ML20237C547
| ML20237C547 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 08/20/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20237C535 | List: |
| References | |
| NUDOCS 9808210262 | |
| Download: ML20237C547 (22) | |
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t UNTED STATES g
j NUCLEAR REGULATORY COMMISSION o,
WASHINGTON, D.c. 30seH001 k.....,d SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.137 TO FACILITY OPERATING LICENSE NO. NPF-2 AND AMENDMENT NO.129 TO FACILITY OPERATING LICENSE NO. NPF-8 SOUTHERN NUCLEAR OPERATING COMPANY. INC.. ET AL.
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_ JOSEPH M. FARLEY NUCLEAR PLANT. UNITS 1 AND 2 j
DOCKET NOS. 50-348 AND 50-364 1
1.0 INTRODUCTION
l By letter dated February 14,1997, as supplemented by letters dated June 20, August 5, September 22, November 19, December 9, December 17, and December 31,1997, j
January 23, February 12, February 26, March 3, March 6, March 16, April 3, April 13, and two letters on April 17,1998, the Southern Nuclear Operating Company, Inc. (SNC) et al.,
submitted a request for changes to the Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2, Technical Specifications (TS) to increase the maximum reactor core power level for facility operation from 2652 megawatts-thermal (MWt) to 2775 MWt. The amendments also approve changes to the TS to implement uprated power operation. The results of the uprate evaluations and analyses were documented in Westinghouse WCAP-14723, "Farley Nuclear Plant Units 1 and 2 Power Uprate Project NSSS Licensing Report," dated January 1997 and the "Farley Nuclear Plant Units 1 and 2 Power Uprate BOP (Balance of Plant] Licensing Report" submitted by SNC with the February 14,1997, request.
The November 19, December 9, December 17, and December 31,1997, January 23, February 12, February 26, March 3, March 6, March 16, April 3, April 13, and two letters on April 17,1998, provided clarifying information that did not change the February 14,1997, application and the initial proposed no significant hazards consideration determination (October 8,1997,62 FR 52588).
2.0 BACKGROUND
FNP Units 1 and 2 are currently licensed for operation at a reactor core power level of 2652 MWt. SNC undertook a program to uprate the FNP units to a maximum reactor core power level of 2775 MWt, approximately a 4.6 percent increase. At the core uprate power, the generator electrical output for each unit will increase approximately 25 megawatts-electrical (MWe). The engineering studies supporting the core uprate have been performed in accordance with guidance contained in Westinghouse Topical Report WCAP-10263, entitled "A Review Plan for Uprating the Licensed Power of a Pressurized Water Reactor Power Plant,"
dated January 1983. This WCAP methodology, although not formally reviewed and approved by the NRC, was followed by North Anna Salem, Indian Point 2, Callaway, Vogtle, and Turkey Point for their core power uprate and these uprates were found acceptable.
9808210262 980820 PDR ADOCK 05000348 P
. SNC's letter of February 14,1997, submitted Westinghouse WCAP-14723, "Farley Nucicar Plant Units 1 and 2 Power Uprate Project NSSS Licensing Report," dated January 1997, which provided supporting documentation and analyses for the proposed changes. SNC stated that the results of the Nuclear Steam Supply System (NSSS) analyses and evaluations demonstrate that the FNP Units 1 and 2 NSSS is in compliance with applicable licensing criteria and requirements and can operate acceptably at the power uprate conditions.
SNC addressed the overall risk associated with the increase in rated thermal power and concluded that there is no impact on the calculated core damage frequency. The conclusion was based on the fact that all of the Final Safety Analysis Report (FSAR) transients were analyzed or evaluated at the uprated conditions with acceptable results and because SNC evaluated the success criteria and operator action failure probabilities used in the current FNP probabilistic risk assessment (PRA) model with no adverse impact.
The FNP power uprate amendments were reviewed with consideration given to the recommendations from the Report of the Maine Yankee Lessons Learned Task Group, dated December 5,1996. This report is documented in SECY-97-042, " Response to OlG Event J
Inquiry 96-04S Regarding Maine Yankee," dated February 18,1997. The Task Group l
concluded that a power uprate review procedure should be developed in light of the Maine Yankee findings. Although a Maine Yankee lessons learned power uprate procedure has not j
been developed, the recommendations of the report were considered in the review of the FNP
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uprate. The main findings centered around the use and applicability of the code methodologies j
used to support the uprated power. SNC has made an effort to verify that the code inputs and i
assumptions are appropriate and applicable to the plant given the uprated conditions and any
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changes (plant modifications and procedural changes) that have occurred since initial licensing.
SNC indicated that all principal codes were used in accordance with the applicable limitations and restrictions. SNC has also provided a summary list of analysis assumptions and of codes used and their review status (i.e., generically approved and new for FNP, generically approved
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and approved for FNP) for consideration in the NRC review. The staff considered all of the l
Maine Yankee Lessons Leamed recommendations and appropriately addressed them in this review. For the few recommendations that were not adopted, the staff provided adequate justification and obtained cognizant management approval.
3.0 ACCIDENT ANALYSES EVALUATION in support of this power uprate, the FNP units were reevaluated by SNC for operation at a rated thermal power of 2775 MWt with respect to safety analyses.
l The transient analyses presented rely heavily on analysis performed in the past to support other NRC-approved licensing actions (overpower delta temperature (OPAT) - overtemperature delta l
l temperature (OTAT) review, VANTAGE-5 fuel conversion, and steam generator level tap i
relocation). SNC stated that these analyses were also used in accordance with the applicable l
code limitations and restrictions. SNC stated that the principal codes and methodologies used are all part of the FNP design basis with the exception of the methodology used to evaluate the
-large break loss-of-coolant accident (LOCA). SNC is using the NRC-approved WCAP-12945-P-I A, " Code Qualification Document for Best Estimate LOCA Analysis," for the analysis of this accident. This methodology received a rigorous NRC review and approval. Sensitivity studies and comparative analysis were presented to the staff as part of the review effort. As a result, i
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1 the staff concludes that a comparative analyses, between the old and new power levels, is not necessary, as recommended in the Report of the Maine Yankee Lessons Learned Task Group.
3.1 LOCA Analvses 3.1.1 Laroe Break LOCA (LBLOCA)
The LBLOCA ana6jsis war. performed, at the uprated conditions, using the NRC-approved Westinghouse Best Estimate Methodology or WCOBRA/ TRAC. This methodology is appropriate for use at FNP because Units 1 and 2 are 3 loop Westinghouse designs for which the topical was apcroved. SNC performed the analysis in accordance with the code limitations and restrictions and as a result, the methodology is acceptabie for use in FNP licensing applications, including reference in the technical specifications and core operating limits report (COLR). Use of this methodology is for the time from event initiation to core quench and not for use in evaluating long-term cooling.
The results of the best estimate analysis indicate that the calculated peak clad temperature (PCT) is 2064*F, the maximum localized oxidation is 12 percent, the maximum hydrogen generation is 0.6 percent, the core remains coolable, and the core remains cool in the long term. These results are acceptable and meet the acceptance criteria of 10 CFR 50.46 of a PCT less than 2200*F, a maximum localized oxidation less than 17 percent, the maximum hydrogen generation less than 1.0 percent, the core remains coolable, and the core remains coolin the long term. As a result, the staff finds this acceptable.
