ML20059M073
| ML20059M073 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 11/10/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059M062 | List: |
| References | |
| NUDOCS 9311180141 | |
| Download: ML20059M073 (16) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 69 TO FACILITY OPERATING LICENSE NO. NPF-42 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482
1.0 INTRODUCTION
By application dated January 5,1993, as supplemented by letter dated October 1,1993, Wolf Creek Nuclear Operating Corporation (the licensee) requested an amendment to Facility Operating License No. NPF-42 for the Wolf Creek Generating Station. The proposed change would revise the Wolf Creek license and Technical Specifications to increase the maximum reactor core power level (rated thermal power) from the present value of 3411 megawatts thermal (MWt) to a revised limit of 3565 MWt. The increase in allowed core power combined with the energy added by the reactor coolant pamps would result in the proposed changes allowing Wolf Creek to operate at a nuclear steam supply system (NSSS) power of 3579 MWt.
In addition to the increased core power level, the licensee has also proposed changes in the allowable operating temperatures of the reactor coolant system. The reductions in reactor coolant hot-leg temperatures are being proposed in order to reduce the potential for stress corrosion cracking of steam generator tubes.
The proposed changes represent an approximate 4.5 percent increase over the current licensed power level. The proposed temperature changes include a planned hot leg temperature reduction of 5 degrees Fahrenheit and a possible 15 degree Fahrenheit reduction which may be pursued in the future. The proposed maximum temperature reduction (15'F) results in secondary conditions that would likely require modifications to the turbine in order to gain the desired increase in electrical output.
In support of the proposed changes, i
the licensee provided the results of analyses and evaluations performed to determine the impact of the changes in power level and operating temperature on the NSSS and balance of plant (B0P). Analysis of the core thermal-hydraulic response to various potential accidents determined that the 5*F hot leg temperature reduction was necessary in order to meet safety limit design criteria. The October 1,1993, letter provided additional information in response to a request by the staff and did not propose additional changes from those described in the Notice of Consideration of Issuance of Amendment to l
Facility Operating License and Opportunity for Hearing published in the Federal Reaister (58 FR 26565 dated May 4,1993).
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2.0 EVALUATION 2.1 Nuclear Steam Supply System (NSSS)
The licensee termed the combination of the higher core power level and reduced reactor coolant hot leg temperature as the proposed rerate conditions of the Wolf Creek Generating Station. The scope of the licensee's review to support the rerate conditions encompassed all aspects of the Wolf Creek NSSS design and operations affected by the proposed changes. NSSS designs were reviewed to verify compliance at the rerated conditions with licensing criteria and standards currently specified in the Wolf Creek operating license. The structural design of the NSSS equipment was reviewed to assure that compliance with industry codes and standards had been maintained at the rerated conditions. The review encompassed the verification that the NSSS components and systems will continue to meet functional requirements specified in the Updated Safety Analysis Report (USAR), Currently approved analytical methods were used for the analyses at the rerated conditions.
In addition to evaluating the ability of the NSSS to operate at the rerated conditions during normal operation, the licensee reanalyzed the design basis transients and accidents which the staff utilizes to determine the adequacy of safety margins. The licensee submitted these analyses in support of License Amendment Number 61 which included revised limits for reactor coolant flow rate, core peaking factors, and administrative processes related to Wolf Creek's seventh operating cycle. Although Amendment 61 did not include the rerated core power level or reduced hot leg temperatures, the supporting analyses for this previously issued amendment included assumptions for the limiting rerate conditions. The staff findings related to Amendment 61 in terms of the design basis transients and accidents have been determined to be bounding for the proposed rerate conditions.
Core Desion On March 30, 1993, the NRC issued License Amendment 61 for Wolf Creek Generating Station authorizing the changes requested by the licensee for the unit's seventh operating cycle. The supporting analyses were performed assuming the limiting rerate conditions and therefore remain applicable for the proposed changes being addressed by this safety evaluation. The specific changes related to changing the plant's operating conditions to the power t
level and temperatures associated with the rerate include changes to the definition of rated thermal power, overtemperature and overpower delta-T setpoints, and maximum reactor coolant average temperature.
Overoressure Protection Pressurizer safety valves are required to be designed with sufficient capacity to prevent the pressurizer pressure from exceeding 110 percent of design pressure following the worst reactor coolant system (RCS) pressure transient.
