWt, accounting for a 1% RTP uncertainty. Similarly, the overtemperature reactor trip

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l Attachment 2 to TXX-98265
! Page 3 of 32                                                                                        l l                                                                                                    i l                function nominal trip setpoint and allowable value, which protect the core safety   j l
limits, were developed to support operation at the 3445 MWt core power level.       l These changes are consistent with the Westinghouee improved Standard Technical Specifications (NUREG-1431, Revision 1).
Ill. ANALYSIS l
The Analysis is presented in three sections. An overview of the Leading Edge         l Flow Meter (LEFM) instrumentation used to measure the main feedwater mass flow and temperature for input into the plant calorimetric measurement is presented in Section A.
l Many of the existing safety analyses, particularly those presented in FSAR          1 Chapter 15, incorporate a core thermal power uncertainty allowance of 2% RTP.       !
With the use of the LEFM, it is proposed to reduce the required power uncertainty    l allowance to 1% RTP. These same analyses support the revised Rated Thermal Power with a power uncertainty allowance of 1% RTP. Th'se analyses were performed in accordance with the approved methodologies described in Technical       l Specification 6.9.1.6b, continue to comply with all applicable event acceptance     l criteria, and thus, will not be discussed further.                                   '
in contrast, many of the remaining safety analyses supporting the design and         ,
operation of CPSES were performed at the nominal,100% RTP condition. These           l analyses were reviewed; the evaluations supporting the Nuclear Steam Supply          ,
System (NSSS) and the Balance of Plant (BOP) are presented in Sections B and         l C, respectively. The Radiological Consequences are presented in section D. In        i general, these evaluations were performed using design conditions representative     '
of a 4.5% increase in the Rated Thermal Power which corresponds to the original CPSES Engineered Safeguards Design (ESD) rating. Recognizing that the evaluation for a larger power increase would encompass the 1% RTP increase and entail essentially the same amount of resources, the 104.5% RTP design condition was selected to support the potential pursuit of additional power increases in the future; however, only a 1% core thermal power increase is requested at this time.
l A.      LEADING EDGE FLOW METER (LEFM)
The LEFM is an ultrasonic flow meter consisting of a common control cabinet in the Control Room and spool pieces located in each unit's 30-inch main feedwater line. The LEFM uses acoustic energy pulses to determine the final feedwater mass flow rate. Transducers that transmit and receive the pulses are mounted in the LEFM spool piece at an angle of 45 degrees to the flow. The sound will travel         I faster when the pulse traverses the pipe with the flow and slower when the pulse traverses the pipe against the flow. The LEFM uses these transient times and time differences between pulses to determine the fluid velocity and temperature.
l              The LEFM computer also uses the calculated temperature and measured fluid pressure to account for changes in the LEFM's spool piece dimensions. The adjusted cross section area, velocity and temperature are then used to determine the final feedwater mass flow rate. The mass flow rate is displayed on the local display panel and transmitted to the plant process computer for calorimetrics. The feedwater mass flow rate is calculated and is used by the plant computer to 1
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l Attachment 2 to TXX-98265 Page 4 of 32 i
determine the reactor thermal output based on an energy balance of the secondary system.
The LEFM(check) is an improved system for use in determining and monitoring feedwater flow in nuclear power plants. The LEFM(check) provides increased safety by providing on-line verification of the accuracy of the feedwater flow and temperature measurements upon which NSSS thermal power determinations are based. In addition, the LEFM(check) provides a significant improvement in accuracy and an increase in reliability of flow and temperature measurements.
These accuracies are valid while the instruments are performing as designed. The on-line verification features of the LEFM4eck) provide the ability to assure on-line that performance is consistent with tiie design basis.
The LEFM(check) system measures the transit times of pulses of ultrasonic energy traveling along chordal acoustic paths through the flowing fluid. This te::hnology provides significantly higher accuracy and reliability than flow instruments which use differential pressure measurements and temperature instruments which use conventional thermocouple or resistance thermometers.
The LEFM(check) provides measurements of feedwater mass flow and temperature yielding a 10.6% RTP uncertainty in thermal power, substantially more accurate than the typical 12% RTP obtained with the conventional venturi-based instrumentation, or the 11.4% RTP uncertainty obtainable with precision, venturi-based instrumentation.
Furthermore, even with a 1% power increase associated with use of the LEFM(check), the probability of operation above 1.02 times the current licensed power level is still lower for the LEFM(check) than for the current instrumentation and no power increase. it is less likely that a plant would operate at greater than 101.8 percent above the current licensed power using the LEFM(check), with a one percent power increase in place, than with the current instrumentation and no power increase.
The LEFM(check) indications of feedwater mass flow and feedwater temperature will be directly substituted for the venturi-based mass flow indication and the RTD temperature indications currently used in the plant calorimetric measurement calculation performed with the plant computer. The plant computer will then calculate enthalpy and thermal power as it does now. As an alternative, the calorimetric power can be manually calculated, using the LEFM(check) indications and following prescribed procedures. In either of these cases, the use of the LEFM(check) will be limited to the calorimetric power determination. However, the venturi-based feedwater flow will continue to be used for feedwater control and other functions that it currently fulfills. Further, the venturi-based indication may be adjusted periodically on the basis of the LEFM(check) indication, so that it serves as a backup in the event that the LEFM(check) system is not available.
The use of the enhanced LEFM equipment supports the reduction in the l               uncertainty in the core thermal power measurement from 2% RTP to 1% RTP.
Based on the use of the Caldon LEFM(check), TU Electric submitted a request for exemption from Appendix K to 10CFR50. Appendix K currently requires that i

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l      Attachment 2 to TXX-98265 Page 5 of 32 l
l Emergency Core Cooling System (ECCS) evaluations be performed assuming reactor operating power to be 1.02 times the licensed core power level. The increase to 101% of the originallicense operating power level, accounting for a l                    1% RTP power uncertainty, continues to bound the original analyses of record for the ECCS and the safety analyses based on the ECCS.
B. NUCLEAR STEAM SUPPLY SYSTEM (NSSS)
The CPSES Power Uprate Project was completed consistent with the methodology                  :
established in WCAP-10263,"A Review Plan for Uprating the Licensed Power of a PWR Power Plant," dated 1983. The methodology establishes the general approach and criteria for uprate projects including the broad categories that must i                   be addressed, such as NSSS performance parameters, design transients, systems, components, accidents and nuclear fuel as well as interfaces between the NSSS and Balance of Plant (BOP) systems. Inherent in this methodology are l                    key points that promote correctness, consistency and licensability. The key points include the use of well-defined analysis input assumptions / parameter values, use            '
of NRC currently approved analytical techniques and use of CPSES currently                      4 applicable licensing criteria and standards. The evaluations and analyses                       !'
described in this document have been completed consistent with this l                    methodology.
!                                                                                                                  J l             B.1    NSSS PARAMETERS INPUT PARAMETERS AND ASSUMPTIONS i
The major inputs used in the development of the NSSS parameters used by Westinghouse in the re-evaluation of the NSSS design are summarized below.
The values der.oted as " current" were used to support the original NSSS design evaluations performed by Westinghouse. Other less conservative values may have been used for other types of analyses.
The NSSS power level for the uprating analysis was established as 3582 MWt (3565 MWt core). This is approximately 4.5% higher than the current NSSS power rating of 3425 MWt (3411 MWt core). The 4.5% increase was selected to allow for the planned 1% increase and provide margin for any future increases, up to 4.5%,
if desired. This core thermal power rating corresponds to the original CPSES ESD rating.
The current NSSS power level assumes a net RCS heat input of 14 MWt (3425 -
3411 MWt) from the reactor coolant pumps. The NSSS power level for the uprating is based on a net RCS heat input of 17 MWt (3582 - 3565 MWt) for the reactor coolant pumps, which more accurately represents the actual net reactor coolant pump heat input addition.
i                    The reactor coolant system flowrate of 97,900 gpm/ loop was revised from the currently value of 95,700 gpm/ loop. This flow was applied for all cases, even i                   those which assumed 0% steam generator tube plugging (SGTP).
Two values of SGTP were assumed: 0% and 10%. The 10% SGTP level supports a peak SGTP level of 15% ln any one steam generator, provided that the average level of plugging of all four steam generators is 510%.

