ML20116G454
| ML20116G454 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 08/05/1996 |
| From: | Link B WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML19311C175 | List: |
| References | |
| VPNPD-96-051, VPNPD-96-51, NUDOCS 9608080108 | |
| Download: ML20116G454 (20) | |
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Wisconsin l Electnc POWER COMPANY 231 W Michigan. PO Box 2046. Mdwoukee. WI 53201-2046 (414)221-2345 VPNPD-96-051 10 CFR 50.4 10 CFR 50.90 August 5,1996 Document Control Desk US NUCLEAR REGULATORY COMMISSION Mail Station PI-137 Washington, DC 20555 Gentlemen:
DOCKETS 50-266 AND 50-301 SUPPLEMENT TO TECHNICAL SPECIFICATIONS CliANGE REOUESTS 188 AND 189 POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 By letters dated June 4,1996, Wisconsin Electric Power Company, Licensee for the Point Beach Nuclear Plant, submitted Technical Specifications Change Requests 188 and 189. These requests proposed amendments to the Point Beach Technical Specifications identified by analyses performed in support of Unit 2 operations following replacement of steam generators this fall. The revisions proposed in Change Request 188 modify Technical Specification Section 15.2.3, " Limiting Safety System Settings and Protective Instrumentation," and Section
.15.5.3, " Design Features - Reactor." The revisions proposed in Change Request 189 modify Sections 15.2.1,
" Safety Limit, Reactor Core," 15.2.3, " Limiting Safety System Settings, Protective Instrumentation," and 15.3.1.G,
" Operational Limitations."
We are providing additional information in support of the safety evaluations provided with these change requests.
The additional information is presented in a safety evaluation entitled " Supplemental Safety Evaluation for Technical Specification Change Requests 188 and 189," attached to this letter.
We are also including additional revisions to the proposed Technical Specifications submitted in our Change Requests 188 and 189, identified during review of evaluations performed in support of the steam generator replacement project. The proposed revisions include changing the Ovenemperature delta T K1 term from the originally proposed value of 1.26 for 2250 psia operation to a value of 1.19. The revisions also include new low-low steam generator water level setting limits for reactor trip and auxiliary feedwater initiation necessary because of the difference in lower level tap location between the existing and replacement steam generators. Edited Technical Specifications pages are attached that show these revisions. The attached supplemental safety evaluation j provides a description of these changes and the basis andjustification for these revisions. These pages, along with those included with our earlier requests, identify the changes evaluated in th supplemental safety evaluation With the exception of the proposed Technical Specification setting for Steam Generator lo-lo level, we have included for your information uncertainty analyses which demonstrate appropriate margins between the safety analysis acceptance criteria and Technical Specification limits are maintained. The uncertainty calculations related 9608080108 960805 PDR ADOCK 05000266 P PDR M 015 C\sf og/,.h %p.4 \
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Document Control Desk August 5,1996 Page 2 to Steam Generator 1o-10 level are being finalized and will be forwarded when complete. The uncertainty ana!ysis related to the Overtemperature and Overpower-delta-T setpoints contains information which is proprietary to the Westinghouse Electric Corporation. Accordingly, we request that this information be withheld from public disclosure. We will comply with the requirements of 10 CFR 2.790 to provide a non-proprietary version of this material together with an aflidavit as soon as the non-proprietary version has been prepared. We will submit the required number of copies of the non-proprietary version of the information and the required affidavit at that time.
Also attached la support of your reviews is the Mechanical Components and System Evaluation which evaluates the effect of the proposed changes in unit operating conditions on plant systems and components. Mark-ups of the safety analyses contained in the Paint Beach Nuclear Plant Final Safety Analysis Report to reflect the analyses and evaluations performed are also included.
We have determined that the proposed amendments do not involve a significant hazards consideration, authorize a significant change in the types or total amounts of any effluent release, or result in any significant increase in individual or cumulative occupational exposure. Therefore, we conclude that the proposed amendments meet the requirements of 10 CFR 51.22(c)(9) and that an environmental impact statement or negative declaration and f environmental impact appraisal need not be prepared. The original "No Significant Hazards" determinations for operation under the proposed Technical Specifications remain applicable.
Our replacenv at outage is presently scheduled to commence on October 5,1996. We request approval of these changes by October 31,1996, with implementation to occur prior to startup of Unit 2 from the outage. This will ensure the amendments are implemented and appropriate training is performed prior to operation.
1 If you require additional infortnation, please contact us. l l
Sincerely,
[/cht. /b Bob Link b Vice President Nuclear Power TGM cc: NRC Resident inspector NRC Regional Administrator PSCW Subscribed and sworn to before me this 9 day of AM,1996.
h ota Pglic, State of Wisconsin My commission expires 10f.27/%
i I
SUPPLEMENTAL SAFETY EVALUATION FOR TECHNICAL SPECIFICATIONS CHANGE REOUESTS 188 AND 189 INTRODUCTION Wisconsin Electric Power Company (Licensee) has applied for amendments to Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant Units 1 and 2. The proposed revisions will modify Technical Specification Sections 15.2.1, " Safety Limit, Reactor Core," ;
15.2.3, " Limiting Safety System Settings, Protective Instrumentation," 15.3.1.G, " Operational Limitations," and Section 15.5.3, " Design Features - Reactor."
Additionally, Wisconsin Electric is proposing revision to Technical Specification Section 15.3.5,
" Instrumentation System," to modify the steam generator low-low level auxiliary feedwater initiation setting limit.