3.1.2 Small Break LOCA (SBLOCA)
SNC stated that the NRC-approved NOTRUMP continues to be used for the SBLOCA analysis, at the uprated conditions. In order to determine the limiting set of conditions, a number of cases were evaluated. The analysis was run at both the high and low average temperatures limits, the Units 1 and 2 flow characteristics (Unit 1 has a barrel / baffle upflow configuration while Unit 2 has a downflow configuration), and for both ZlRLO and zircaloy clad fuel. SNC stated that the analysis assumes the limiting single failure of a loss of one train of the emergency core cooling system (ECCS) with an assumed loss of offsite power (LOOP) at the time of the reactor trip. SNC stated that the delay from the time of event initiation to the time of the trip in the mode'ing is inconsequential and is not less limiting than assuming the LOOP occurs at the time of the event. The results indicate that the limiting PCT is 1968* F for the Unit 1, low Tm, ZlRLO clad 3-inch break. The highest localized oxication is 5.84 percent, the total oxidation remains less than 1 percent, the core remains coolable, and long-term cooling is maintained. As a result, the analysis of the SBLOCA is acceptable.
3.1.3 Hot Lea Switchover SNC stated that a calculation has been performed to determine the new hot leg switchover (HLSO) time and minimum hot leg recirculation flow based on an uprated core power of 2775 MWt. The new HLSO time is 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The new hot leg recirculation minimum flow for the worst break and single failure is 89.1 lbm/sec. This hot leg recirculation minimum flow has
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been shown to be available. SNC concluded that with the above HLSO time and flow rate, the core geometry will remain coolable.
3.1.4 Post-LOCA Lona-Term Coolina SNC stated that an analysis of long-term cooling was performed at the uprated power. SNC has performed an evaluation to determine the effects of power uprating to post-LOCA long-term cooling, it is concluded that the new average temperature range has a very small effect on the post-LOCA sump boron concentration. Therefore, the core will remain subcritical post-LOCA and that decay heat can be removed for the extended period of time required by the long lived radioactivity remaining. The revised post-LOCA long-term coa cooling boron limit curve is used to qualify the fuel on a cycle-by-cycle basis during the fuel reload process.
3.2 Non-LOCA Transient Analvses SNC stated that the non-LOCA analysis was performed using codes that have been approved both generically and for FNP. They were used in accordance with applicable limitations and restrictions. A number of other NRC-approved license amendments were referenced because the previously approved analysis assumed uprated conditions. These include the transition to VANTAGE-5 fuel, the revision of the OPAT and OTAT setpoints with the implementation of the relaxed axial offset control, and the steam generator level tap relocation license amendments.
SNC stated that these analyses were also performsd in accordance with the applict ole code limitations and restrictions. The analysis associated with the revision of the OPAT and OTAT setpoints included all of the changes associated with this amendment. The analysis associated with the VANTAGE-5 fuel transition included the uprated power, the ZlRLO fuel, and the reduced reactor coolant system (P.CS) flow; however, it did not include all other conforming uprated related changes, including the reduction in ECCS flow. The transients that continue to rely on these analyses have been evaluated to verify continued acceptability. The staff finds this acceptable.
1 SNC stated that where applicable, the NRC-approved revised thermal design procedure (RTDP) was used in the evaluation of the departure from nucleate boiling (DNB). A reduced cora flow of 86,000 gpm per loop was employed in the analysis, which is associated with a minimum measured flow of 87,800 gpm ano 20 percent steam generator (SG) tube plugging was assumed. SNC stated that for each transient, both a stuck rod and the limiting single failure were assumed. The limiting single failure for each transient was selected on an event-specific basis. For the SG tube rupture, however, the single failure of the steam driven auxiliary feedwater pump was chosen, even though this is not the most limiting single failure. SNC stated that its design basis does not require the consideration of a single failure in the analysis of the SG tube rupture and the original analysis did not assume a single failure. The consideration of this single failure is more conservative than was required by the original design basis. ' As a result, the consideration of the failure of the steam driven auxiliary feedwater pump, a less limiting single failure, is now acceptable. Additionally, SNC stated that where an LOOP was assumed, the timing was chosen consistently with the original design basis assumptions.
These assumptions are acceptable. A brief discussion of each individual transient is presented below.
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. 3.2.1 Uncontrolled Rod Cluster Control Assembly (RCCA) Bank Withdrawal from a Suberitical Condition The uncontrolled RCCA bank withdrawal from a subcritical condition is analyzed to ensure that the core and the RCS are not adversely affected. This analysis was performed with acceptable results at the uprated conditions ' ring the VANTAGE-5 fuel conversion. This has been demonstrated because the results of the analysis show that the minimum DNB ratio (DNBR) remains greater than the safety analysis finat and that the maximum fuel temperatures predicted to occur are much less than those required for clad damage or fuel melting to occur.
The effect of the power uprate on this event is therefore, acceptable.
3.2.2 Uncontrolled Rod Cluster Control Assembiv Bank Withdrawal at Power The analysis of the uncontrolled RCCA bank withdrawal at power case was analyzed while
. revising the OPAT and OTAT setpoints, at the uprated conditions, with acceptable results and l
NRC approval as addressed in Farley License Amendment No.121 for Unit No.1 and License j
Amendment No.113 for Unit No. 2, issued on September 3,1996.
3.2.3 Rod Cluster Control Assembiv Misationment The dropped RCCA and the statically misaligned assembly were analyzed at the uprated l
conditions while transitioning to VANTAGE-5 fuel. The analysis resulted in the calculated DNB meeting the design basis. Although the rod control system has been modified, affecting the control system response, SNC performed an analysis to demonstrate the acceptability of the rod control parameters since this modification. These calculations confirmed that the DNB design basis continues to be met. As a result, the staff finds this acceptable.
3.2.4 Uncontrolled Boron Dilution The uncontrolled boron dilution was analyzed while revising the OPAT and OTAT setpoints, at l
the uprated conditions, with acceptable results and NRC approval as addressed in Farley License Amendment No.121 for Unit No.1 and License Amendment No.113 for Unit No. 2, issued on September 3,1996.
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3.2.5 Partial Loss of Forced Reactor Coolant Svstem Flow
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The transient was analyzed and evaluated at uprated conditions. The event was analyzed l
L using the RTDP and concluded that the DNBR design basis is met, the pressure of the primary and secondary remain within the design limits, and fuel centerline melt is not predicted. As a result, the staff finds this acceptable.
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3.2.6 Startuo of an inactive Reactor Coolant Looo The FNP TSs do not allow operation with a loop out of service, therefore this is not analyzed and it is acceptable.
. 3.2.11 Accidental Deoressurization of RCS The accidental depressurization of RCS was analyzed while revising the OPAT and OTAT 1
setpoints, at the uprated conditions, with acceptable results and NRC approval as addressed in l
Farley License Amendment No.121 for Unit No.1 and License Amendment No.113 for Unit No. 2, issued on September 10,1996.