l    Attachment 2 to TXX-98265 l    Page 6 of 32 l                  A range of full power normal operating T vo 4 from 585.7 degrees F to 592.7 degrees F was selected for the analyses. This range represents a 13.5 degrees F variation from the current design TAVG of 589.2 degrees F and will allow for greate' operating flexibility.
A full power feedwater temperature range of 390 degrees F to 444.6 degrees F was selected for the analyses. This range supports greater operating flexibility by allowing the plant to reduce feedwater temperature to maintain 100% thermal power in the event that the Steem Generator outlet steam prassure is insufficient to      '
satisfy the turbine volumetric flow limit.
l          DISCUSSION OF PARAMETER CASES Table IV-1 summarizes the NSSS parameter cases which were developed and used as the basis for the uprating project. The original design parameters are also shown for ccmparison purposes. A description of the 3 uprated cases follows.
Case 1 represents the uprated conditions with the reactor lower reactor vessel average temperature of 585.7 degrees F and a 10% SGTP level. It yields the lowest possible initial primary side temperatures for the analyses as well as the l                 minimum secondary side steam generator temperature, pressure and flow.
Case 2 represents the uprated conditions with the higher reactor vessel average temperature of 592.7 degrees F and a 10% SGTP level. It yields the highest possible initial primary side temperatures for the analyses.
Case 3 represents the uprated conditions with the higher reactor vessel average temperature of 592.7 degrees F and a 0% SGTP level. It yields the highest possible initial steam temperature, steam pressure, and steam flow.
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l l      Attachment 2 to TXX-98265 t-Page 7 of 32 Table IV-1 NSSS Revised Design Parameters l           BASIC COMPONENTS Reactor Vessel, ID, in.                      173                              Isolation Valves                                   No Core                                                                          Number of Loops                                    4
,                   Number of Assemblies                 193                              Steam Gener.itor i                   Rod Array                            17x17                                 Model                                         D4 (Ul)/D5 (U2)
Rod OD, in.                         0.374                                  shcIl Design Pres ure, psia                    9300 Number of Grids                     8                                 Reactor Coolant Pump Active Fuel Length, in.              144                                   Model/ Weir                                   93A/Yes Number of Control Rods                      53                                     Pump Motor, hp                                7000                       j Frequency, llz                                60 UPRATING OPERATING DESIGN PARAMETERS                         N LOOP CASE 1(2)                    CASE 2(2)              CASE 3(2)
NSSS Power, %                                       100        100                       100                   100 MWt                                         3425      3582                      3582                   3582 10' BTU /hr                               i I,687   12,222                     12,222                12,222 i           Reactor Power, MWt                                  3411      3565                      3565                   3565 j                     10' BTU /hr                               11,639     12,164                    12,164                12,164                                       :
Assumed Flow, Loop gpm                              95,700    97,900                    97,900                 97,900 Reactor 10* Ib/hr                           142.0      146.1                     144.6                 144.6 Reactor Coolant Pressure, psia                      2250      2250                       2250                  2250 Core Bypass, %                                      5.8       5.8                        5.8                   5.8 Reactor Coolant Temperature, degrees F Core Outlet                                 622.1     619.4                     626.0                  626.0 Vessel Outlet                               618.8     616.1                     622.7                  522.7 Core Average                                592.5     589.1                      596.2                 596.2 Vessel Average                              589.2     585.7(3)                   592.7(3)              592.7(3)                                     j Vessel / Core inlet                         559.6     555.3                      562.7                 562.7 Steam Generator Outlet                      559.3     555.0                      562.4                 562.4 Steam Generator Steam Temperature, degrees F                $44.6     533.6/537.2                541.9/545.4           545.7/549.2 Steam Pressure, psia                        1000      913(IV941(1)               978(1)/1007(1)        1009(l)/1039(1)
Steam Flow,10' lb/hr total                  15.14      15.89/14.77               15.93/14.81           15.96/14.83 Feed Temperature, degrees F                 440       444.6/390                 444.6/390              444.6/390 Moisture, % max.                            0.25      0.25                       0.25                  0.25                                         I Tube Plugging, %                            0          10                        10                    0 Zero Load Temperature, degrees F                    557       557                        557                   557 i
HYDRAULIC DESIGN PARAMETERS Mechanical Design Flow, gpm                         105,000 Minimum Measured Flow, gpm tc+.at                   394,000 FOOTNOTES:
(1)            Steam pressures listed are applicable to Unit 2 ;enerators, add 7 psi to reDect steam pressure for Unit I generators.
 ,.         (2)            For system and component analysis only.                                                                                                      !
(3)            Tm; range +/-3.5 degrees F.
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l to TXX-98265 Page 8 of 32 B.2      DESIGN TRANSIENTS NSSS DESIGN TRANSIENTS The revised design conditions in Table IV-1 and the NSSS design transients applicable to the uprated conditions serve as primary inputs to the evaluation and analysis of the st us and fatigue ana.;ses of the NSSS systems and components.
Current primary anJ secondary design transieMs were reviewed in order to determine their continued applicability to upra' d conditions.
Primary Side Transients The review of the primary side design conditions lit.,ted in Table IV-1 indicates that the full power values of vessel outlet, vessel inlet and vessel average temperatures (hot leg cold leg, and loop average temperatures) vary by less than 4 degrees F from the previously applicable design values. Given the conservative assumptions used to develop the current design transients (e.g., initial conditions, unavailability of control systems during certain transients), a 4 degrees F change in primary side full power temperatures is considered insignificant during all transient conditions. Therefore, the revised conditions have negligible impact on the primary side design transients, and the prr'iously applicable NSSS design transients for the primary side continue to apply, without modification, at the uprated conditions.
Secondary Side Transients With regard to secondary design parameters, the revised design conditions in Table IV-1 indicate that the plant may operate with lower full power values for steam temperature, steam pressure and feedwater temperature. Lower nominal steam temperatures (e.g., from 544.6 to 533.6 degrees F) and pressures (e.g.,
from 1000 psia to 913 psia) result in larger changes from initial conditions than those currently reflected in the current NSSS design transients. Similarly, a 50 degrees F reduction in feedwater temperature (i.e., from 440 to 390 degrees F) may result in operating conditions that are more limiting than the ones determined at higher feedwater temperatures for the components evaluated at uprated conditions. Therefore, some of the existing secondary side design transients were modified to reflect the lower nominal feedwater and steam temperatures. These modified design transients were used in the steam generator analyses.
AUXILIARY EQUll' MENT DESIGN TRANSIENTS The review of the NSSS auxiliary equipment design transients was based on a comparison between the revised operating conditions in Table IV-1 and the parameters which make up the current auxiliary equipment design transients.
A review of the current auxiliary equipment transients determined that the only transients impacted by the revised conditions are those temperature transients impacted by e change in Teoto. Those transients are currently based on an assumed full lead NSSS worst case Teoto of 560 degrees F and have been modified to reflect the new worst case value of 563 degrees F.
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Attachment 2 to TXX-98265 Page 9 of 32 CONCLUSIONS The revised design conditions did not require a change to the primary side design transients. However, the revised condition did require modification of several secondary side transients which have been considered in the steam generator design discussed later, in addition, the auxiliary transient temperature curves were changed to accommodate the Teoto increase from 560 degrees F to 563 degrees F and have been considered in the design of the NSSS auxiliary equipment discussed later.
B.3    NSSS SYSTEMS This section presents the results of the evaluations and analyses performed in the   l NSSS systems area to support the revised design conditions in Table IV-1. The
,                systems addressed in this chapter include Fluid Systems and NSOS/ BOP interface Systems. The results and conclusions of each analysis are presented within each subsection.
NSSS FLUID SYSTEMS Reactor Coolant System The Reactor Coolant System (RCS) consists of four heat transfer loops connected in parallel to the reactor vessel. Each loop contains a reactor coolant pump (RCP), which circulates the water through the loops and reactor vessel, and a steam generator (SG), where heat is transferred to the main steam system (MSS).
In addition, the RCS contains a pressurizer which controls the RCS pressure          l through electrical heaters, water sprays, power operated relief valves (PORVs) and spring loaded safety / relief valves. The steam discharged from the PORVs and safety / relief valves flows through interconnecting piping to the pressurizer relief tank (PRT).
Assessments were performed to demonstrate that the RCS design basis functions        ,
could still be met at the revised design conditions.
Pressurizer spray flow capability was calculated considering the T,vo range of       4 585.7 to 592.7 degrees F. The calculations demonstrate that the minimum              !
required spray flow of 900 gpm can be achieved over the entire range of anticipated RCS process conditions.
Also, the maximum expected Taor at uprated conditions is 622.7 degrees F. This temperature is well within the RCS toop design temperature of 650 degrees F.
With respect to the PRT, the revised T,vo range will change the nominal full load pressurizer steam volume at uprated conditions. In general, the reference nominal pressurizer level is coordinated with RCS T ivo such that an increase in T,vo raises the nominal pressurizer reference level condition. With respect to the PRT discharge analysis, a lower RCS T 4vo condition is more limiting than a higher RCS T AVG Condition, since pressurizer level is lower and steam volume is larger. It was determined that the current PRT level setpoint was still sufficient to accommodate the expected increase in steam volume from the reduced T,vo from 589.2*F to 585.7 degrees F.

