ML20059B120
| ML20059B120 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 10/21/1993 |
| From: | Gody A Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9310280040 | |
| Download: ML20059B120 (26) | |
Text
_
r Docket Nos. 50-266 and October 21, 1993 i
50-301 4
LICENSEE:
Wisconsin Electric Power Company FACILITY:
Point Beach Nuclear Power Plant, Units 1 and 2
SUBJECT:
REACTOR COOLANT SYSTEM FLOW TECHNICAL SPECIFICATION CHANGE e
MEETING
SUMMARY
1 On October 14, 1993, the staff met with representatives of the licensee and Westinghouse Electric Corporation at One White Flint North, Rockville, t
Maryland. The purpose of the meeting was to discuss questions raised during t
the NRC review of the Point Beach license amendment request dated June II, 1993. lists the meeting participants and observers. documents questions raised during the meeting. is a copy of information submitted to the staff as part of the meeting.
Following introductions, and a brief summary of the purpose of the meeting, the licensee provided a summary of the proposed Technical Specification i
changes and why they are needed for Unit 2.
The licensee indicated that due to Unit 2 steam generator tube plugging, the reactor coolant system flow rate may not meet current Technical Specification limits upon.startup from the current refueling outage. The proposed Technical Specification change reduces the minimum reactor coolant system raw measured total flow rate, changes the overtemperature and overpower setpoints, and charees tbc Reactor Core Safety
~
Limits for Unit 2.
The licensee's safety evaluation addresses the following:
(1) Non-LOCA [ loss-of-coolant accident] Transients, (2) LOCA Transients, (3)
Steam Generator Tube Rupture, and (4) Systems and Components Integrity i
Evaluations. The licensee acknowledged that the primary focus of the meeting was to address NRC questions on the Point Beach System and Component Integrity Evaluations for te proposed operating temperatures of the plant.
The licensee and their contracted Westinghouse representative presented additional information in the revised system and component ' evaluation. The i
additional information concerned primarily a more focused evaluation of critical components that were the most limiting in terms of calculated fatigue usage and impact on plant shutdown capability over the design life of the pl ant.
Original Signed By:
$k Och oSobb266 Anthony T. Gody, Jr., Project Manager p
PDR g Project Directorate III-3 Division of Reactor Projects III/IV/V
Enclosures:
As stated g g g{g cc: See next page y
DISTRIBUTION g p dRoe c
Docket 0 File-OGC J. Gavula, RIII
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NRC & Local PDRs JZwolinski ACRS(10)
K. Jury, SRI PD3-3 Reading JHannon TMurley/FMiraglia t
Region III, DRP JPartlow AGody Plant File j
g'i; Name MR ok AGody, Jr JHannon W
Office LA/PD3-3 PM/PD3-3 PD/PD3 Date
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UNITED STATES
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j NUCLEAR REGULATORY COMMISSION g
WASHlNGTON, D.C. 20555-0001
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October 21, 1993 Docket Nos. 50-266 and 50-301 LICENSEE:
Wisconsin Electric Power Company Point Beach Nuclear Power Plant, Units I and 2 FACILITY:
t
SUBJECT:
REACTOR COOLANT SYSTEM FLOW TECHNICAL SPECIFICATION CHANGE-MEETING
SUMMARY
On October 14, 1993, the staff met with representatives of the licensee and Westinghouse Electric Corporation at One White Flint North, Rockville, Maryl and. The purpose of the meeting was to discuss questions raised during the NRC review of the Point Beach license amendment request dated June 11, 1993. lists the meeting participants and observers. documents questions raised during the meeting. is a copy of information submitted to the staff as part of the meeting.
Following introductions, and a brief summary of the purpose of the meeting, the licensee provided a summary of the proposed Technical Specification changes and why they are needed for Unit 2.
The licensee indicated that due to Unit 2 steam generator tube plugging, the reactor coolant system flow rate may not meet current Technical Specification limits upon startup from the current refueling outage.
The proposed Technical Specification change reduces the minimum reactor coolant system raw measured total flow rate, changes the overtemperature and overpower setpoints, and changes the Reactor Core Safety Limits for Unit 2.
The licensee's safety evaluation addresses the following:
(1) Non-LOCA [ loss-of-coolant accident] Transients, (2) LOCA Transients, (3)
Steam Generator Tube Rupture, and (4) Systems and Components Integrity Evaluations.
The licensee acknowledged that th_e primary focus of the meeting was to address NRC questions on the Point Beach System and Component Integrity Evaluations for the proposed operating temperatures of the plant.
The licensee and their contracted Westinghouse representative presented additional information in the revised system and component evaluation. The additional information concerned primarily a more focused evaluation of critical components that were the most limiting in terms of calculated fatigue usage and impact on plant shutdown capability over the design life of the pl ant.
M dy t.
Anthony T. Gody, Jr., Proj ct Manager Project Directorate III-3 Division of Reactor Projects III/IV/V
Enclosures:
As stated cc:
See next page
i
. Mr. Robert E. Link Point Beach Nuclear Plant.
Wisconsin Electric Power Company
-Unit Nos. I and 2
-ij cc:
i' Mr. Robert E. Link, Vice President Nuclear. Power Department j
Wisconsin Electric Power Company 231 West Michigan' Street, Room P379 Milwaukee, Wisconsin 53201 Ernest L. Blake, Jr.
Shaw, Pittman, Potts & Trowbridge j
a ing C 2b37 Mr. Gregory J. Maxfield, Manager Point Beach Nuclear Plant Wisconsin Electric Power Company 6610 Nuclear Road Two Rivers, Wisconsin 54241 Town Chairman Town of Two Creeks Route 3 Two Rivers, Wisconsin 54241 l
Chairman Public Service Commission
[
of Wisconsin Hills Farms State Office Building
-Madison, Wisconsin 53702 i
Regional Administrator, ' Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road i
Glen Ellyn, Illinois 60137 Resident Inspector's Office U.S. Nuclear Regulatory Commission j
6612 Nuclear Road Two Rivers, Wisconsin 54241 t
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h s-ENCLOSURE I-
-Reactor Coolant System Flow Rate Reduction Technical Specification Chance Reouest Meetino Participants and Observers Name Affiliation J. Hannon NRC/PD III-3 Anthony T. Gody, Jr.
