ML20148Q309

Safety Evaluation Supporting Amend 35 to License NPF-30

ML20148Q309
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Site:	Callaway
Issue date:	03/30/1988
From:	Office of Nuclear Reactor Regulation
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ML20148Q298	List: ML20148Q295 ML20148Q309 ML20148Q321
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NUDOCS 8804120399
Download: ML20148Q309 (14)

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• g SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 35 TO FACILITY OPERATING LICENSE NO. NPF-30 UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 DOCKET NO. STN 50-483

1.0 INTRODUCTION

By letter dated March 31, 1987, Unfor Electric Company requested an anendment to Facility Operating License No. NPF-30 for the Callaway Plant.

Specific-ally, the amendment would revise the Callaway license and Technical Specifica-tions to increase the rated core power level from the present specification of 3411 megawatts thermal (MWt) to a specification of 3565 MWt. The proposed changes would allow Callaway to operate at a Nuclear Steam Supply System (NSSS) power of 3579 MWt.

The proposed change, would represent an approximate 4.5 percent increase over the presently licerned core power rating of 3411 MWt.

In support of this 154 MWt uprating, Callamy was reevaluated for operation at the Engineered Safety Features Design Rating of 3565 MWt core power and 3579 MWt NSSS power.

In su (BOP)pport of the application, the licensee provided NSSS and balance-of-plant upratin; licensing repor*s.

Further information was provided in response to NRC reques :s. Als; provided were proposed technical specification changes to support the uprat1Fg.

2.0 EYJLUATION 2.1 NUCLEAR STEAM Sd3 PLY SYSTEM (NSSS)

The scope of the licens h's review to support the proposed core uprating encompassed all aspects ot' the Callaway NSSS design and operation affected by the increase.

NSSS det igt.1 were reviewed to verify compliance at the increased power rating with licenMng criteria and standards currently specified in the Callaway operating license.

In addition, a review was conducted by the licensee to identify any potential unreviewed safety questions that might occur as a result of the increased power level in accordance with 10 CFR 50.59. The i

structural design of NSSS equipnent was reviewed to assure that compliance had been maintained at the increased power level with industry codes and standards that applied when the equipment was originally built.

In addition, the review encompassed the verification that NSSS components ar.d systerns will continue to neet functional requirements specified in the Final Safety 1,nalysis Report (FSAR) at the increased power level. Currently approved NRC analytical techniques were used for analyses performed at the increased power level.

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. Also,thedefinitionofNSSS/BalanceofPlant(80P) safety-relateditterfaces were reviewed for any impact at the increased power level.

Based on the scope of review as outlined above, the licensee states that Callaway is capable, in its present design configuration, of operating at the proposed core power level of 3565 MWt and an NSSS power level of 3579 MWt without violating any of the design criteria or safety limits specified in the Callaway FSAR and as currently required in Facility Operating License NPF-30 for Callaway.

In addition to evaluating the ability of the plant to perform at the new power level under steady state conditions, the licensee reevaluated all the design basis transients and accidents which the NRC staff utilizes to determine that adequate safety margins are maintained.

These analyses were perfonned by Westinghouse using computer codes which have been previously reviewed and approved by the NRC staff. Those events which might challenge the core Depar-ture from Nucleate Boiling Ratio (DNBR) limits wers evaluated using the Westing-house Improved Thennal Design Procedurr.

Steady state instrument errors were considered in establishing the initial conditions, including the addition of 2 percent to the initial power to account for calorimetric error.

Core Design By letters dated November 14, 1985 and March 31, 1987, the licensee described Optimized Fuel Assembly (OFA) and Westinghouse Vantage 5 (V-5) fuel loadings for the Callaway core, respectively, in addition to Low Parasitic (LOPAR) fuel remaining in the core. These descriptions considered operation at 3565 MWt, except for Technical Specifications 1.10 through 1.41, Figure 2.1-1, and Table 2.2-1 dealing with revised definitions, revised core safety limits, and revised notes associated with the overtemperature and overpressure delta T trips (includingrevisedsetpoints). These previous descriptions were approved by the staff and, except in the technical specifications identified ebv/a, remin appl *. cable and acceptable for the uprated power.