3.2.12 Inadvertent Ooeration of Emeraency Core Coolina System Durina Ppwer Ooeration The inadvertent ECCS is analyzed to assure that the primary and secondary pressure limits are not exceeded, that the DNBR limits are not exceeded, and that the event does not progress to i
a more severe event. The event was analyzed and meets this criteria; however, to prevent the water solid relief from the primary code safety relief valves and the potential for the valve sticking open creating a more severe transient, SNC credits the opening of the PORVs. l'he l
staff has reviewed this and determined that this is acceptable because there is sufficient time for the operators to take action and open the PORV block valve if it is closed. Additionally, I
although the automatic actuation of the PORVis not considered safety-related, the actuation f
circuits are routed separately; there are two separate transmitters which are the same as those
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used in Class 1E applications, powered from 1E power supplies, with 1E relays. Therefore, the t
PORV is considered highly reliable and its use is acceptable.
3.2.13 Inadvertent Loadina of a Fuel Assembiv into an Imorocer Position SNC performed an analysis to verify that operation at the uprated power does not affect the ability of the instrumentation to detect the incorrect loading of an assembly. The evaluation concluded that the current analysis is still applicable and the conclusions remain valid, if a pin or rod were to be incorrectly loaded, the damage would be limited to that pin, if an assembly were incorrectly loaded, it would be detected by the incore detectors. The staff finds this acceptable.
3.2.14 Comolete Loss of Forced Reactor Coolant Flow
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The complete loss of flow was analped for the power uprate using the RTDP. For the limiting case run, the reactor is assumed to trip on the loss of RCS flow. The results indicate that the DNBR does not decrease below the limit, fuel centerline melt is not predicted, and the primary and secondary pressure also remain below the limit. As a result, the acceptance criteria are met.
3.2.15 Sinale Rod Cluster Control Assembiv Withdrawal at Full Power I
The single rod withdraw transient was analyzed during the VANTAGE-5 fuel conversion with NRC approval as addressed in Farley License Amendment No. 91 for Unit No.1 and License i
Amendment No. 84 for Unit No. 2, issued on December 30,1991. An evaluation was I
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- generation post-LOCA due to radiolytic decomposition of core and sump solution willincrease by approximately 5 percent proportional to the increase in reactor power level.
SNC indicated that the increase in hydrogen generation rate due to power uprate is determined to have a negligible effect on the post-accident hydrogen mixing system. The analysis performed for uprated power showed that with no recombiner in service, the hydrogen I
concentration wiil not exceed 4 percent by volume for 17 days following a LOCA. Piacing a hydrogen recombiner in service prior to the 18th day following a LOCA will maintain containment hydrogen levels below the lower flammability limit of 4 percent. Based on the above review, the staff finds that the power uprate will not impact the post-LOCA hydrogen control system.
3.4.2 Comoliance with 10 CFR Part 50. Aooendix R SNC stated that the power uprate evaluation did not identify changes to design or operating conditions that adversely impact the post-fire safe shutdown capability in accordance with Appendix R. SNC also stated that there were no physical plant configuration changes or combustible load changes. The staff finds this acceptable. The staff may review, during a future onsite inspection, the supporting information for SNC's letter of March 6,1998.
3.4.3 Station Blackout (SBO)
SNC performed evaluations of the impact resulting from plant operationr at the proposed uprated power level on system response and coping capabilities for SBO events. SNC stated that current design basis temperature profiles in areas housing SBO-required equipment remain-bounded for an SBO in an uprated plant.
With this evaluation, the staff has also reviewed whether this power uprate would affect current General Design Criterion (GDC)-17 and SBO requirements. Although the power uprate resulted in a small electrical load increase of the reactor coolant pumps and condensate pumps l
on non-Class 1E 4160 V buses, SNC's assessment has not indicated that pcwer uprate would affect bus loadings and voltages to such a degree that the electrical onsite ar offsite power system configuration would need to be modified. Therefore, the staff finds that the power uprate continues to satisfy the current GDC-17 requirement. SNC has also reviewed the SBO coping analysis which is a function of offsite power design, emergency power configuration, and emergency diesel generator target reliability per Regulatory Guide 1.155, " Station j
Blackout." SNC finds that none of the SBO coping criteria are affected by power uprate. On j
this basis, the staff concludes that power uprate at FNP would not affect current SBO coping duration and would continue to meet the GDC-17 requirements.
J Based on its review and the experience gained from the review of power uprate applications for l
similar PWR plants, the staff finds that the impact on system response and coping capabilities I
for an SBO event resulting from plant operations at the proposed uprated power level is
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insignificant.
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. which exists, that an MSLB may induce leakage in the tubes that have been allowed to remain in service due to application of ARC. Consequently, when an analysis is performed of the consequences of an MSLB accident concurrent with a loss of offsite power, SNC assumes that the accident induces leakage in those tubes of the affected SG which has experienced the pressures associated with a steamline break and to which ARC has been applied.
l On October 29,1997, Amendments 132 and 124 were issued to FNP Units 1 and 2, l
respectively, which implemented the most recent ARC criteria. This amendment approved a l
l primary-to-secondary leak rate of 23.8 gpm and primary coolant activity levels of dose
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equivalent 8'l of 9 pCi/g for the maximum instantaneous value and 0.15 pCi/g for the 48-hour j
value. On March 16,1998, SNC submitted a revised MSLB analysis that utilized the l
assumptions of Amendments 132 and 124 but modified the assumptions to incorporate l
changes associated with the power uprate such as steam released from the intact SGs.
For the power uprate amendment, the reevaluation of the MSLB involved two cases. One case assumed the accident occurred following an iodine spike, referred to as the preexisting spike case. The second case assumed that the MSLB resulted in the initiation of an iodine spike, referred to as the accident-initiated spike. In both cases, a 150 gpd/SG, primary-to-secondary l-leak was assumed for the intact SGs. For case one, it was assumed that a preexisting iodine spike had occurred prior to the steamline break. Reactor coolant concentration was assumed to be at the TS Figure 3.4-1 full power limit of 9 pCi/g of dose equivalent l.
The second case assumed the steamline break initiated a concurrent iodine spike. The reactor coolant concentration was assumed to be at the existing TS 48-hour limit of 0.15 pCi/g dose equivalent '8'I. The secondary system activity was assumed to be at the TS limit of 0.1 pCi/g dose equivalent l. Concurrent with the MSLB, an iodine spike was assumed to occur that releases iodine from the fuel gap to the reactor coolant at a rate that is 500 times the normal iodine release rate. No failed fuel was assumed to occur as a result of the MSLB.
For both analyses, it was assumed that the 23.8 gpm of primary-to-secondary tube leak occurred in the faulted SG untilit was isolated. After isolation of the faulted SG, it was assumed that primary-to-secondary leakage to the intact SGs would continue at a rate of 150 gpd/SG. Because offsite power is assumed to be lost, the main condenser was unavailable for steam dump and cooling of the reactor core must occur through the use of the atmospheric relief valves. After 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, no further steam release or activity release was
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assumed to occur due to the steamline break.