For purposes of analytical justification, this event is specified to be a turbine trip.
No credit is taken for operation of reactor coolant system relief valves, steam line relief valves, steam dump valves, pressurizer level or pressure control systems, or direct reactor trip on turbine trip. The
Safety Evaluation Report for Wolf Creek (NUREG-0881) was based upon the criteria provided in the Standard Review Plan (NUREG-0800) in that the safety valve capacity had been shown to be adequate assuming a reactor trip occurred on the second safety-grade signal received following the turbine trip.
In the licensee's submittal, the analysis which demonstrated adequate overpressure protection credited the pressurizer high pressure reactor trip which is the first trip signal generated except for the anticipatory reactor trip on turbine trip. This assumption was consistent with the analysis presented in the original USAR. However, in order to resolve the discrepancy between the various licensing basis documents, the licensee has committed to submit information in a future update of the USAR which will document the plant specific analyses related to overpressure protection and adequacy of relief capacity assuming the rerated conditions and reactor trip upon receipt of the second safety-grade signal. The licensee has indicated that assurance that the existing overpressure protection is adequate to meet the more conservative analytical criteria is provided by sensitivity analyses performed for similar plants which have concluded that the safety valve capacity is not exceeded if the reactor trip is delayed until actuation of the second safety-grade signal.
In addition, the original sizing calculation for the pressurizer safety valves was based on a reactor power equivalent to 102 percent of the proposed rerate power level.
Based upon the licensee's evaluations, related analyses for similar plants, and the licensee's commitment to submit confirmatory documentation, the staff concludes that compliance with General Design Criteria (GDC) 15, Reactor Coolant System Design, and other applicable regulatory requirements is maintained during operation at the rerate conditions.
The licensee stated that the low temperature overpressure protection function was not affected by the proposed rerate changes. This conclusion is based upon the fact that the low temperature overpressure protection function is only used during shutdown conditions which are not significantly affected by the proposed changes in operating conditions.
The staff finds this conclusion acceptable.
Auxiliary Feedwater and Residual Heat Removal The staff review and approval of the Wolf Creek auxiliary feedwater system (AFS) design is given in NUREG-0881, Safety Evaluation Report (SER), Section 10.4.9 by reference to NUREG-0830, SER for Callaway. These SERs describe the limiting transients identifying worst single failure assumptions and minimum flow requirements for the Wolf Creek AFS design. The rerate conditions were found not to change the limiting scenarios for single failure and required flow. The adequacy of revised AFS flow capacity was demonstrated in the supporting analyses submitted for Amendment 61. The licensee determined that the condensate storage tank volume required by technical specifications is adequate to support AFS operation at the rerate power level for the design basis plant cooldown to RHR entry conditions and for the station blackout coping analysis. The staff concludes that the performance of the AFS remains adequate to provide the required cooling functions following the change to the rerate conditions.
I The licensee indicated that the original 16-hour plant cooldown time from residual heat removal (RHR) initiation at 350*F to 140*F would be increased to 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> for the rerated conditions. Although the time to cool down to 200*F with only a single RHR train available increases from 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br />, the system capabilities continue to comply with the guidance provided in Branch Technical Position RSB 5-1 and 10 CFR Part 50, Appendix R.
Therefore, the staff concludes that the impact of the rerate conditions on RHR performance is acceptable.
Emeroency Core Coolino System (ECCS)
From the licensee's study, no adverse impact to ECCS operability or vulnerability to single failure due to the rerated conditions was identified.
The licensee submitted revised ECCS performance analyses in support of Amendment 61 which justified various changes associated with Cycle 7 operation. The licensee performed large and small break analyses at the limiting rerate conditions and determined that all acceptance criteria continued to be satisfied. The NRC staff has reviewed the licensee's analyses and concludes that the ECCS analyses referenced in support of the rerate conditions continues to be in compliance with 10 CFR 50.46 and Appendix X.
The Wolf Creek ECCS is, therefore, acceptable for operation at the rerated conditions.
Accident Analyses The licensee states that all events in USAR Chapter 15 were reanalyzed or reevaluated considering the rerate conditions. The details of these analyses were submitted in support of Amendment 61 which justified various parameter changes associated with Cycle 7 operation.