Overpressure Protection It is r', quired that pressurizer safety valves be designed with suf ficient capacicy to prevent the pressurizer pressure from exceeding 110% oesign pressure following the worst reactor coolant system pressure transient.

For purposes of analytical justification, this event is specified to be a 100'; load rejection resulting from a turbine trip with concurrent loss of main feedwater. No credit is taken for operation of reactor coolant system relief valves, steam line relief valves, steam dump system, pressurizer level control system, pressurizer spray, or direct reactor trip on turbine trip.

React w scram is initiated by the second safety-grade signal from the reactor protection system.

For operation at 3411 MWt core power, the Callaway safety valve capacity was found to be acceptable based on reference to analyses contained in WCAP-7769.

However, the applicability of those analyses was not justified for the uprated power.

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. By letter dated October 23, 1987, the licensee provided the results of analyses similar to those referenced in WCAP-7769 for the above described licensing-basis event which were performed specifically for Callaway operating at 3650.7 MWt (102% of the Engineered Safety Features Design Rating - 3579 MWt). The results of these analyses demonstrate that the combined capacity of two of the three pres-surizer safety valves is adequate to meet the perfonnance criterion identified above if credit is taken for the first safety-grade signal (high pressurizer pressure) from the reactor protection system.

Further, 70 percent of the total safety valve capacity is needed to meet regulatory criteria if any of four subsequent trips (overtemperature delta-T, high neutron flux, high pressurizer level, low-low steam generat6r water level) initiate reactor scram. The licensee also stated that steam generator safety valve flew did not exceed nominal plant steam flow at any time during the analyzed transients.

Based on the above information, the staff concludes that the pressurizer safety valve capacity at Callaway is adequate for operation at core powers up to 3565 MWt (NSSS power - 3579 MWt).

Auxiliary Feedwater and Residual Heat Removal The staff review and approval of the Callaway Auxiliary Feedwater System (AFS) design is given in NUREG-0830 Safety Evaluation Report (SER), Section 10.4.9.

By letter dated October 23, 1987, the licensee stated that analyses supporting Callaway 0FA and Vantage-5 fuel reloads includeo reanalyses of station blackout, loss of nonnal feedwater, and feedwater line break events considering the uprated power. These events are the limiting transients identifying worst single failure assumptions and minimum AFS flow requirement for the Callaway AFS design. The licensee stated that, as a result of these analyses, identification of limiting scenarios for single failure and required flow were not changed. The adequary of the AFS flow capacity was also demonstrated in these analyses, which were approved by the staff in 0FA and Vantage-5 fuel reload safety evaluations. The staff concludes that, since the Callaway AFS flow capacity exceeds cooling requirements for the uprated power, cooldown time to residual heat removal (RHR) system cut-in conditions would not be significantly affected.

The licensee indicated that the original 16-hour olant cooldown time from RHR cut-in at 350*F to 140*F would be increased to 19.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> for the uprated power.

The licensee also stated that required RHR heat exchanger tubeside and shellside flow rates would not change and that safety requirements would continue to be satisfied.

Based on the above, the staff agrees with the licensee that the impact of the power uprating on RHR performance would not be significant.

Emergency Core Cooling System (ECCS)

From the licensce's study, no adverse impact to ECCS operability or vulnerability to single failure resultant from the power uprating was identified.

ECCS perfomance for the uprated power was demonstrated in analyses supporting the Vantage-5 reload. These analyses were reviewed by the staff and found to demonstrate compliance with 10 CFR 50.46(b) and Appendix K.

Therefore,the Callaway ECCS is adequate for the uprated power.

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. $ 1 dent Analysis The licensee indicated that all FSAR Chapter 15 events were reanalyzed or reevaluated considering the uprated power as part of documentation to support 0FA and Vantage 5 reload cores. The staff reviewed these analyses and concluded that appropriate safety criteria are met.