EVALUATION We are planning to replace the steam generators in PBNP Unit 2 during the Fall,1996 refueling outage. Westinghouse has performed the safety analyses and evaluations applicable to both units, in support of operation of Unit 2 with the replacement steam generators. These analyses and evaluations include; large and small break LOCA, LOCA forces, non-LOCA events, LOCA long and short term mass and energy release, stcam generator tube rupture, and systems and ,
components. The safety analyses and evaluations examined the operating conditions for Unit 1, and operation with the new steam generators for Unit 2, including an average coolant temperature ;
(T y) range between 557 F and 573.9 F, an operating pressure of either 2000 or 2250 psia, a !
reduced thermal design flow, a slightly larger primary volume for Unit 2, and a slightly smaller I secondary volume for Unit 2.
Evaluations of each of the proposed Technical Specification changes is provided as follows:
Iligh Pressurizer Pressure Reactor Trip The high pressurizer pressure reactor trip, TS 15.2.3.1.B(2), is being changed such that the present setting limit ofs2385 psig will be applicable for 2250 psia primary system pressure operation. A new limit of s2210 psig for 2000 psia primary system pressure operation is also i being proposed.
The high pressurizer pressure reactor trip protection function is provided to trip the reactor to prevent overpressurization of the reactor coolant system. The high pressurizer pressure reactor l trip is one of the reactor protection functions utilized in the analysis of the loss ofload accident as described in FSAR 14.1.9. Two cases analyzed for the loss ofload accident, cases c. and d.
(see attached FSAR Mark-up 14.1.9), use the analysis trip point of 2425 psia (2410 psig) for l
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2250 psia operation. The resultant maximum RCS pressure is 2740 psia, which is below the analysis limit of 2748.5 psia.
The analysis trip points for high pressurizer pressure are shown in the attached FSAR mark-up, Table 14-3. The Technical Specification setting limit of s2385 psig for 2250 psia operation provides adequate allowance for the analysis trip point of 2410 psig (2425 psia). The Technical Specification setting limit of s2210 psig for 2000 psia operation provides adequate allowance for the analysis trip point of 2235 psig (2250 psia). The minimum difference between the Technical Specification setting limit and the analysis trip point is 25 psig. The analysis of pressurizer pressure instrument inaccuracy is provided as an attachment to this safety evaluation. This analysis shows that these Technical Specification setting limits provide adequate difference to account for possible inaccuracies in the protection function. These analysis trip limits are also used in the determination and evaluation ofinstrument ranges for the Overtemperature and Overpower delta T trip functions.
The setting limit of s2385 psig for 2250 psia primary system pressure operation is not being changed, it is only being made specifically applicable to 2250 psia operation. The new setting limit of s2210 psig for 2000 psia primary system pressure operation is substantially lower and more conservative with respect to preventing overpressurization than the presently allowable value of s2385 psig.
Low Pressurizer Pressure Reactor Trip The low pressurizer pressure reactor trip, TS 15.2.3.1.B(3), is being changed such that the setting limit of 21905 psig will be applicable for 2250 psia primary system pressure operaticn. A limit of 21800 psig for 2000 psia primary system pressure operation is also proposed.
The low pressurizer pressure reactor trip provides core protection from departure from nucleate boiling caused by transients that reduce primary system pressure. Additionally, this protection function can actuate reactor protection in an anticipatory manner for accidents in which the primary system pressure is expected to also actuate the low pressurizer pressure safety injection protection function. These accidents include small break LOCA, steam generator tube rupture, and steam line break. 1 l
The analysis trip points for low pressurizer pressure are shown in the attached FSAR mark-up, ;
Table 14-3. The Technical Specification setting limit of 21905 psig for 2250 psia operation provides adequate allowance for the analysis trip point of 1865 psig (1880 psia). The Technical Specification setting limit of 21800 psig for 2000 psia operation provides adequate allowance for the analysis trip point of 1760 psig (1775 psia). The minimum difference between the Technical Specification setting limit and the analysis trip point is 40 psig. The analysis of pressurizer pressure instrument inaccuracy is provided as an attachment to this safety evaluation. This analysis shows that these Technical Specification setting limits provide adequate difference to l
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1 account for possible inaccuracies in the protection function. These analysis trip limits are also ,
used in the determination and evaluation ofinstrument ranges for the Overtemperature and l Overpower delta T trip functions. l l
The new setting limits of 21905 psig for 2250 psia primary system pressure operation and 21800 psig for 2000 psia primary system pressure operation are higher and more conservative from preventing DNB than the presently allowable values of 21865 and 21790 psig.
Reactor Coolant System Volume The nominal reactor coolant system volume provided in the design features section of the Technical Specifications is being changed from a liquid volume of 6040 cubic feet to a total volume (both liquid and steam) of 6500 cubic feet for Unit I and 6643 cubic feet for Unit 2.
The actual RCS volume for Unit 1 is not changing. The pressurizer steam space volume is being included into the stated volume. The nominal reactor coolant system volume for Unit 2 is increasing due to higher volume associated with the new steam generators.
These changes do not result in any significant FSAR accident analysis changes, because the smaller coolant volume associated with Unit 1 is typically more limiting and as stated previously, the Unit I coolant volume is not being changed. A m:.nor change to the boron dilution accident was included in FSAR 14.1.4 to account for lower steam generator tube plugging levels. The evaluation of the containment integrity analysis (FSAR 14.3.4) results in minimal increase in maximum calculated containment pressure due to the slight increase in RCS volume. The evaluation shows that the peak containment pressure could be increased by approximate 0.77 psi.
This results in an estimated peak containment pressure of 53.34 psig, which is still substantially less than the design pressure limit of 60 psig.