Included in the revised accident analysis assumptions is a steam generator plugging level of 10 percent. The staff reviewed these analyses and concluded that the appropriate safety criteria continue to be met. The methodologies used by the licensee to analyze the USAR Chapter 15 accidents and transients have been previously reviewed and approved by the staff.
The licensee provided the evaluation of the effect of the rerate on accident radiological consequences with the supporting documentation for Amendment 61.
The original licensing basis accident analysis source terms for Wolf Creek were conservatively based on an assumed core power level equal to the proposed rerated core power level of 3565 MWt. The original licensing basis source terms assumed a twelve month operating cycle. As part of the analytical efforts related to the rerate and cycle 7 operation (Amendment 61), the radiological source terms were determined with an assumed operating cycle of eighteen months to reflect current fuel cycle designs.
Except for the changes related to cycle length, the accident doses were re-calculated using methodology and assumptions which are consistent with the original licensing basis. The calculated doses for the exclusion area, low population zone, and control room were found to change slightly from those presented in the original licensing basis but in all cases remained well below applicable regulatory limits.
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5-l 2.2 Safety Related Coolina Water Systems Service Water and Ultimate Heat Sink i
The Wolf Creek station service water system consists of the service water system (SWS) and the essential service water system (ESWS). The SWS is used during normal operation and normal shutdown conditions. The ESPS removes heat i
from plant components which require cooling for safe shutdown of the reactor 1
or following a design basis accident. Major components supplied by the ESWS include the component cooling water (CCW) heat exchangers, containment air coolers, diesel generator cooler, and various engineered safety features room coolers. The ESWS can also provide makeup water to the spent fuel pool and CCW system as well as serve as the backup water supply to the auxiliary t
feedwater system.
The ESWS intake and discharge is the plant's main cooling lake which also serves as the ultimate heat sink (UHS).
The UHS for Wolf Creek consists of a normally submerged seismic category I cooling pond which is formed by a dam built into the main cooling lake.
The licensee evaluated the ESWS and UHS for the rerated conditions under the most limited postulated post-accident scenarios. New analyses of ESWS heat inputs, including the rerate related' changes, determined that no significant i
increases occurred which would result in the plant being outside original s
bounding conditions or analysis results. Analysis of UHS capabilities were i
originally performed assuming the rerated power level. Other changes associated with the rerate were determined to have an insignificant impact on i
UHS temperatures.
Thus the analysis demonstrates that the UHS and ESWS have adequate cooling capability to satisfy design and regulatory requirements and i
are therefore acceptable for operation at the rerated conditions.
Component Coolina Water i
The component cooling water (CCW) system provides cooling water to selected auxiliary components during normal operation and provides cooling water to several engineered safety feature systems (ESFS) during design basis i
accidents. The system is a closed loop system which serves as an intermediate i
barrier between the SWS or ESWS and potentially radioactive systems in order to eliminate the possibility of an uncontrolled release of radioactivity.
l Nonessential loads are isolated in the event of a safety injection signal.
Essential loads of the CCW system include the residual heat removal (RHR) heat t
exchangers, ECCS pump coolers, spent fuel pool heat exchangers, and reactor i
coolant pump coolers.
J The licensee evaluated the effects of the proposed rerated conditions on the CCW system. The licensee found that normal and accident heat loads following the proposed rerate would be bounded by existing analyses.
Based upon the licensee's evaluations of the CCW and related systems, the staff finds that the system will continue to fulfill its functions under the rerated plant conditions.
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s Soent Fuel Pool Coolina The licensee evaluated the spent fuel pool cooling (SFPC) system to determine the ef fects of the increased rated thermal power on the capability of the SFPC system to maintain fuel pool temperatures within acceptable limits. The evaluation included scenarios related to a normal refueling offload and a full core offload.
In each scenario, only a single train of SFPC was assumed to be operating. The total heat load was determined in accordance with the current licensing basis analysis.
The licensee's analyses concluded that the bulk spent fuel pool temperature would remain below 135'F for the normal offload and below 160*F for the full t
core offload case. The staff concludes that the SFPC system has adequate capacity to support the additional heat loads associated with the rerated conditions and continues to satisfy applicable design and regulatory requirements.