NRC/PD III-3 Terence Chan NRC/EMEB Cheng-Ih Wu NRC/EMEB Harry Balukjian NRC/SRXB Brad Maurer Westinghouse Gary Krieser WEPCo Curtis Castell WEPCo Rick J. Khort WEPCo Tom Malinowski WEPCo 1
)
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e
a ENCLOSURE 2 Reactor Coolant System Flow Rate Reduction Technical Specification Chanae Reauest Meetina Reauest For Additional Information and Submittal Clarification questions:
(1)
Indicate in your submittal that the most limiting component / system with respect to component / system fatigue is the pressurizer spray nozzle as you stated in your presentation and what the "end-of-life" fatigue usage factor results are.
l (2)
With respect to component / system fatigue and stress, indicate in your conclusions that each system has been evaluated for all service conditions as applicable.
(3)
With respect to LOCA hydraulic forces analysis, at what temperature was the analysis performed for the reactor coolant piping?
(4)
Provide an indication of significance of this change in average coolant temperature with respect to the LOCA hydraulic force analysis of record and indicate if any components are significantly affected.
(5)
With respect to the information submitted as part of this meeting (Enclosure 3), address the significance of the difference in AI in the previously analyzed Series 44 steam generator and the Point $hach g
Series 44 steam generators.
i (6)
Is the 4 *F reduction in Reactor Coolant Pump / Motor operating temperature stated in the Westinghouse evaluation (Enclosure 3) correct?
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I ENCLOSURE 3 I
t Reactor Coolant System Flow Rate Reduction Technical Specification Chanac Reauest Meetina 8
Meetinc Aaenda Summary of Technical Specification Chance Reauast Reactor Coolant System Averaae Temperature Chance System and Component Evaluations f
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i Agenda: RCS Flow Tech Spec Change Meeting with NRC October 14,1993
Introductions
Purpose of Meeting (NRC-Gody, WEPCO-Krieser) i Summary of Tech Spec Change Request #160 OVEPCO-Castell)
Summary of Additional System and Components Information OVestinghouse)
NRC Questions i
Conclusion and Summary of Meeting l
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4 Summary of TSCR #160 The first flow measurement after Unit 2 startup last cycle was 181,873 gpm,73 gpm above the limit.
Plugging tubes during this outage for Unit 2 could cause the RCS flowrate to be below the 181,800 gpm measurement limit.
Analysis and Evaluations performed for reducing the RCS flow rate limit by 2600 gpm.
TSCR #160 submitted June 11, 1993.
TS 15.3.1.G:
Unit 2 RCS Flow Rate Limit reduced from 181,800 gpm to 179,200 gpm, about 1.4%
TS 15.2.1:
Unit 2 Reactor Core Safety Limits changed for reduction in flow.
TS 15.2.3:
T' term in overpower and overtemperature AT changed from s 573.9 F to s 570 F due to the change in Reactor Core Safety Limits.
TSCR #160 Safety Evaluation:
Non-LOCA Transients LOCA Steam Generator Tube Rupture Systems and components NRC Questions and WEPCO answers to be documented via a letter. Question of the Systems and Components evaluation and the average RCS operating temperature is subject of this meeting.
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RCS Average Temperature In 1992, during SG Replacement meetings, WEPCO and Westinghouse discovered a discrepancy between Temperatures used in System and Component analyses and actual operating temperature for PBNP.
Evaluation by Westinghouse provides an assessment that concludes that continued compliance with design standards will be maintained.
NRC has asked for additional information.
Conference Call: Westinghouse, WEPCO, and NRC on October 7,1993.
Westinghouse performs additional evaluations October 8-13, 1993.
i
d d
a To:
Allen Hansen & Tony Gody, Project hianagers for PBNP From:
Curt Castell, Senior Engineer Licensing
Subject:
ADDITIONAL SYSTEh! AND COMPONENT EVALUATION f
INFORMATION Date:
October 13,1993 Attached is a copy of the additional information provided by Westinghouse for the 570F average reactor coolant temperature evaluation for reactor coolant system and component evaluations for Point Beach. Tom Malnowski, Rick Kohrt and I from WEPCO along with Brad Maurer from Westinghouse will be available to meet with you and discuss this information at NRC headquarters at 1:00 pm on Thursday, October 14,1993.
The additional information is in the form of a revised Justification for Continued Operation (JCO) based on the original JCO contained in Westinghouse letter #MK92-0005-SAC from S.
A. Craft to R. K. Hanneman, Wisconsin Electric, dated October 21,1992. This JCO is applicable from October 21,1992 through October 1996 because it is currently planned that other evaluations will be available by October 1996 as part analyses and evaluations for the replacement of steam generators in Unit 2.
This JCO is in accordance with the PBNP FSAR, which states on Page 4.1-11:
To provide the necessary high degree of integrity for the equipment in the Reactor Coolant System, the transient conditions selected for equipment fatigue evaluation are based on a conservative estimate of the magnitude and frequency of the temperature and pressure transients resulting from normal operation, and normal and abnonnal load transients. To a large extent, the specific transient operating condition considered for equipment fatigue analyses are based upon engineering judgment and experience. Those transients are chosen which are representative of transients to be expected during plant operation and which are sufficiently severe or frequent to be of possible significance to component cyclic behavior.
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OCT 13 '93 22:03 FROM OFL LICENSIbG TD IJEP OFFICE PAGE.002/017 3
Revision 1 POIhT BEACII UNTIS 1 AND 2 OPERA' HON AT RCS AVERAGE TEMPERATURE OP 570'F ne purpose of this document is to address differences in the actual operaticg %. hue of the Point t
Beach Units and the temperatures at which component analyses and safety /acemient analyses were l performed. %is document provides the basis for operation of the Point Beach Units through October 1 1996.
l l Westinghouse performed a review of systems and components doc "-ike in the Fall of 1992.
l Wisconsin Electric Power Company (WEPCO) submitted a TehMr=1 Specification change request in June i of 1993 for Unit 2 to reduce reactor molant system raw measured total flow rate by 2@0 gpm. The 1 basis used to address the acceptability of the proposed Technical Specification change was the Justification l for continued Operation (JCO) completed in the Fail of 1992 (Westinghouse lecer MK92-0005 SAC dated 1 10!21/92). ne NRC had questions on the syste:n and component sections of the JCO and has requested l additional information. His revised JCO provides additional information in the revised component
', evaluation section.
LNTRODUCTION As a result of dM=tnns between Wisconsin Electric Power Company (WEPCO) and Westmghouse relative to steam generator rephwe, it has been determmed that differences exist between the actual operating conditions at Point Beach Units 1 and 2 and the conditions used in the eMem analyses and in analysis of systems and components. A review of historical correspondence by Westmghouse and discussions with WEPCO indicated that Point Beach Unit I went into operation with a Reactor Coolant Syste.m (RCS) average temperature (T,,) of 582.5'F and Unit 2 went imo operation with a T, of 573'F.