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Attachment 2 to TXX-98265 i Page 10 of 32 Chemical and Volume Control System l
The Chemical Volume and Control System (CVCS) provides for boric acid addition, chemical additions for corrosion control, reactor coolant clean-up and
:               degasification, reactor coolant make-up, reprocessing of water letdown from the RCS, and RCP seal water injection. During plant operation, reactor coolant flows through the shell side of the regenerative heat exchanger and then through a letdown orifice. The regenerative heat exchanger reduces the temperature of the reactor coolant and the letdown orifice reduces the pressure. The cooled, low pressure water leaves the reactor containment and enters the auxiliary building. A second temperature reduction occurs in the tube side of the letdown heat exchanger followed by a second pressure reduction due to the low pressure letdown valve. After passing through one of the mixed bed demineralizers, where ionic impurities are removed, coolant flows through the reactor coolant filter and enters the Volume Control Tank (VCT).
In the assessment of CVCS operation at revised RCS operating temperatures, the maximum expected RCS Teoto must be less than or equal to the applicable CVCS design temperature and less than or equal to the heat exchanger design inlet operating temperature. The former cnterion supports the functional operability of the system and its components. The latter criterion supports the conclusion that the current heat exchanger design operating conditions remain bounding.
With regard to the CVCS thermal performance, the Teoto of 562.7 degrees F is slightly higher than the (esign system inlet temperature of 560 degrees F. This slight increase in hect cxchanger inlet temperature can be accommodated by the system. Although the letdown temperature at the outlet of the regenerative heat exchanger wi!I increase slightly, the letdown heat exchanger will reduce the letdown system temperature to its design value. The increase in heat transfer to the component cooling wate r system will be less than 1% due to '% increase in regenerative heat exchanger outlet temperature. The excess l#own path is used to process excess effluents associated with fluid expansion during plant heatup and thus, is unaffected by the revised Tcoto at full power conditions. The excess letdown heat exchanger outlet flow is throttled to maintain the desired outlet temperature and efflux. Therefore, operation of this system is unaffected by the temperature change.
Safety injection System The Safety Injection System (SIS) is an Engineered Safeguards System used to mitigate the effects of postulated design basis events. The basic functions of this system include providing short- and long-term core cooling, and maintaining core shutdown reactivity margin. The SIS is also referred to as the emergency core cooling system (ECCS).
The SIS is comprised of three subsystems. The passive portion of the system is the four accumulator vessels which are connected to each of the RCS cold leg l               pipes. Each accumulator contains borated water under pressure (nitrogen cover
;               gas). The borated water automatically injects into the RCS when the pressure l               within the RCS drops below the operating pressure of each of the accumulators.
The ' active' part of the SIS injects borated water into the reactor following a break in either the reactor or steam systems in order to cool the core and prevent an

l l
; Attachment 2 to TXX-98265 Pzge 11 of 32 uncontrolled return to criticality. Two safety injection (SI) pumps and two residual l                 heat removal (RHR) pumps take suction from the Refueling Water Storage Tank (RWST) and deliver borated water to four cold leg connections via the accumulator discharge lines. In addition, two centrifugal charging pumps take suction from the RWST on Si actuation and provide flow to the RCS via separate Si connections on
;                 each cold leg. This arrangement of SI pumps can provide safety injection flow at l                 any RCS pressure up to the set pressure of the pressurizer safety valves.
The revised design conditions have no direct effect on the overall performance capability of the SIS. These systems will continue to deliver flow at the design l
basis RCS and containment pressures since there are no changes in the RCS operating pressure.
l i                 Other NSSS Systems                                                                              l l
l                 Other NSSS systems which were reviewed and found to be acceptable for power                     '
l                 uprate w9re the Boron Recycle System, the Boron Thermal Regeneration System,                    1 and the Gaseous Waste Processing System.
                                                                                                                  )
l                 Cold Overpressure Mitigation System (COMS)
COMS is designed to protect the RCS from overpressure events when the RCS is i                 below a temperature of approximately 350 degrees F. Changes to full power l                 operating parameters, such as NSSS power, TAVG, and RCS pressure do not                         l l                 impact COMS. An increase in steam generator tube plugging to 10%, which                         l l                 reduces the RCS volume, can have a smallimpact on the peak RCS pressure
!                 experienced during the design basis mass injection event. An evaluation i                 confirmed that the current mass injection event is still valid for the revised design conditions.
NSSSI BOP INTERFACE SYSTEMS l                 The following Balance-of-Plant (BOP) fluid systems were reviewed for cornpliance with Westinghouse Nuclear Steam Supply Systems (NSSS)/ BOP interface guidelines. These guidelines do not necessarily reflect design requirements; some l                 are provided to ensure the assumptions of the NSSS analyses remain valid, and                   l l                 the remainder are intended to facilitate operation of the plant.                                !
A comparison of the revised design conditions with the current design conditions previously evaluated for systems and components indicates differences that could                l impact the performance of these systems.                                                        l 1
Main Steam System
;                 The uprating coupled with the potential reduction in full-load steam pressure to the average minimum value of 913 psia impacts the main steam line pressure drop.
At the revised design conditions, the steam line pressure drop could increase by as much as 21.8 percent due to the increased steam flow and lower steam i                 density.
The following summarizes the evaiuation of the major steam system components relative to the power uprate conditions. The major components of the Main Steam i
l l
to TXX-98265 Pcge 12 of 32 System (MSS) are the Steam Generator Main Steam Safety Valves (MSSVs), the SG Atmospheric Relief Valves (ARVs), and the Main Steam Isolation Valves (MSIVs).
Steam Generator Main Steam Safety Valves The MSSVs must have sufficient capacity so that main steam pressure does not exceed 110 percent of the MSS design pressure (the maximum pressure allowed by the ASME B&PV Code). Each operating unit at Comanche Paak has twenty MSSVs with a total capacity of 18.19 x 105 lb/hr. Therefore, br. sed on the revised design conditions, the capacity of the installed MSSVs meets the required sizing criterion.
Steam Generator Atmospheric Relief Valves (ARVs)
The primary function of the ARVs is to provide a means for decay heat removal and plant cooldown by discharging steam to the atmosphere when either the condenser, the condenser circulating water pumps, or steam dump to the condenser is not available. Under such circumstances, the ARVs in conjunction with the Auxiliary Feedwater System (AFWS) permit the plant to be cooled down from the pressure setpoint of the lowest-set MSSVs to the point where the Residual Heat Removal System (RHRS) can be placed in service. During cooldown, the ARVs are either automatically or manually controlled. Each ARV controller automatically compares steam line pressure to the pressure setpoint, which is manually set by the plant operator.
In the event of a steam generator tube rupture event in conjunction with loss of offsite power, the ARVs are used to cool down the RCS to a temperature that permits equalization of the primary and secondary pressures at a pressure below the lowest-set MSSV. RCS cooldown and depressurization are required to preclude steam generator overfill and to terminate activity release to the atmosphere.
For the revised design conditions, each steam generator ARV is required to have a capacity at least equal to 62,150 lbs/hr at 100 psia inlet pressure. At these conditions, this capacity permits a plant cooldown to RHRS operating conditions in 5 hours (at a cooldown rate of 50 degrees F/hr) assuming a minimum of 2 hours at hot standby. This sizing is compatible with normal cooldown capability and minimizes the water supply required by the AFWS. This is based on one train of auxiliary feedwater (AFW) operating and flow going through three SGs. The design capacity of the installed ARVs meets the required sizing criterion.
Main Steam isolation Valves and Main Steam isolation Bypass Valves The MSIVs are located outside the containment and downstream of the MSSVs.
The valves function to prevent the uncontrolled blowdown of more than one steam generator and to minimize the RCS cooldown and containment pressure to within acceptable limits following a main steam line break. To accomplish this function, the original design requirements specified that the MSIVs must be capable of closure within 5 seconds of receipt of a closure signal against steam break flow conditions in either the forward or reverse direction.