The power uprato amendment request presented a new value for the quantity of steam released from the intact SGs during an MSLB. The quantity presented in the power uprate amendment was approximately 25 percent less than the value which was presented in support of the latest ARC amendments. This decrease in steaming rate for the intact SGs incorporated a plant-specific value for Farley whereas the previous value had been a generic Westinghouse value. A rigorous calculation of the dose consequences of an MSLB would include the contribution from the steaming of the intact SGs. However, the dose contribution from this source is minimal. The staff's analysis, which assessed the latest ARC amendment request, l
assumed, in the calculation of doses, that any primary-to-secondary leakage to the intact SGs l
was released to the environment without a partition factor. This is a conservative assumption.
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l l l zone exceeded the acceptance criterion of Appendix A to SRP 15.4.8. The staff finds this l
departure from the acceptance criterion acceptable because:
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The doses are below the guideline values of 10 CFR Part 100; i
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The consequences were calculated on the basis of fuel failures that are predicted by conservative models; and 3.
The SRP acceptance criterion itself is conservative.
I All other doses were found to meet the acceptance criterion of SRP 6.4 and SRP 15.4.8.
3.5.1.9 Small Break LOCA
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SNC calculated the potential consequences of an SBLOCA. In this assessment, SNC assumed l
the very same assumptions as those for an LBLOCA with two exceptions. The first exception I
was the source term that was assumed to consist of 100 percent failure of the fuel cladding that results in release of 100 percent of the gap activity. The second exception assumed that the sprays were not activated. The resultant doses were assumed to occur only from containment leakage with no inclusion of the potential consequences of either a mini-purge or H venting.
2 The staff performed a similar calculation. The staffs assessment assumed that the particulate form of iodine would undergo removal via plateout. The doses are presented in Tables 3.5.2-1 l
through 3.5.2-3. The doses were found acceptable.
3.5.2 Conclusions The staff has assessed those accidents for which the power uprate would have an impact upon the offsite and control room operator doses. These doses are presented in Tables 3.5.2-1 through 3.5.2-3. The staffs results demonstrated that, for those accidents that are impacted by the power uprate, the doses would not exceed the dose guidelines present'f contained in the Standard Review Plan,10 CFR Part 100, or GDC-19 of 10 CFR Part 50, Appendix A, for either offsite locations or control room operators. Therefore, the staff finds the proposed power uprate acceptable.
4.0 SYSTEMS. STRUCTURES. AND COMPONENTS EVALUATION In its evaluation, the staff refers to several terms in the submittal. These terms are defined below:
ASME Code - editions of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Sections lil or XI, as appropriate, through the 1988 Addenda and the 1989 Edition.
ASTM Standard Procedures or Practices - procedures or practices for testing or analysis developed by the American Society for Testing and Materials.
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Charov-V notch test or Charov-V test - a fracture toughness test that involves impacting a small notched impact specimen (usually in bar form) with an impact pendulum and measuring the fracture energy of the specimen.
Charov transition curve or Charov-V curve - a graphic presentation of Charpy-V test data, l
including absorbed energy, and fracture appearance. The curve includes the lower shelf energy (< 5 percent shear), the transition region, and the upper shelf energy
(> 95 percent shear).
f EQL "end of life,' the scheduled date of expiration of the operating license for a licensed l
nuclear power generation facil"y.
r EOL fluence - the best-estimate neutron fluence projected for a specific reactor pressure vessel (RPV) beltline material a the clad-base-metal interface on the inside surface of the RPV at the location where ths material receives the highest fluence on the expiration date of the operating license.
Pressurized thermal shock (PTS) event - an event or transient in pressurized water
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reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed I
by significant pressure in the reactor vessel.
Reactor oressure vessel (RPV) beltline - the region of the RPV (shell material including welds, heat-affected zone, and plates or forging materials) that directly surrounds the effective height of the active core and adjacent regions of the RPV that are predicted to experience significant neutron irradiation damage (a consideration in selecting materials).
RIpis - the reference temperature, RTuo7, evaluated for the EOL fluence for each of the RPV beltline material, using the procedures of Paragraph (c) of the revised PTS rule, i
10 CFR 50.61 (Ref. 22). For materials in the RPV beltline, the RTers must account for l
changes in the reference temperature as a result of neutron irradiation damage.
l ARIsor - the transition temperature shift, or change in RTuor, due to neutron irradiation effects, which is evaluated as the difference in the 30 ft-lb (41 J) Index temperatures from the average Charpy-V curves measured before and after irradiation.
Uooer shelf enerav (USE) - the average value for all Charpy-V test specimens (normally three) whose test temperature is above the upper end of the transition region of the Charpy-V curve; for specimens tested in sets of three at each test temperature, the set having the highest average may be regarded as defining the USE of the material.
4.1 Reactor Vessel Intearity To determine the acceptability of the power uprate on the integrity of the reactor vessel, the staff evaluated the following:
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Appendix H. Therefore, the staff concludes that SNC's proposed changes to the material surveillance programs for FNP Units 1 and 2 are acceptable.
4.2 Reactor Vessel SNC reported that the power increase will result in changing the design parameters given in Table 2.1-1 of Reference 36. Table 2.1-2 provides various cases that were developed for use l
in the power uprate analysis.
SNC evaluated the reactor vessel considering the worst load sets of operating parameters and design transients and a new set of LOCA loads, which was identified for the uprated power condition. The regions of the reactor vessel affected by the power uprate include outlet and inlet nozzles, the RPV (main closure head flange, studs, and vessel shell), control rod drive mechanisms (CRDM) housing, bottom head to shell juncture, core support pads and the instrumentation tubes. SNC evaluated the maximum ranges of stresses and cumulative fatigue l
usage factors for the critical components at the core power uprated conditions. The evaluation was performed in accordance with the ASME Code, Section lil,1968 Edition, with addenda through the Summer 1970 to assure compliance with the code of record.
The calculated maximum stresses and the maximum cumulative fatigue usage factors (CUF) for the reactor vessel criticallocations are provided in Tab % A of Reference 6. The results i1dicate that the maximum stresses are within the allowable limits for FNP Units 1 and 2, and the CUFs remain below the allowable ASME Code limit of 1.0.
On the basis of its review, the staff concludes that the current design of the reactor vessel is in compliance with codes and standards under which the plant was licensed, for the upra5d power conditions.
4.3 Reactor Core Sucoort Structures and VesselInternals By letter dated August 5,1997, SNC provided additional information requested by the staff, with regard to the evaluation of the reactor vessel core support and internal structures. The limiting
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reactor intemal components evaluated include the lower core plate, core barrel, baffle plate, l
baffle / barrel region bolts, and the upper core plate.
SNC evaluated the upper and lower internals considering the worst case set of operating parameters provided in Table 2.1-2 of Reference 36. The evaluation was performed in accordance with the original design-basis criteria for the FNP reactor internals, which had been
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previously reviewed by the staff. For the lower core plate reanalysis, the evaluation was l
performed using Section Ill, Subsection NG of the ASME Code,1989 Edition with 1990 l
Addenda. The maximum calculated stresses and cumulative fatigue factors for the limiting j
internal components at the power uprate conditions are identified on page 45 of Reference 6.
l The maximum stresses for components were below the Code allowable limits, except the upper l
core plate and the baffle / barrel bolts. For the upper core plate, the maximum combined primary and secondary stress exceeded the 3Sm limit. A simpt, ad elastic-plastic analysis was performed in accordance with ASME NG-3228.3. The reanalysis indicated that the fatigue usage factor was 0.08, which is less than the limit of 1.0, and that the combined primary and
. secondary stress intensity, excluding thermal bending stresses, was less than 3Sm. No stress was given for the baffle / barrel bolts. SNC indicated that the baffle / barrel bolts were originally qualified by test to loads associated with the existing design basis conditions, and these bolts i
are acceptable because the existing design basis condition is bounding for the proposed power l
epdate condition. The maximum CUF was calculated to be 0.917 for the baffle / barrel bolts at l
- the proposed power uprate condition.
it is noted that baffle / barrel bolts degradation was reported in European PWR plants. The l
Westinghouse Owners Group (WOG) has identified a lead plant for inspection in the fall of 1998 for a possible baffle / barrel bolt cracking. SNC should assess the results of the lead plant's inspection for its applicability to the baffle / barrel bolts at FNP and update its evaluation and conclusion as necessary.