His temperamre was reduced in order to limit the steam generator outlet pressure to a value of 821 psia at full load to preserve structural integrity of the turbine. His steam pressure limit still remains applicable. WEPCO personnel imiic*wi that the Units are actually operating at a T.,, of 570'F based on a 1971 letterN from Westinghouse.
l Westingbouse performed a review of accident analysis and systems and components documentation. The i
review has determined the following:
Most of the current accident analyses of record have been performed at an RCS T, of 573.9'F.
ne component E-specs and stress analyses reflected various levels of not and Tcold for the RCS nomimi operstmg parameters. He range of parameters contamed in the documentation does not correspond to a T, of either 573.9'F or 570.0*F. The majority of the analyses reflect operation at higner average, temperatures.
Plam operat)on at lower than design temperamres yields increased hydraulic forces and could yieldI increased thermal stresses dudng operational transients on the primary loop, vessel, mternals, fuel, pressuruer, reactor coolantpumps and motors, and steam genera: ors as a result ofincreased subcooling, higher fluid densities, and larger transient t=vmahde changes. De component analyses must be reviewed to estimate the effect of lower temperature operation. In addition, utile lower ternperature cu m z w.=
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OCT 13 '93 22:03 FROM OPL LICENSING TO WEP OFFICE PAGE.003/017 a
f Revision 1 operazion is normally a beneSt for accident analysis, some MM-* may have worse results if a lower T, is utillred. Thus, an as-mmt of the accide:ns is also nemmy to address operanon at 570'F.
EVALUATION His evaluation is performed to assess the effect of operation of the Point Beach Unhs at a T, of 570*F.
This m-mmt provides justification for contimed operation of the Units.
ACCIDENT ANALYSIS LOCA-RELATED ANALYSES f
ESAR Oanter 14 3 2: Larre Break LOCA De cunent Large Break IJOCA (LBLOCA) analysis of record for the Point Beach Units was pufvaued using the NRC-approved ECOBRAfrRAC Two-Loop Upper PlenumInjection Best-Brim =te Large Breek LOCA Evaluation Modelta and resulted in a peak cladding %. hue (PCT) of 2028'F, for a double <nded cold leg guillotine (DECLG) break wi:h a discharge coefficient (Co) of 0.4 and based on a nominal RCS T, of 575'F. He sensitivity study reported in WCAP-10924R indicated that a lower core average temperature reduced blowdown PCT. The lower blowdown peak is expected to result in a lower reflood peak as well. Rus, the current analysis is conservative, and operation of the Point Beach Units at an RCS T, of 570.0*F is deemed acceptable with respect to LBLOCA.
FSAR Chanter 14.3.1: Small Break LOCA ne current Small Break LOCA (SBLOCA) analysis of record for the Point Beach Units was performed using the NRC-approved Small Break LOCA ECCS Evaluation Model with NOTRUhU" whidi resuhed in the most lizmting PCT of 809'F for a 4-inch breaF. nat analysis assumed an RCS T, of 570.0*F.
Therefore, conti::ued operation of Point Beach Unb is deemed acceptable for the small break LOCA analysis.
FSAR Chanter 14.3 7 Post-IDCA Lone Term Core Cooline Subcriticality Renuim.mmt ne Westmghouse licensing position for sa:isfying the requirernems of 10CFR Part 50 Sealon 50.46 Paragraph (b) Item (5) "Long Term Cooling" is defined in WCAP-8339M. He Westinghouse commitmem is that the reaaor will remnb shut down by borated ECCS water residing in the sump following a LOCAts. Since credit for the control rods is not taken for large break LOCA, the borated ECCS water provided by the accumulators and the RWST must have a concentranon that, when mixed with other sources of borated and non&> rated water, will result in the reactor core remaining subcntical assuming all control rods are out. A redaction in T to 570.0'F will have an insignificant effect on the RCS water mass and no effect on the Easuvs,y Core Cooling System water mass and boron concentrations assumed for this calculation. Herefore, the current calculation trmains valid for a reduction in T, to $70.0'F for the Point Beach Units and continued operation is deemed acceptable.
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DCT 13 '93 22:04 FROM OPL LICENSDG TO IJEP OFFI PAGE.004/017 e
Revision 1 FSAR Chapter 5.2.2: Boron Preceitation Drme Lone Term Cooline The post-LOCA Boron precipitation long-tum cort cooling requirement ensures no boren precipitation in the reactor vessel following boiling in the core. Since the Point Beach Units have simultaneous injection from the residual heat removal safety injection system into the upper ple:mm and the high head safety igjection system into the cold legs, this requirement is met by requiring alternate igiection after a LOCA Re time is for alternate igjection after a LOCA depends on powerlevel, and the RCS, RWST, accunnlator, and other wmer volmnes and boron comw* rations. A reduction in T, to 570.0'F will have an insignificant effect on the RCS water mass and will have no effect on the volumes and boron concentrations assumed for the RCS, RWST, accumulators, and other water sources. Herefore, there will be no effect on the post-LOCA altenate injection requirement of 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> for the Point Beach Unhs and continued operation is deemed acceptable.
FSAR Chapter 14 3 2: Post-LOCA Ianc-Term Cooline Minimum Flow Post-LOCA long-term core cooling mmimum flow is determined to ecsure adequate flow for large break and small break at the time of recirculation switchover. Reducing the T, to 570.0*F will have no effect i
on the inputs for this calculation. Therefore, this reduction in T, will have no effect on the post-LOCA long<erm core cooling mimmnm flow for the Point Beach Units and continued operstion is deemed acceptable.
Condusion of LOCA Assessment He effect of reducing the RCS T, to 570.0'F from 573.9'F on the LOCA-related anal m for Point 3
Beach Units has been evaluated by Westmghouse. He potmtial effea of this change on the FSAR analysis results for each of the LOCA-related accidents has been mmM, and shows in all cases that the efTecs of the change is not expected to result in exceedmg any of the following design or regulato
+
limits:
1.
The calculated peak fuel dement cladding temperanrre is below the requirements of 2200'F.
2.
The amoum of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.
3.
He cladding temperature transient is terminated at a time vehen the core geometry is still amenable to cooling. The localized cladding oxidation limit of 17 percent is not exceeded during or after quenching.
4.
The core remnim amenable to cooling during and after the break.
5.
The core temperature is reduced and decay beat is removed for an extended period of time, as required by the long-lived radioactivity remaning in the core.
