ML052440398

Supplemental Information Regarding License Amendment Application to Support Mark B-HTP Fuel Design for Cycle 15 (License Amendment Request (LAR) 05-0002; TAC No. MC6888)

ML052440398
Person / Time
Site:	Davis Besse
Issue date:	08/28/2005
From:	Bezilla M FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
3166, TAC MC6888
Download: ML052440398 (33)

Text

-1FENOC FENOC 5501 North State Route 2 FirstEnergy Nuclear Operating Company Oak Harbor, Ohio 43449 Mark B. Bezilla 419-321-7676 Vice President - Nuclear Fax:419321-7582 Docket Number 50-346 10 CFR 2.390 10 CFR 50.90 License Number NPF-3 Serial Number 3166 August 28, 2005 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

Davis-Besse Nuclear Power Station Supplemental Information Regarding License Amendment Application to Support Mark B-HTP Fuel Design for Cycle 15 (License Amendment Request (LAR) 05-0002; TAC No. MC6888)

Ladies and Gentlemen:

This letter responds to two NRC requests for additional information (RAI) regarding LAR 05-0002.

By letter dated May 2, 2005 (Serial Number 3131), the FirstEnergy Nuclear Operating Company (FENOC) submitted an application for amendment of the Operating License, Appendix A, Technical Specifications (TS) for the Davis-Besse Nuclear Power Station (DBNPS). The proposed amendment would revise TS Section 2.1.1, "Safety Limits -

Reactor Core," and TS Section 2.2.1, "Limiting Safety System Settings - Reactor Protection System Setpoints" to support use of the Framatome Mark B-HTP Fuel design for Cycle 15, which is scheduled to begin following refueling in March 2006.

On June 9, 2005, and June 15, 2005, FENOC received informal requests from the NRC staff for additional information regarding the license amendment application.

Enclosure I provides the proprietary response to the June 9, 2005 request. Enclosure 2 provides the response to the June 15, 2005 request. Enclosures 3 and 4 provide the non-proprietary versions of Enclosures I and 2, respectively. Enclosure 5 provides specific responses to the concerns and request for additional information presented in the NRC letter to the Nuclear Energy Institute dated March 31, 2005, concerning trip setpoints and allowable values (ADAMS ML050870008). Enclosure 6 provides an affidavit supporting the proprietary status of Enclosures 1 and 2. A list of regulatory commitments made in this letter is included in Enclosure 7.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 Page 2 Certain information included in Enclosures 1 and 2 is confidential proprietary information which FENOC requests be withheld from public disclosure pursuant to 10 CFR 2.390. Redacted versions of Enclosures I and 2 are provided in Enclosures 3 and 4, respectively. Proprietary information is denoted by I brackets ].

Public disclosure of this information is likely to cause harm to the competitive position of AREVA/Framatome ANP. Information identified as proprietary in Enclosures I through 4 was originally identified as proprietary in the AREVA/Framatome ANP Engineering Information Record 51-5069986, Revision 0, "RAI Response Information for Davis Besse Cycle 15 TS Change LAR," dated August 12, 2005. An affidavit supporting the request for withholding this information is provided in Enclosure 6, as required by 10 CFR 2.390(b)(iii). Although the affidavit provided in Enclosure 6 was prepared for AREVA/Framatome ANP Engineering Information Record 51-5069986, the 51-5069986 document was prepared exclusively for the purpose of supplying information to be used in this letter, and therefore the text identified as proprietary in this letter is equivalent to corresponding text described in the affidavit.

The supplemental information provided in this letter does not affect the conclusion of the license amendment application that the proposed changes do not involve a significant hazards consideration and do not have an adverse effect on nuclear safety.

As described in Enclosure 5, FENOC will propose an additional TS change to add a footnote to TS Table 4.3-1, "Reactor Protection System Instrumentation Surveillance Requirements," applicable to Functional Unit 7, "RC pressure-temperature," regarding the as-left instrument setting. This change will be consistent with the NRC position described in the March 31, 2005 letter from James A. Lyons, NRC, to Mr. Alex Marion, Nuclear Energy Institute. The proposed change will require review by the Davis-Besse Plant Operations Review Committee and Company Nuclear Review Board and will be submitted under a separate letter.

Should you have any questions or require additional information, please contact Mr.

Henry L. Hegrat, Supervisor - Fleet Licensing, at (330) 315-6944.

