ML20058J093

Provides Responses to 931119 RAI Re in Which DPC Proposed Changes to TS to Reduce Required Minimum Measured RCS Flow Rate

ML20058J093
Person / Time
Site:	Mcguire, Catawba, McGuire
Issue date:	12/03/1993
From:	Tuckman M DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9312130388
Download: ML20058J093 (14)

Text

{{#Wiki_filter:I I DukeIbwer Company M.S TLtxurv PO Box KK/.; Senior VicePresident CharMte, NC28201 WX NudearGeneration (704)382-2200 Ofhce {704)3824360Far V DUKEPOWER December 3, 1993 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Document Control Desk

Subject:

McGuire Nuclear Station Docket Numbers 50-369 and -370 Catawba Nuclear Station Docket Numbers 50-413 and -414 Technical Specification Change to Reduce Required Minimum Measured Reactor Coolant System Flow; Responses to NRC Questions By letter dated October 25, 1993, Duke Power Company proposed changes to Tecnical Specifications to reduce the required minimum measured reactor coolant system flow rate. The NRC, by letter dated November 19,

1993, requested additional' information to support the review cf the amendment request.

Attached are Duke's responses to the questions of the November 19, 1993 letter. If any additional information is required, please call 3cott Gewehr at (704) 382-7581. Very truly yours, 0, [d h cl % A% M. S. Tuckman 0900[O gb ,f 9312130388 931203 j PDR ADOCK 05000369 p PDR u n.u v nman

U. S.' Nuclear Regulatory Commission December 3, 1993 Page 2 cc: Mr. V..Nerses, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 14H25, OWFN Washington, D. C. 20555 Mr. R. E. Martin, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 14H25, OWFN Washington, D. C. _20555 Mr. S. D. Ebneter, Regional Administrator U.S. Nuclear Regulatory Commission - Region II 101 Marietta Street, lui - Suite 2900 Atlanta, Georgia 30323 G. F. Maxwell Senior Resident Inspector McGuire Nuclear Station R. J. Freudenberger Senior Resident Inspector Catawba Nuclear Station

4 i ATTACHMENT i

1. The application proposes to modify Technical specification (TS) 2.1.1 by replacing the definition of the Safety Limit with one from the B&W Standard Technical Specifications (STS) in NUREG-1430 and to modify TS Figure 2.1.1 by retitling it, to further define the acceptability of areas of operation on the figure, and by recalculating the lines on the figure to l

reflect the reduced assumption on core flow rate. Only a minimum of justification has been provided for the proposed change to the form of this TS. Significant additional information would be necessary to address apparent inconsistencies and potential misinterpretations. e These would include the following as applicable to Catawba and McGuire: Q: a. Appropriateness of the B&W STS definition for TS 2.1.1 without the B&W STS type figure for the Catawba and McGuire protection methodology. R: a. The intent of TS 2.1.1 is to meet the requirements of 10 CFR 50.36 (c) (1), which states that the technical specifications will include limits to "... reasonably protect the j integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity." The current text of TS 2.1.1 refers to TS Figure 2.1-1, which presents a l family of pressure curves. The family of pressure curves in TS Figure 2.1-1 is comprised i of a DNB line segment and a hot leg boiling line segment. The hot leg boiling line segments exist on the figure because they are more restrictive than the DNB limits and centerline fuel melt (CFM) limits within certain power level intervals. However, since safety limits, within the definition of 10 CFR 50.36, are limits to protect barriers to the release of radioactivity, hot leg boiling is not a safety limit. As opposed to DNB and CFM, hot leg boiling does not defeat a barrier to the release of radioactivity, and therefore hot leg boiling is not actually a safety limit. As stated on page 2-2 of WCAP-8745, " Design Bases for the Thermal-Overpower AT and Thermal Overtemperature AT Trip Functions", the hot leg boiling limit is imposed to assure that AT is proportional to core power. Further describing the hot leg boiling limit, the WCAP states in the same paragraph that "This limit is not a core protection limit". Therefore, to better comply with the requirement of 10 CFR 50.36 (c) (1), the text of TS 2.1.1.is changed to reflect i the true safety limits, i.e., DNB and CFM. The use of the B&W STS definition for TS 2.2.1 is appropriate because reference to the B&W STS is made solely for the purpose of demonstrating that extensive dialogue has l already occurred between the NRC and an industry group on the subject of this content of TS 2.2.1 meeting the intent of 10 CFR 50.36 (c) (1). The proposed revisions to the i McGuire and Catawba Technical Specifications are completely analogous to the B&W l STS in that the actual safety limits are given in the TS text, and a permissible transient operating region is graphically represented in the TS figure. The use of DNB and CFM as safety limits is appropriate for any B&W or Westinghouse PWR. ) Q: b. Appropriateness of the 5080 degrees F value as TS limit on fuel centerline temperature. j