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OCT 13 '93 22:05 FROM tPL LICENSil{1 TO WEP OFFICE PAGE.005/017 f
f b
-i Revision 1 f
1 NON-IACA TRANSIENTS I
frnnact to Non-IACA Licentia* R=de Operating at a reduced T.,, has a beneficial effect on DNBR. De lower T, results in additional margin to DNB and will not adversely affect the calculated #'
DNBR for any transicut. He increased fluid subcooling and mass flow rate increases the critical heat flux, thereby reducing DNBR.
4 He reduced T, will have no adverse effect on calent=*M peak pressures for any non-LOCA FSAR event. The calculated peak pressures are primarily el*minM by RCS heat load, primary to secondary j
heat removal capacity, and safety valve setpoints and capacity. These functions and processes are not adversely affected by a reduced T.
{
A criterion of *no wa:er relief through the pressurizer safety valves" is applied by Westinghouse to all I
ANS Condition 2 non-LOCA events. The reduced T,,, operation can adversely affect the peak pressurizer water level reached for certain Cordition2 heatup events. De events of concern with respect to peak pressurizer level are Ims of Normal Feedwater (FSAR 14.1.10) and Station Blackout (FSAR 14.1.11). These analyses already support a reduced T, of 570.0*F.
i The reduced T,,, operation will not adversely affect the calentatM peak fuel and cladding %.mres.
De lower fluid ternperatures will emharn hear transfer d=Taminics and will tend to result in lower fuel and cladding tempentures for all non-LOCA events.
NON-LOCA EVENT SPECIFIC DISCUSSIONS 7
PSAR 14.1.1 Uncon: rolled RCCA Withdrawal from a Suberitical Condition i
his event is analyzed for hot zero power (HZP) initial conditions. The average RCS %miare as well i
as all other initial conditions at HZP are not affeded by the reduced full power T.,,.
He results and conclusions of the FSAR for this event remain valid for reduced T, operation.
FSAR 14.1.7. Uncontrolled RCCA Withdrawal at Power -
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Ris event is analyzed primarily for DNB considerations, and as such, reduced T,,, operation is a benefit.
However, pressurizer filling is often predicted for small reactivity insertion rates. In accordance with standard Westinghouse methodology, the current FSAR analysis for this event does not credit the high !
pressurtzer level trip fanction. Normally, the CVCS letdown system would be expected to reduce RCS inventory during any heatup transient. In this case, the high pressurizer level trip setpoint would not be reached as soon as predicted by the LOFIRAN code, which cannot model CVCS letdown. Rus, takin i
credit for this trip would not be conservative. If the CVCS were not operable, pnssurizer level would increase until the trip sapoint were reached, at which time the transient would be termin=tM by the Reactor Protection Systan. Based on the above, pressurizer filling cannot occur during this event. The i
i results and conclusions of this section of the FSAR bound operation at a reduced T.,,,
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DCT 13 '93 22:05 FROM OPL LICENSING TD IJEP OFFICE PAGE.006/017 a
Revision 1 FSAR 14.1.3 Rod Cluster Contml Assembiv (RCCA) Dron
%!s event is limited primarily by DNBR. DNBR values are calculated by the Westinghouse Cvam dal Nuclear Fuels division using generic statepoints generated as part of the Westmgbouse Owner's Group program. DNBR is calm 1*d at the T, used to generate the gewic statepoints, thus actual T, never enters into the analysis. These statepoints are confirmed every cycle as part of the Reload Safety Evaluation process.
PSAR 14.1.4: Chemical and Volume Control System Malfunction
%e reduced T, can only affect the Boron Dilution at Power case. All other cases assume HZP or shutdown conditions. A reduced T, results in a greater RCS mass inventory due to the increased density. For a given CVCS dilution flowra:e, RCS dilution will occur more slowly, yielding more time beforeloss of shutdown margin. The results and conclusions of this section of the FSAR bound operation a reduced T,.
FSAR 14.1.5: Startup of an Inactive Reactor Coolant Iron The limiting criterion for this event is minimum DNBR. A reduced T, mereases the margin to DNB.
He results and conclusions of this section of the FSAR bound operation at a reduced T,.
FSAR 14.1.6: Reduction in Feedwarrr Enthalov Incident His event is analyzed at both full power and zero power conditions. Only the fuU power cases could t
be affected by the reduced T,. He limiting criterion for this event is mW;m DNBR. A reduced T, increases the margin to DNB. De results and conclusions of this section of the FSAR bound operation at a reduced T,.
ESAR 14.1.7: Excessive Imd increate Incident He limiting criterion for this event is mM m DNBR. A reduced T, increases the margin to DNB.
He results and conclusions of this sectaon of the FSAR bound operation at a reduced T,.
FSAR 14.1.8: Irss of Reactor Coolant Fim he Complete and Partial Loss of Flow events are primarily limited by minimum DNBR. A redac increases the margin to DNB. Le results and conclusions of this section of the FSAR bound operation at a reduced T,.
The Locked Rotor evcat is primarDy limnM by peak cladding +~ mme and peak RCS pressure. In addition, the percentage of rods in DNB is calmimM. A reduced T, increases the marge to DNB and enhances heat transfer characterzstics between the fuel cladding and the coolant. The reduced temi operation wH1 not adversely affect the peak pressure, peak cladding temperature, and the calculated percentage of rods in DNB for this event. The results and conclusions reached in this secbon of the FSAR remain valid for reduced T, operation.
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OCT 13 '93 22:06 FROM OPL L! NS!!G TO IJEP OFFIG PAGE.007/017 i
Revision 1 FSAR 14.1.9: Ioss of External Elearical Load he 11mzung critala for this evart are peak RCS pressure, peak main steam system pressure, and minmmm DNBR. A r~fncut T, increases the margin to DNB and thus the results and conclusions reached in the FSAR analysis with respect to minimum DNBR will not be adversely affected. Peak RCS and main steam system pressures are derarmined primarily by the pressurizer and main steam safety valve setpoints and capacitxs, and are insensidve to T,.
The results and conclusions of this section of the FSAR remain valid for reduced T-avg operation.
FSAR 14.1.10: Loss of Normal Fe%-
De Loss of Normal Feedwater analysis for both Point Beach Units currently supports a T, of 570.0'F.
FSAR 14.1.11: Irss of AC Power to the Strion Auriliaries he Loss of AC Power to the Station Auxiliaries analysis for both Point Beach Units currently supports a T,,, of 570.0'F.