Further, SNC reviewed the potential for a flow-induced vibration on the guide tubes and the upper support column at the uprated power level. The evaluation indicated that the existing analysis provides sufficient margins to accommodate the increase (by approximately 1.9 percent) in the flow-induced vibration loads due to the power uprate.
On the basis of the above evaluation, the staff concludes that the reactor internal components at FNP Units 1 and 2 will remain within the allowable limits of stress and fatigue usage factor l
for operation at the proposed uprated power conditions.
4.4 Control Rod Drive Mechanisms SNC evaluated the adequacy of the CRDMs by reviewing the FNP current Model L106A CRDM l
design specifications and stress report to compare the design-basis input parameters against the operating conditions at the uprated core power. The components reviewed include the full length (F/L) L-106A CRDMs, the Capped Latch Housing Assembly and the Royal Industries l
Part Length (P/L) CRDMs employed at FNP. The maximum temperature for FNP power uprate is 613*F, which is below the design temperature of 650*F. The pressure remains the same at 2250 psia. On the basis of this evaluation, SNC concluded that the original design basis thermal and structural analyses are bounding for the core power uprate.
On the basis of its review, the staff concludes that the current design of CRDMs is in compliance with licensing basis codes and standards for the uprated power conditions.
4.5 Steam Generators SNC evaluated the SGs by comparing the power uprate conditions with the design parameters of the Westinghouse Model 51 SGs at FNP. The comparison shown in Table 2.1-2 of Reference 36 indicates that critical design system parameters such as the primary and secondary side pressures, as well as the vessel outlet and secondary side temperatures, are not significantly affected by the uprated power cor:ditions. The variation in the primary-to-secondary pressure differential is within about 8 percent. SNC indicated that there are no significant changes to the design transients as a resuK of the core power uprate. The evaluation was performed in accordance with the requirements of the ASME Code, Section 111, 1971 Edition.
l
. The maximum stresses and cumulative fatigue usage factors for the critical SG components are provided in Table B, page 53 of Reference 6. The results indicate that the maximum calculated stresses are below the Code allowable limits except for the divider plate. For the divider plate, SNC performed a plastic analysis considering the actual material stress-strain relation and the stress redistribution, in accordance with the NB-3228.1(b) of the 1971 ASME Code. The analysis indicated that the maximum strain and the calculated CUF are within the allowable range, and that the divider plate will remain in compliance with licensing basis codes and standards for the uprated power condition. The revised CUF values for other SG components excluding the U-bend tubes are less than the Code limit of unity except for the secondary manway bolts at which the CUF was calculated to be 1.18. SNC indicated that it plans to replace the secondary manway bolts prior to 34 years of service so as to remain within compliance with the fatigue limit requirement.
SNC assessed the potential for flow-induced vibration for the small radius U-bend tubes at the power uprate condition. The evaluation concluded that changes to vibration levels and cross flow velocity due to power uprate were within the acceptable range. The U-bend fatigue evaluation was updated using the FNP existing methodology in WCAP-11875 (Reference 32).
The results indicate that the fatigue usage of the U-bend may exceed the acceptance limit of l
1.0 in 13.7 years after the implementation of the power uprate at FNP Units 1 and 2 in 1998 outages. SNC concluded that the U-bend may need to be monitored and may require some l
type of corrective action at that time, as necessary. SNC noted in Section 5.7.3 of Reference 36 that following the implementation of power uprate, the outlet pressure and the steam flow for each steam generator can be documented on a cycle-specific basis for use in any future update of the U-bend fatigue evaluation.
On the basis of its review, the staff finds that the current FNP Units 1 and 2 SGs are acceptable for the proposed core power uprate up to 13.7 years following its implementation at both Units 1 and 2. It should be note _d that SNC currently is planning to replace the Unit 1 SGs l
during the spring 2000 refueling outage and Unit 2 SGs during the spring 2001 refueling outage. Upon replacement, the issue regarding exceeding the cumulative fatigue usage factor for the secondary manway bolts and U-bend tubes will no longer exist.
4.5.1 SG Tube Inteority 4.5.1.1 SG Tube Degradation To minimize the effect of the power uprate on tube degradation, SNC intends to maintain the same T% (the hot leg temperature) before and after power uprate, with a 0.5 'F allowance to account for measurement variations and uncertainties. Industry experience has shown that, in general, a high To correlates with increased tube degradation. Therefore, limiting Tm to the preuprate value should prevent the rate of tube degradation from increasing after the power uprate. However, the power uprate will slightly reduce the steam pressure and the corresponding saturation temperature, changes that may increase tube degradation.
SNC evaluated the effect of power uprate on tube degradation mechanisms, including primary water stress corrosion cracking (PWSCC), and outside diameter stress corrosion cracking
. (ODSCC), in various regions of tubing. The staff's assessment of the mechanisms is discussed below.
4.5.1.1.1 PWSCC of tubina SNC stated that the main contributors to PWSCC are residual, pressure, and thermal stresses.
SNC stated that the power uprate would increase only the through-wall pressure stress. This stress w'uld increase because the primary-to-secondary differential pressure would increase from about 1435 psi to 1463 psi. The resulting pressure stress will increase the PWSCC kinetics by 2 percent to 2.5 percent. SNC concluded that considering the uncertainties in estimating the various contributions to PWSCC, the increase in PWSCC kinetics is insignificant; therefore, the effect of the power uprate on PWSCC is insignificant. The staff finds that an increase of 2 percent to 2.5 percent in PWSCC kinetics is insignificant and, therefore, the impact of the power uprate on PWSCC of tubing is insignificant.
4.5.1.1.2 ODSCC of tubina SNC evaluated the regions where ODSCC has occurred, such as in the tube support plate crevices, the sludge pile at the top of the tubesheet, and the free span regions of the tubing.
SNC found that the beneficial effect of lowering secondary temperature tends to be stronger than the deleterious effect of increasing the applied stress in the tube support plate crevices and on the free span regions. Therefore, the power uprate will minimally affect the tube support plate crevices and free span regions. However, SNC predicted the expected ODSCC rate in the sludge pile to slightly increase because its analysis did not take credit for the lower secondary temperatures in the sludge pile. SNC inspects the parts of the tubes in the sludge pile during each refueling outage and would readily detect any increase in tube degradation there. Based on SNC's assessment, the staff finds that the effect of power uprate on ODSCC of tubing would be minimal.