The statements contained in this submittal, including its associated enclosures, are correct to the best of my knowledge and belief. I am authorized by the FirstEnergy Nuclear Operating Company to make this submittal. I declare under penalty of perjury that the foregoing is true and correct.

Executed on: 39 c4 By: Ma 3V / $ P t MSH

Docket Number 50-346 License Number NPF-3 Serial Number 3166 Page 3 : Proprietary Response to June 9, 2005 Request for Additional Information : Proprietary Response to June 15, 2005 Request for Additional Information : Non-Proprietary Response to June 9, 2005 Request for Additional Information : Non-Proprietary Response to June 15, 2005 Request for Additional Information : Response to Concerns and RAI presented in the NRC letter to NEI, dated March 31, 2005, concerning trip setpoints and allowable values (ADAMS ML050870008). : Affidavit Supporting Proprietary Status of Enclosures 1 and 2 : Commitment List cc: J. L. Caldwell, Regional Administrator, NRC Region III W. A. Macon, DB-1 NRC/NRR Project Manager N. Dragani, Executive Director, Ohio Emergency Management Agency, State of Ohio (NRC Liaison); w/o Enclosure 1, Enclosure 2 C. S. Thomas, DB-1 NRC Senior Resident Inspector Utility Radiological Safety Board; w/o Enclosure 1, Enclosure 2

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 9, 2005 - non-proprietary Page 1 of 11 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION DATED JUNE 9,2005 REGARDING LICENSE AMENDMENT REQUEST 05-0002 FOR DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1

1. There is no description in the May 2, 2005, submittalregardingthe updated core thermnal-Ihydraulicanalysisperformedfor the reactorcore safety limit in the TS Figure 2.1-1 to support the use ofMark B-HTPfuel assemblies, exceptfor the statements that (I) the necessaryDNB margin is necessary to offset the transitionDNB pqnaltyfor Cycle 15 and subsequent cycles with co-existence of the Mark-B HTPfiel assemblies and the existingfuel assemblies in the core; and (2) the safety limit curve is based on the design hot channelfactors with potentialfilel densificationz andfiuel rod bowving effects.

A. Confirmn whether the core thermal-hydrazulicanalysis is performed using the statisticalcore design methodology. If so, provide a derivation of the SCD DNBR limit with the BHTP correlationto accountfor the uncertainties of the parameters included in the SCD.

RESPONSE L.A The core thermal-hydraulic analyses, performed to determine the DNB performance of Davis-Besse Cycle 15, utilized the Statistical Core Design (SCD) methodology that is discussed and approved inBAW-10187P-A, Statistical CoreDesignfor B&W-Designed 177FA Plants (Reference 1). The SCD methodology was

.incorporated into the approved Safety Criteria and Methodology topical report as Appendix D in BAW-1 01 79P-A, Safety Criteriaand MethodologyforAcceptable Cycle Reload Analyses (Reference 2).

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 9, 2005 - non-proprietary Page 2 of I1 The SCD DNBR limit, commonly referred to as the Statistical Design Limit (SDL),

was derived in accordance with the methodology discussed in BAW-10187P-A, Statistical Core Designfor B& W-Designed 1 77FA Plants, by first defining variables and their associated uncertainties. This included the distribution characteristic (uniform or normal) of each uncertainty. Table 1.A-I provides a list of the variable uncertainties that have been incorporated into the SDL determination for Davis-Besse Cycle 15. A Response Surface Model was established that provides a DNB response as a function of the independent variables. Next, a Monte Carlo propagation of uncertainties was performed at various typical and limiting core states to determine the respective core state SDLs from the coefficient of variations. The maximum coefficient of variation for tested core states defined the hot pin SDL. Core-wide DNB protection was established by determining the corresponding SDLs for various core radial peaking distributions. The maximum SDL determined for the core-wide assessment defined the core-wide SDL. The larger of the hot pin SDL and core-wide SDL became the SDL for DNB protection of the core. The SDL determined for Davis Besse Cycle 15, using Table l.A-1 variables and uncertainties, was found to be I I BHTP. For analysis application, the variables in Table I.A-1 were analyzed without the corresponding uncertainties, using the LYNXT core thermal-hydraulic code (BAW-10156A, LYNXT- Core Transient Thernmal-HydrauilicProgram)to obtain a DNBR prediction.

To provide an accomodation for cycle-specific DNB margin needs, a Thermal Design Limit (TDL) of I l was established for Davis-Besse Cycle 15 (see Response 1.D) that is sufficiently greater than the SDL. The TDL is used as the DNBR criterion (limit) for assessing the acceptability of DNB predictions for core protection.