Reactor Core Safety Limits The reactor core safety limits have been recalculated and the separate limits that were esta'slished for Unit 2 at reduced RCS flow are no longer necessary. The recalculation of the reactor core i safety limits was performed to correct a previously evaluated DNB analysis discrepancy and l include a slightly higher pressure for the highest pressure safety limit, which is consistent with the high pressurizer pressure trip point used in the accident analyses. Otherwise, the core safety limits are essentially unchanged.
l Overtemperature and Overpower Delta T Reactor Trip The Overtemperature and Overpower delta T reactor trip setting limits are being changed to provide adequate protection over the full range of expected reactor coolant system operation.
FSAR figure 14-1, in the attached FSAR mark-up, shows these reactor protection functions graphically. Figure 14-1 shows the Overtemperature and Overpower delta T trips in relation to the core safety limits.
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l The Overtemperature delta T reactor trip is one of the reactor protection functions utilized in the analysis of a uncontrolled rod withdrawal at power as described in FSAR 14.1.2 and analysis of the loss ofload accident as described in FSAR 14.1.9. One of the two cases analyzed for the rod withdrawal accident, case B. (see attached FSAR Mark-up 14.1.2) credits the Overtemperature
, delta T trip to provide reactor protection. Two cases analyzed for the loss ofload accident, cases
- a. and b. (see attached FSAR Mark-up 14.1.9), credit the Overtemperature delta T trip to
- provide reactor protection.
l During additional reviews of the Overtemperature delta T trip setting limit, it was determined that the K1 term should be further reduced from the initially proposed value of $1.26 to sl.19 for 2250 psia operation. This reduction is considered necessary based on Westinghouse analyses that compare a maximum value of the Ovenemperature trip to the range of the delta T instrumentation. It was determined by Westinghouse that the maximum trip value was higher than desired. This reduction in the Technical Specifications setting limit for Overtemperature will place the maximum trip value within the desired range of the delta T instrumentation. A revised Technical Specification mark-up of page 15.2.3-2, which shows the new K1 value, is provided as an attachment to this safety evaluation.
The Overpower delta T reactor trip function is not utilized as the reactor protection initiator in any of the FSAR Chapter 14 accident analyses for PBNP. The Overpower delta T trip function is discussed in FSAR accident analyses for uncontrolled rod withdrawal (14.1.2), excessive load increase (14.1.7), steam generator tube rupture (14.2.4), and mpture of steam pipe (14.2.5). As stated previously, FSAR figure 14-1 (see attached FSAR mark-up) shows the Overpower reactor protection limit in relation to the core safety limits.
The proposed Ovenemperature and Overpower delta T setting limits provide adequate reactor protection over the required ranges that are applicable for these functions. The Technical Specifications basis states that the Overpower delta T setting limit prevents power density anywhere in the core from exceeding 108% of design power density, and includes corrections for change in density and heat capacity of the water with temperature, and dynamic compensation for piping delays from the loop temperature indicators. The setting limit includes allowance for instrument errors. The Technical Specification basis also :;tates that the Overtemperature delta T setting limit provides core protection against DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided only that the transient is slow with respect to piping transient delays from the core to the temperature detectors (about 4 seconds),
and pressure is within the range between the high and low pressure reactor trips. The proposed setting limits for overpower and Ovenemperature delta T continue to meet these requirements.
Therefore, safe operation of both units will be maintained by utilizing these proposed setting limits.
Full Power Average Reactor Coolant System Temperature Operating Range The proposed full power average RCS temperature operating range is 573.9 F 2 Tm 2 557 F.
The RCS average temperature operating range provides operational flexibility for reactor
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operation. The reactor coolant system is presently normally operated at a full power average RCS temperature of 570 F.
Analyses and evaluations have been performed that show all the applicable acceptance criteria continue to be met with this proposed operating range for full power average RCS temperature.
The upper limit of the full power RCS average temperature operating range (573.9 F) is consistent with the previously analyzed average reactor coolant system temperature. Hence, the lower limit (557 F) is the main difference between previous analyses and eva!uations and those that have been completed to support this change. The following is a summary of the results of the analyses and evaluations completed that support the proposed full power average RCS temperature range:
Non-LOCA Events The attached table FSAR Table 14-1 summarizes the initial conditions used for each of the non-LOCA analyses. All accidents have been evaluated. Only one accident analysis was determined to be affected. FSAR 14.1.2 Uncontrolled RCCA Withdrawal at Power, was revised to include the lower full power average temperature. The analysis results in the attached FSAR mark-up show that the acceptance criteria for minimum DNBR continues to be met for this accident.
Sensitivity analyses were performed for the loss ofload analysis, FSAR 14.1.9, and the locked rotor analysis, FSAR 14.1.8, that showed that higher average temperature is the appropriate initial condition for these analysis. Evaluation of the reduction in feedwater enthalpy did not result in any changes to the accident analyses as presented in the FSAR.
Large Break LOCA (FSAR 614.3.2)
To evaluate the low limit for the full power average temperature, a recently completed analysis for another two-loop plant that supported a reduction in full power Txvo was used. This analysis conservatively estimates the PCT penalty for a given reduction in Tavo. For Point Beach, the effect of the reduction of full power average temperature from the analysis average temperature is conservatively estimated as a 70 F increase in the PCT. Combination of this PCT increase with the analysis result of 2028 F and other PCT evaluations results in an evaluated PCT of 2137 F.
This result remains below the limit of 2200 F. Therefore, the large break LOCA analysis remains acceptable.
Small Break LOCA (FS AR 614.3.1)
A sensitivity analysis performed for another two-loop Westinghouse PWR shows that a lower full power average reactor coolant system temperature results in lower PCT for small break loss-of-coolant accidents. Therefore, no PCT assessment is necessary to account for the proposed full power average RCS temperature range.