The staff finds that these analyses continue to be applicable and acceptable for the uprated power.

In addition, the licensee addressed other topics related to accident analyses and power uprating.

The licensee addressed potential rod withdrawal events during Mode 4 operation by identifying that the reactor would be tripped by the source range high neutron flux trip, which is required to be in operation during Mode 4 by technical specifications. The staff finds this acceptable.

The licensee addressed postulated loss-of-coolant accidents during Mode 4 operation by referencing the Westinghouse Owners Group (WOG) program to deal with shutdown loss-of-coolant accidents (LOCA's) on a generic basis.

In the near term, the licensee is implenenting Interim Shutdown LOCA Guidance developed through the WOG program. To address the longer term resolution of this issue, the licensee comitted to continue to support the WOG efforts toward a generic resolution and to implement portions of the final resolution which are applicable to Callaway. The staff finds this acceptable.

Steam generator tube rupture (SGTR) events were discussed in the staff SER and in Supplement 61 Safety Evaluation Reports (SSER's) 3 and 4 The SER considered radiological consequences of a SGTR and concluded that the consequences, calculated assuming what is now the uprated power, are acceptable.

SSER'e 3 and 4 consider the SGTR event scenario description and conclude that the design basis scenario description is acceptable, pending review of a confirmatory reanal/ sis.

Of significance in this scenario is an expeditious cooldown from full power conditions to cressure and temperature equilibrium between the primary system anu the shell side ci the ruptured steam generator, using the unruptured steam generators fed by the auxiliary feedwater system for the cooldown.

As identified in the discussion of auxiliary feedwater system, the staff concludes that the flow capacity of the AFS continucs to exceed cooldown requirements for the 1

uprated power, and that cooldown times would not be significantly affected.

Therefore, it is concluded that the conclusions of the staff SER and SSER's 3 and 4 concerning SGTR continue to apply for the uproted power. The licenste also identified that the reanalyses described in the SGTR submittals of 1986 and 1987, to address the SSER 3 license condition, consider cperation at the uprated power. These confirnatory submittals are under ongoing staff review.

From the licensee's submittals, it is apparent that the uprated NSSS thermal power (3579 MWt) is about 17 MWt greater than the calculated valves-wide-open (VWO) rating of the turbine.

The licensee states that the apparent disparity would not result in an increased likelihood of challenges to safety systems.

The reconciliation of this apparent turbine rating discrepancy is, discussed in the turbine missiles section of this report.

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. With regard to its impact on accident analyses, the staff concludes that the design of safety systems is adequate to assure safety, shculd a power mismatch between NSSS and turbine occur.

If operating experience were to contradict the licensee's expectations regarding the likelihood of challenges to safety systems, this would be reflected in licensee reports per 10 CFR 50.72 and 50.73, with appropriate followup. Therefore, the staff concludes that this apparent power mismatch is acceptable within the context of this discussion.

The licensee identified that, though many accident analyses were perfomed assuming 15% steam generator tube plugging, some of the accident analyses referenced in the uprating request were performed assuming 10% steam generator tube plugging. Therefore, the licensing basis for the uprated power continues to be 10% maximum steam generator tube plugging.

Based on the above discussions, the staff finds the accident analyses acceptable to support operation at the uprated power with a limit of 10Fsteam generator tube plugging.

NSSS Sumary Based on its review and referencing past reviews of 0FA and Vantage-5 fuel core reloads within the scope of systems areas discussed above, the staff finds the proposed Callaway power uprating to 3565 MWt core power (3579 MWt NSSS power) acceptable for steam generator plugging up to a maximum of 10 percent.

2.2 BALANCE OF PLANT (B0P) SYSTEMS The licensee stated that the BOP systems and components were analyzed by performing a feasibility study to determine the impact of power uprating on their performance capability and to identify any required modifications. The analysis was performed using the VWO heat balance data (15.85 E6 lb/hr main steam flow and main feed flow). The systems reviewed were the non-safety-related secondary side power generating systems.