Table l.A-I Uncertainty Parameters Used to Derive the Statistical Design Limit for Davis-B esse Cycle 15 1.

2.

3.

4 ._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

5.

6.

8.=

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 9, 2005 - non-proprietary Page 3 of 11 B. Provide a description of the magnitude and the derivation of the transitioncore penalty assignedin the analysis.

RESPONSE L.B The DNBR transition core penalty has been determined for Davis-Besse based on the number of Mark-B-HTP fuel assemblies residing in the core. For Cycle 15 there will be 76 Mark-B-HTP fuel assemblies in the core and the transition core DNB penalty will be approximately I l DNB points, where I DNB point = 0.01. Figure L.B-1 provides the relationship of the transition core penalty versus the number of Mark-B-HTP fuel assemblies in the core. The transition core penalty will be applied to the DNB margin for predictions made using a full core of Mark-B-HTP fuel.

The transition core penalty relationship was developed by first quantifying the DNB performance for a full core of Mark-B-HTP fuel. The DNB performance was analyzed for statepoint conditions from the core safety limits, for the limiting Condition M/II DNB transient, and for the range of axial power shapes that are used for establishing the core safety limits and core operating limits. Next, a transition core model was examined in which a certain number of the Mark-B-HTP fuel assemblies were placed into the core of the resident fuel design in a conservative manner to allow flow diversion out of the limiting Mark-B-HTP fuel assembly.

Placement of the limiting Mark-B-HTP fuel assembly at the center of the core and surrounding it with the resident fuel design, with the remaining Mark-B-HTP fuel assemblies placed on the core periphery, results in the lowest DNB prediction for the limiting Mark-B-HTP fuel assembly in the transition core. This DNB behavior is a result of the Mark-B-HTP fuel assembly having a higher pressure drop than the resident fuel design, thereby creating a higher flow diversion out of the Mark-B-HTP fuel.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 9, 2005 - non-proprietary Page 4 of 11 The transition core model DNB performance was analyzed for the same range of statepoint conditions, transients, and range of axial power shapes studied for the full core model. The process was repeated for transition cores containing from 1 to 129 Mark-B-HTP fuel assemblies. The largest DNBR difference between the limiting Mark-B-HTP fuel rod in a full core model (of Mark-B-HTP fuel) and a specific transition core model for all of the statepoints, transients, and axial power shapes was defined as the transition core penalty for the specific transition core model.

The tabulation of this largest DNBR difference for the transition core containing 1 to 129 Mark-B-HTP fuel assemblies resulted in the line presented in Figure l.B-1.

The conservatism added by surrounding the limiting Mark-B-HTP fuel assembly with the lower pressure drop fuel design was quantified for the transition core model containing 85 Mark-B-HTP fuel assemblies. Note in Figure 11.B-l that use of a typical checkerboard reload pattern for the fresh fuel results in an approximately

[ I DNB point smaller transition core penalty.

As stated above, the transition core penalty for Davis-Besse Cycle 15, using 76 Mark-B-HTP fuel assemblies, will be [ ] DNB points. The transition core penalty will be smaller in subsequent cycles as the number of Mark-B-HTP fuel assemblies increases in the core.

C. Providea breakdown of DNBR margins included to accommodate the effects offuel denpsit cation and rod bow for the Mark B-HTPfuel design, and discuss how these values are derived.

RESPONSE L.C The impact of fuel densification is discussed in Section 6.2.7 in Reference 2. The thermal-hydraulic core models for the Davis-Besse DNB predictions use a "cold" nominal fuel stack height. The thermal expansion effect on the fuel stack height is greater than the densification effect, therefore, the application of a "cold" nominal fuel stack in the DNB core models yields a conservative heat flux for DNB predictions. Since the accommodation of the fuel densification effect is within the core model basis, there is no need for additional DNBR margin to offset the effect.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 9, 2005 - non-proprietary Page 5 of 11 Fuel rod bow is discussed in Section 6.2.6 in Reference 2. As noted in Section 6.2.6 of Reference 2, a

] Since accommodation of the fuel rod bow effect is captured in the development of the SDL, there is no need for additional DNBR margin to offset the effect.

D. In consideration offuel deisification,rod bow transitioncore penalty and SCD, and with the ininimwn DNBR limit of 1. 132 for the BHTP correlation,what is the safety analysis DNBR limit usedfor the core safety limit curve and the safety analyses of the design basis transients and accidents?