LOCA Forces Evaluation (FS AR 614.3.3)
The core and internals integrity analysis is performed to show that the core will remain coolable and intact after being subjected to the hydraulic forces associated a major reactor coolant system rupture. An analysis was performed that envelopes the proposed full power average temperature range. The full power average temperature range used for this analysis was 557 F to 581.3 F to bound all past and proposed operating conditions. The results of this analysis show that the acceptance criteria continue to be satisfied.
Long Term LOCA Mass and Enerny Release and Containment Integrity Evaluation (FSAR sl4.3.4)
The most limiting analyses for containment integrity are based on higher full power average RCS temperatures. As stated above, the proposed maximum of the full power average RCS temperature range is consistent with the current full power RCS average temperature. Therefore, the proposed full power RCS average temperature range does not affect the containment integrity analysis.
Shon Term LOCA Mass and Energy Release and Subcompartment Evaluation The lower full power RCS average temperature could slightly increase the mass and energy deposition analysis for subcompartment pressurization. The current basis for evaluation of the containment internal structures is a postulated pipe break in the reactor coolant loop piping with an area equivalent to the cross-sectional area of a reactor coolant loop pipe. Based on the use of
" Leak-Before-Break" (LBB) methodology, it is not necessary to consider RCS loop piping breaks for this analysis. The smaller RCS loop piping connections such as the surge line, RHR line, and l accumulator line are considered. It has been assessed that the lower mass and energy release rates !
associated with the smaller pipe diameters, including the effect oflower full power RCS average l temperature, would result in less significant subcompartment pressurization )
Steam Generator Tube Ruoture (FSAR 614.2.4)
The steam generator tube rupture accident was reanalyzed. The reanalysis included input assumptions that envelope the proposed range of full power RCS average temperature. The results of these analyses are provided in the attached FSAR mark-up of FSAR 14.2.4. As shown in the results of the attached analysis, the doses to the public as a result of a steam generator tube rupture are less than the permissible limits of 10 CFR 100.
Mechanical Components and Systems Evaluation In 1993, it was determined that increasing steam generator tube plugging levels in Unit 2 would necessitate a reduction in the RCS raw measured total flow rate Technical Specification requirement for Unit 2. Technical Specification change request 160 was submitted to the NRC via letter dated June 11,1993. During processing of Technical Specification change request 160 (Amendments 142 and 146), information was provided to the NRC that explained a discrepancy
between the original full power average RCS temperature used in these analyses and the actual full power average RCS temperature at which PBNP was being operated. Additionally, evaluations were submitted to the NRC that justified operation of both units and it was stated that this discrepancy would be corrected during analyses being performed to support the l replacement of steam generators at PBNP.
1 The analyses that support the correction of the full power average reactor coolant system temperature discrepancy have been completed. These evaluation cover the full power average temperature range from 557 F to 581 F to cover all past and future operating conditions. The results of these analyses are summarized in the attached report, " Mechanical Components and Systems Evaluation." As shown in the attached report, all applicable acceptance criteria continue to be satisfied for replacement steam generators in Unit 2 and under the proposed Technical Spe-ifications for both units at PBND.
Low-Low Steam Generator Levi Seuing Limit Changes The Technical Specifications for low-low steam generator level are provided in Technical Specifications 15.2.3.1 for the reactor trip function and Technical Specifications Table 15.3.5-1 item 7 for auxiliary feedwater initiation. The current Technical Specifications setting limit for these protection fimetions is 5%, which is based on an analysis trip point of 0% plus 5% error allowance. The recently completed analyses for the replacement steam generators use an analysis trip point of 10% for these protection functions. Using an error allowance of 10% results in the proposed Technical Specifications setting limit of 20% for these protection functions. This allowance accounts for instrument uncertainties as well as the affect of flow and subcooling of the fluid in the steam generator downcomer.
The analyses that utilize these setting limits are the loss of normal feedwater (FSAR Q14.1.10) and loss of AC power to the station auxiliaries (FSAR 14.1.11). The attached FSAR mark-up for these analyses shows that the new setting limit is acceptable and that all accident analysis acceptance criteria is satisfied.
The proposed Technical Specifications maintain the 5% setting limit for Unit I until such time that the narrow range level instrumentation is modified to be consistent with the Unit 2 replacement steam generator narrow range level instrumentation. The 10% steam generator low-low level analysis trip point for use with the lower steam generator narrow range level tap locations is a lower trip point than the 0% analysis trip point with the current narrow range tap locations. The 10% analysis trip point corresponds to approximately 343.2 inches above the tubesheet. The 0%
trip point corresponds to 385.5 inches above the tubesheet. Therefore, the accident analyses that utilize the 10% trip point provide more conservative results than would be achieved if the analyses were performed using the 0% trip point.
The uncenainty analysis that shows the 10% error allowance is adequate will be provided when finalized.
f Therefore, the proposed steam generator low-low level setting limit of 20% for the lower narrow
- range tap location provides appropriate Technical Specifications requirements for actuation of the
{ reactor trip and auxiliary feedwater system in response to falling steam generator level.
j CONCLUSION 1
4 i Based on the results of analyses and evaluations summarized in this shfety evaluation and its I
attachments, the proposed Technical Specifications associated with Technical Specifications j change requests 188 and 189 and this supplement will provide appropriate requirements for safe ,
4 operation of both units.
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i MECIIANICAL COMPONENTS AND SYSTEMS EVALUATION NSSS Fluid Systems
- NSSS fluid systems were reviewed to confirm continued compliance with industry codes and standards, regulatory requirements, and applicable performance and design basis requirements.