Included in the licensee's analysis were safety-related portions of the main feedwater and main steam system, steam generatorblow-downsystem(SGBS),componentcoolingwatersystem(CCWS),

auxiliary feedwater system (AFS), fuel building heating, ventilation and air conditioning (HVAC) system, and spent fuel pool (SFP) cooling and cleanup system.

The performance of the above BOP systems was evaluated at the uprated power level by using the new primary side NSSS data (15.96 E6 lb/hr main steam flow and main feed flow, and 446*F main feedwater temperature) furnished by Westing-house. The licensee stated that the impact on containment pressures and temperatures following a postulated design basis main steam line break was evaluated and its effect on equipment qualification was verified.

The flooding analysis in safety-related areas of the plant as a result of a postulated pipe break was reevaluated due to the increase in flow rates in the main feedwater, condensate, and main steam systems. The turbine-generator system was also evaluated to confirm its integrity and performance at the increased steam flow and to verify that the original turbine missile analysis remains Valid.

The following items concerning the proposed power uprating were or are addressed separately by the staff, and are, therefore, excluded from this discussion:

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. (a) The SFP cooling capability, fuel building HVAC perfonnance and radioactive effluent releases and solid waste generation at the uprated power were addressed in the staff's Cycle 3 Reload Amendment (and safety evaluation) dated October 9, 1987.

(b) Themainsteamlinebreak(MSLB)outsidecontainmentanalysis(with superheated steam) at the uprated power was addressed in the staff's safety evaluation issued on February 18, 1988.

(c) The turbine missile analysis for the power uprating is discussed in the turbine missiles section of this report.

The licensee's analysis of S0P system perfonnance provided the following findings concerning the uprated power level of 3579 MWt NSSS power:

(a) The capability of the safety-related portion of the main feedwater system will not be affected and will continue to perfonn its safety function because the uprated power con-ditions are bounded by the existing main feedwater system design. The licensee's analysis of the pressure / temperature rating conditions for the system piping confirmed that pres-sure boundary integrity will not be affected.

In addition, the main feedwater isolation valve closure time is not affected by the increased feedwater flow.

(b) The capability of the steam generator blowdown system to remove impurities from the secondary side is unaf fected by the increased main steam flowrate during nonnal operation based on the existing system design.

(c) The reactor water makeup system (RMWS) capability to provide ceaerated water for makeup and flushing operations throughout the NSSS auxiliaries, the radwaste systems, and fuel pool cooling and cleanup system is not challenged because the existing I

system design is based on the worst case demand which bounds the uprating demands.

(d) The licensee confinned that safety-related equipment will not be affected by changes in the flooding analysis due to the power uprating. Outsice containment, the safety-related areas which contain system piping operating at increased flowrates consist of the main feedwater isolation valve compartments and main steam isolation valve compartments. The flood level in these rooms as a result of postulated breaks in the main steam and feedwater piping are not impacted by the increased flowrates in these systems due to the uprated conditions since the existing drainage capability in these arers is sufficient to prevent exceedingtheacceptablefloodlevellimitofthreefeet(3'0")

i assumed in the FSAR flooding analysis.

Inside containment, the resulting increased flood level due to the increased main steam and feedwater flowrates at the uprated conditions for a postulated main steam or feedwater line break would be bounded by the existing LOCA analyses.

(e) The AFS is unaffected since the AFS flowrate for design basis accident decay heat removal in the existing Westinghouse analysis, which served as the sizing basis, is based on the uprated NSSS power level of 3579 MWt.

(f) As evaluated by Westinghouse, the CCWS is capable of removing the slightly increased heat loads from safety-related equipment which it serves. However, an additional 3.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> is recuired to cool the primary plant from 350'F to 140'F once cold shut-down is initiated through the RHR system. This extended cool-i down period was previously addressed in the NSSS section.

(g) For main steam line breaks inside containment, the pressure and temperature will remain within the bounds of the peak pressure and temperature used in the evaluation of containment performance.

In addition, the equipment qualification envelope 1

for harsh environments inside containment bounds the slight increase in energy release from a postulated MSLP due to the power uprating. The licensee confirmed that containment environ-mental qualification of equipment inside containment is not affected.