2.3 Balance of Plant Turbine Generator The turbine overspeed protection system reduces the risk of generation of turbine missiles that could impact operation of safety-related structures, systems, or components. The licensee's evaluation concluded that the maximum steam generator outlet steam flow for the various possible operating conditions would not be significantly greater than the valves wide open (VWO) steam flows assumed in the original turbine overspeed analyses. The bases for the turbine overspeed protection system are not considered to be adversely impacted by the proposed rerate conditions. Since the probability of turbine missile generation is not significantly changed by operation at the rerated conditions, the staff concludes that the existing turbine overspeed protection is adequate for the rerated conditions.
It should be noted that although bounded by the above discussion regarding steam flows and turbine overspeed protection, the licensee does not anticipate that the maximum proposed temperature reduction (15'F) can be achieved with the desired electrical output without modification of the turbine generator.
The main steam system dissipates energy generated by the reactor core to the turbine generator and auxiliary steam loads, the main condenser via the steam dump valves, or to the atmosphere via atmospheric relief valves or main steam safety valves.
Isolation of the main steam system is achieved by the main steam isolation valves and main steam bypass isolation valves.
The licensee evaluated the capability of the main steam system components to perform their design functions under the proposed rerate conditions. The licensee determined that the existing setpoints and capacity of the main steam safety valves are adequate to prevent exceeding 110 percent of design pressure of the main steam system under the most limiting transient. The setpoint and
'e capacity of the atmospheric relief valves were found to remain adequate to control the design load shed of 10 percent rated themal power.
In addition, the atmespheric relief valves were found to have adequate capacity to achieve a 50*F per hour cooldown if the main condenser was unavailable. The main steam isolation valves were evaluated to ensure the valves will continue to perform their isolation function under the maximum differential pressure conditions and within the time limits assumed in the safety analysis.
The staff concludes that the existing main steam system components are adequate to perform their safety functions under the rerated plant conditions.
Main Feedwater The main feedwater system delivers feedwater, at the required pressure and temperature, to the four steam generators.
The safety-related portions of the system ensure isolation capability and provide a path to permit the addition of auxiliary feedwater for reactor cooldown following design basis transients.
The licensee's evaluation shows that the existing design basis for the main feedwater isolation valves and main feedwater bypass isolation valves is not significantly affected by operation at the rerate conditions. The piping configurations associated with the feedwater and auxiliary feedwater systems do not change as a result of the rerate conditions.
The ability of the auxiliary feedwater system to perform its heat removal function was addressed by the licensee. The staff finds that the safety functions of the feedwater system will continue to be satisfied during operation at the rerate conditions.
Other Power Conversion Systems The licensee's analyses indicate that power conversion systems and components (e.g., the steam dump valves, extraction steam system, turbine generator, main condenser, condensate system, feedwater heaters, heater drain system, and circulating water system) satisfy their power generation design bases for the operation as proposed except that the turbine may require modification to allow full power operation at the minimum proposed reactor coolaat temperature. Since the power generation design bases do not include safety functions, these systems and components do not require review by the staff.
2.4 Containment Analyses The licensee has perfomed containment integrity analyses at the rerated conditions to ensure that the maximum pressure inr ue containment following a design basis accident remains below the containment buildir.g design pressure of 60 psig. The calculated peak pressure is also used as a basis for the containment leak rate test pressure to ensure that dose limits would be met in the event of a release of radioactivity to containment.
The licensee indicated that the containment functional analyses included the assumption of most limiting single active failure and the modelling of minimum
's and maximum safety' injection flows. Bounding initial temperatures and pressures for analyses were selected to envelope the limiting conditions of operation. The licensee indicated that although the current licensed NSSS power level is 3425 MWt, containment pressure analyses were initially performed assuming an NSSS power of 3579 MWt. However, improved analytical models, developed since the original licensing analyses, have been utilized in the evaluations performed for the proposed rerate.
Loss of Coolant Accidents (LOCA)
As in the current licensing basis, the licensee's rerate analyses determined the containment temperature and pressure response for a postulated double ended reactor coolant pump suction break (DEPSG) for both minimum and maximum safety injection cases and a postulated double ended hot leg break (DEHLG) which results in the maximum mass and energy release rates during the blowdown phase of the accident.
For the DEHLG case, the Wolf Creek rerate analyses resulted in a containment peak pressure of 42.6 psig. The existing licensing basis analyses for Wolf i
Creek resulted in a peak pressure of 41.7 psig.