c. Appropriateness of the burnup correction factor.

R: b and c. The original source of the fuel melt temperature and its burnup dependency is WCAP4065, " Melting Point of Irradiated Uranium-Dioxide", Westinghouse Electric

e Corporation, February 1965. Duke's application of this data is through the use of TACO 2 fuel performance code as given in BAW-10141P-A, " Fuel Pin Performance Analysis," Babcock and Wilcox Company, Revision 1, June 1983. The melt temperature and its burnup dependency are contained within the TACO 2 software. The melt temperature of 5080 'F is given in Chapter 4.4 of both the McGuire and Catawba FSARs. Although Chapter 4.4 of both the McGuire and Catawba FSARs states that the burnup dependence of the CFM temperature is 5.8 x 10-' 'F per MWD /MTU, which is based upon the results given in WCAP-6065, the approved TACO 2 methodology, given in DPC-NE-2001-A,

• Fuel Mechanical Reload Analysis Llethodology for Mark-BW Fuel",

Revision I approved October 1990, utilizes a more conservative burnup dependence of 6.5 x 10~' 'F per MWD /MTU, which is used for the proposed safety limit value. Q: d. Potential confusion introduced by the addition of a new box in the lower part of Figure 2.1-1 labeled "DNB Parameters Technical Specification" which is a function of only a constant, Tavg and power while the function and applicability of the remaining three dependent variable limits in the figure are unchanged. Please explain further how this proposed new box in the lower part of Figure 2.1-1 is to be interpreted for use. R: d. The addition of the box in the lower part of TS Figure 2.1-1 is intended to dispel confusion due to the seeming contradiction of the existing TS Figure 2.1-1 with the values given in TS Table 3.2-1 regarding the region of acceptable plant operation. The existing T5 Figure 2.1-1 format could easily be misinterpreted to allow normal operation at any condition below the family of pressure curves of RCS Tavg versus Power, when in fact, the region of acceptable normal operation is limited by the parameters given in TS Table i 3.2-1. This confusion arises because the existing figure does not distinguish acceptable normal operation from acceptable transient operation. Recognizing this deficiency in the - i format of the existing figure, the new Westinghouse Standard Technical Specifications (WSTS) include a horizontal line, below the family of pressure curves, which constitutes the maximum allowable temperature for normal operation. This WSTS figure is attached for convenience, in the proposed McGuire and Catawba figures, this horizontal line is the top of the box, and the temperature value is taken from TS Table 3.2-1. In addition, a vertical line which constitutes the right side of the box has been added to reflect maximum normal ful' power operation. The box is labeled "DNB Parameters Technical Specification" because that is the title of TS Table 3.2-1 from which the maximum allowable temperature for normal operation is taken. Q: e. With respect to the changes proposed to BASES page B 2-1 which state that the revisions to Figure 2.1-1 are more restrictive than the actual safety limit curves, please provide background information on how much more restrictive they are and on why this is so. I R: e. The changes on the Catawba and McGuire TS pages B 2-1 and B 2-2 reflect the fact i that the family of pressure curves of RCS Tavg versus Power given in TS Figure 2.1-1 are the curves used in the determination of the OTAT and OPAT trip equation constants (as described below in the response to question 6). Since the family of pressure curves in TS Figure 2.1-1 are used to determine RPS setpoints. the curves incorporate both instrument uncertainty and the DNBR correlation uncertainty, and additional margin, which makes the curves more restrictive than the actual safety limit curves. The actual safety limit curves would not include any uncertainty or margin, and would demonstrate