(h) The licensee confirmed that B0P systems have the capability to maintain plant operation at the uprated power level without modification to their existing design.

The staff has reviewed the FSAR and licensee submittals in order to verify that safety-related B0P system perfonnance capability as analyzed bounds the changes in design basis accident assumptions created by the increased main steam and feedwater system conditions. The staff has confinned that safety-related BOP cystem design capability, flooding protection and equipment quali-fication are bounded for the proposed power uprating and, therefore, modifica-tions to them are not required.

Based on the above, the staff concludes that the proposed Callaway licensing amendment concerning the core power uprating from 3411 MWt to 3565 MWt is within the existing safety-related B0P system design capability for design basis accident mitigation and, therefore, the staff's previous approval against the applicable licensing criteria for the main steam and feedwater system CCWS, AFS, RMWS, SGBS, flooding protection, containment performance,

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and equipment qualification remains valid.

The staff, therefore finds the 80P systems acceptable for the proposed power uprating.

l 2.3 REACTOR VESSEL The reactor vessel must be designed to Section III of the ASME Code and satisfy the "Pressurized Thermal Shock" (PTS) rule in 10 CFR 50.61. To meet Section III of the ASME Code, the stresses calculated from reactor vessel design transients must satisfy the stress limits in the ASME Code.

The licensee has detennined that the uprating did not change the original design transients and verified that the existing reactor vessel stress analysis bounds the uprated condition.

Since the FSAR Section 5.2 indicates that the reactor vessel met Section III of the ASME Code, the information provided by the licensee indicates that the reactor vessel remains in compliance with Section III of the ASME Code.

The PTS rule requires that at the end-of-life of the reactor vessel, the projected reference temperature (calculated by the method given in 10 CFR 50.61(b)(2),RT value for th2 materials in the reactor vessel beltline be lessthanthesNhe)ningcriterionin10CFR50.61(b)(2).

The RT value is t

dependentupontheinitialreferencetemperature,marginsforuncNkaintyin the initial reference temperature and calculational procedures, the amounts of nickel and copper in the material, and the neutron fluence at the end-of-life of the reactor vessel. Of these properties, only neutron fluence is affected by uprating the core to a higher power level. All other properties are inde-pendent of the core power level.

In a letter dated January 21, 1986, the licensee provided the staff with the l

infonnation needed to determine whether the Callaway reactor vessel beltline material would neet the screening criterion in 10 CFR 50.61(b)(2). The neutron i

fluence estimates were based on the core power level prior to uprating.

In a letter dated December 15, 1986, the staff evaluated the infomation submitted by the licensee and determined ti ; the calculated RT values for the reactor vessel beltline materials were below the screebg criterion at the end-of-life of the reactor vessel.

The information provided by the licensee in their letters dated March 31 and October 2,1987 indicates that the licensee has revised the neutron fluence calculation method and the estimated end-of-life neutron fluence for the i

reactor vessel. The staff has reviewed this information and determined that the revised neutron fluence calculation nethod is acceptable and the revised neutron fluence will not result in the reactor vessel's projected RT value exceeding the screening criterion in 10 CFR 50.61(b)(2). Hence,aftehTS considering the uprated core power level, the Callaway reactor vessel remains in compliance with the PTS rule.

Since the Callaway reactor vessel is in compliance with the ASME Code and the PTS rule,10 CFR 50.61, the proposed plant uprating will not affect the integrity of the Callaway reactor vessel.

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2.4 TURBINE MISSILES By letter dated November 13, 1987, the licensee responded to the staff's request for additional information rogarding the turbine missile review for the Callaway Plant power uprating. The request was related to (1) the effect of changes in steam conditions due to the uprating on stress corrosion cracking of low pressure turbine wheels, (2) the probability of turbine overspeed protection system failure under uprating conaitions, and (3) the potential 1

increass in the probability of turbine missile generation.