For the DEPSG cases, the rerate analyses resulted in peak containment pressures of 43.9 psig and 41.0 psig for minimum and maximum safety injection respectively. The i
corresponding analysis results for existing licensing basis analyses are 47.3 psig and 46.0 psig. The licensee indicated that the reduced calculated peak pressures for the rerate conditions compared to the existing analyses are due to model enhancements.
The differences between the existing licensing basis methodology and the rerate analyses include mass and energy releases calculated using the Westinghouse 1979 evaluation model instead of the 1975 model and modelling of steam / water mixing in the broken loop.
The current technical specification containment leak rate pressure (Pa) is 48 psig. Although the rerate analysis results determined a reduced peak' containment pressure, the licensee has not proposed to reduce the value of Pa as part of the rerate request.
In addition to the review of the peak i
containment pressures, the licensee demonstrated that pressure decreased to less than 50 percent of the peak value within the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> follow %g the postulated LOCA. The LOCA temperature response was found to remain bounded by the main steamline break analysis results. The staff has reviewed the licensee's evaluations and determined that the licensee has adequately demonstrated that the containment will satisfy its design functions under the rerate conditions.
1 Main Steam Line Breaks The licensee's rerate evaluation included determination of containment temperature and pressure response for four cases selected from the existing analysis spectrum of main steamline break sizes and initial power levels.
The cases were chosen to demonstrate that the existing analyses bound the rerate conditions. The evaluation was determined to adequately demonstrate that the existing analyses conservatively bound the possible rerate operating conditions.
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s l Subcompartment Analysis The licensee indicated that the existing mass and energy releases remain bounding for the re&ctor cavity and steam generator subcompartments if leak-before-break is credited. The limitin, case for the pressurizer subcompartments was and remains an assumed break of the surge line. The licensee determined that the rerate conditions would result in an increased mass and energy release to the pressurizer subcompartments. However, significant margins remain between the design limits for the pressurizer subcompartments and the calculated maximum pressures associated with the rerate analyses.
Based on the licensee's evaluation, the staff concludes that the rerating is acceptable in terms of subcompartment pressure loading analyses.
Post-LOCA Hydrocen The licensee indicated that the Wolf Creek containment post-LOCA hydrogen generation analyses were reviewed to determine any impact due to the rerate.
The LOCA analysis determined that the hydrogen generation from zirconium / water reactions remains less than the hypothetical 1 percent assumed. The current hydrogen generation from radiolysis is based on a core power of 3636 MWt, which is 102 percent of the rerated power level. For corrosion sources, the hydrogen generation was determined based on corrosion rates corresponding ti the temperature profiles calculated for the containment under post-LOCA conditions. Considering the various sources, the licensee's analysis determined that the current hydrogen generation analysis remains bounding for the rerate conditions. Therefore, the staff concludes that the hydrogen control systems and related hydrogen generation analysis are not affected by the rerate conditions.
2.5 Plant Structural Analyses The licensee has performed evaluations of the effects of the proposed rerate conditions upon the structural integrity of the NSS5 sd B0P pressure boundary systems, including the system piping, components, related supports, the reactor vessel and internals, steam generators, control rod drive mechanisms, reactor coolant pumps, and pressurizer.
Reactor Vessel and Internals The licensee assessed the adequacy of the reactor vessel by determining the stress and fatigue usage effects resulting from operation at the rerated conditions throughout the period of the current operating license. The calculated stresses and fatigue usage factors for the reactor vessel components were found to be within the allowable design limits.
The licensee assessed the adequacy of the reactor internal components for the rerate conditions. The assessment included ana'.es foi a LOCA, flow induced vibration, seismic, thermal transients, and s:: : and fatigue analysis for reactor internal components. The licensee stateu that the structural adequacy of the reactor internals is not affected by tpration at the rerate
s conditions. The staff has reviewed the licensee's assessmant and concludes that the integrity of the reactor vessel and internal components is not affected by the rerate conditions.
The licensee has also addressed reactor vessel structural integrity with respect to fracture touqhness requirements for protection against pressurized i
thermal shock.