The systems reviewed were the Reactor Coolant System (RCS), Chemical and Volume Control System (CVCS), Residual Heat Removal System (RHRS), Safety Injection System / Containment .-
Spray System (SIS / CSS), Sampling System (SS) and the Component Cooling Water System
(CCWS). The review was performed based on the range of NSSS performance parameters consistent with the RSG and proposed Technical Specifications.
A comparison of the proposed range of NSSS performance parameters with the reference performance parameters previously evaluated for systems and components indicates differences that could impact the performance of the above fluid systems. For example, the proposed analyzed steam generator tube plugging level of 25 percent would result in a reduction in the RCS full-load thermal design flow of 4.2 percent (from 89,000 gpm to 85,200 gpm per loop). A reduction in primary loop flow also decreases vessel pressure drop and the available driving head for pressurizer spray.
The evaluations of the above fluid systems relative to compliance with industry codes and standards, regulatory requirements, applicable performance and design basis requirements are delineated below.
Reactor Coolant System PressurizerSajhty Valves The Point Beach Units i and 2 pressurizer safety valves are required to have adequate capacity to ensure that the RCS pressure does not exceed 110 percent of system design pressure (the maximum pressure allowed by the American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME B&PV) Code) for the worst-case loss-of-heat-sink event.
It was determined that the pressurizer safety valves are not affected by the proposed Technical Specification changes The power level remains unchanged so that the operation of these valves with the RSGs is unchanged. The design of the pressurizer safety valves, associated piping and relief tank is also not changed.
Power Operated Relief Valves The Point Beach Units 1 and 2 power-operated relief valves (PORVs) are required to have adequate capacity to prevent pressurizer pressure from reaching the high-pressure reactor trip setpoint for the design basis 50 percent turbine load rejection transient. Based on the composite range of NSSS performance parameters approved for systems and components, a component sizing evaluation was performed and the results indicated that the present PORV design capacity (358,000 lbs/hr or 179,000 lbs/hr per valve at 2350 psia) is adequate to meet the above design basis performance requirement. It can also be concluded that the supporting fluid systems design
I
! of the inlet and discharge piping from the PORVs is also adequate for the composite range of l uprated NSSS performance parameters approved for systems and components, since this system piping is designed based on the design capacity of the PORVs.
PressurizerSpray Valves The Point Beach Units 1 and 2 pressurizer spray valves are required to have adequate capacity to maintain the pressurizer pressure below the pressurizer PORVs actuating set pressure of ,
2350 psia following a 11. percent step load decrease from full power. Based on the uprated NSSS performance parameters approved for systems and components, a component sizing evaluation was performed and the results indicate that the present spray valve design capacity is adequate to meet the above design basis performance requirement.
The spray valves and spray lines are sized to pass the design spray valve flow rate with a pressure ,
drop equal to the pressure drop from the spray flow scoop on the cold leg to the pressurizer surge l line connection on the hot leg. The design of the pressurizer PORVs, spray valves, and associated piping is adequate for the range of operating parameters for the RSGs.
Resistance Temperature Detector Bypass Delay Time The Point Beach Units 1 and 2 RCS fast-response temperature detectors that provide temperature signals to the Reactor Control and Protection System are mounted in manifolds located in small bypass loops around the steam generator and the reactor coolant pump (RCP) of each loop.
l The design of the Reactor Control and Protection System requires that the fluid transport time from the reactor coolant loops to the last RTD in the RTD manifolds be less than or equal to 1.0 second. This limits the bypass loop piping size and length and bypass flows to particular values.
The bypass loops are sized to pass sufficient flow rates to meet this fluid transport delay time based on the available pressure drop in the main coo! ant loops. In 1984, the bypass loops were modified to eliminate several bypass loop isolation valves. This modification resulted in an increase in flow through the bypass loops and a corresponding decrease in fluid transport times.
With a given bypass loop configuration, the flow through the hot-leg bypass loops is primarily a function of the pressure drop acioss the steam generators and the flow through the cold-leg bypass loops is primarily controlled by the operating head of the RCPs. The replacement steam generators for Unit 2, assuming no tube plugging, will result in a decrease in primary loop frictional resistance / pressure drop and a corresponding decrease in RTD bypass flows, a fluid systems hydraulic analysis was performed to confirm that the reduced flow rates through the bypass loops will result in a fluid transport delay time ofless than or equal to 1.0 second. The results of this analysis concluded that the existing margin in fluid transport times was more than adequate to offset the decrease in bypass flows and corresponding increase in fluid transport delay times caused by the replacement steam generators.
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Chemical and Volume Control System A review of CVCS operation was performed. The review included the letdown and charging flows, heat loads on the heat exchangers and boric acid storage requirements.
The CVCS review reached the following conclusions:
- The proposed changes in RCS operating temperatures at either approved RCS pressure will have negligible impact on letdown flows.
. At an RCS operating pressure of 2000 psia, the CVCS letdown flow is expected to decrease by about 7 percent. All CVCS design basis requirements can be met at this reduced letdown capability.
. The heat loads on the CVCS heat exchangers are either equal to or less than the design basis heat loads based on the range of RCS operating parameters.
Residual lleat Removal System The RHRS is designed to reduce the temperature of the RCS to 140 F within 20 hrs after reactor shutdown when the service water temperature is 70 F. The RHRS is designed to be placed in service 4 hrs after reactor shutdown, when the RCS pressure and temperature are 400 psig and 350 F, respectively. The initial phase of plant cooldown is accomplished by employing the steam generators and the Steam Dump System (SDS). I l
It has been concluded that the RHRS will still function as designed for operation because the l
proposed Technical Specifications do not alter any of the bases ofRHR operation. '
Safety Irjection System / Containment Spray System The fluid systems calculations and assumptions supporting the safety injection flow data used in the latest loss-of-coolant accident (LOCA) and non-LOCA transient analyses of record were reviewed and confirmed to be conservative for the accident and transient analysis evaluations that support the proposed Technical Specifications.