Attachment 2 to TXX-98265 j Page 13 of 32 l               Rapid closure of the MSIVs following postulated steam line breaks causes a l               significant differential pressure across the valve seats and a thrust load on the main steam system piping and piping supports in the area of the MSIVs. The worst cases for differential pressure increase and thrust loads are controlled by the steam line break area (i.e., mass flow rate and moisture content), throat area of the steam generator flow restrictors, valve seat bore, and no load operating pressure. Since these variables and no-load operating pressure are not impacted by the revised design conditions, the design loads and associated stresses resulting from rapid closure of the MSlVs will not change. Consequently, the revised design conditions have no significant impact on the interface requirements for the MSIVs.
The MSIV bypass valves are used to warm up the main steam lines and equalize pressure across the MSIVs prior to opening the MSIVs. The MSIV bypass valves perform their function at no-load and low power conditions where the revised design conditions have no significant impact on main steam conditions (e.g., steam flow and steem pressure). Consequently, the revised design conditions have no significant impact on the interface requirements for the MSIV bypass valves.
Steam Dump System The steam dump system creates an artificial steam load by dumping steam from upstream of the turbine valves to the main condenser. The Westinghouse sizing criterion recommends that the steam dump system (valves and pipe) be capable of discharging 40 percent of the rated steam flow at full-load steam pressure to permit the NSSS to withstand an externalload reduction of up to 50 percent of plant rated electrical load without a reactor trip. To prevent a trip, this transient requires all NSSS control systems to be in automatic, including the Reactor Control System, which accommodates 10% of the load reduction. A steam dump capacity of 40 percent of rated steam flow at fullload steam pressure also prevents MSSV lifting following a reactor trip from full power.
The steam dump system is composed of twelve condenser steam dump valves.
The total capacity for all twelve valves exceeds the Westinghouse sizing criterion of 40 percent of rated steam flow.
NSSS operation within the revised design conditions at lower steam generator pressures and higher steam flows will result in a reduced steam dump capability.
An evaluation for these conditions indicates that the total steam dump capacity still exceeds the required sizing criterion.
In addition, to provide effective control of flow on large step load reductions or plant trip, the steam dump valves are required to go from full-closed to full-open in 3 seconds at any pressure between 50 psiless than fullload pressure and steam generator design pressure. The dump valves are also required to modulate to i               control flow. These requirements remain applicable for the revised design l               conditions.
L i

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Attachment 2 to TXX-98265 Page 14 of 32 l               Condensate and Feedwater System The Condensate and Feedwater System (C&FS) must automatically maintain steam generator water levels during steady-state and transient operations. The revised design conditions will impact both feedwater volumetric flow and system pressure drop.
The major components of the C&FS are the Main feedwater Line Flow Restricting
;               Orifices, Feedwater Isolation Valves, the Feedwater Control Valves, and the Condensate and Feedwater System Pumps.
Main Feedwater Line Flow Restricting Orifices The main feedwater line to each steam generator incorporates a Feedwater Bypass System (FBS). The function of the FBS is to minimize the potential occurrence of water hammer in the steam generators, and to mitigate flow induced tube vibration in the steam generators. The FBS includes the feedwater split flow bypass line which connects the main feedwater line to the feedwater preheater bypass line inside containment. The feedwater preheater bypass line delivers flow                      '
into the steam generator auxiliary feedwater nozzle. The main purpose of the feedwater split flow bypass line is to provide a flow split between the main                          ,
feedwater nozzle and the auxiliary feedwater nozzle at 100 percent of rated flow.                      l The required flow split between the auxiliary nozzle and the main feedwater nozzle is facilitated by a flow restricting orifice installed in each main feedwater line just downstream of the feedwater split flow bypass line connection.
The maximum permissible flow to the steam generator main nozzles is 3.48 x 108 lb/hr at full load. A high flow alarm is provided to facilitate operations and the alarm is presently set at 3.39 x 10'lb/hr based on the main nozzle flow limit less allowance for flow measurement uncertainty, Note, since the alarm set point correction for flow measurement uncertainty increases as total feedwater flow increases, the alarm set point will decrease to 3.376 x 10' lb/hr based on the l
highest feedwater flow for the revised design conditions.
An evaluation of the fullload main feedwater flow split for the range of revised design conditions indicates that the main nozzle flow rate will fall below the revised alarm set point with the existing orifices. Therefore, the installed orifices should preclude the steady state feedwater flow into the main nozzles from exceeding the specified limit.
Feedwater Isolation Valves /Feedwater Control Valves The feedwater isolation valves (FIVs) are located outside containment and downstream of the feedwater control valves (FCVs). The valves function in conjunction with the primary isolation signals to the FCVs and backup trip signals to the feedwater pumps to provide redundant isolation of feedwater flow to the i
steam generators following a steam line break or a malfunction in the steam i               generator level control system. Isolation of feedwater flow is required to prevent
.               containment overpressurization and excessive reactor coolant system cooldowns.
,               To accomplish this function, the FCVs and the backup FIVs must be capable of i
f

1  to TXX-98265                                                                            l Pcge 15 of 32 l
l l
fast closure, that is within about 5 seconds following receipt of any feedwater        ,
isolation signal.                                                                      l l
l The quick-closure requirements imposed on the FIVs and the backup FCVs causes dynamic pressure changes that may be of large magnitude and must be considered in the design of the valves and associated piping. The worst loads occur following a steam line break from no load conditions with the conservative       l assumption that all feedwater pumps are in service providing maximum flow              ;
following the break. Since these conservative assumptions are not impacted by          .
the revised design conditions, the design loads and associated stresses resulting      I from rapid claure of these valves will not change.
1 Condensate and Feedwater System (C&FS) Pumps The C&FS available head in conjunction with the FCV characteristics must provide sufficient margin for feed control to ensure adequate flow to the steam generators during steady-state end transient operation. A continuous steady feed flow should be maintained at all loads. To assure stable feedwater control, with variable speed feedwater pumps, the pressure drop across the FCVs at rated flow (100 percent power) should be approximately equal to the dynamic losses from the feed pump discharge through the steam generator. In addition, adequate margin should be available in the FCVs at fullload conditions to permit a C&FS delivery of 103 percent of rated flow with a 75 psi pressure increase above the full load          i pressure with the FCVs fully open. Because actual operating conditions may change over time due to tube plugging, etc., the FCV and pump speed controller may be recalibrated as necessary to optimize performance.
To provide effective control of flow during normal operation, the FCVs are required to stroke open or closed in 20 seconds over the anticipated inlet pressure control     ,
range (approximately 0-1600 psig). Additionally, rapid closure of the FCVs is          l required in 5 seconds after receipt of a trip close signalin order to mitigate certain transients and accidents. These requirements are still applicable at the uprated conditions. Other C&FS evaluations, including the feedwater and condensate pumps, are contained in the Balance of Plant section.
Auxiliary Feedwater System (AFWS)
The AFWS supplies feedwater to the secondary side of the steam generators at times when the normal feedwater system is not available, thereby maintaining the heat sink of the steam gerierators. The system provides feedwater to the SGs during normal unit startup, hot standby, and cooldown operations and also functions as an Engineered Safeguards System. In the latter function, the AFWS is directly relied upon to prevent core damage and system overpressurization in the event of transients and accidents such as a loss of normal feedwater or a secondary system pipe break.

l Attachment 2 to TXX-98265                                                                                                      ,
Page 16 of 32                                                                                                                   l l
Auxillary Feedwater Storage Requirements                                                            ;
l The AFWS pumps are normally aligned to take suction from the condensate                             1 l                                storage tank (CST). To fulfill the Engineered Safety Features (ESF) design                          i functions, sufficient feedwater must be available during transient or accident conditions to enable the plant to be placed in a safe shutdown condition.
Sufficient CST useable inventory must be available to bring the unit from full powar to hot standby conditions, maintain the plant at hot standby for 4 hours, and I
then cooldown the RCS to the residual heat removat system cut-in temperature (350 degrees F)in 5 hours. in addition, the CPSES ncensing bases requires that the minimum useable inventory must also be adequate to support 18 hrs of decay heat removal at no load plant conditions. In light of these requirements, the current analysis concluded that the tank must accommodate a minimum useable i                                inventory of 249,100 gals.
1 An analysis was performed at the revised design conditions which demonstrated that the existing minimum useable inventory of 249,100 gals is adequate.
l Steam Generator Blowdown System The Steam Generator Blowdown System is used in conjunction with the Chemical Feed and Sampling Systems to control the chemical composition of the steam generator shell water within the specified limits. The blowdown system also controls the buildup of solids in the steam generator water.
l                                Two blowdown locations are provided for each Steam Generator, a tube sheet connection and a shell connection. The tube sheet maximum allowable blowdown flow rate decreases with power. For the subject 1% power increase, allowable
,                                tube sheet blowdown flow declines by approximately 1%. However, the total l
allowable blowdown flow from each steam generator can be maintained constant since the shell blowdown flow can be increased by the same amount. Therefore, the ability of the system to control the rate of addition of dissolved solids to the                1 l                                secondary systems by condenser leakage or makeup water is not impacted by the                        1 l                                revised design conditions. Also, the reduction in tube sheet blowdown capability is not significant in terms of the ability of the system to control the rate of generation of particles by eiosion-corrosion effects within the secondary systems, since the required blowdown rate from the tube sheet to control particulate is well within the                 i allowable tube sheet blowdown flow rate.
The actual required blowdown flow rates during plant operation will not be significantly impacted by the revised design conditions, since neither the rate of                   i addition of dissolved solids or the rate of addition of particulates into the steam                  l generators will be significantly impacted.                                                           l l