Recent licensee submittals haae provided the analytical results of the most recently withdrawn surveillance capsule; response to Generic Letter 92-01, " Reactor Vessel Structural Integrity," ano proposed changes to the reactor coolant pressure and temperature limits. The licensee assumed the rerated conditions for future operation in the determination of the revised pressure and temperature limits. The staff has determined that no immediate concerns related to reactor vessel embrittlement and compliance with regulations are introduced by the proposed rerate at Wolf Creek and will address the issue more thoroughly during reviews of the above submittals.
Control Rod Drive Mechanisms The licensee evaluated the adequacy of the Control Rod Drive Mechanisms (CRDMs) by comparing the design bases input parameters with the operating conditions for the proposed rerate. The licensee stated that the rerate conditions would have an insignificant impact on the original design bases analyses for the CRDMs. The staff has reviewed the licensee's evaluation and i
concurs with the licensee's conclusion that the current design of the CRDMs would not be impacted by the rerate.
The licensee performed analyses of Wolf Creek's steam generators fo operation at the proposed rerate conditinns. The evaluations were performed according to the requirements of the ASME Code,Section III,1971 to 1973 Editions.
The calculated stress intensities were found to be within the allowable limits for all locations. The fatigue usage factors were also found to be-acceptable. However, the increased cumulative usage in the secondary manway bolts shortened their fatigue life from 20 years to 18 years for the planned rerate conditions and to 15.8 years for the maximum temperature is uction d
rerate conditions.
The Westinghouse Model F steam generators at Wolf Creek have thermally treated Inconel-600 tubes with an outside diameter of 0.688 inches and a wall thickness of 0.040 inches. As a result of the proposed changes in operating conditions, several design parameters of interest to steam generato-tube structural performance would change. These include reactor coolant temperatures, the differential pressure across the steam generator tubes, and the flow rates through the steam generators. The licensee has analyzed these design parameter changes associated with the rerate for the effect on the structural integrity of the steam generator tubing. The evaluations performed included cases bounding the potential range of rerate conditions. The licensee performed evaluations of the effect of the rerate on the minimum required steam generator tube wall thickness, the number of steam generator tubes susceptible to anti-vibration bar (AVB) wear, and the propensity of the steam generator tubing to various forms of corrosion degradation.
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The licensee determined the minimum acceptable steam generator tube wall thickness for the rerated conditions to be 0.016 inches (40 percent of the nominal wall thickness) using the criteria of Regulatory Guide 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes." The present minimum acceptable tube wall thickness for Wolf Creek is 0.014 inches (35 percent of the nominal wall thickness). This increase in minimum steam generator tube wall thickness is due to the increased differential pressure across the tubes for the rerate conditions. The technical specification plugging limit is an imperfection depth of 40 percent (minimum undegraded wall thickness of 60 percent) and does not need to be amended for the rerate conditions.
The licensee also performed an analysis of the number of additional steam generator tubes that may be affected by wear at the AVB supports as a result -
of operation at the rerate conditions. This analysis determined that the number of tubes which may require plugging due to AVB wear could increase slightly as a result of operation at the rerate conditions.
The estimated increase in AVB wear for the rerated conditions was determined using a three step approach.
First, the change in stability of the tubes with respect to fluid elastic vibration was determined. A probabalistic approach was then used to estimate the proportion of tubes which would become unstable as a result of possible fluid conditions at the AVB supports.
Finally, the increase in the number of tubes subject to wear was estimated using field data for the distribution of wear as a function of time.
Since the analysis determined that there was a slight increase in the potential for tube wear, the licensee plans to perform increased inspections.
The licensee committed to follow Westinghouse recommendations for inspection of the steam generators following the implementation of the plant rerate. The surveillances will consist of inspecting all steam generator tubes in Rows 25 or greater during the three refueling outages following the rerate. A significant fraction of these inspections will be completed during the first or second refueling outage following the rerate. The purpose of these inspections is to assess wear rates under the rerate conditions. The staff notes that the growth rates are used in the calculation of the technical specification tube plugging limit. The licensee has also committed to notify the NRC of any significant increase in the historically observed steam generator tube wear rates at Wolf Creek.
The licensee performed an evaluation of the effect of the rerating on the corrosion propensity for the steam generators. The analysis technique did not calculate absolute corrosion rates associated with the various corrosion mechanisms but did calculate the relative rate changes resulting from operation at the rerate conditions. The Wolf Creek steam generators were determined to be most susceptible to pitting and ODSCC and least susceptible to denting corrosion.