4.5.1.1.3 Degradation of U-bends in small radius tubes bv stress corrosion crackina SNC stated that it stress-relieved the small radius U-bends to reduce their residual stresses.
The increased wall pressure stress and the reduced residual stress cancel each other out so that the impact of the power uprate in the small radius U-bends is negligible. To ensure stress corrosion cracking does not increase in the small radius U-bends, SNC plans to inspect all U-bends in rows 1 and 2 at the refueling outage after the power uprate for Unit 2. Based on SNC's assessment, the staff finds that the impact of the power uprate on stress corrosion cracking in the small radius U-bends is insignificant.
4.5.1.1.4 Tubes susceptible to wear from anti-vibration bars
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SNC stated that it replaced the anti-vibration bars on steam generator tubing in Units 1 and 2.
Since replacing them, SNC has not observed any tube wear typically caused by anti-vibration bars. SNC stated that the power uprate is not expected to cause wear from the replaced anti-(
vibration bars. Based on SNC's observation, the staff finds that the power uprate should not i
significantly affect the tube wear by anti-vibration bars.
I
1
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l differential and thermal stress that are experienced by the pressurizer. The RCS pressure is unchanged for the FNP power uprate. There is an insignificant change in the design transients regarding type and number cf occurrences during the 40 years of plant operation, for FNP
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power uprate as seen in a comparison table on page 57 of Reference 6. The minimum cold leg l
temperature decreased by 13'F, with respect to the original design conditions. The original fatigue analyses were updated to account for the uprated power conditions. The CUFs at the critical locations are provided in a table on page 58 of Reference 6. SNC indicated that the maximum calculated stresses at the entical locations are unchanged except for the surge line nozzle. The maximum Mress intensity at the surge line nozzle was recalculated and was found to be less than the allowable Code limit. The maximum CUFs at the limiting locations are 0.94 for the spray nozzle and 0.78 for the pressurizer upper head and shell.
l The performance u the pressurizer safety valves (PSVs) and the PORVs are dependent on the pressurizer operating pressure and temperature, which are unchanged for the power uprate at FNP. The transient ana'ysis for the power uprate was performed assuming a maximum tolerance of lift setting, as specified in Sections 3/4.4.2 and 3/4.4.3 of the FNP Technical Specifications. The PSVs and PORVs setpoints, rated capacities, and corresponding dynamic loads imposed on tb7 piping and supports due to valve operation do not change as a result of the proposed powt: uprate.
On the basis of the above review, the staff finds that the exist:ng pressurizer and components remain adequate for the plant operation at the proposed uprated core power.
4.8 Chemical and Volume Control System (CVCS)
The main role of thu CVCS la to maintain RCS water inventory, boron concentration, and water chemistry. To perform these functions, the CVCS must meet the following requirements: (1) the parts of the system that constitute the reactor coolant pressure boundary can withstand the expected RCS conditions, (2) boration meets the design requirements for reactivity control, and (3) with the exception of RCP seal injection line, the system can be automatically isolated during all events requiring its isolation. The proposed power uprate will not affect the CVCS isolation function, but it will have some effect on the integrity of the RCS pressure boundary and on the boration of the RCS. SNC therefore, evaluated the effect of the uprate on the performance of the CVCS.
4.8.1 RCS Temperature During power uprating, a change in reactor coolant temperature may affect the integrity of the primary coolant boundary during normal operation and during thermal transients. However, SNC's evaluation concluded that, after the power uprate, the design basis RCS cold leg temperature of 541.1*F will be well below the system design temperature of 650*F and below the maximum RCS inlet operating temperature of 547'F, established in the CVCS overall design. Furthermore, the regerierative, letdown, excess letdown, and seal water heat exchangers will also operate within acceptable limits, and the load on the component cooling
[
water system will not be excessive. The CVCS will, therefore, preserve its design functions.
SNC also analyzed the performance of the CVCS during conditions resulting frcm thermal i
. 4.10.2 Comoonent Coolina Water System (CCNS)
The CCWS is a closed loop system that serves as an intermediate barrier between the service water system and systems that contain radioactive or potentially radioactive fluids in order to eliminate the possibility of an uncontrolled release of radioactivity. It provides cooling water to various safety and nonsafety systems during all phases of normal plant operation, including startup through cold shutdown and refueling, as well as following an SBO event, LOCA, or MSLB accident. The CCWS heat loads resulting from plant operations at the proposed uprated power level willincrease slightly. SNC performed evaluations of the effects of these increases in heat loads on CCWS and stated that the additional heat loads result in minor temperature increases on the CCWS for normal and accident scenarios and that the CCWS has the capacity to a:commodate the additional heat loads and resultant temperature increases. No increase in CCWS flow rates is required to handle the additional heat loads.
Based on the staff's review and the experience gained from its review of power uprate applications for similar PWR plants, the staff finds that plant operations at the proposed uprated power level do not change the design aspects and operations of the CCWS and have an insignificant or no impact on the CCWS. Therefore, the staff concludes that the FNP CCWS is acceptable for operations at the proposed uprated power level.
4.10.3 Soent Fuel Pool Coolinc System (SFPCS)
The SFPCS is designed to remove the decay heat released from the spent fuel assemblies stored in the spent fuel pool (SFP); to maintain the SFP water temperature at or below the
)
maximum operating temperature limit of 150*F durir,j plant operations and refueling; and to j
maintain its cooling function during and after a seismic event. The SFPCS heat loads will J
increase slightly resulting from plant operations at the proposed power level.
I SNC performed SFP heatup analyses for plant operations at the proposed power level. For the bounding (emergency full core off-load)' case, the SFPCS hect load increases from 34.77 x 105 Btu /hr to 37.0 x 108 Blu/hr, and the corresponding calculated peak SFP temperature with one SFP cooling train in service increases from less than 170*F to 175'F.
l SNC stated that FNP operates the SFP cooling system with one SFP cooling train in service and the rema:ning train on standby. Plant administrative procedures," Outage Nuclear Safety" and " Outage Planning Manual" have provisions to ensure that the backup SFP cooling train is available prior to the outage. Plant procedures also require fuel handling operations be suspended upon receipt of the high SFP temperature alarm which is set at 130*F. Plant procedures provide the controls necessary to ensure that the maximum operating temperature limit of 150'F will not be exceeded. Nevertheless, SNC performed evaluations (with demineralized isolated and fuel handling operations prohibited) to demonstrate the acceptability 3 The maximum SFPCS heat load for the end-of-cycle full core off-load during routine refueling is 36.5 x 10' Blu/hr.
l The corresponding calculated peak SFP temperature is 174 F with one SFP cooling train in service.
__J
. of SFPCS operation, SFP liner and concrete struc'ure at temperatures below 180'F. SNC concluded that no changes to the SFP cooling or cleaning systems are required to support plant operations at the proposed power uprate level.
In the unlikely event that there is a complete loss of SFPCS cooling capability, the SFP water temperature will rise and eventually'will reach boiling temperature. SNC performed analysis to demonstrate the time to boil and the boil off rate based on the heat load for the emergency full core off-load scenario. The calculated mininium time from the pool high temperature (130*F) alarm caused by loss-of-pool cooling until the pool boils is 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and the maximum boil-off rate is 76.4 gpm. However, makeno water from the primary water storage tank can be inNeted within the required time utilizing either of two 165 gpm reactor water makeup pumpc. The primary water storage tank, reactor water makeup pumps, and piping are seismic Category 1.