Attachment 2 to TXX-98265 Page 17 of 32 CONCLUSIONS The following is a brief summary of the NSSS/ BOP interface evaluation conclusions.
Main Steam System Operation at reduced steam pressures and corresponding higher pressure drops will have a negative effect on plant heat rate.
The capacity of the installed MSSVs meets the Westinghouse sizing criterion for the proposed range of NSSS operating conditions.
i The capacity of the installed ARVs meets the Westinghouse sizing criterion for the proposed range of NSSS operating conditions.
The MSIVs and MSIV bypass valves are not adversely impacted by the uprating.
Steam Dump System The capacity of the steam dump system continues to comply with the Westinghouse sizing criterion for the proposed range of NSSS operating conditions.
Condensate and Feedwater System For the range of uprated NSSS operating conditions, the Unit 2 main feedwater flow restriction orifices provide an acceptable flow split between the steam generator main nozzles and auxiliary nozzles at full load.
Auxiliary Feedwater System The minimum flow requirements of the AFWS are dictated by accident analysis which are unaffected by the 1% uprate. Therefore, the AFWS performance remains acceptable at the uprated operating condition.
The CST minimum useable inventory of 249,100 gals remains adequate and complies with the plant design bases requirements for the range of uprated NSSS operating conditions.
Steam Generator Blowdown System The actual required blowdown flow rates during plant operation are not significantly impacted by power uprate, since neither the rate of addition of

i i
I Att chment 2 to TXX-98265 j Prg)18 of 32 dissolved solids nor the rate of addition of particulates into the steam generators is significantly impacted by power uprate.
B.4    NSSS COMPONENTS REACTOR VESSEL The reactor vessel (RV) was evaluated at the revised design conditions with respect to: the structural acceptability of the vessel; and the reactor vessel integrity in terms of the impact due to neutron fluence.
Reactor Vessel Structural Evaluation Evaluations were performed to assess the effects of the revised design conditions in Table IV-1 on the most limiting vessel locations with regard to ranges of stress intensity and fatigue usage factors in each of the regions, as identified in the reactor vessel stress reports and addenda. The evaluations considered a limiting worst case set of operating parameters from among the high temperature conditions, the low temperature conditions,'and the original design basis. As a result, all of the design conditions in Table IV-1 are fully enveloped by the evaluations, and reactor vessel operation in accordance with the parameters for the remainder of the current operating licenses is justified.
The revised design conditions, in particular the Tw and Tem increase the maximum ranges of stress intensity and fatigue usage for the primary outlet and inlet nozzles. These increased stress intensities remain within the acceptance criterion of 3Sm,and the maximum cumulative fatigue usage factors continue to remain below the acceptance limit of 1.00.
Reactor Vessel Integrity - Neutron Irradiation Several analyses are performed to determine the impact that neutron irradiation has on the reactor vesselintegrity. The most critical area is the beltline region of the reactor vessel since it is predicted to be most susceptible to neutron damage.
These analyses include a surveillance capsule withdrawal schedule, heat-up and cooldown pressure-temperature limit curves, pressurized thermal shock calculations and upper shelf energy evaluations. All of these analyses and evaluations can be affected by changes in the neutron fluences and operating temperatures and pressures. A 4.5% power uprating would be expected to increase the neutron fluences due to the increased power distributions. However, it was determined that based on the CPSES operation, the calculated fluences, used in the current vessel design, bound the fluences expected to occur with the revised design conditions. In addition, the changes in the RCS temperature are
,               well within the temperature ranges assumed in these analyses. Thus, the revised l               design conditions do not invalidate these analyses.

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Attachment 2 to TXX-98265 Page 19 of 32 l
l l            REACTOR INTERNALS l
The reactor internals support and orient the fuel and control rod assemblies, absorb control rod assembly dynamic loads and transmit these and other loads to the reactor vessel. The internals also direct flow through the fuel assemblies, provide adequate cooling to various internal structures and support the in-core instrumentation. The increase in thermal design flow and changes in the RCS                        ;
I temperatures, reported in Table IV-1, produce changes in the boundary conditions experienced by the reactor internals components. Also, increases in core thermal power may increase nuclear heating rates in the lower core plate, upper core plate and baffle-barrel former region.
l                   Core Bypass Flou Calculation Bypass flow corresponds to the amount of reactor coolant flow that bypasses the core region and is not considered effective in the core heat transfer process. The principal core bypass flows are the baffle barrel region, vessel head cooling spray nozzles, vessel outlet nozzle gap, baffle plate cavity gap and the thimble tubes.
Since variations in the size of some of the bypass flow paths, such as gaps at the vessel core barrel and outlet nozzles, occur during manufacturing or change due to different fuel assembly designs or to changes in the RCS conditions, plant specific as-built dimensions. Therefore, analyses are performed to determine core bypass flow values to either demonstrate that the design bypass flow limit is not exceeded or to determine a revised design core bypass flow.
The analysis results indicate that the bypass flow remains less than the 5.8%
assumed in the development of the revised design conditions and, therefore are acceptable.
RCCA Drop Time Analyses                                                                            l The RCCAs represent the interface between the fuel assemblies and the other internal components. The Technical Specifications require that the RCCA drop time be less than or equal to 2.4 seconds. The revised design conditions, in particular the reduced Tcold, can increase the drop time due to the increased fluid density. An evaluation was performed to demonstrate continued compliance with the current technical specification value at the revised design conditions. Further, the RCCA drop times are explicitly confirmed, by measurement after each refueling outage, to meet the times assumed in the accident analyses.
Hydraulic Lift Forces The reactor internals hold-down spring is essentially a large diameter belleville type spring of rectangular cross section. The purpone of this spring is to maintain a net clamping force between the reactor vessel head flange and upper internals flange and the reactor vessel shell flange and the core barrel flange of the internals, An evaluation was performed to determine hydraulic lift forces on the various reactor internal components to ensure that the reactor internals assembly

i                                                                                                    l l
; Attachment 2 to TXX-98265 Paga 20 of 32 l               remains seated and stable for all conditions. Increases in mechanical design flow    l l
or changes in operating temperatures can reduce the clamping force. It was           I determined that, with the revised design conditions, the reactor internals assembly would remain seated and stable.
Mechanical Evaluations l               The revised design conditions do not affect the current design bases for seismic and LOCA loads. Thus, it was not necessary to re-evaluate the structural affects from seismic OBE and SSE loads and the LOCA hydraulic and dynamic loads.
With regards to flow and pump induced vibration, the current analysis is based on a mechanical design flow which was not impacted by the revised design conditions. The revised design conditions slightly alter the Teoto and T sor fluid densities which will slightly change the forces induced by flow. However, these changes are insignificant when compared to the current design temperature ranges. Thus, the impact of the revised design conditions on the mechanical loads is acceptable.
Structural Evaluations Structural evaluations are required to demonstrate that the structural integrity of the reactor components is not adversely affected by the change in RCS conditions and transients and/or by secondary effects of the change on reactor thermal hydraulic or structural performance. The presence of heat generated in reactor internal components, along with the various fluid temperatures, results in thermal gradients within and between components. These thermal gradients results in thermal stresses and thermal growth which must be accounted for in the design and analysis of various components. The core support structures affected by the revised design conditions (Table IV-1) are discussed in the following sections.
The primary inputs relevant to the evaluations are the revised design conditions in Table IV-1 and the gamma heating rates and the NSSS design transients which remain bounding for the revised design conditions. The gamma heating rates were increased to account for an increase in core power to 3565 MWt.
The reactor internals components subjected to heat generation effects (either directly or indirectly) are the upper and lower core plates, the lower core support, the core baffle plates, the former plates, the core barrel, the neutron panel, the baffle-former bolts and the barrel-former bolts. Note, however, that due to relatively low heat generation rates the upper core plate, the lower core support, and neutron panels experience little, if any, temperature rise over the surrounding reactor coolant.
Baffle Barrel Region Evaluations t
The baffle-barrel regions consist of a core barrelinto which baffle plates are installed, supported by bolting interconnecting former plates to the baffle and core barrel.