FS AR 14.2.5 Rtmtare of a Steam Pine This event is analyzed at Hot Zao Power conditions and is not affected by the reduced full power T,.
FSAR 14.2.6 RCCA Eiection
%e limumg criteria for this event are fuel enthalpy, fuel melt, and peak cladding temperature. A reduced T, has a beneficial efTect in that it promotes better heat transfer between the fuel and core fInid.
Initial fuel and cladding temperatures are siso reduced. De results and conclusions of this secuon of the FSAR remam bounding for reduced T, operation.
C&aclusions of the Non-LOCA Evmts Assetement It can be concluded that cominned operanon of Point Beach at a nominal T.,, of 570.0'F is justified No current non-LOCA FSAR analyses are adversely impaNad by reduced T, operation.
+
STEAM GENERATOR TUBE RUPTURE The steam generator tube rupune (SGTR) analysis presented in the original Point Beach Units 1 and 2 FSAR has been updated to support a fuel upgrade with increased peaking factors. He resuhs of the updated SUIR analysis were presented in WCAP-11872 and the FSAR has been revised to reflect the updated SUTR analysis. The updated SGTR analysis was performed for operation with a reactor cDolan average temperature of 573.9'F and a steam generator tube plugging level of 25 %. The offsite radiation doses resulting from the updated SGTR analysis are:
Site Boundary - 0.55 rem thyroid and 0.117 rem whole body Low Population Zone - 0.065 rem thyroid and 0.0131 dole body e:zemrmss 6 of 16
OCT 13 '93 22:07 FROM OPL LICENS!tG TO UEP OFFICE PAGE.008/017 4-Revision 1 These results demonstrate that the radiological consequences for an SGTR would be signi&indyless than the permissiblelimbs of 10CFR100.
Ahhough an analysis has notbeen pufviuse to determine the impact on the SGTR analysis for operation at a reactor coolant average temperature of 570*F, it is expected that the difference in average temperature will only have a minor impact on the SGTR radiological consequences. Based on previous SOTR sensitivity studies, it is expected that the reduction in average temperature will result in a small increase in the primary to secondary break flow and a small decrease in the steam release to the attnesphere from the ruptured steam generamr. The net efect on the calculated offsite radiation doses is expected to be a slight decrease in the thyroid doses and a slight increase in the whole body doses.
Since the offsite radiation doses for the updated SGTR analysis in the FSAR are signi&mndy below the limits in 10CFR100, it is concluded that the SGTR radiological consequences will remain acceptable for operation at a 570*F T.,.
LOCA CONTAINMENT MASS AND ENERGY RELEASE I
Containment mtegrity analyses are peformed to demonstrate the acceptability of the co=d=#=
safeguards systems to mitigate the consequences of a hypothetical 1arge break LOCA. De current design basis mass and energy releases were performed at a higher temperature than the current Point Beach operatmg temperature. These results have been evaluated relative to the operation at an average RCS ternpera:ure of $70*F. It has been concluded that the current mass and energy releases, and therefore, the containment analysis remam boundmg.
Subcompartment analyses are performed to ensure that the walls of a subcompartment can maie=in their structural integrity during the short pressure pulse that accompanies a high energy line rupture within the subcompartment It has been concluded that the current releases remain bounding, whenever the pmalties associated whh the reduction in RCS temperature to 570*F and the benefits of RCS loop leak-before-break are considered. The current subcompartment analyses therefore also remain bounding.
COMPONENT ANALYSES An *emment of the design transients and component fatigue was performed for operation at an RCS average tap.uiure of 570*F. A review of components Equipment Speci6 cations and stress analyses was conducted as a result of the request for Steam Generator replacement at Unit 2. The review indicated that the components were evaluated for vanous levcis of T<old and T-hot that correspond to T, from 581*F to 587'F.
The following assessment provides the justi5 cation for continued operation (JOO) at the reduced RCS 1 temperamre through October 1996. Point Beach Unit I achieved commercial operation on 12/21/70 and Unit 2 a chieved commercial operation on 10/102.
EVALUATION The Point Beach Units have been operated at an RCS temperature lower than design basis temperamre.
This situation yields increased hydraulle forces and increased thermal stresses on the primary loop, ve sternals, fuel, pressurtzer, reactor coolant pumps, control rod drive marhems and steam generators e n traw
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a Revision 1 as a result of increased subcooling, higher fluid densities, and larger ternperature diffcreatials during postulated plant transients.
Fetirue and Usare Factors l Continned operation of the Point Beach Unks at their current operating tvaims is justified based
{ on the number of fatigue cycles they will have accumulated through October 1996. TheMesign transient specifications allow for a conservatively large number of plant operations and events. It is highly unlikely that the Point Beach Units will experience more than a fraction of the allocated fatigue cycles through 1996, and, regardless of the average temperature referenced in the Equly.s Specifications and stress calculations, h is equally unlikely that the pressure boundary or any critical component will fail to perform its miended fwdnn because of stress fatigue. For example,14,500 (once each day of 1 operation) loading and 14,500 unloading tramimdare allowed. He cooldown and heatup transients are l specified between 15% and 100% power at a rate of 5%/ minute The Point Beach Units were heavily involved in load follow maneuvers during the early 1970s. Signhady less load follow activity has occurred since that time. Extrapolating the load follow experience data indicates that neither of the Point Beach Units will have performed 1000 (less than 1/10 of the allowed) load follow cycles through 1996.
Furthermore, theirload follow maneuvers seldom exceed 35% power and their ramp rates average less than 0.5%/ min. As another example, it is doubtful if either Point Beach Units will come close to the allocated 400 reactor trips through 1996.
In 1986 and 1987, Wisconsin EIcctric initiated a Tranumt and Fatigue Cycle Monitoring Program for the Point Beach plants as part of a life extension study. Under this program, Westinghouse performed a detailed review of historical operating records for the plant, and used this information to:
Determine the historical operating transient set the plant had undergone from initial cold hydrotest through July,1986.
Establish which componems in the plant are most susceptible to fatigue over the plant operating period.
Perform a fatigue analysis on susceptible components to compare actual fatigue accumnistion for the historical transients with that predicted under the design basis.
He results of the Transient and Fatigue Cycle Monitoring Program showed that on an overall basis the actual transients were less severe, and were em*dng at a lower rate than was assumed in the original design analyses (roughly half the rate). WCAP-11501, " Transient Monitormg Program for Wisconsin Electric Power Company, Point Beach Units 1 aml 2, Phase 1 Final Report" deme < the results of this Program.
l Westinghouse has updated the Point Beach historical transient set as part of the pressurizer surge line l plant-specific applicability study.