The factors that directly or' indirectly cause stress corrosion cracking in the low pressure turbire wheels are steam pressure and temperature, mass flow rate, steam moisture content, water chemistry, oxygen level, and turbine speed.

The licensee reported that the changes in these factors are negligible due to uprating. The only noticeable change that the staff can determine is a 0.5%

increase in the steam flow rate. To safeguard against any potential cracks in the turbine wheels, the wheels will be ultrasonically inspected every 6 years, which is consistent with other nuclear plant inspection schedules.

The licensee stated that the turbine overspeed protection system will not be modified for the uprating. The staff's concern was whether the existing controls (e.g. valves and trips) are capable of responding to a turbine over-speed transient such that the turbine speed would remain under control.

According to the Callaway FSAR, the tur51ne overspeed protection system was designed to control overspeed up to 114%.

General Electric, the turbine vendor, analyzed the system for additional mechanical control failures which would result in 120% overspeed. The limiting case, however, assumed complete failure of all the control systems and a 180% overspeed event. The probability associated with a 180% overspeed event was estimated to be 1.5 E-7. Because of a small incre6se in the steam flow rate, the staff believes that the power uprating will not affect the probability of a turbine runaway event nor the overall prebability of turbine missile generation.

2.5 PLANT STRUCTURAL AND THERMAL DESIGN The NSSS review consisted of comparing the existing NSSS design with the performance requirements at the uprated power level.

Review of the reactor coolant pump and control rod drive rechanism design indicated that operating conditions for 3579 KWt operation are bounded by the original thermal and structural design analyses.

The stress analyses for the reactor coolant loop, reactor coolant loop bypass, and pressurizer surge line piping were reviewed for the 3579 MWt operating conditions. The review demonstrated that piping stresses and support loads at the uprated conditions remain in compliance with requirements of all applicable i

design criteria as defined in the FSAR.

1 Review of the pressurizer design verified that the operating conditions for 3579 Kdt NSSS operation are bounded by the original thermal and structural design analyses. The existing pressurizer safety valves and power operited relief valves were also found to be adequate for operation at the uprated conditions.

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. Modifications to the model F series steam generator stress report have been made using a set of operating parameters which bound conditions for the Callaway Plant at 3579 MWt NSSS operation. The evaluation indicated that the model F series steam generator stress report satisfies all applicable ASME Code require-ments when updated to the enveloping set of plant operating parameters.

As indicated previously, on reactor coolant piping, uprating of the Callaway Plant to 3579 MWt has a negligible impact on the Westinghouse supplied portion of the reactor coolant system supports.

For the Callaway auxiliary sistems components supplied by Westinghouse, evalua-tion was again made based on a comparison of the original design requirements to the systems design requirements at the uprated conditions.

In each case, the conditions used in the original design enveloped those required for opera-tion of the Callaway Plant at 3579 MWt NSSS.

Finally, the original Callaway reactor internal design was found to remain in compliance with the requirements of all applicable design criteria, as cefined in the FSAR, based on the following findings:

(a) The potential for flow-indi.ced vibrations is not increased due to power uprating.

(b) Stresses and fatigue usage factors for components in the baffle-barrel-I former region of the reactor internals are bounded by the original analysis.

I (c) Stresses and fatigue usage factors for the upper and lower core plates are bounded by analyses performed for other plants using the same core plate design.

The staff concurs with the above evaluations performed by the licensee and I

finds the original NSSS equipment and component designs to be acceptable for power uprating to 3579 FWt NSSS.

The B0P review consisted of comparing the existing BOP design with the perform-i ance requirements at the uprated power level and determining if modifications i

to the plant were required.

The licensee evaluated performance using the VWO heat balance but with the modified pressure, temperature, and flow from various cases of steam generator tube plugging and throttle pressure. These cases reflect extreme conditions which might be encountered should the full 3565 MWt (3579 MWt NSSS) be reached. The licensee has determined that at 3579 MWt the temperature of main feedwater will increase by only 1.5'F and the main steam pressure may decrease.