RESPONSE 1.D The safety analysis DNBR limit for Davis-Besse is I ] BHTP.

This limit, more commonly known as the Thermal Design Limit (TDL), is sufficiently high to cover the BHTP correlation design limit (1.132), the impact of the statistical treatment of the state variable uncertainties I l, and the transition core penalty I 1,yet is small enough to yield acceptable transient analyses results.

Figure 1.D- 1 shows the DNB margin between the TDL [ ] and the SDL I ] of approximately I ] DNB points. The I ] DNB points will be used to offset the I I DNB point transition core penalty, with the remaining I ] DNB points to be available to offset cycle-specific needs.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 9, 2005 - non-proprietary Page 6 of II

2. The staff's revieivfound that the Reactor ProtectionSystem (RPS) "Allowable Valufe" of Functional Unit 7, "RCpressure-temperature"tripfiunction in TS Table 2.1-1 is essentially the same as the "RCPressure Temperature Trip" curve in TS Figure2.1-1.

RESPONSE 2 This is true. The equation listed in TS Table 2.2-1 defines the line that is plotted in TS Figure 2.1-1. There maybe a small difference if the data points on Figure 2.1-1 are used to verify the equation that defines the Reactor Coolant Pressure-Temperature Trip. This is due to the number of significant digits that are listed on the figure. It should be noted that the box that defines the ACCEPTABLE OPERATION window represents the "Allowable Values" for the pressure- and temperature-related reactor trip functions. The changes proposed by the May 2, 2005 license amendment application are only required due to the implementation of the Mark B-HTP fuel design. The Mark B-HTP fuel design incorporates a more robust intermediate spacer grid design that has a higher resistance to flow. This translates to a more restrictive pressure-temperature limit that is used to define this trip.

A. Clarify whether the various tripfunctions (i.e., RC High PressureTrip, RCLow PressureTrip, RC High Temperature Trip, and RC Pressure-TemperatureTrip) in TS Figure 2.1-1 are intended to be the Analytical Limits, actual Trip Setpoint, or the Allowable Valutes of the Reactor ProtectionSystem 'sfor each of these trip functions.

RESPONSE 2.A Each of the trip functions identified in TS Table 2.2-1 and TS Figure 2.1-1 represent the "Allowable Values". On TS Figure 2.1-1, the line identified as the Safety Limit represents the "Analytical Limit" for the Reactor Coolant (RC)

Pressure-Temperature trip.

B. Describe the derivationand the values of the "AnalyticalLimit" utsed in the safety analyses of the design basis transientsfor the "RC Pressure-TemlperatureTrip,"

and the RPS "Trip Setpoint"for the same tripfiunction.

RESPONSE 2.B A description of the Variable Low Reactor Coolant Pressure Trip is provided in Section 7.6 of BAW-101 79, Revision 1, Safety Criteriaand Methodologyfor Acceptable Cycle Reload Analyses (Reference 2). A general description is provided

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 9, 2005 - non-proprietary Page 7 of 11 below. To have a common understanding, however, the following definitions are presented:

Analytical Limit - This is the value that is used in the accident/safety analysis.

When these limits are modeled, the consequences of the transient being analyzed will successfully meet the acceptance criteria for the specified event as listed in the Updated Safety Analysis Report.

Allowable Value - The Allowable Value is derived from the Analytical Limit by incorporating instrument string uncertainties. These uncertainties include hardware and process measurement error for the individual components in the instrument string for a given trip function. Random errors are combined statistically whereas the non-random errors are combined by linear addition.

This ensures that if the reactor is tripped on, or before, the Analytical Limit is reached, the protective systems will ensure that the plant conditions will not exceed what has been previously analyzed and found acceptable.

Trip Setpoint - The trip setpoint is also referred to as the "Field Setpoint". The Field Setpoint is the actual reactor trip setpoint for the reactor protection system. This setpoint is more restrictive than the "Allowable Value" and accounts for the ability to set a specific value for the setpoint. This includes setpoint "setability" of the setpoint, setpoint drift (as-left/as-found), measurement and test equipment (MTE) uncertainty, and calibration errors.

The Safety Limit shown on TS Figure 2.1-1 corresponds to the "Analytical Limit" for the Reactor Coolant Pressure-Temperature Trip function. This line corresponds to the hot leg pressure/core exit temperature DNB limits for steady-state operation for the limiting RCP combination at the respective maximum allowed core overpower condition. This line defines the pressure-temperature combination that will result in a minimum DNBR equal to, or greater than the Thermal Design Limit (TDL) for the appropriate Critical Heat Flux correlation. The region above and to the left of this line will result in a minimum DNBR greater than the TDL.