This work is completed and thi results available show that the 1
4 corclusions of the 1986-1987 program are still valid and the rate of accumulation from the 1986-1992 l period was lower than for the period from initial hydro to 1986.
1 l
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Revision 1 Also, the operation of nuclear plants at --gamme up to 25'F lower than the origmal design temperamre has been addressed for a number of Units via implementation of T-hot reduedon programs.
In all instances, when the appropriate analyses were pafumai, continued compliance with all applicable l industry codes / standards and regulatory requirmv nts was confirmed.
j Mentification of Critical Comtr 133 l
l Fatigue due to thermal and pressure fluctuations during plant operation is one of the degradation l mech =i-affecting the life and flexibility of operation of nuclear power plants. Fatigue can ocent in l many locations throughout the plant and can be caused by many loading conditions. De components and l locations idectified below are the most limiting in terms of calculated fatigue usage and impact on plant l shutdown capability over the design life of Point Beach Unit 1 & 2. Both the NSSS components and l Auxiliary Systems components were included in this assessment.
l l With the exception of the Cass I piping, all components listed below have a calculated fatigue usage l factor of greater than 0.30 for forty years and some values range as high as 0.85. As Point Beach Unit l 1 & 2 are covered by ANSI B-31.1 requiremero, ne design basis fatigue evaluations are required for 1 Oass I piping for this plant. However, the Cass I equivalent piping is addressed in the jusdfication for
] continued operation. Also, Oass I valves were not included because of the extensive mamtenance
] performed on them and the less limiting role fatigue plays for these components, l
l Re critical locadens for evaluation ate:
l l Steam Generntor i
1 Tubes l Tubesheet Region (Top of the Tubesheet/ Tube 40-hbesheet Weld)
~
] Feedwater Nozzle 1
l Pressuriret i
l Spray Nozzle i Lower Head l Upper Shell i
1 l Reactor Vessel I
l Head Flange, Vessel Flange, Oosure Studs j CRDM Housings l Safety Injection Nozzle I
}
l Reactor Vessel Intemals l
l Upper ed lower core pla:es l Upper core alignment pins l 14wer core support columns t
i cu*m**=
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Revision 1 l l_ower radial restramts l Core barrel outlet nozzles l Baffle plates I
l Pmine l
l Pressurzzer Surge Line I
l Re effect of Tavg reduction on these aMWnal components are also addressed below: Control Rod j Drive mechanism, Puel, and the Reactor Coolant Pump.
I l System Commnent Evaluarlon Methodolorv I
l In general, the methodology used in the justification for condmwi operation of Point Beach Units 1 and i 2 establishes the crhical areas in which fatigue failure may be a macern. Operation of Point Beach at j the reduced Tavg temperature of 570'F is based on a review of-e i, eat stress reports for Point Beach.
c l Once the critical areas are IAandfied and the associmed fatigue usage factors are derarmined, the results l of not reduction analyses and stretch rating programs complettd at other plants are utilized where i appropriate to determine the effect of reduced Tavg operation of Point Beach on these factors.
l l Re primary criteria used to determine applicability of results of the analyses conducted to dare for use l in a comparative study with Point Beach are:
l 1.
The components are similar in design (materials and structural considerations) l 2.
Plant operating conditions are similnr.
l 3.
%e fat:gue evaluation methodology used is similar to that which has been used at Point Beach.
l l Based on a comparative study, the expected fadgue usage values are calculated for the critical areas and
[ compared to the 1.0 fatigue usage factor gaps limit.
l Fatigue is generally understood as the gradual deterioration of a material which is subjected to repeated l loads. Instead of a single cyclic stress, a component is subjected to a certain stress for al cycles, another i
stress for n, cycles and so forth. He ASME Boiler and Pressure Vessel Code for Nuclear Vessels, 1 Section III,1968 Edition Article 4 - Design, used in the design of Point Beach coiapchma, states that i the cumulative fatigue usage shall not exceal a value of 1.0. His means that the allowable number of t
I cycles always exceeds the design cycles. As stated in the Point Beach FSAR, this certainty assure l of the components against fatigue failure.
I l Additionally, h is important to note that a fatigue usage faaor of 1.0 does not necessarily mean that a l fatigue failure is imminmt. De design fatigue curves used in the code are based primarily on strain-l controlled fatigue tests of small polished speiment. The design stress values were obtamed from the best i fit curves by applying a factor of 2.0 on stress or a factor of 20 on cycles, whichever was more i
conservative at each point. Rese factors are intended to cover such effects such as environment, size l effect and scener of the data,
ra I
l Regarding any uncenainty in the fatigue usage values determmed using the comparative smdies for the I different components, as discossed above, to date Point Beach Units 1 and 2 have operated for ewe. men 10 of 16
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Revision 1 l significantly less than the 40 year designlife and, thus, has experiericed significantlyless than the number 1 of assumed transients. Herenre, sufficient margin exists to accomm* any small expected difference i in the plant design transier a a result of reduced Tavg operation.
I l The comparisons performed for the Point Beach coary+- are as follows:
Control Rod Drive Mechniem For Control Rod Drive Mechmieme (CRDMs), operation of the Point Beach Units at a lower RCS temperature than specified by design, is equivalent to a not Reduction Program.
He CRDMs of another Westinghouse plant were evaluated for a not Reductiog Program and found to be bounded by the existing pressure boundary analysis. ne vessel outlet temperature (Thot) was reduced from 610.9'F to 595.0'F.
The Point Beach Urdts use the same model CRDMs as the other Westinghouse plant, so they share the same generic pressure boundary analysis. Since Point Beach has reduced not from 602.8*F to about 1 598.9'F, then similarly, the Point Beach CRDMs are bonMM by the existmg analysis.
Pressurizer l An assessment was performed of the Point Beach Pressurizer Spray Nozzle and Upper Head to Shell l Junction for operation at average RCS temperantre of 570 degree F.
A review of the Series 84 l Pressurizer Stress Report (Section 3.2, Spray Nozzle Analysis and Section 3.6, Upper Head to Shell l Junction Analysis) showed that for the transients during which sprayr are spedfied, the thermal analysis l conservatively assumed a minimum differential t% aeue, delta T, of 125 degrees F between the spray I nozzle and incoming Teold water. For the proposed Tavg of 570 degrees F, Teold is 540.6 degrees F l resulting in a spray nozzle to Teold differential of delta T = 653.0 - 540.6 = 112.4 degrees F. This l ddta T for the proposed operation is smaller than one used in the stress analyses. The stress and fatigue l analyses for the Series 84 pressurizer envelope the Tavg of 570 degree operation. Herefore, there l should not be any change in the fatigue usage of the spray nozzle and uprer head to shdl junction.