For other systems, the changes in pressures and temperatures will also be negligible. The increase in flow in the systems, if any, will also be too small to impact pipe stresses of systems for which transient analysis was done. Considering the insignificant changes in system operating conditions and the conservatism in stress analyses and the pipe support design, the staff concurs with the licensee's conclusion that no pipe stress reanalysis will be required due to uprating.

. In addition, the following hazard analyses were also reviewed by the licensee and found to be acceptable by the staff:

(a)

II/I Analysis The existing stress analyses of the II/I piping systems will not be affected because no significant changes in their operating conditions will occur. Also, no modification to the piping geometry of any kind will be made. Therefore, uprating will not have any impact on the II/I evaluation.

(b) Pipe Break and Jet Analysis The changes in pressures and temperatures of safety-related main steam and main feedwater piping are too insignificant to have any effect on the existing break locations. Current jet impingement analyses are unaffected as these analyses are generally conservative, and the increases in the flow rates are less than 1 percent. The operating conditions of other safety-related high energy systems evaluated will not change due to uprating. Therefore, the pipe break locations and jet analyses will not be affected.

For main steam line breaks inside containment, it was found that the maximum temperatures and pressures will remain within the peak pressure and temperature used in the evaluation of containment perfornance.

In addition, equipnent inside containment was reviewed, and it was confirmed that the previous environmental qualification will not be affected.

(c) Moderate Energy Crack None of the moderate energy lines will experience any significant change in their operating conditions due to uprating. Therefore, the "No Crack Zones" as well as the evaluation done for the moderate energy cracks will not be affected.

Based on the above, for all the secondary-side systems reviewed, it was concluded by the licensee and concurred with by the staff that they have the capability to function properly at the power level of 3579 MWt NSSS power without any rodifications to the existing design.

2.6 NUCLEAR, PROCESS AND POST-ACCIDENT SAMPLING SYSTEMS The uprating is expected to produce some changes in temperatures and pressures in different plant systems and in a higher post-accident containment atmosphere l

activity. Since nuclear, process and post-accident sampling systems take fluid samples from these systems, the licensee was required to demonstrate satisfactory perforrance of the sampling systems under uprated pl&nt conditions.

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B The licensee has detennined that increasing of core power from 3411 MWt to 3565 MWt would raise the primary coolant temparature by approximately 2*F.

However, since this would not provide sufficiently large temperature differential for transferring additional heat energy to the secondary coolant system, the temperature of steam will be reduced by about 7.5'F and the corresponding pressure by 50 psi.

The effect of core power increase on the post-accident containment thermal environment will be relattvely small, since the core power will be increased by only 4.5 percent. The pot.ential activity in the containment atmosphere due to release of noble gases and iodine will increase.

The primary coolant system is sampled by the nuclear and post-accident sampling systems, and 2'F temperature rise could be easily acconnodated by these systems without causing degradation of their perfonnance. Similarly, secondary system fluids sampled by the process sampling system would, for the same reason, cause no problems.

In sampling the post-accident containment atmosphere the slight change in thermal conditions would have an insignificant effect on the post-accident sampling. Therefore, all the systems sampling the primary and secondary coolant and the containment atmosphere will meet the requirements of General Design Criteria 13 and 64. They will also conform to the reconnendations of StandardReviewPlan(SRP)Section9.3.2. An increase of the reactor power by 4.5 percent would proportionally increase activity of the samples. The staff has independently determined that the increase in activity will not be significant and that the sampling systems will meet the requirements of General Design Criterion 19.

On the basis of its evaluation, the staff concludes that the proposed uprating of the core power in Callaway Plant does not significantly affect the perfonnance 6

of the nuclear, process and post-accident sampling systems and their operation i

remains consistent with General Design Criteria 13, 19 and 64 and with SRP l

Section 9.3.2 and, therefore, is acceptable.

2.7 ENVIRONMENTAL CONSEQUENCES Radiological The licensee's analyses of radiological source terms for normal operations and accidents are reported in Attachment 1 of their March 31, 1987 submittal. The i

licensee's submittal and the staff's previous Safety Evaluation Report (SER),

dated October 1981, were reviewed. The power level used in the staff's previous analysis of source terms for normal operations was 3565 MWt. The staff concludes that the radwaste-management systems remain capable of n,eeting the requirements 4

of 10 CFR Part 50, Appendix 1, and the Annex to Appendix 1.