For the proposed rerate conditions, the impacts on the various corrosion mechanisms were determined to be that pitting and ODSCC propensity remained basically unchanged, PWSCC is reduced, and the propensity for denting is increased.
It should be noted that factors such as chemistry which are not directly affected by the rerate continue to significantly influence the corrosion rates of steam generator tubes.
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. i Based upon the information presented by the licensee, the staff finds that the Wolf Creek steam generators are acceptable for operation under the conditions related to the rerate.
Reactor Coolant Pumos The licensee has evaluated the impact of the conditions resulting from the proposed rerate on the design analyses of the reactor coolant pumps. The licensee determined that the proposed changes in operating temperatures have an insignificant effect on the thermal analysis performed on the reactor coolant pump design.
Based on the review of the rerated transients, the licensee confirmed that the original design analyses remain bounding. The reactor coolant pump motors were evaluated and the increased loads associated with the reduced coolant temperatures were considered to be acceptable. The NRC staff has reviewed the licensee's assessment and finds that the reactor coolant pumps are acceptable for operation at the rerated conditions.
Pressurizer i
The licensee has evaluated the pressurizer equipment specification and stress I
report relative to the proposed rerate conditions and revised NSSS transients.
ThE evaluation determined that the generic transients used in the original anaiyses remain bounding. The evaluation included consideration of changes in hot leg and pressurizer spray temperatures. Although these parameters change as a result of the rerate, the temperature differentials and related thermal stresses are enveloped by the original design analyses. The licensee's evaluations determined that the pressurizer control features such as safety valves, relief valves, spray valves, and heaters remained adequate for operation at the rerated conditions. Based on its review, the staff concludes that the pressurizer remains acceptable for operation at the rerated conditions.
NSSS Pipino. Components. and SuDDorts I
The licensee evaluated the following components and supports for the rerated operating conditions:
the RCS piping and supports, the primary equipment nozzles, the primary equipment supports, and the Class 1 auxiliary (branch) lines connected to the RCS piping. The evaluation compared the existing design bases with the performance requirements at the rerated conditions with respect to the design system parameters, transients, LOCA forcing functions, and the dynamic LOCA reactor vessel movements used in the original structural analyses. The licensee's adoption of leak-before-break design criteria ensures that the original design analyses, based on large guillotine breaks, remain bounding for the proposed rerate conditions. The gap conditions of the primary equipment support system have been assumed to change by a negligible amount and therefore the seismic loadings on the equipment would not be significantly affected by the lower operating temperatures. The licensee also i
factored the analyses and evaluations related to the " noise events" into the rerate program. The licensee's evaluations also addressed piping and equipment fatigue and surge line thermal stratification issues.
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The licensee also evaidated the auxiliary NSSS components. The evaluation was performed by comparing the original design bases and qualification require-ments with those for the rerated conditions.
The licensee found that, for each component evaluated, the original design enveloped those for the rerate.
Based on its review, the staff agrees with the licensee's conclusion that the existing NSSS piping and supports, primary equipment nozzles, primary equipment supports, branch lines, and auxiliary NSSS components remain in compliance with the design bases criteria given in the USAR with respect to the rerate operating conditions. These components and supports are, therefore, acceptable for operation as proposed.
B0P Pioina. Components. and Supports The licensee evaluated the adequacy of the B0P piping systems by comparing the existing design basis conditions with those for the proposed rerated conditions.
From this review, the licensee determined that most of the original design analyses bounded the conditions associated with the rerate.
l For those cases in which an increase in forces was calculated, the licensee confirmed that piping stresses and support loads remained within acceptable limits.
The licensee also reviewed the design bases pipe break analyses to evaluate the effects of the rerate conditions upon pipe break locations, jet thrust and i
impingement forces, and the design of pipe whip restraints.
In all cases, the licensee determined that original analyses bases remained conservative with respect to the rerate operating conditions.
Based on the review of the information provided by the licensee, the staff finds that the changes associated with the rerate conditions would either remain bounded by the original design analyses or have been evaluated and the related effects have been found to be acceptable.