the actual conditions at which DNB or centerline fuel melt wouM c ir. In addition, as stated in the response to question La above, the family of pressure cu.e in TS Figure 2.1-1 are comprised, in part, of hot leg boiling limit lines, and hot leg boiling is not a safety limit. Q: f. In Attachment II of the application, it states that DNBR and centerline fuel temperature i limits have been added to TS 2.1.1 as Limiting Conditions for Operation (LCO),- replacing a reference to Figure 2.1-1, Reactor Core Safety Limit since this is considered to more accurately reflect the requirements of 10 CFR 50.36. Explain what particular requirements of 10 CFR 50.36 are referred to. R: f. Please see the response to question 1.a above. Q: Also, please provide an explicit discussion of any new information that would indicate that the subject parameters (DNB, fuel melt limitt, coolant enthalpy and RPS trip i setpoints) would not be controlled consistent with the design basis for the plants based i upon the current form for TS 2.1.1. P R: The subject parameters (DNB, fuel melt limits, coolant enthalpy and RPS trip setpoints) would be controlled in the same manner as in the past. The design basis for the plants is not being changed with the proposed revisions to TS 2.1.1. Instead, the proposed 3 revisions to TS 2.1.1 more accurately reflect the method by which the subject parameters (DNB, fuel melt limits, coolant enthalpy and RPS trip setpoints) have been and will continue to be controlled. The most significant proposed change to TS 2.1.1 stems from the recognition that it is not possible, solely by the use of TS Figure 2.1-1, to determine whether a safety limit has been violated. TS Figure 2.1-1 does not, and cannot within practicality, incorporate the impact of transient power distributions on transient DNBR. The proposed revision to TS 2.1-1 distinguishes the actual safety limits of DNB and CFM from the transient operating region described in TS Figure 2.1-1, which is used to establish the RPS setpoints to prevent DNB and CFM. 2. Q: Discuss the hardware changes mentioned as being necessary to accommodate revised RPS trip setpoints. l R: The hardware changes necessary to accommodate the revised RPS trip setpoints are limited to the calibration of the existing equipment and implementation of the new setpoint settings. These hardware changes are the same as in previous setpoint changes. No new equipment is necessary to accommodate the revised RPS trip setpoints. 3. Q: Discuss the error found in the OPAT values including its root cause and its effect on margins to trip setpoints. 3 R: The cause of the error found in OPAT values was an omission of the recalculation of the allowable value (AV) of the OPAT trip function (Note 4 of both the Catawba and McGuire TS Table 2.2-1). The AV can be viewed as the amount that the trip setpoint can drift due to random uncertainties and not violate the assumptions in the safety analysis. The AV is the smaller of two " trigger" values, which are typically designated T1 and T2. These trigger values are sums of the different measurement uncertainties associated with the trip function. Tl is the sum of the uncertainties associated with the electronic process f )