In addition, makeup water can be supplied from the refueling water storage tank and the demineralized water system.
In addition, the amount of fission product release and the chemical and radionuclear composition of pool water will not change appreciably after the power uprate and the existing cleanup system will perform adequately. When one train is in operation, the temperature of bulk water may exceed 140*F, the upper limit for the domineralizer resin. However, the system has a provision to alarm the operators whenever water reaches 130*F, so that they can manually initiate appropriate actions to prevent damage to the cleanup system. The staff finds that with this provision, the existing SFP cleanup system will be adequate for maintaining purity of the SFP water after the power uprate.
Based on its review, the staff finds the SFPCS acceptable for plant operations at the proposed power uprate level.
i 4.11 Main Turbirg SNC performed evaluations on turbine operations with respect to design acceptance criteria to verify the mec;.anical integrity under the conditions imposed by plant operations at the proposed uprated power level. Results of the evaluations showed that there would be no increase in the probability of turbine overspeed. Therefore, the turbine could continue to be operated safely at the proposed upiated power levels.
Based the staffs review and the experience gained from its review of power uprate applications for similar PWR plants, the staff finds that operation of the turbine at the proposed uprated power levelis acceptable.
4.12 fjiah Enerav Line Break (HELB) Outside Containment System operating parameters for uprate were evaluated against the system pressure and/or temperature parameters used in the existing plant bases to demonstrate the acceptability for HELB effects. Core uprate will not change the bounding temperature and pressure used as the
. basis for pipe break analyses. SNC stated that design basis analyses remain bounding for all HELB events.
f Based on the staff's review and the experience gained from its review of power uprate applications for similar PWR plants, the staff concludes that plant operations at the proposed uprated power level have an insignificant or no impact on the consequences (e.g.,
environmental pressure and/or temperature parameters, etc.) resulting from HELB outside containment.
4.13 Safetv-Related Eauioment Qualification EQ)
The effects of all changes due to plant operations at the proposed uprated power level on the design of mechanical components as well as EQ of safety-related electrical equipment were
{
evaluated. The temperatures, pressures, and in some cases flows, in certain systems would be affected slightly by plant operations at the proposed uprated power level. However, these changes in temperatures, pressures, and flows are bounded by the original design of components. The existing parameters used for evaluating mechanical components inside and l
outside containment remain bounding for the conditions resulting from plant operations at the proposed power level.
As part of the safety-related electrical equipment qualification evaivation for the power uprate, SNC reviewed: (1) the radiological dose limit for safety-related electrical equipment located in a harsh environment (e.g., containment), (2) the composite temperature and pressure curves for safety-related electrical equipment qualification, and (3) components that were qualified based on calculated surface temperatures.
(a) Radiological doses SNC's review of the radiological doses for the power uprate showed that many of the original design-basis doses were bounding. For safety related electrical equipment not bounded by the original design basis, radiological doses at uprate conditions were compared with the dose threshold limits used for the individual components or individual items of equipment. SNC finds that the comparison showed sufficient margin available to accommodate the increased uprate dose without compromising the equipment qualification.
The staff raised a question that the dose should be bounded by the test report values, not by the dose threshold limits. By letter dated August 5,1997, SNC clarified that the terminology of dose threshold limits and test report values have the same meaning in the context of equipment qualification evaluatior prepared for power uprate.
SNC concludes that the radiological cumulative dose or dose rate was either enveloped by the results of previous design-basis radiological analysis or was within the threshold limit for which the individual components or individual items of equipment were qualified.
(b) Containment pressure and temperature evaluation SNC's comparison of the composite uprate temperature profile to the existing composite 1
temperature profile indicated that the uprate maximum accident temperature is approximately i
. separate initiative outside the scope of this evaluation, the staff will continue to review the adequacy of an equivalency evaluation using Arrhenius methodology.
(c) Surface temperature analyses Comparison of the MSLB cases used in the current design-basis surface temperature analyses with the corresponding uprate MSLB cases indicated that the surface temperature results are bounding and are not affected by the power uprate. SNC also compared the design-basis post-LOCA containment temperature profile with the uprate post-LOCA containment temperature profile and finds that the original temperature curve was bounding and that the power uprate does not affect the surface temperature.
SNC finds that maximum temperature for the MSLB cases used in the design-basis surface temperature analyses are greater than or comparable with the corresponding uprate MSLB cases.
Based on its review, the staff concludes that plant operation at the proposed uprated power level will have an insignificant or no impact on the safety-related mechanical components inside l
or outside the containment. In addition, the existing safety-related electrical equipment qualification is not affected by the power uprate and remains bounded for the power uprate, and therefore, is acceptable.
4.14 Safetv/ Relief Valves SNC indicated that relief valve setpoints have been assumed to be 103 oercent of the lift setting in the analysis for the power uprate. The relief valve setpoints, rated capacities, and corresponding dynamic loads due to valve operation imposed on the piping and adjacent structures do not change as a result of uprate. On the basis of its review, the staff finds the safety and relief valves will continue to perform their function at the power uprate condition.
4.15 Reactor Trio System /Encineered Safetv Feature Actuation System Instrumentation Trio Setooints and Allowable Values SNC proposed one change to Trip Setpoint in TS Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints, and two changes to Trip Setpoint in TS Table 3.3-4, Engineered Safety Feature Actuation System Instrumentation Trip Setpoints. The change to Table 2.2-1 will lower the Trip Setpoint for Functional Unit 20.C, Power Range Neutron Flux, P-8 from 35 percent to 30 percent of rated power. The first change to Trip Setpoint in TS Table 3.3-4 will decrease the Trip Setpoint for Functional Unit 5.a. Turbine Trip and Feed Water Isolation from Steam Generator Water Level-High-High, from 79.2 percent to 78.5 percent of narrow range instrument span. The second change to Trip Setpoint in TS Table 3.3-4 will increase the Trip Setpoint for Functional Unit 8.b, Engineering Safety Feature Actuation System Interlocks for Low-Low Tavg, P-12, from 544*F to 545'F.
SNC provided justifications for these changes in its submittal dated September 22,1997, which employed uncertainty calculations using NRC previously approved methodology in WCAP-13751, " Westinghouse Setpoint Methodology for Protection Systems, Southern Nuclear I
-._m_m_
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- 7.1.2 New Anoendix C. " Additional Conditic s." to Ooeratino License The staff will impose three new license conditions for each unit, to be located in a new l
Appendix C of the respective operating license. The new license conditions are needed in order to grant approval of the power uprate license amendments. By letter dated April 17, 1998, SNC proposed the following license conditions for Unit 1: (1) SNC shall complete classroom and simulator training for operations crews prior to Unit 2 entering Mode 2 from the spring 1998 refueling outage; (2) SNC shall complete final simulator modifications in accordance with ANSI /ANS 3.5-1985 and review results of the Cycle 16 startup testing to determine any potential effects on operator training within 2 years after restart from the Unit 1 fall 1998 refueling outage; and (3) SNC shall provide an SGTR radiological consequences analysis that incorporates a flashing fraction, which is appropriate for the Unit 1 design prior to the Unit 1 SG replacement outage in spring 2000.