The heat loads placed on the residual heat removal heat exchanger during the recirculation phase of tafety injection, were evaluated as part of the Component Cooling Water System evaluation.
'I he fluid systems performance used in the analysis of record was confirmed to be conservative for the applicable LOCA and non-LOCA evaluations.
Sampling System The proposed change in RCS operating temperatures will have no impact on SS flows. During sampling, the flow into the SS is controlled by manual throttle valves downstream of the sample heat exchangers and the control range of these throttle valves-will permit the operator to maintain design flows through the SS over the proposed range of RCS operating temperatures at either of the approved RCS nominal operating pressures. The heat load on the steam space sample heat exchanger is higher at an RCS pressure of 2000 psia versus 2250 psia due to the higher steam enthalpy at 2000 psia. However, adequate margin is available to maintain acceptable process conditions during steam space sampling at the lower pressure. All other sample heat exchanger heat loads are either equal to or less than the design heat loads.
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4 Compenent Cooling Water System The CCWS is designed to remove residual and sensible heat from the RCS via the RHRS during i plant cooldown; cool the letdown flow to the CVCS during power operation; provide cooling to
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dissipate waste heat from various plant components; and provide cooling to safeguards loads after an accident. Heat loads on the CCWS are imposed by primag and auxiliary equipment The post- '
- accident heat load imposed on the CCWS is not impacted by the proposed Technical
. Specifications.
i Primary Components
- As part of the analyses performed, component fatigue was evaluated for an extension of plant life i to 60 calendar years. Based on actual plant transient information, the existing set of transient j cycles for 40 years covers the extension of fatigue usage to 60 years, with the exception of l l certain hydrostatic tests. An increased number of hydrostatic tests was included in these fatigue l evaluations. This was not necessary for the analysis of Reactor Vessel Integrity which determines the acceptability of the vessels based on application of the PTS rule for the remainder of tne current license.
i, j Reactor Coolant Pumps I The Point Beach RCPs are Model 93 which are pre-code stamping. The transient analyses i that support the replacement steam generator program were evaluated. Temperature and
! pressure changes relative to the normal conditions we-re considered. In some cases the changes
- represented lower stresses and hence did not affect the stress intensity range. In some cases, the i changes were found to be insignificant in accordance with the ASME definition. Thus fatigue usage was not affected.
1 The conditions associated with the rei acement ! steam generators and the proposed technical specifications are considered acceptable for the Model 93 RCPs with respect to Code structural integrity. The new NSSS performance parameters and the new design transients are considered acceptable.
The RCP motor performance was evaluated by calculating the new loads resulting from the revised parameters. The results show a new hot loop motor load of 5473 HP and a new cold loop motor load of 7173 HP. The Point Beach RCP motors have a nameplate rating of 6000 HP.
Since the new loads are less than the nameplate rating of the motors, (the cold loop rating is taken to be 125% of the hot loop rating if not shown separately), they are acceptable using the revised parameters.
Pressurizer The results of the pressurizer analysis for conditions associated with the replacement steam generators and the proposed technical specifications indicate that the pressurizer components meet the ASME Code, Section HI stress analysis and fatigue analysis requirements plus all other loadings given in the applicable design specifications. Analysis results also demonstrate that there is no impact on the primay stresses and primary and secondary stresses presented in the pressurizer stress reports and the surge nozzle analysis.
I i
l Reactor Coolant Piping and Supports l The parameters associated with the replacement steam generators (RSG) and proposed Technical l Specifications were reviewed for potential impact on the following components: reactor coolant )
loop (RCL) piping, primary equipment nozzles, primary equipment supports, and the pressurizer j' surge line piping. The temperature and weight and center-of-gravity changes associated with the RSG were evaluated. The changes in the thermal design transients were factored into the pressurizer surge line thermal stratification analysis because of the fatigue aspects of that evaluation. ,
i The acceptance criteria for the loop piping stress evaluation is contained in the B31.1 Power Piping Code. Except for a small change in nonnal operating thermal stresses, the main loop piping and supports and equipment nozzles have the same loads and stress levels as existed previous to the RSG program.
The primary loop piping LBB loadings were evaluated in WCAP-14439. That document ;
indicates that the loadings associated with the RSG program are acceptable for the criteria l defined.
1 Since the RSG is a little heavier than the original SG, the RCL support loads for deadweight and !
seismic conditions were affected. Since the pipe rupture design basis is a static type design, the l RSG had no effect on the pipe rupture design aspects of the RCL supports, except that the SG I columns had a higher deadweight load to combine with pipe rupture. The support assessment for the effect of the RSG considered load combinations that included OBE or DBE. These seismic loadings from the Unit I analysis were appropriate to use because of the similarities in the Units 1 l and 2 RSV's. The SG columns were also addressed for the higher deadweight loading and pipe i rupture combination. The results of the assessment were such that all load combinations met appropriate allowables.
The results of the evaluation for the pressurizer surge line stratification showed that the RSG conditions changed the fatigue usage factor at the location of highest usage factor by only a negligible amount. The calculated change in loadings on the pressurizer nozzle due to stratification for the RSG conditions was not considered significant. 'Me change in nozzle loadings was considered insignificant because the original loadings on the pressurizer nozzle were conservative envelopes that lumped various transients under a small number of bounding thermal cases.