I l Based on the above assessment, it is concluded that for Tavg of 570 degrees F operation, the spray nozzle i
and the upper head to shelljunction meet the ASME Code,Section III stress analysis and fatigue analysis
! requirements.
End The grid load assessment for the 14X14 optimnM fod assembly design in Point Beach I and 2 was evaluated. There will be an had increase of 4% in peak LOCA bydraulic forces due to a decrease in average temperanne of 3.95'F. He increase in hydraulic forces could result in the grid impad force increase ne IDCA hydraulic forces have no direct impact on seismic loads. Based on a conservative 10% grid load increase factor, the combined grid loads resulting from a seismic event and a pipe break is below the grid strength. Thus, it is Westmghouse's judgement that the fud assembly structural integrity will not be affected emomm 11 of 16
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e.
I Revision 1 Steam Ceerators l The fatigue evaluation of the Point Beach steam generator components at reduced Tavg tuuw.uae is l based on a similar one perfonned for other steam gerwMois of the same series (Series 44) in another
! Westinghouse plant. The evaluation performed for that plant earvelopes that for Point Beach. The steam l parameters applicable for these two plants are:
I l
Previously Anahzed Point Beach l
Series 44 Unit Series 44 Unit I
l T Average 547.7 degree F 70 degree F I
/
l P Steam 768 psia 785 psia i
l P Prunary 2250 psia 2250 psia 6
1 l
Differential Pressure 1482 psia 1465 psia j
(enveloping)
I l The parameters for the previously analyzed plant are such that the pressure differential between the l primary and the secondary side is grea:er than that for Point Beach I
l Based on the analyses of Point Beach steam generator components, it is Westinghouse'sjudgement that l
l their continued operation isjustified tmder reduced Tavg conditions through October 1996. In the fatigue l analysis the reduced Tavg steam parameters were considered, while it is assumed that the other transient l conditions remained unchanged. Both the Code stress and fatigue linuts are acceptable for contmud I operation through October 1996. All stresses are within the Code allowable limits. None of the values l exceed the mnimnm Code allowable limit of 1.0.
Reactor hsel Internals t
l Fatigue of the Point Beach reactor internals due to a lower operating temperature of 570'F as compared l to E-Spec operatmg temperamres ranging from 581*F to 587'F was invesugated.
l l The conclusions reached for the Point Beach reactor vessel internals concermng fatigue were based on i various T-hot reduction fatigue evaluations performed on mimamus Westinghouse plants. For example, i the evaluations for a typical plant with an appronmea 20 degree F T-hot reduction was analyzed for the j various transients in both the cold and hot legs. The effect on the reactor internals components fatigue l strength was evaluated. For this evahtstion the core average temperature of 591,8 degree F was reduced
} to 572.2 degree F, thereby yleIding a T-hot redtsction program of 19.6 degree F for the core average l temperature.
I
[ The results of these analyses show that the hot leg Duid trmimt when compared to the original condition I transients showed no significant increase in delta T. The delta T showed some improvements for some l of the transients while others Many restw unchanged. However, the cold leg transients compared ct e x w m 12 of 16
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t Revision 1 l to the original condition transients show some increase in ddta T for the large step load decrease and the l loss of power transient mnditions.
l l The corresponding increase in stresses for these cases were evaluated for the critical reactor vessel l internal components which consisted of upper and lower core plates, upper core plate alignment pins, l lower support columns, upper support columns, lower radial restraints, core barrel outlet nozzles, and l the baffle plates. "Ibe critical components were chosen based on the criteria of having the marimum l accumulated fatigue usage factor (which was less than or equal to 0.39 for the core barrel outlet nozzle).
l The stress results showed a small increase in stresses (less than 7%) and the corresponding impact on the l fatigue usage factors was insignW"* (less than 1 %) and the accumulated fangue usage factor remained l less than unity.
I l These comparisons of results do not consider a similar two loop plact, but the design, methodology and l the materials used in these plams are similar to Point Beach and meet the ialustry codes and standards I even though some of these standards were non existent at the time when Point Beach was originally l designed.
I i
l Based on these comparisons to other similar WeAPm plants, it is our conclusion that the reduced l Tavg conditions at the Poim Beach plant would not have adverse effect on the reactor internal components
) of the Point Beach plant.
LOCA Hydraulie Forces Analysis Two analyses are currently don =~*ad for Point Beach. The first analysis dommentui the LOCA hydranlic forces analysis of the upflow conversion program at these units, while the second analysis was performed in support of the reload of Opt =irad Fuel Assemblies (OFA). These analyses were based upon full Double-Ended Guillotine (DEG) breaks a the biological shield for the first analysis on upflow conversion and the accumulator line break for the qualification of the OFA fuel in the second analysis.
Each of these analyses assumed a reactor coolant average tempmhue of 573.9*F consistent with full power operation and a corresponding cold leg hmg Ahue of $44.8'F. Point Beach operation at a full power average temperature of 570.0'F corresponds to a cold leg temperature on the order of 540.0*F.
As such, justification for continued operation with a cold leg difference of appronmatdy 5'F from that of the analyses is required. From known sensitivities,'this small temperature difference is expected to result in an increase in the magnitude of the peak LOCA hydraulic forcing functions of 4 %. Based upon l previous work and engineering judgemme, the continued operation of the Point Beach units with a deviation of 3.9'F in RCS T, is acceptable from the perspective of the 1DCA hydraulic forces analyses.
l Thisjudgement is applicable to the effects of the LOCA hydraulic forces on the reactor coolant loop as well.
Reactor Vem!
The Point Beach Units 1 and 2 reactor vessel stress reports were authored by Babcock and Wilcox Co.
(B&W) and Combustion aginming, Inc. (CE), respectively. The Point Beach Unit I reactor vessel stress report reflects the nonnal operating temures and design transiems from the Westinghouse reactor vessel equipment specification including a normal operating inlet t=g ure (Toold) of 559.5*F and a normal operating outlet temperamre (Ibot) of 614.5'F. With these operating temperatures and the ewsm m m 13 of 16 l
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I i
1 Revision I l
l l
corresponding design tr*mimtt, the mnimmn cumuladve fangue usage factor for 40 years of operadon i
is reported to be.790 in the safety injection nozzies. h reactor vessel closure snid usage factor is reported to be only 0.23. m Point Beach Unit 2 reactor vessel stress report do-analy4s considering Tcold at 552.5"F and bt at 610.0'F. Therefore, the Unit 2 reactor vessel was to be analyzed for a Tavg of 581'F. h mnimmn cumuladve usage factor reported is 0.715 at the external support bracket He mnim-Unit 2 closure stud usage faaor is reported to be 0.484.