In addition, by letter dated April 17,1998, SNC proposed the following license conditions for
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Unit 2: (1) SNC shall complete classroom and simulator training for operations crews and l
temporary simulator modifications prior to Unit 2 entering Mode 2 from the spring 1998 refueling j
outage; (2) SNC shall review the results of the Cycle 13 startup testing to determine any j
potential effects on operator training and incorporate these changes into licensed operator trainity prior to Unit 1 startup from the fall 1998 refueling outage; and (3) SNC shall provide an SGTK ; radiological consequences analysis that incorporates a flashing fraction, which is appropriate for the Unit 2 design prior to the Unit 2 SG replacement outage in spring 2001.
i 7.1.3 TS 1.25. Definitiorii Rated Therrnal Power The definition for rated thermal power is being changed to increase the rated power from 2652 MWt to 2775 MWt. SNC has provided the results of its reanalyses or evaluation including
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LOCA and Non-LOCA transients and accidents, containment response, radiological consequences, NSSS and BOP systems and components to support the operation of FNP Units 1 and 2 at the uprated power level. The staff has reviewed SNC's submittal and concludes that both FNP units can safely operate at a core power of 2775 MWt.
7.1.4 TS Table 2.2-1. Reactor Trio System Instrumentations Trio Setooints TS Table 2.2-1 was modified to change the setpoints and allowable values for the trip setpoints.
The allowable values were based on the FNP instrumentation and were modified to support the amendment request. All applicable LOCA and non-LOCA transient analyses were performed using the new setpoints with acceptable results. The Power Range, Neutron Flux allowable value for the low setpoint was changed from 26 percent to 25.4 percent, and the high allowable value was changed from 110 percent to 109.4 percent. The Power Range, Neutron Flux, High Positive and Negative rate allowable values were changed from 5.5 percent to 5.4 percent. The Pressurizer Pressure values for the low alloweble value was changed from 1855 psi to 1862 psi l
and the high allowable value was changed from 2395 psig to 2388 psig. The Pressurizer Water Level-High allowable valve was changed from 93 percent to 92.4 percent. The Loss of Flow allowable value of 88.5 percent of 89,290 gpm was changed to 89.7 percent of 88,100 gpm.
The Steam Generator Water Level-low-low allowable value of 23.3 percent was changed to 24.6 percent. The Reactor Trip System Interlocks for Low Power, Power Range Neutron Flux,
. reduced from 112 percent of total flow to 105.8 percent total flow. SNC serformed an analysis of overpressure protection and the limiting overpressure transients to verify that there continues to be sufficient secondary relieving capacity. The staff has reviewed this change and concludes that operation at the proposed uprated power level will have an insignificant impact on the main steam system and its associated components. As a result, the changes are acceptable.
7.1.11 TS Table 5.7-1 SNC proposed to increase the number of secondary system hydrostatic tests limit to 10 to allow a relaxation of the current limit of 5 specified in Table 5.7-1/' Component Cyclic or Transient Limits" of the FNP TSs. SNC indicated that the power uprate fatigue analysis assumed a maximum number of occurrences of 10 for the secondary site hydrostatic tests, as shown in a table on page 52 of SNC's August 5,1997, response to the staff's request for additional information. The staff has reviewed SNC's power uprate evaluation and finds that the proposed change is acceptable.
7.1.12 TS 5.4. Desian Features.
The value given for the total water volume and steam volume in the RCS is being changed from 9723 to 9829 ft'. The new value reflects the current calculated total water and steam volume of l
the RCS at 567.2*F, The new calculated RCS fluid volume is supported by the overall uprate program that included the reanalysis or evaluation of the LOCA, non-LOCA, thermal-hyoraulic, and nuclear aspects of the NSSS and BOP structures, systems, and components. As a result, the staff finds this acceptable.
7.1.13 TS 6.9. Administrative Gontrols The references to WCAP-10266-P-A. Rev. 2, "The 1981 Version of the Westinghouse Evaluation Model Using BASH Code," is being deleted and replaced with a reference to the approved version of WCAP-12945-P-A, " Code Qualification Document for Best Estimate LOCA Analysis" (W-Proprietary), March 1998. Because the new code has been reviewed and approved for LOCA analysis and the code is used in accordance with alllimitations and restrictions, this is acceptable. A reference to the Westinghouse fuel design Topical Report, WCAP-12610-P-A, " VANTAGE + Fuel Assembly Reference Report," April 1995 (W-Proprietary),
using ZlRLO cisdding is also being added to the TS. This is acceptable because ZlRLO clad fuelis being used at FNP and the referenced topical report is approved for FNP. The use of these methodologies will ensure that values for cycle-specific parameters are determined such that all applicable limits of the plant's safety analysis are met.
8.0 STATE CONSULTATION
i In accordance with the Commission's regulations, the State of Alabama official was notified of l
the proposed issuance of the amendments. The State official had no comments.
i 1
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9
o Table 4.1.1-1.
Comparison of NRC and SNC Determined End of Life (EOL) Upper-Shelf Energy (USE) Values for the Beltline Plate and Weld Materials in the FNP Unit 1 Reactor Pressure Vessel' Plate No.
Cu Content il4T Unirradiated Decrease NRC EOL SNC Projected (wt.%)
Fluence USE (ft-lb) "8 Decrease in USE (ft.
USE 1/47 USE (ft-Ib)"'
{E19 n/cm")
in USE m Ib)'
8 il4T (ft-Ib)
EOL USE Decrease I"
0 0.13 2.705 99.0 27.5 27.2 72 28 71 g6903 2 I
0 0.12 2.705 87.0 26.3 22.9 64 27 64
, 33 0.14 2.705 86.0 23.5
- 20.2
- 65
- 24
- 65
- g,
0 0.14 2.705 86.0 28.8 24.8 61 29 61 B 9 Int Shell Axial Welds 0.258 0.841 149.0 23.1
- 34.4
- 115*
23
- 115*
19-894 A/B C rc Wold 0.205 2.705 104.0 43.5 45.2 59 43 59 Low. Shell Axial Welds 0.197 0.841 82.5 32.5 26.8 56 34 54 20-894 A/B Footnotes:
1.
All projections for the EOL USE values were performed in accordance with the methodology of Position 1.2 of Section C of Regulatory Guide 1.99, Revision 2, " Radiation Embnttlement of Reactor Vessel Materials," except as noted in Footnote No.
j 3.
1 I
2.
Calculational methods for calculating the neutron fluence at the 1/4 thickness (1/4T) location of the reactor pressure vessel (RPV) wan are as defined in Equation (3) of Regulatory Position 1.1 in Regulatory Guide 1.99. Revision 2," Radiation Embnttlement of Renctor Vessel Materials " or in the equivalent equation in the revised rule 10 CFR 50.61. " Fracture l
Toughness Requirements for Protection Against Pressurized Thermal Shock."
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3.
Projections for the EOL USE values using surveillance data were performed in accordance with the methodology of Regulatory Position 2.2 of Regulatory Guide 1.99, Revision 2," Radiation Embnttlement of Reactor Vesse! Matenals."
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