y 1 1 equipment. T2 is the total allowance less the sum of the sensor uncertainties. Since the j total allowance is the setpoint value assumed in the safety analysis less the TS setpoint value, a change in either the safety analysis value or TS value will affect the total allowance, and therefore the value of T2, and potentially affect the TS AV. When the AT trip constants were changed to the current values, the AVs were not recalculated and the previous AVs were retained. Upon calculation of the AT trip constants to reflect the impact of the reduction of minimum measured flow, it was realized that these retained AVs might not be conservative. It has been determined that the McGuire and Catawba OTAT AV was conservative, but the McGuire and Catawba OPAT AV was not conservative during this time. The non-conservatism in the OPAT trip setpoint due to the AV was at most 0.9 % Rated Thermal Power (RTP). However, administrative controls exist which cause the setpoint to be reset if the setpoint has drifted by an amount significantly less than either trigger. The i likelihood of this error affecting margin to trip is minimal because multiple channels of the OPAT trip function would have to simultaneously drift beyond the administrative l limits and beyond the correct AV to impair the function of the OPAT trip. Furthermore, since operation of the OPAT is not required for acceptable results in any FSAR Chapter 15 analysis, the impact of this error is minimal. To ensure that this error is not duplicated, a review of all reactor trip function setpoints and AVs is underway, and l training on this subject is scheduled. The correct value of the OPAT trigger for the OPAT trip equation constants based on a minimum measured flow of 385000 gpm is 2.2 % span (3.3 % RTP) for Catawba and 3.4 % RTP for McGuire. 4 Q: Clarify the RCS flow value used for the LOCA analyses. R: As discussed in the Duke Power response to NRC requests for additional information regarding the proposed changes to the Catawba ECCS surveillance requirements, large break and small break LOCA reanalyses were recently performed for McGuire and Catawba Nuclear Stations. The reanalyses were performed by Westinghouse using the NRC approved LOCA methodologies for McGuire and Catawba. An initial RCS flow of 374,000 gpm was assumed for the LOCA reanalyses. The peak clad temperatures from these reanalyses are well below the 2200 *F 10 CFR 50.46 acceptance criterion. The Duke Power responses to NRC requests for additional information are given in the following. Letter from D. L. Rehn, Duke Power Company, to U. S. Nuclear Regulatory Commission, titled " Response to Request for AdditionalInformation, CLA Water Volume and ECCS Subsystem Surveillance Requirements," November 15, 1993, for Docket Numbers 50-413 and 50-414. Letter from D. L. Rehn, Duke Power Company, to U. S. Nuclear Regulatory Commission, titled " Response to Request for AdditionalInformation, CLA Water Volume and ECCS Subsystem Surveillance Requirements," November 22,1993, for Docket Numbers 50-413 and 50-414 1 5. Q: Provide a corrected version of the current FSAR Table 15-4 for the forthcoming Catawba Unit 1 Cycle 8. v =

R: The corrected version of FSAR Table 15-4 which will be included in the next FSAR update is attached. l 6. Q: For the various changes in Table 2.2-1 (examples: K, K, K, K., K. and q, - q,) provide 4 2 3 information on the reasons for the changes and the approved method used to arrive at the new values. Also provide an explanation of the method and reason for the change made ^ in Notes 2 and 4 for the channel maximum trip setpoint. R: The changes to Table 2.2-1 are as follows: TS Table 2.2-1 Item 11. Item 11 in both the Catawba and McGuire TS refers to the minimum measured flow. The 95500 gpm value is the single loop equivalent of the four loop minimum measured flow of 382000 gpm, just as the 96,250 gpm value is the single loop equivalent of the four loop minimum measured flow of 385000 gpm. TS Table 2.2-1 Notes 1 and 3 (Catawba) and Notes I and 2 (McGuire). The 1 following description was submitted to the NRC via the April 26,1993 letter from T. C. McMeekin, Duke Power Company, to U. S. Nuclear Regulatory Commission, titled " Supplement to Technical Specification Amendment Relocation of Cycle-Specific Limits to Core Operating Limits Report (COLR)" for Docket Numbers 50-369 and 50-370 for McGuire Nuclear Station and Docket Numbers 50-413 and 50414 for Catawba Nuclear Station. It is repeated here for convenience. The methodology described below is how Duke Power Company currently arrives at values for the OTAT and OPAT parameters. This is one of many equally valid methods for determining these parameters. Once a new preliminary set of overtemperature and oveipower setpoint equation parameters is selected, they must be evaluated by reanalyzing the appropriate transient analyses with the new setpoint parameters. The transient analyses utilized to validate these new setpoints are performed using the NRC approved methodology documented in l Duke Power Company topical reports DPC-NE-3002-A, "FSAR Chapter 15 System Transient Analysis Methodology," DPC-NE-3001-A, "McGuire/ Catawba Nuclear Station Multidimensional Reactor Transients and Safety Analysis Physics l Parameters Methodology" and DPC-NE-3000, " Thermal-Hydraulic Transient Analysis Methodology," approved by the NRC for McGuire/ Catawba use in. November 1991. Once the analysis is performed and the new setpoint constants demonstrate they are capable of protecting the plant during the appropriate transients, the new setpoint parameters may be used. In other words, there are many possible methods for selecting these setpoint parameters and regardless of the method used they are not considered valid parameters until they are proven capable of protecting the plant under transient conditions with the NRC approved methodology described in the topicals above. The OPAT parameter K, is not being relocated to the COLR since it currently is not calculated as part of the reload design methodology described below. The purpose of the OTAT trip function is to protect the reactor core against DNB s 4 < ~