l Attachment 2 to TXX-98265 Page 21 of 32 The baffle to former bolts restrain the motion of the baffle plates that surround the core. These bolts are subjected to primary loads consisting of deadweight, hydraulic pressure differentials, seismic loads, as well as secondary loads consisting of preload, and thermalloads resulting from RCS temperatures and gamma heating rates. The baffle to former bolt thermalloads are induced by differences in the average metal temperature between the core barrel and baffle plate. In addition to providing structural restraint, me baffles also channel and direct coolant flow such that a "coolable core geometry can be maintained."
The thermal stresses in the core barrel shell in the core active region are primarily due to temperature gradients through the thickness of the core barrel shell. These temperature gradients are caused by the fluid temperatures between the inside and outside surfaces and the contribution of gamma heating,                                  l To demonstrate that the baffle-barrel metal temperatures remain bounded by the generic analysis, a comparison of core power in the fuel assemblies at the periphery was made. The results indicated that the current structural and thermal analysis of record for the baffle-barrel region remain bounding for the revised design conditions. Thus, the baffle-barrel region is structurally adequate for the revised design conditions.
Lower Core Plate Structural Analysis The lower core plate is a perforated circular plate that supports and positions the fuel assemblies. The plate contains numerous holes for fluid flow through the plate. Flow is provided to each fuel assembly and the baffle-barrel region flow exits through the holes. The plate is bolted at the periphery to a ring welded to the inside diameter of the core barrel. The center span of the plate is supported by            ,
the lower support columns which are attached at the lower end to the lower                  I support plate.                                                                              l Temperature differences between components of the lower support assembly induce thermal stresses in the lower core plate. In addition, due to the lower core plate's proximity to the core and thermal expansion of fuel rods at power, the heat generation rates in the lower core plate due to gamma heating result in a                   '
significant temperature increase in this component. Thermal expansion of the tower core plate is restricted by the lower support columns, lower support plate and core barrel. These restraining items are exposed to the inlet temperature and have heat generation rates much lower than those found in the lower core plate.
Structural evaluations were performed to demonstrate that the structural integrity of the lower core plate was not adversely affected by the revised design conditions. These evaluations determine that the fatigue usage remains less than 1.0 and the plate is structurally adequate at the revised design conditions.
Upper Core Plate Structural Analysis The upper core plate positions the upper ends of the fuel assemolies and the lower ends of the control rod guide tubes, thus serving as the transitioning member for the control rods in entry and retraction from the fuel assemblies. It also controls coolant flow exiting from the fuel assemblies and serves as a boundary between the core and the upper plenum. The upper core plate is 4

l l
  'sttachment 2 to TXX-98265 '
j Pcge 22 of 32 restrained from vertical movement by the upper support columns which are attached to the upper support plate assembly. Lateral movement is restrained by four equally spaced core plate alignment pins.
The normal and upset stresses in the upper core plate are mainly due to hydraulic, seismic, and thermalloads. The total thermal stresses are due to secondary membrane stress and surface skin stress. Evaluations were performed to determine the impact that the revised design conditions had on the structural integrity of the upper core plate. As a result of the evaluation, it was concluded that the fatigue usage remains less than 1.0 and the plate is structurally adequate for the revised design conditions.
CONTROL ROD DRIVE MECHANISMS (CRDMS)
The pressure boundary portions of the CRDMs are exposed to the vessel / core inlet fluid. The conditions in Table IV-1 Indicate that the maximum increase in vessel / core inlet temperature was from 559.6 degrees F to 562.7 degrees F. An analysis was performed to determine the impact that the revised design conditions had on the fatigue usage of the CRDM components. The results indicate that the stress and fatigue usage are still within ASME Code limits.
REACTOR COOLANT LOOP PIPING AND SUPPORTS The revised design conditions were reviewed for impact on the existing design basis analysis for the reactor coolant loop piping, primary equipment nozzles, primary equipment supports and the pressurizer surge line piping. The temperature changes associated with the revised conditions could change in the loads in the components. Also, the changes in temperatures could impact the applicable stresses and fatigue usage values since these temperature serve as the initial conditions in some of the design transients.
The evaluation of the reactor coolant loop piping, the primary equipment nozzles, the primary equipment supports and the pressurizer surge line piping indicated that the loads remain bounded by the values in the existing analyses. Also, the stresses and fatigue usage values for the reactor coolant loop piping, the primary equipment nozzles and the pressurizer surge line piping remain below the allowable limit.
Leak-Before-Break (LBB) Analysis The current LBB evaluation was performed for the primary loops to provide technical justification for eliminating large primary loop pipe rupture as the structural design basis. In order to demonstrate the elimination of RCS primary loop pipe breaks, the following objectives were achieved:
Demonstrate that margin exists between the " critical" crack size and a postulated crack which yields a detectable leak rate.
I                      Demonstrate that th. a is sufficient margin between the leakage through a l                      postulated crack and the leak detection capability.

                   -.                  -~ ~ -_.-           - --                  --          --           -- . - .
Attachment 2 to TXX-98;'65 Page 23 of 32 Demon 1trate margin on applied load.
Demonstrate that fatigue crack growth is negligible.
1 An evaluation was performed which determined the impact of the revised design conditions on the LBB margins is negligible, and the LBB conclusions remain unchanged.
REACTOR COOLANT PUMPS (RCPs)
Structural Analysis The pressure boundary portions of the RCPs are exposed to the steam generator outlet fluid. Table IV-1 Indicate that the maximum change in steam generator outlet temperature was increased from 559.3 to 562.4 degrees F. An analysis was performed to determine the impact of the revised design conditions on the stresses and fatigue usage of the CRDM components. The results indicated that the stress and fatigue usage remain within ASME Code limits.
RCP Motor Analysis The RCP motors were evaluated for the I'miting case loads based on the revised design conditions for continuous operation at revised hot loop rating, operation at revised cold loop rating, starting, and loads on thrust bearings, it was determined that for operation at the revised design conditions, the RCPs continue to comply with their app" cable hot and cold loop operating ratings. Thus, the RCPs are able to accelerate at the resultant loads for the limiting case design conditions and the thrust bearings do not exceed their load ratings.                                            I STEAM GENERATORS The following sections describe the Unit 2 model D5 steam generator analyses and evaluations performed for the revised operating conditions.
Structural Integrity The bases for the existing structural and fatigue analyses of the steam generators are contained in the Model D5 Steam Generator Stress Reports. An evaluation was performed to demonstrate that continue to comply with the ASME limits is maintained for the revised design conditions. This evaluation considered the most critical components with regard to stress and fatigue usage. The primary inputs for the evaluation are the revised design conditions in Table IV-1 and the secondary side decign transients required to accommodate the changes in steam pressure, feedwater temperature and steam pressure. Table IV-1 Indicates that the reactor coolant side pressure remained unchanged at 2250 psia, while steam pressure could decrease from 1000 psia to a minimum value of 913 psia or increase to a maximum value of 1039 psia. Also, the steam temperature could decrease from a current design value of 544.6 degrees F to a value of 533.6 degrees F or increase to a maximum value of 594.2 degrees F and the feedwater temperature decreased from a design value of 440 degrees F to a design range of
,                     390 degrees F and 445 degrees F.
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Attachment 2 to TXX-98265 Page 24 of 32 The evaluation incorporated these inputs by developing scaHng factors necessary to calculate the increased stress and fatigue usage. The results indicate that all applicable stresses and fatigue usage values remain within the allowable limits.
Thus, the evaluation demonstrated that the critical components of the steam generators continue to comply with the requirements of the ASME Code at the revised design condition.
Thermal-Hydraulle Performance Secondary side steam generator performance characteristics such as circulation, moisture carryover, hydrodynamic stability heat flux and others are affected by increases in thermal power, and steam pressure. Steam pressure is, in turn, 4
utermined by the power as well as the primary side temperature, tube plugging level and feedwater temperature. The magnitude and importance of changes in the secondary side thermal hydraulic performance characteristics at the uprated power, with increased plugging, reduced primary side temperatures and a feedwater temperature range are assessed be' t Circulation Ratio / Bundle Liquid Flow The circulation ratio is a measure of tube bundle liquid flow in relation to steam flow and is primarily a function of steam flow. The bundle liquid flow minimizes the accumulation of contaminants on the tube sheet and in the bundle. There is a slight decrease in the bundle flow which also has a minimal affect on its function. Thus, the bundle flows are still adequate.
Hydrodynamic Stability - Damping Factor The hydrodynamic stability of a steam generator is characterized by the damping factor. A negative value of this parameter indicates a stable unit; i.e., small perturbations of steam pressure or circulation ratio will diminish rather than grow in amplitude causing instability in the control systems.
The Unit 2 damping factors remain highly negative, at a level comparable to
,                           the current design, for all cases. Thus, the steam generators remain hydro-dynamically stable for all uprated cases.
Heat Flux - Nucleate Bolling Limits The maximum heat flux transmitted from the primary to the secondary side is evaluated to determine that it remains within the applicable nucleate boiling limits. The peak heat flux increases with power and tube plugging. The increased total heat load is paesed through the same bundle heat transfer area, increasing the heat flux. For increased tube plugging (10%), the same heat load is passed through a smaller heat transfer area, also increasing the heat flux. It was determined that the peak heat flux increases slightly; however, the flux remains well within nucleate boiling limits and is comparable to values for a number of steam generators currently operating.
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Attachment 2 to TXX-98265 Page 25 of 32 Secondary Side Pressure Loss Unit 2 steam pressure losses are equal to or less than the current design pressure loss as a result of the reduced water level and circulation ratio at higher power levels offsetting the increase in steam flow. For the low feed temperature cases (390 degrees F), the lower steam flow causes the total pressure loss to be less than the design condition value. Thus, the revised pressure loss has a negligible affect on feedwater system operation.
Moisture Carryover Evaluations were performed to predict the amount of moisture carryover from the steam generators at the revised design conditions. Increases in steam flow, and reductions in steam temperature and pressure can increase the moisture carryover. The evaluations concluded that the moisture carryover is predicted to remain less than the current design value of 0.25% with plant operation at the revised design conditions.
PRESSURIZER The pressurizer limiting locations from a structural standpoint are the surge nozzle, the spray nozzle, and the upper shell at the point of spray impingement.
The pressurizer limiting operating condition occurs when the RCS pressure is high and the RCS hot leg temperature (Tsor) and cold leg temperature (T eoto) are low.
The pressurizer structural evaluation was performed by comparing the key inputs in the current pressurizer stress report with the revised design conditions in Table IV-1. The results indicated that the design conditions used in the current analysis remain bounding for the revised design conditions.
NSSS AUXILIARY EQUIPMENT The NSSS auxiliary equipment includes the heat exchangers, pumps, valves and tanks. These components were evaluated for the Teoto increase from 560 degrees F to 563 degrees F. The results indicated that the components continue to comply with the current design criteria since the fatigue usage values for each component remain below the allowable limit.
B.5    MASS ENERGY RELEASES AND LOCA HYDRAULIC FORCING FUNCTIONS MASS ENERGY RELEASES The LOCA (short and long term) and Steam Line Break (inside and outside) Mass and Energy Releases has been reviewed with respect to the 1% power uprating.
Results of these reviews have determined that a calorimetric uncertainty of 2%
was incorporated into the analysis. Providing that the NRC approves the exemption of the 2% calorimetric uncertainty to 1%, and operating conditions do not change, the LOCA and Steam Line Break Mass and Energy releases originally calculated remain bounding.