Based on this data from the reactor vessel stress report review, there is at least 0.21 margin to the 1.0 fatigue usage factor limit at every location in the reetnr vessels. Review of dome-terion on Thot
] reduction evaluations for 4-loop reactor vessels (reduction of appronmntdy 20 degrees F) indicates that incremental increases in the usage factors due to the design transient changes are relatively small (less
[ than 0.10).
I l Therefore, no reactor vessel fatigue concern exists based on obvious design margins. When the l conserva: Isms in the design transient t-g.hae variations and number of occurrences are considered, j the margms of safety are even larger.
Two other reactor vessels on Westmghouse plants have 132 inch I.D. 2-loop vessels similar in design to the Point Beach reactor vessels. However, they were designed for a Toold of 532.2'F and a Thot of 596*F. Thus, the design Tavg is 564*F. Furthermore, the design transiem specifications are the same as those for Point Beach Unit 2. With this smaller Tavg, the other reaaor vessels not only exhibit cumulative fatigue usage factors less than 1.0, but have been shown to possess design margin for a 20 yearlife extension.
g l Based on the above, it is Wesdnghouse's engineeringjudgement that the Point Beach reactor vessels will l operate safely with reduced RCS i.c.wgratres.
I Reactor Coolant I.oon Pinine aid Suonorts j The Poim Beach piping (both the primary loop and the Class 1 auxiliary lines) was originally designed 1 to the ASMEJANSI B31.1 Power Piping Code. The reduced operating teap imes in the hot and cold
[ legs will generate lower thermal loadings than those in the exisung design basis. Because of the vint j of these plants, a fatigue analysis of the loop and Class 1 piping was not required in the original desig i
basis. W B31.1 Code does not have a formal fatigue evaluation methodology.
1 Continued operation of the Point Beach Units at their current operanag temperatures is justified b on work performed for other plants that implemented Thot reduction programs. The hot and cold leg temperature differences applicable for Point Beach are less significant than typical temperature differences i successfully evaluated and qualified for other plants. It is, therefore, Westinghouse's conclusion that the loop piping and supports are not si.mihtly affected by the lower operatmg temperatures at Point Beach Surre Line/ Pressurizer Stratifiemlon
] When the thermal stratification issue surfaced in 1988 for the pr+% surge line, the method used to l address this new loading was primarily a fatigue evalunion. For this reason, the pressurizer l is the only piping system m Point Beach with a fatigue analysis.
- Ibis evaluation was documente atterme 14 of 16
DCT 13 '93 22:12 FROM OPL LICENS!tG TO IJEP OFFICE PAGE.016/017 1,
Revision 1 l specifically for Point Beach Units 1 and 2 in WCAP-13509 (" Structural Evaluation of Point Beach Units l 1 and 2 Pressuruer Surge Lines Considering the Effects of Thertnal Stratification")in October,1992.
l 1bose results and toethodology were reviewed and cover the pcennal impact of the reduced Tavg 1 operatim.
Reactor Coolant Pumn/MotDI A 4*F reduction in pump operating imaiuie (steam generator outlet im mure) has negligible affect on the analysis of the Reactor Coolant Pump pressure boundary components. hefore, the pump structural integrity is not impacted by Point Beach operation at the subject reduced Tavg.
The motor rating is 6000 HP. The original hot loop pump load was estimated to be 4926 HP. Based on previous terating analysis, it is not expected that a decrease of 3.9'F in hot loop reactor coolant twaue with no accorganymg flow change would consume this operating margm. Based on this I
judgement, continued ope a: ion of the motors at the reduced RCS coolant temperature is acceptable and pump coastdown is not irpe+ed CONCLUSION OF COi.CONENT EVALUA* DON i It is Westinghouse's engineeringjudgement that the Point Beach NSSS hardware remaim in compliance l with all applicable requirements through October 1996. It is shown based on comparative studies for critical areas identified in the NSSS cogra that fatigue usage factors remMn below 1.0. Regarding any uncertainty in the fatigue usage values determined using the comparative smdies, it is judged that both
! Point Beach Units I and 2 have operated for significantly less than the 40 year design life with rates of l transiems less than originally assumed, and thus, have experienced significantlyless than the number of l design transients. 'Ibus, sutYicient margin exists to umw=da:e any small difference in plant design j
I transients as a result of the reduced Tavg operation not curremly addressed in this analysis.
Itis d
Westinghouse's engmeering judgement that the Point Beech Units may safely operate until through October 1996.
CONCLUSION An assessment of the impact of operanon of a T, of 570*F on the Point Beach Units 1 and 2 design has been completed. The assessment considered accident analysis and the effects of these operadng condition on the mechanical design of componems. It was concluded that continued operation of Point Beach Units i
1 and 2 through October 1996 is justzfied.
l 1
REFERENCES 1.
Letter PBW-WMP-2333 D. F. Rinald, Westmghouse, to G. A. Reed, WEPCo,1/21n1.
i i
2.
Dederer, S. I., Hochreiter, L. E., Schwarz, W. R., Stucker, D. L., Tsai, C. K., and Young, M.
Y., Westinchouse Larre-Break LOCA Best-Estimate Methodolorv. Volume 2:
Atelication to Two-loon PWRs Eouipped with Urmer Plenum Iniection, WCAP-10924-P-A, Vol. 2, Rev. 2, and 1
Addenda, December 1988.
i an*mn=
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l p
Revision 1 3.
WCAP-10054-P-A (Proprietary), WCAP-10081 (Non-rivpihy), I.ce, H., et al., " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code", August 1985.
4.
WCAP-8339 (Non-Proprietary), Bordelon, F. M., et. al., "Westingbouse ECCS Evaluation Model
- Snmmary", June 1974.
5.
Westinghouse Tecimical Bu!!ctia NSID-TB-8648, "Pou-IDCA Long Term Cooling: Boron Requirements", October 31,1986.
6.
NRC letter to Wisconsin Electric Power Company, Nos. 50-266/301, December 24,1975.
7.
Criteria of the ASME Boiler and Pressure vessel Code for the Design by Analysis in Sections III and VIII, Division 2, The American Society of Mechanical Engineers,1969 t
1 h
catraum i
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- TOTAL PAGE.017 **