t and hot leg boiling for any combination of power, pressure and temperature during normal operation and transient conditions. The parameter values for K,, K,, K,, K., K., and f,(Al) breakpoints and slopes are calculated using as inputs the DNB core limit lines at different pressures, axial offset versus power limits, and various nominal operating condition parameters. For steady state conditions and a reference power shape these inputs are used to calculate the overtemperature and overpower AT setpoints based on the following constraints: Thermal overtemperature limits, which provide protect!r against DNB and hot leg boiling. Pressurizer low pressure and high pressure safety limits, which limit the range of pressures over which the overtemperature AT and overpower AT trips must function. 1 The locus of conditions where the steam generator safety valves open, which places a limit on the primary side temperature based on the steam generator design pressure and the primary to secondary heat transfer capacity.

• Thermal overpower limit, which protects against centerline fuel melt.

Multiple K, and K. pairs and corresponding K,, K,, and K,s are calculated such that they meet the above constraints. Using engineeringjudgment and plant operating experience a set of Ks is then chosen from these allowable sets. The chosen set is then evaluated in the transient analysis using the methodology described in topical reports DPC-NE-3002-A, "FSAR Chapter 15 System Transient Analysis Methodology and DPC-NE-3000, " Thermal-Hydraulic Transient Analysis Methodology," to determine whether the K parameter values are capab'e of protecting the plant during the appropriate transients. t The thermal overtemperature limits described above are calculated for zero axial imbalance. Therefore, once the setpoint constants are calculated, the f,(AI) trip 3 reset function for the OTAT equation is determined using two axial offset versus power envelopes (typically 100% and 118% power) supplied by nuclear design analyses. This function is determined as described in the technical justification using the methodology described in Chapter 4 of the NRC approved report DPC-NE-2011-P-A. A value of imbalance, Al, and a point on the DNB line for a given pressure which is not bounded by the exit boiling line, steam generator safety valve line, or OPAT setpoint equation is selected. This point is compared with the OTAT setpoint and the amount the setpoint must be lowered, if at all, to bound this point is calculated. This process is repeated for this Al for the other non-bounded DNB points at this and other pressures. The largest reduction in the OTAT setpoint equation required to bound the imbalance corrected DNB points becomes the f,(AI) enalty for this particular value of AI. This process is repeated for a range of vis that will envelope all the expected skewed axial power distributions. The f,(AI) breakpoints and slopes are then selected in a manner such that they bound the calculated f,(AI) penalties which J