The parameters associated with the RSG program have been evaluated for impact on the RCL piping , the primary equipment nozzles, the primary equipment supports, and the pressurizer surge line. The evaluation indicates that all components meet appropriate allowables. The evaluation for the stated components concludes that the RSG program has no adverse effect on the ability of these components to operate until the scheduled end of plant operation.
l Auxiliary Components l The purpose of this evaluation is to determine the effect of the RSG and proposed Technical l Specification conditions on the auxiliary equipment provided by Westinghouse, i.e. auxiliary 1
l l
valves, pumps, tanks and heat exchangers. The primary acceptance criteria is that the auxiliary equipment transients, as applicable to the original design basis of the equipment, remain unchanged or still bounding. If the transients are not affected, any original design analyses performed on the auxiliary equipment would remain applicable.
The original design and qualification requirements for the auxiliary valves were reviewed as described by the equipment specifications applicable to Point Beach Units 1 and 2. It was determined for this evaluation the existing auxiliary system design transients bound the steam generator replacement and proposed Technical Specifications conditions. Therefore the auxiliary component reviews in general were to determine the impact of the steam generator replacement-conditions, if any, on the auxiliary equipment.
In addition it has been assumed that any equipment maintenance or replacement was accomplished in accordance with the original equipment design requirements.
It was concluded that the RSG and proposed Technical Specifications parameters do not affect are bounded by the original Point Beach Units 1 and 2 design parameters for the auxiliary equipment. Therefore, the original design analyses remain applicable for the qualification of the auxiliary mechanical equipment.
Leak Before Break Methodology The original structural design basis of the reactor coolant system for the Point Bet ch Units 1 and 2 required consideration of dynamic effects resulting from pipe break and that protective measures for such breaks be incorporated into the design. NRC and industry initiatives resulted in demonstrating that Leak-Before-Break (LBB) criteria can be applied to reactor coolant system piping based on fracture mechanics technology and material toughness. Generic analyses by Westinghouse for the application of LBB for specific plants was documented in response to Unresolved Safety Issue A-2 and. approved for Point Beach in a NRC letter dated May 6,1986.
This use of LBB technology allows the dynamic effects of postulated ruptures in primary coolant loop piping to be excluded from the design basis.
The loadings, operating pressure and temperature parameters that were used in the analysis envelope the Unit 2 Steam Generator replacement and proposed Technical Specifications for both units. Mechanical properties were determined at the various operating temperatures expected.
Based on plant specific analysis, the LBB conditions are satisfied for the Point Beach Units 1 and 2 primary loop piping. All recommended margins are satisfied. It is therefore concluded that dynamic effects of reactor coolant system primary loop pipe breaks need not be considered in the structural design basis of the Point Beach Units 1 and 2 for the RSG and proposed Technical Specifications conditions.
Steam Generators All three steam generator models were reviewed with respect to the RSG and proposed Technical Specifications NSSS performance parameters.
The evaluations for the Point Beach Model 44 steam generator components show that the structural integrity of the steam generator components are maintained. The fatigue calculations were performed for 40 years of operation. Specified conditions are accommodated without exceeding ASME Code limits.
The evaluations for the Point Beach Unit 1 Model 44F steam generator components show that the structuralintegrity of the steam generator components would be maintained. The fatigue calculations were performed for augmented load cycles postulated for 60 years of operations.
Increased load cycles were accommodated without exceeding the Code limits.
The thermal hydraulic operating characteristics of the Point Beach Unit I Model 44F steam generators are within acceptable ranges for anticipated RSG and proposed Technical Specification conditions. The Model 44 steam generators currently in Unit 2 are acceptable for the conditions defined for the RSG and proposed Technical Specifications. The replacement steam generator structural and thermal-hydraulic design performance is acceptable.
Reactor Vessel Structural Integrity Evaluations were performed for the various regions of the Point Beach I and 2 reactor vessels to determine the stress and fatigue usage effects of NSSS operation at the revised operating conditions of the RSG and proposed Technical Specifications throughout the current plant operating licenses and for up to 60 years with renewed operating licenses. The evaluations assess the effects of the revised design transients on the most limiting locations with regard to ranges of stress intensity and fatigue usage factors in each of the regions as identified in the reactor vessel stress report and addendum.
The RSG and proposed Technical Specifications affects several of the maximum ranges of stress intensity reported in the Point Beach I and 2 reactor vessel stress reports. The evaluations show that for the limiting locations, most of the maximum ranges are unchanged when the revised operating parameters, design transients and design interface loads are incorporated. The exceptions are the outlet nozzles, the closure studs, the CRDM housings, the inlet nozzles and the bottom head instrumentation tubes. The maximum cumulative fatigue usage factors at all of the limiting locations were determined to be less than 1.0. The number of pressure tests covered by the hydrostatic test transient were increased to account for 60 calendar years of reactor operation with license renewal.
In addition to the revised operating parameters and design transients for the RSG and proposed Technical Specifications, a new set of LOCA loads at the reactor vessel / reactor internals interfaces was identified. The revised interface loads were evaluated by comparing them with the corresponding Faulted Condition reactor vessel / reactor internals interface loadings which were justified for application to the Point Beach 1 and 2 reactor vessels. All of the LOCA loads associated with the RSG and proposed Technical Specifications were found to be less than the corresponding loadings and are therefore, acceptable for application to the reactor vessel.
The reactor vessel evaluation for RSG and proposed Technical Specifications conditions showed that the operation is acceptable in accordance with both the 1965 Edition of Section III of the ASME Boiler and Pressure Vessel Code for the Unit I reactor vessel and the 1968 Edition of
i j Section III of the ASME Boiler and Pressure Vessel Code with Addenda through the Winter of 1968 for the Unit 2 reactor vessel for the remainder of the current plant licenses and for license
- extension up to 60 years.