Attachment 2 to TXX-98265 Page 26 of 32 LOCA HYDRAULIC FORCING FUNCTIONS This section discusses the impact that the revised design conditions (Table IV-1) have on the LOCA hydraulic forcing functions. The purpose of a LOCA hydraulic forcing functions analysis is to generate the hydraulic forcing functions and hydraulic loads that occur on Reactor Coolant System (RCS) components as a result of a postulated loss-of-coolant accident (LOCA). These forcing functions and loads are considered in the structural design of the NSSS components. In general, LOCA hydraulic forces increase with an increase in RCS coolant density and, consequently, LOCA hydraulic forces increase for lower RCS temperatures.
The hydraulic forcing functions and loads that occur as a result of a postulated LOCA are calculated assuming a limiting break location and break area. The most limiting auxiliary line breaks are the 4" pressure spray line break on the cold leg and the 6" safety injection line break on the hot leg.
Description o' inalyses/ Evaluations Performed An evaluation was performed to demonstrate that the current LOCA hydraulic forcing functions remain bounding for the revised design conditions in Table IV-1.
Table IV-1 indicates that the maximum reduction in Teoto occurred from 559.6 degrees F to 555.3 degrees F. This temperature change was determined to increase the LOCA forces (due to the density increase) by less than 3% of their current values. The frequency of the LOCA forces pressure transient remained essentially unchanged and the amplitude of the LOCA decompression wave remained smaller when compared to the current values.
The 3% increase in forces was offset by a more accurate model of the loop at the break location. The more accurate model provides for up to a 30% decrease in the current forces. Thus, the current LOCA forces remain bounding for the revised t               design conditions.
C.      BALANCE OF PLANT A detailed evaluation was also completed which demonstrated that the Unit 2 existing non-NSSS systems, structures, and components and related programs continue to remain in compliance with all CPSE3 applicabla industry and regulatory requirements at a core thermal power of 3445 MWt.
The primary source of input for this determination was a detailed review of all the Design Basis Documents (DBDs) supplemented with expert opinions from personnel familiar with nuclear power plant design and CPSES, in particular, in order to identify and evaluate each system and program impacted by the power uprate.
The following systems were identified as being affected by the uprate: main steam, feedwater, steam generator blowdown, auxiliary feedwater, extraction steam, heater drains, condensate, condensate polishing, circulating water, turbine plant cooling, secondary sampling, spent fuel pool cooling, residual heat removal, component cooling, station service water, and combustible gas control systems.

I  to TXX-98265 Page 27 of 32 These affected systems were evaluated based on the revised NSSS parameters (revised RCS temperatures, and revised steam generator temperature and steam flow rate) and the 3445 MWt heat balance. The evaluation process addressed:
thermal-hydraulic performance, piping and support qualification, instrumentation and control functionality, equipment performance, and impact on the existing          !
consequences of a pipe break.                                                         I The programs, components, structures, and non-NSSS system issues identified as affected by the uprate were station blackout, equipment qualification, fire safe shutdown analysis, the concrete temperature at containment penetrations, fuel design, spent fuel storage, HVAC systems, radiological ansiyses, operating permits, the turbine / generator, isophase bus, main transformers, and switchyard.
All of these areas were reviewed for the revised thermal-hydraulics parameters        '
(temperature, pressure, flow rate), decay heat, and radioactivity resulting from the power uprate. It was determined that no hardware or operational . modifications are required.
Other than design documentation changes and severalinstrumentation and control setpoint-related changes, the only impact on plant operation is the reduction of approximately 1% in the maximum normal (tube sheet) steam generator blowdown mass flow rate (described previously herein). However, total blowdown flow can be maintained by increasing the shell blowdown flow by the equivalent amount.
This change will pose no limitation on the power the unit can generate and only marginally limits chemistry control flexibility. This reduction is necessary to preclude excessive erosion of the steam generator blowdown pipe.
D. RADIOLOGICAL CONSEQUENCES Calculations have been performed to determine potential impact on the radiological consequences from a 1% reactor power level uprating to 3,445 MWt.
The calculations and principal conclusions obtained are described below.
Description of Calculations Calculations address radiation source terms for four general categories of potential radiological consequences:
post-accident (LOCA) doses affecting equipment qualification (EO) of safety-related equipment, post-accident (LOCA) dose rates affecting vital area accessibility, post-accident dose rates in the control room er offsite following a postulated LOCA or a postulated fuel-handling accident, and offsite doses from normal-operating effluent releases.