r 1 were determined from the two axial offset envelopes. The purpose of the OPAT trip function is to provide protection against fuel i center-line melt (CFM) during normal operation and Condition Il transients. The AT trip setpoint for this trip function is typically set at 118 %FP and is deterrnined as described above. The trip reset portion of this trip l function, f,(AI), is designed to lower the trip setpoint when measured imbalances exceed predetermined values. Once highly skewed power distributions lead to high kw/ft values, a f,(AI) trip function can be developed to prevent CFM limits j from being exceeded at large imbalances, or to increase the available margin to the CFM limit for highly skewed power distributions. Current core designs do not challenge the CFM limits and therefore a f,(AI) i penalty is not required. However, from an operational and design standpoint, it is desirable to eliminate from consideration power distributions with high imbalances. Therefore, a f,(AI) trip reset function was established to trip the t reactor at high imbalances. The breal 0ints and slopes of this function were l 7 arbitrarily chosen to limit the power distributions that need to be considered during the design of the reactor core and to increase the margin to the CFM limit and therefore reducing the probability of the CFM Technical Specification { surveillance limits being violated. } In the event that it would to be necessary to establish a f,(AI) trip reset function j because CFM limits were being exceeded, one possible method of determining l this trip reset function would be to develop a kw/ft versus imbalance envelope based on the analysis of Condition II transients such that this envelope would j conservatively bound expected transient peaks. This envelope would next be used to determine the f,(AI) penalty as a function of imbalance by comparing the CFM i kwift limit against the maximum expected peak for a given imbalance. The f,(AI) i trip reset function would then be developed such that the breakpoints and slopes i bouu:1 the f,(AI) penalties developed from the kw/ft versus imbalance envelope which is generated in a manner similar to the f,(41) reset function described above. The dynamic terms (t,, t,, t,, t, t,, t.) in the OTAT and OPAT setpoint equations compensate for inherent instrument delays and piping lags between the I reactor core and the temperature sensors. Lead-lag and rate-lag compensations are required for the following reasons: { To offset measured RTD instrumentation time delays. To ensure the protection system response time is within the limits required by the accident analyses. In addition, the dynamic terms are used as noise filters and to decrease the likelihood of an unnecessary reactor trip following a large load rejection. Models have been created to examine the effects of different sets of t values used

w in the lead-lag, lag, and rate lag functions of the overtemperature and overpower equations. These models are the same as those given in EPRI NP-1850-CCM-A, i "RETRAN A Program for Transient Thermal-Hydraulic Analysis of Complex -l Fluid Flow Systems." Using these models the t values are selected in a manner such that the optimum response of the OTAT and OPAT setpoints to changes in plant variables is obtained while satisfying the transient analyses acceptance { criteria. The acceptability of the chosen t values is determined utilizing these same mathematical models, which are also c(mtained in the transient analysis models described in DPC-NE-3000, " Thermal-Hydraulic Transient Analysis Methodology."

• TS Table 2.2-1 Note 2 (Catawba) and Note 3 (McGuire). As stated above in the i

response to question 3, a change in the trip setpoint might change the TS allowable value (AV). He AV of the OTAT trip setpoint remains conservative with the revised OTAT trip equation constants. Previously, the units of the AV were expressed in % span in the Catawba TS, but since the text of Note 2 gave the units of the AV in %, it could be unclear whether the units were intended to be % span or % RTP. To alleviate any ambiguity, the AV is converted from l units of % span to % RTP, i.e., from 3.0 % span to 4.5 % RTP. Although the flow reduction change has not been proposed for Catawba Unit 2, the conversion in AV units to % RTP is proposed for consistency. In addition, the existing l OTAT AV for McGuire is overly conservative for the OTAT setpoint constants l based on 382000 gpm, and therefore an increase in the AV to 4.4 % RTP is proposed. A summary of the AV values is provided below. e TS Table 2.2-1 Note 4. The existing AV of both the McGuire and Catawba OPAT trip setpoint is not conservative with the revised OPAT trip equation constants based on a reduced minimum measured flow. For the revised OPAT trip equation constants, an AV of 2.0% span for Catawba is required. However, i as stated above in the discussion on the changes to Note 2 (Catawba), the units of i the OPDT AV at Catawba are changed to % RTP. Therefore the proposed AV for Unit 1 is 3.0 % RTP. As stated in the response to Question 3 above, the correct value for the Catawba OPAT AV based on a minimum i meast' red flow of 385000 gpm is 2.2 % span. This value is also l converted to units of % RTP,i.e., the proposed AV for Catawba Unit 2 is 3.3 % RTP. For McGuire, the required OPAT AV for the trip equation constants based on 382000 gpm is 3.0 % RTP. A summary of the AV i values is provided below. Current TS Required AV Required AV Proposed AV i AV for 3E5,000 for 382,000 l gpm gpm } f CNS 1 3.0 % span 3.0 % span 3.0 % span 4.5 % RTP OTAT l F f h