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- Fatigue and stress evaluations were performed based on the new thermal transients that show the
! reactor vessel head adapter plug design complies with the description in the applicable equipment j specification and the requirements of Section III of the ASME Code.
Reactor VesselIntegrity i Reactor vessel integrity is impacted by any changes in plant parameters that affect neutron fluence j levels or temperature / pressure transients. The changes in neutron fluence resulting from the j proposed RSG and proposed Technical Specifications have been evaluated to determine the i impact on reactor vessel integrity for Point Beach. This assessment included evaluations of the
) beltline region material upper shelf energy (USE) values (excluding weld metal), current
- surveillance capsule withdrawal schedules, applicability of the current plant heatup and cooldown
{
! pressure-temperature limit curves, applicability of the Emergency Response Guideline (ERG) ;
j limits, and RTers., values for the Pressurized Thermal Shock (PTS) Rule (10CFR Part 50.61).
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It is concluded that the Unit 2 RSG and proposed Technical Specifications for Point Beach Units j
- I and 2 will not have significant impact on the reactor vessel integrity based on the following l
reasons:
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1 j e The USE values of all beltline region plates and forgings are expected to remain above
?
- 50 ft-lb through the life of the Units 1 and 2 vessels.
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- The current heatup and cooldown pressure-temperature limit curves will remain applicable.
l Reactor Internals i
Evaluations have been performed to assess the effect of the new RCS conditions (due to the i
RSG and proposed Technical Specifications) on the reactor pressure vessel / internals system at b Point Beach Units. Operating a plant at conditions (power, pressure, temperature, flow) other i than those considered in the original design requires that the reactor vessel system / fuel interface !
! be thoroughly addressed in order to assure compatibility and that the structural integrity of the
! reactor vessel / internals / fuel system is not adversely . In addition, thermal-hydraulic analyses are
} required to determine plant specific core bypass flows, pressure drops and upper head l l temperatures in order to provide input to the Loss of Coolant Accident (LOCA) and non-LOCA safety analyses as well as Nuclear Steam Supply System (NSSS) performance evaluations.
I l i The results of these analyses are summarized below:
I i e The vessel pressure drops, bypass flows and hydraulic lift forces are not significantly
) affected by the proposed RCS conditions.
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e The design core bypass flow value of 6.5 % of the total vessel flow can be maintained at Point Beach Units. !
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- The current RCCA drop time Technical Specification limit of 2.2 seconds remains i applicable. l
. The stmetural integrity of the reactor internals is maintained with the new reactor coolant system conditions.
Control Rod Drive Mechanisms This section addresses the acceptability of the RSG and proposed Technical Specifications NSSS performance parameters for the Westinghouse Control Rod Drive Mechanisms (CRDMs). The units have L-106A CRDMs, full length (F/L) manufactured by Westinghouse Electro-Mechanical Division and 12X135 part length (P/L) mechanisms manufactured by Royal Industries.
The RSG and proposed Technical Specifications NSSS performance parameters were compared to the existing CRDM E-Spec values and/or to the generic analysis values.
The CRDM parameters are given by the hot leg data which is the vessel outlet data. The temperature is shown to increase from the present 602.8 F to as much as 611.3 F. This is only an 8.5 F change and is insignificant since this is only a 1.4% increase. The F/L CRDM Code generic analyses were done for normal operation at 650 F, hence the higher temperatures are still bounded by the generic analysis. The P/L CRDM also considered 650 F in the 3Sm stress range evaluation.
The only RSG and proposed Technical specifications transients to be resolved as being acceptable are those which exceed either the original E-Spec values or the generic CRDM Code evaluation.
If the RSG and proposed Technical Specifications pressure changes (APs) and temperature changes (ATs) and the maximum pressure and temperature of a transient are less than presently accepted, they are obviously acceptable. Pressure and temperature decreases do not affect the stress intensity range and the magnitudes and cycle count must be significant to affect a Code fatigue waiver or fatigue analysis. P/L CRDMs are no longer actively used, however, the housings remain. The components of the P/L housings arejustified to be acceptable by geometric similarity to the F/L housings and by previous analysis.
A comparison of the RSG and proposed Technical Specifications transients to either the initial E-Spec values or the generic analysis shows the RSG and proposed Technical Specifications transients to be acceptable for Point Beach Units 1 and 2. Since the RSG and proposed Technical Specifications parameters and transients do not significantly affect the present Code analysis or fatigue waiver / evaluation, there is no expected increase in the CRDM fatigue usage due to the additional transients or the increase in hydro-test cycles.
Fuel Structural Design The 14 x 14 Optimized Fuel Assembly (OFA) was evaluated for the safe shutdown earthquake (SSE) and LOCA conditions. Steam generator replacement has no impact on the seismic
evaluation. The LOCA analysis used the LBB criterion. The LOCA grid load was based on an accumulator line break which is the most severe pipe rupture.
The maximum grid loads from both SSE and LOCA events were combined using the square root sum of the squares (SRSS) method. It was determined that the 14x14 OFA fuel assembly design is structurally acceptable for Point Beach Units 1 and 2.
Conclusion NSSS fluid systems continue to comply with all industry codes and standards, regulatory requirements, and applicable performance and design basis requirements for the range ofNSSS performance parameters consistent with the RSG and proposed Technical Specifications.
Evaluation of NSSS component fatigue at conditions associated with the RSG and proposed Technical Specifications shows acceptable results through the currently licensed life of the plant.
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