                                   - . . - . - - _             . _       _ _ = - - . . - _ . _ _ - . .- -- .
Attachment 2 to TXX-98265 Page 28 of 32 in each case, assessment of potential impact, in comparison to current results, is dependent upon potential changes in fission product inventory at core eld-of-life (EOL) and typical application of the inventory data to dose rate or integrsted dose computations.
Fission Product inventory To assess the impact, additional core fission product inventory calculations were performed assuming full cycle operation at either 3445 MWt or 3565 MWt, with corresponding core average burnups of 32,332 and 33,202 MWD /MTU, respectively. In both cases, some noble gases (i.e., Kr-85, Xe-131m, and
,                 Xe-135), halogens (i.e., Br-82, and 1-130), and other nuclides (i.e., H-3, Sr-89,
!                 Y-91, Z-95, Nb-95, Z-97, AG-111, SB-125, Ba-140, CE-141, CE-143, Pr-143, Nd-147, Pm-147, and Pm-148) are somewhat higher than the current values.
l                 Post-accident Gamma Ray Source Strengths The calculated fission product inventories are appropriately converted to obtain energy-dependent gamma-ray source strengths at various times after shutdown j                 for comparison with current values. Using TID.-14844 release fractions, consideration is given to post-accident containment air (inside), reactor coolant (pressurized), sump water (depressurized), and containment air (outside). For the 1% uprating, some containment air source strengths (>1.0 Mev) are somewhat higher than current values in the post-accident time interval of 30 minutes to 24 hours. During this time interval, the containment air source strengths at lower energies, and reactor coolant and sump water source strengths, tend to be somewhat lower than current values. At later times, six months to a year, some lower energy source strengths exceed current values in all categories. Therefore, l
application to typical dose rate computations (discussed below) are necessary to determine the impact on radiological consequences.                                            I Post-accident Dose Rates and Doses from Contained Sources l
Using the calculated gamma-ray source strengths in typical sample applications indicates that post-accident dose rates from reactor coolant or sump water sources during the first few months are lower than current values. Therefore, existing postulated dose rates affecting vital area accessibility remain bounding.
However, dose rates from containment air in recirculating systems outside containment (assumed only for EQ purposes) tend to exceed current values during l                 the time interval of 30 minutes to 2 hours (by < 8% for 101% power).
Nonetheless, this difference in dose rate (for a short period of time) does not result in accumulated doses exceeding current limits.
The post-accident dose rates at later times (6 months to a year) are increased above current dose rates in all source-term categories. However, integration of dose-rates (from t = 0 to t = 1 year) for typical sample cases demonstrates that accumulated doses at all times out to one year remain lower than current values.
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Post-accident Doses from Released Sources investigation of potentialimpact on radiological consequences from Containment release (due to a LOCA) or Fuel Building release (due to a fuel handling accident )
determined that current calculations remain bounding. Most important fission product activities remain below current values for the 1% power uprating. As noted above, the LOCA containment source has increased quantities of Kr-85, Xe-131m, and Xe-135. The first two nuclideo make negligible contributions to postulated                 l control room and offsite whole body doses, and the increased contribution from Xe-135 (< 2 percent)is offset by lower cont" utions from other noble gases, primarily Xe-133 and Xe-135m. For the cu             the fuel handling accident, only Xe-131m and Xe-135m have increased activ              However, both of these nuclides make negligible contributions to postulated control room and offsite whole body doses for this accident. In all cases, the important radiciodine activities (I-131 through I-135) affecting thyroid inhalation doses remain below current values; therefore, the current remain bounding for the uprating.
Doses from Normal Effluent Releases                                                            I For the 1% uprating, calculations demonstrate that offsite doses from normal effluent releases remain significantly below referenced bounding results, which are within 10CFR50 Appendix Ilimits.
Conclusions Based on the above, the following is concluded:
Calculated fission product activities which increase as a result of uprating to 101%
of 3411 MWt do not contribute significantly to radiological consequences.
: 1.      Post-accident gamma-ray source strengths which increase (for some energies and post-shutdown times) due to a 101% uprating do not result in increased accumulated doses.
: 2.      Postulated post-accident vital area accessibility and EQ doses remain within existing current values for a uprating to 101%.
: 3.      Control Room or offsite postulated doses due to release from a LOCA or fuel handling accident are not impacted by an uprating to 101%.
: 4.      Offsite doses due to normal plant effluent releases remain below 10CFR50 Appendix I limits for an uprating to 101%.

Attachment 2 to TXX-98265 Pege 30 of 32 IV     SIGNIFICANT HAZARDS CONSIDERATIONS ANALYSIS TU Electric has evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on the three standards set forth in 10CFR50.92(c). The following information is provided to address the three questions required for the 10 CFR 50.92 evaluation.
: 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?
The comprehensive analytical efforts performed to support the proposed                 i change included a review of the NSSS systems and components that could                 i be affected by this change. All systems and components will function as                !
designed and the applicable performance requirements have been evaluated               !
and found to be acceptable.
The primary loop components (reactor vessel, reactor internals, CRDMs, loop piping and supports, reactor coolant pump, steam generator and pressurizer) continue to comply with their applicable structural limits and will continue to perform their intended design functions. Thus, there is no                 l increase in the probability of a structural failure of these components. The Leak Before Break analysis conclusions remain valid and thus the limiting break sizes determined in this analysis remain bounding.
All of the NSSS systems will still perform their intended design functions during normal and accident conditions. The pressurizer spray flow remains above its design value. Thus, the control system design analyses which credit the flow do not need to be modified for changes in this flow. The auxiliary systems and components continue to meet their applicable structurallimits and will continue to perform their intended design functions.
Thus, there is n.o increase in the probability of a structural failure of these components. All of the NSSS/ BOP interface systems will continue to perform their intended design functions. The steam generator safety valves will provide adequate relief capacity to maintain the steam generator pressures within design limits. The atmospheric relief valves and steam dump system valves meet design sizing requirements at the uprated power level. The current LOCA hydraulic forcing functions are still bounding for the proposed 1% increase in power.
Because the integrity of the plant will not be affected by operation at the uprated condition and it can be concluded that all structures, systems, and components required to mitigate a transient remain capable of fulfilling their intended function, the effects on the remainder of the safety analyses can be assessed. The reduction in the uncertainty allowance provided for the power j                       calorimetric measurement allows many current safety analyses to be used, without change, to support operation at a core power of 3445 MWt. As such, all FSAR Chapter 15 accident analyses continue to demonstrate compliance with the relevant event acceptance criteria. Those analyses performed to assess the effects of mass and energy releases remain valid. The source

1 Attachment 2 to TXX-98265 Pcge 31 of 32 terms used to assess radiological consequences have been reviewed and               i determined to bound operation at the 1% uprated condition.                          '
Based on the forgoing, it is concluded that the proposed changes do not result in an increase in the probability or consequences of any accident previously analyzed.
: 2.      Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?
No new accident scenarios, failure mechanisms or single failures are introduced as a result of the proposed changes. All systems, structures, and components previously required for the mitigation of a transient remain capable of fulfilling their intended design function. The proposed changes have no adverse effects on any safety-related system or component and do not challenge the performance or integrity of any safety related system.
Therefore, it is concluded that the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluation is not created.
: 3.      Do the proposed changes involve a significant reduction in a margin                 .
of safety?                                                                          l Operation at the uprated power condition does not involve a significant reduction in a margin of safety. Extensive analyses of the primary fission product barriers have concluded that all relevant design criteria remain satisfied, both from the standpoint of the integrity of the primary fission product barrier and from the standpoint of compliance with the regulated acceptance criteria. As appropriate, all evaluations have been performed using methods that have either been reviewed and approved by the NRC in Section 6.9.1.6b of the CPSES Technical Specifications, or that are in compliance with all regulatory review guidance and standards. Therefore, it is concluded that the proposed changes do not involve a significant reduction in a margin of safety.
V. ENVIRONMENTAL EVALUATION TU Electric has determined that the proposed amendment would change l                 requirements with respect to the installation or use of a facility component located within the restricted area, as defined in 10CFR20, or would change an inspection j                or surveillance requirement. TU Electric has evaluated the proposed changes and l                 has determined that the changes do not involve (i) a significant hazards i                consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in l
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l- _ Attachment 2 to TXX-98265 i    Page 32 of 32..
                  .10CFR51.22(c)(9). Therefore, pursuant to 10CFR51.22(b), an environmental assessment of proposed change is not required.
REFERENCES WCAP 10263, "A Review Plan for Uprating the Licensed Power of a Pressurized Water Reactor Power Plant," January, 1983.
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AFFECTED PAGES OF NPF 89 (UNIT 2 OPERATING LICENSE)
NPF-89: Page 3                                          ,
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Attachment 3 to TXX-98265 Page1of1 l
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!        (3)    TU Electric, pursuant to the Act and 10 CFR Part 70, to receive, possess and use
;               at any time, special nuclear material as reactor fuel, in accordance with the l               limitations for storage and amounts required for reactor operation, and described l
in the Final Safety Analysis Report, as supplemented and amended; l
(4) TU Electric, pursuant to the Act and 10 CFR Parts 30,40 and 70, to receive, possess, and use, at any time, any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor l               instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) TU Electric, pursuant to the Act and 10 CFR Parts 30,40 and 70, to receive, possess, and use in amounts as required, any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) TU Electric, pursuant to the Act and 10 CFR Parts 30,40 and 70, to possess, but i
not sepaiate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)    Maximum Power Level TU Electric is authorized to operate the facility at reactor core power levels not in excess of 3444 3445 megawatts thermalin accordance with the conditions specified herein.                                                                     :
(2)    Technical Soecifications and Environmental Protection Plan                            j
                                                                                                      \
The Technical Specifications contained in Appendix A and the Environmental            l Protection Plan contained in Appendix B, are hereby incorporated into this license.
TU Electric shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.                                 j (3)    Antitrust Conditions Applicants as defined in Appendix C shall comply with the antitrust conditions delineated in Appendix C to this license; Appendix C is hereby incorporated into this license.
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