i CNS 2 3.0 % span 3.0 % span 3.0 % span 4.5 % RTP OTAT CNS 1 2.8 % span 2.2 % span 2.0 % span 3.0 % RTP OPAT i CNS 2 2.8 % span 2.2 % span 2.0 % span 3.3 % RTP OPAT MNS OTAT 3.6 % RTP 4.7 % RTP 4.4 % RTP 4.4 % RTP MNS OPAT 4.2 % RTP 3.4 % RTP 3.0 % RTP 3.0 % RTP 7. Figure 3.2-1, Reactor Coolant System Total Flow Rate Versus Rated Thermal Power, has been modified, with the RCS flow rate for 100% power being changed from 385,000 gpm to 382,000 gpm. There are further changes in RCS flow rate at other fractions of Thermal Power. In Attachment il you state that plugged steam generator tubes and hot leg streaming affect accurate mear.urement of flow and as a result it will be difficult to ensure meeting the minimum flow requirement (Table 2.2-1 Item 12, as annotated) required by the TS to maintain 100% power operation. Therefore you propose reduction in i minimum measured flow for McGuire Units 1 and 2 and for Catawba Unit 1. Please provide the following information: Q: 1) What are the analyses that have been performed to justify the reduction in minimum RCS flow at 100% thermal power? Are these analyses made using approved methods and if so what are these methods? Note that this application does not identify for any of the proposed TS changes the methodology used to evaluate the change. R: 1) For FSAR Chapter 15 transients in which a low initial flow assumption is conservative, the following FSAR Chapter 15 analyses have been explicitly analyzed with an initial RCS flow assumption which is less than or equal to a minimum measured f!ow of 382000 gpm: .I 15.1.5 Steam System Piping Failure j 15.2.3hTurbine Trip - Peak Primary Pressure 15.2.8 Feedwater System Pipe Bre.tk 15.3.1 Partial Loss of Reactor Coolant Flow 15.3.2 Complete Loss of Reactor Coolant Flow 15.3.3 Locked Rotor 15.4.1 Uncontrolled Bank Withdrawal from Suberitical

i e 15.4.2 Uncontrolled Bank Withdrawal at Power 15.4.3 Rod Assembly Misoperation 15.4.8 Rod Ejection 15.6.3 Steam Generator Tube Rupture 15.6.5 Loss of Coolant Accident The LOCA analyses were performed by Westinghouse using the following approved -{ LOCA methodologies for McGuire and Catawba: l Kabadi, J. N., et al, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," WCAP-10266P-A, Rev. 2, March 1987. i N. Lee, et al,

• Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054P-A, August 1985.

The other analyses were performed by Duke Power using the approved methods described in the following topical reports: i DPC-NE-3000P-A, Rev.1, " Thermal-Ilydraulic Transient Analysis Methodology," November 1991. DPC-NE-300lP-A, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," November 1991. DPC-NE-3002-A, "FSAR Chapter 15 System Transient Analysis Methodology," November 1991. l In addition, the approved transient analysis methodology in Duke Pcwer Topical Report DPC-NE-3002-A states that the following events are bounded by ot'aer more limiting events and are not analyzed. Therefore, the results of these FSAR Chapter 15 events are not affected by a change in the minimum measured flow: 15.1.1 Reduction in Feedwater Temperature 15.1.4 Inadvertent Opening of a Steam Generator Relief Valve r 15.2.2 Loss of External Load 15.2.4 Inadvertent Closure of Main Steam Isolation Valves 15.2.5 Loss of Condenser Vacuum and Events Causing Turbine Trip 15.2.6 Loss of Non-emergency AC Power 15.2.7 Loss of Normal Feedwater Flow 15.3.4 Reactor Coolant Pump Shaft Break l 15.5.1 Inadvertent Operation of ECCS 15 5.2 Increase in Reactor Coolant Inventory Q: 2) For the changes made in RCS flow at reduced thermal power, how are the RCS flow values determined for reduced thermal power? If sensitivity values are used for the effect of DNB please provide the method used with references. R: 2) The RCS flow values for reduced power are determined using the same 2% power per l t