2.6 Miscellaneous Systems and Proarams i
Environmental Oualification The licensee evaluated the effects of the proposed changes on qualified equipment. The radiological dose related to equipment qualification were evaluated by the licensee and determined to be acceptable. Conservatisms in the original equipment specifications allowed for a source term assumption of 50 percent of the total available Cesium immediately following a loss-of-coolant accident. For those cases in which the increased doses associated with the rerate were above the original analyses, the licensee confirmed that a revision to the design basis to a 1 percent Cesium source term would ensure equipment qualification. The calculated environmental temperature and pressure profiles for the limiting LOCA and MSLB cases based on the proposed rerate conditions are enveloped by the current Wolf Creek equipment qualification analyses for equipment located inside containment. The licensee's evaluation of mass and energy releases in the MSIV area, including blowdown of superheated steam under conditions associated with the rerate,
's determined that the maximum calculated temperatures are bounded by the existing design basis analyses.
Since the equipment qualification parameters affected by the proposed changes remain bounded by the values determined by the licensee's current analyses, the staff concludes that the effect of the proposed rerate on equipment qualification is minimal'and therefore operation of the plant at the rerated-conditions is acceptable.
flectrical Distribution System The staff has reviewed the information provided by the licensee regarding the main generator, transformers, onsite and offsite electrical distribution systems, and miscellaneous power systems.
The electrical system changes described included only modifications to high voltage transformers to handle the increased electrical output. The staff would not expect any significant system load changes as a result of the proposed rerate conditions.
Accordingly, the staff has concluded that the licensee's discussions in these areas are acceptable.
Radioactive Waste Systems The Wolf Creek radwaste systems were originally evaluated and accepted by the NRC staff based upon an assumed core power level of 3565 MWt. This power level was used to determine the source term for gaseous and liquid effluents and the waste volume. Therefore, operation at the rerate conditions would not change the analyses results or the staff's conclusions in the Wolf Creek SER (NUREG-0881). Accordingly, the radwaste systems continue to be acceptable for control of radioactive wastes for operation as proposed.
Heatino. Ventilation. and Air Conditionino (HVAC) Systems The control room emergency ventilation system is equipped with isolation, pressurization and filtering components to limit the dose to control room personnel following a design basis act.ident and to maintain a habitable l
environment in the control room. The original design basis analysis was based upon a source term for a core power level of 3565 MWt which is the same as the proposed rerate conditions. Therefore, the staff's conclusion in the Wolf Creek SER (NUREG-0881) remain valid for the proposed changes.
The licensee has also addressed other requirements on HVAC systems including the required heat removal for the support of engineered safety features equipment. The proposed rerate conditions were found to have negligible impact on the requirements for or performance of HVAC systems. The staff finds the licensee's conclusions acceptable.
Internal Floodino The licensee evaluated the potential impact of the rerate conditions on the threat of internal flooding. The evaluation of piping systems considered the changes in operating conditions including pressures and flow rates.
Based
s s-upon a review of the pipe break outflows for primary and secondary systems, the evaluation determined that the results from the existing flooding analyses bound the analyses for the rerate conditions.
Therefore, the staff concludes that the present flooding analyses continue to be valid and that flood levels would not increase as a result of operation at the rerate conditions.
2.7 Proposed License and Technical Specification Chances In_ order to allow the operation of Wolf Creek Generating Station at the proposed rerate conditions, the licensee proposed several changes to the Facility Operating License and associated Technical Specifications. The proposed changes consist of:
1.
A change in the Facility Operating License maximum power level from "3411 megawatts thermal (100% power)" to "3565 megawatts thermal (100% power)".
Changes to the uncertainty allowances and nominal T, Trip function 2.
associated with the overpower and overtemperature delta-temperature setpoints in Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints.
3.
Changes the maximum indicated reactor coolant system T,, from 592.5'f to 585.0*F in Table 3.2-1, DNB Parameters.
The staff's review of the proposed changes to the Operating License and Technical Specifications determined that the changes are consistent with the design analyses discussed previously. The staff finds the proposed changes to be acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Kansas State official was notified of the proposed issuance of the amendment. The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact was published in the Federal Reaister on October 25, 1993 (58 FR 55086)
In this finding, the Commission determined that issuance of this amendment would not have a significant effect on the quality of the human environment.
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5.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such i
activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: W. Reckley I
Date:
November 10, 1993 i
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