The power level used in the staff's previous accident analyses was the uprated value of 3565 MWt. The staff concludes that the radiological consequences of all the design j

basis accidents are either bounded by the values already in the FSAR, or are a j

small fraction of the limits in 10 CFR Part 100.

. i hon-Radiological This minimal change in power level will yield changes in output parameters such as water use, discharge temperature, fish impingement, etc., that are essentially unmeasurable or within the normal measurement ano calculational tolerances. Any associated environmental effects will likewise be unmeasurable i

and, therefore, acceptable. Also, the eriginal licensing evaluations for the plant, including the hRC Environmental Enluations (Reference NUREG-75/011, 1

3/75, Section 1.1), were based on an NSSS thermal power level of 3579 MWt.

Therefore, the proposed uprating remains within the bounds.of the original environmental analyses.

2.8 ELECTRIC SYSTEM DESIGN The staff has reviewed the information provided by the licensee regarding the t

main generator system, excitation and voltage regulation, startup transformer, lower medium voltage system, standby generators, load shedding and sequencing, miscellaneous power systems and the switchyard.

There are no electrical system design changes or significant load changes indicated by the licensee. Additionally, the staff would not expect any significant system load changes as a result of the proposed uprating. Accordingly, the staff finds the licensee's discussions in these areas to be acceptable.

2.9 TECHNICAL SPECIFICATION CHANGES The licensee identified several technical specification changes related to power uprating. Most of these had been previously identifiec, reviewed by the staff, and approved for 0FA and Vantage-5 fuel reloads. These continue to be acceptable, except as replaced in the uprating submittal. The uprating submittal identified four additional changes, as follows:

(a) Pages I and II Index - Delete the term "Design Thermal Power" and renumber the subsequent defined tenns. With the uprating, "design" and "rated" thermal powers are equal; the redundancy is eliminated by using the term "Rated Thermal Power" in the technical specifications wherever either term had been previously used. The staff finds this clerical change acceptable.

(b) Pages 1-2 through 1-7 Sections 1.10 through 1.41, Definitions -

The definition of "Design Thermal Power" is deleted and all subsequent definitions are renumbered.

This clerical change is acceptable as discussed above.

(c) Page 2-2, Figure 2.1-1, Rated Core Safety Limit - Four Loops in Operation. This figure is revised to reflect Reactor Core Safety Limits applicable to core thennal power of 3565 MWt.

This figure was revised using the staff approved trethodology identified in the Vantage-5 Reload Safety Evaluation (SE).

1 This methodology continues to be applicable at the uprated power. Therefore, the staff finds the revisions acceptable, j

i

. (d) Pages 2-7 through 2-10, Table 2.2-1 Table Notations 1 through 4.

Replace the term "Design Thermal Power" with theterm"RatedThermalPower"consistentwithitems(a) and(b)above. Also, revise numerical values in the notes as necessary to reflect overpressure and over-temperature delta T setpoint parameters appropriate to the uprated core thermal power of 3565 MWt. These values were revised using the staff-approved methodology identified in the Vantage-5 Reload SE.

i This methodology continues to be applicable at the uprated power. Therefore, the staff finds these revisions acceptable.

3.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 50.32, the Commission has determined that the issuance of I

this amendment will have no significant impact on the environnent (53 FR 5331).

4.0 CONCLUSION

The staff 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; and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the conmon defense and security or to the health and safety of the public.

5.0 ACKNOWLEDGEMENT Principal Contributors:

F. Orr, PDI-4 J. Raval, SPLB H. Walker, SPLB B. Elliot, EMTB J. Tsao, EMTB A. Lee, EMEB K. Parczewski, ECEB E. Branagan, PRPB G. 5taley, ESGB J. Knox, SELB T. Alexion PDIII-3 Dated: MAR 3 01993 m