l i ~ l 1% flow reduction factor used in the existing TS figure. The 2 to 1 power-flow tradeoff i was approved in Amendment 34 to NPF-35 for Catawba Unit I and Amendment 25 to j NPF-52 for Catawba Unit 2 via SER dated November 24,1987. The 2 to 1 power-flow tradeoff was approved in Amendment 28 to NPF-9 for McGuire Unit I and Amendment 9 to NPF-17 for McGuire Unit 2 via SER dated February 4,1984. Q: 3) Please explain what fraction of the change is for the effect of plugged steam generator l tubes and what fraction is for hot leg streaming. R: 3) The amount of the proposed reduction in minimum measured flow is not made for the purpose of accounting for a specific amount of measured flow reduction due to steam l jl generator tube plugging or hot leg streaming. Rather, these two effects are mentioned in the submittal to identify, in a qualitative sense, the main contributors to the decline of the measured flow that has been experienced at McGuire and Catawba. The amount of the { proposed flow reduction is based upon the value of the RCS flow assumption in the safety l analyses. De flow assumption was chosen somewhat arbitrarily when the safety analyses j were performed, with the expectation that measured flow would remain above that value j in the future. t I 8. Q: Please provide background information on the changes on TS page B 2-2 on the top of the page and also on the insert. For the insert you mention that when the combination of reactor power and axial power imbalance is not within tolerance, tlie OPAT trip function will provide the necessary fuel pin centerline temperature protection. Explain the j mechanism for measuring this and its equivalence to fuel centerline temperature, is the l imbalance measurement by computer? Has this method been approved before in a similar 'l application? If so, provide the reference. + ? R: For the changes to the Catawba and McGuire TS pages B 2-1 and B 2-2, please see the response to question 1.e above. For the insert which mentions the measurement of reactor power and axial power imbalance, reactor power is measured by the' AT instrumentation and the axial power J imbalance is measured by the Al instrumentation. The equivalence to fuel l centerline temperature is given above in the description of the determination of i the OTAT and OPAT trip equation constants in the response to question 6. The j e use of the f,(al) function was approved in the McGuire 1 Cycle 8 reload SER dated November 11,1991, and in the Catawba 1 Cycle 7 reload SER dated September 14, l 1992. 9. Q: Near the middle of page 2 of Attachment II, for the section on Thermal Hydraulic l Design, FSAR Section 4.4, it states that "The reduced flow rate resulted in a slight reduction of the margin in the core DNB limits " Please indicate what the slight reduction f in margin amounted to and what the amount of margin was before. R: The section on Thermal-Hydraulic Design, FSAR Section 4.4, should have stated that the l reduced flow rate resulted in a slight reduction in the core DNB limits (the allowable i temperatures are slightly lower as expected for a lower flow rate). There is no reduction in DNB margin. The core DNB limits are still based on the Design DNBR limit of 1.55 as explained in DPC-NE-2004P-A, "McGuire and Catawba Nuclear Stations Core k, +

4 +- ) i Thermal-Hydraulb Methodology Using VIPRE-01," Duke Power Company, December 1991. The DNB design basis is met using the Design DNBR limit as discussed in the bases for Technical Specification 2.1.1. 10. Q: Starting near the middle of page 3 of Attachment II, Accident Analyses, FSAR Chapter 15, a number of analyses that were reanalyzed are discussed. Please provide information and a listing of the codes used for these analyses and indicate if they are approved methods. l R: Please see the response given to the first part of question 7. The listing of codes used m the FSAR Chapter 15 analyses are given in FSAR Table 15-3, " Summary of Computer j Codes and Methodologies Used in Accident Analyses". The corrected version of FSAR Table 15-3 which will be included in the next FSAR update is attached for convenience. l 1 I i - i t i J I I - t 4 0 ...,}}