The string errors for the individual components (hardware and process measurement errors) are combined and added to the Safety Limit to define the "Allowable Value," as described in the response to Question 2.C in this enclosure.

The Trip Setpoint is derived from the Allowable Value by adding uncertainties for MTE, drift, setability, and calibration.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 9, 2005 - non-proprietary Page 8 of I1 C. Provide a detailed description of the instnrnentationuncertaintiesrelatedto the "RCPressure-TemperatureTrip " and the analyses to derive the RPS "Trip Setpoint" and the "Allowable Value"for this tripfunction.

RESPONSE 2.C As described above, a series of core-exit pressure! core-exit temperature points are calculated for different operating RCP combinations, at the maximum allowed core over-power limit for the pump combination, which results in a minimum DNBR at the TDL for the fuel design being used. The minimum DNBR is calculated using the appropriate NRC-approved CHF correlation. Since the DNBR is a stronger function of temperature, there is a slight "bowing" in the curve with the end-points being more restrictive. The core-exit pressure is adjusted to the hot leg pressure tap by accounting for the dynamic pressure drop and elevation change. This curve represents the Safety Limit on TS Figure 2.1-1.

Since the two end-points are the most restrictive, these points are used to determine the "Allowable Value". By adding the hardware and process measurement errors to these points, the "Allowable Value" can be determined. The component errors are provided by the component vendors. Figure 2.C-I shows the components included in the instrument string. The random errors for the individual components were statistically combined whereas the non-random errors were combined by linear addition resulting in a total combined error of I ] psi. The "Trip Setpoint" is determined by adding the combined uncertainty for setpoint drift, MTE and calibration tolerance (10.72 psi) to the "Allowable Value". (Note that the calculation for the field setpoint for this trip function is being updated. This work is expected to be complete by December 30, 2005.)

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 9, 2005 - non-proprietary Page 9 of I1 The calibration tolerance (10.72 psi) consists of three elements, which were added algebraically: Drift (3.6 psi) , measurement and test equipment error (1.12 psi), and additional tolerance (6 psi). Drift was calculated by comparison of as-found vs. as-left data in accordance with Instrument Society of America (ISA) Recommended Practice (RP) 67.04 Part II, Committee Draft 10, August, 1992, Met hodologiesfor the Deterniination of Setpointsfor Nuclear Safety Related Inistrntnzeetationl. [The computation methods in this draft are equivalent to those described in the 1994 revision of the RP.] It should be noted that the as-found/as-left data may include the combined effects of reference accuracy, inherent drift, measurement and test equipment, humidity, vibration, normal radiation, normal temperature, and power supply variations during the time period of the surveillance. A value for drift obtained by this methodology is conservative as it will include some or all of these other uncertainties and the allowance that is determined will bound 95% of future observations with a 95% confidence.

Field data from monthly channel functional checks was used. The change in the output over an approximately six-month period (without calibration adjustment or module replacement) was calculated. Each value was then 'normalized' to 180 days. A "95%/95%" confidence/tolerance value was developed using statistical analysis. Note: Measurements during refueling outages are made differently, and the data was excluded for consistency. Drift values were determined using the formula:

Xn=,, x I+ Ks where Xn,, = the maximum expected value, x = the sample mean, K = a value from Statisticsfor Nuclear Engineers and Scientists (Reference 4), and s = the standard deviation of the sample.

The calculated drift value was then scaled by 125% to account for the 125%

maximum surveillance interval allowed by TS Surveillance Requirement 4.0.2. The data was then verified to fit a normal distribution. Methods used to perform the drift analysis and data distribution verification are described in detail in Davis-Besse calculation C-ICE-58.01-005 (Reference 5), which is available for inspection.

Details of the component error calculation are contained in the AREVA /

Framatome ANP proprietary document 51-5057725-00, DB Low PressureReactor Trip Setpoint Evaluation (Reference 7). This document can also be made available for inspection if required.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 9, 2005 - non-proprietary Page lOofil Figure 2.C-1 RPS Variable Low Reactor Coolant Pressure Trip String r------------------_

7 r------------------ -

it rI o f I aibae SeaaeyEoyRfeigO Ln hne uclnlTs L_I dl inea -igna 1 L LI Variable Low Pressure Trip Calculation Boundary

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 9, 2005 - non-proprietary Page 11 of 11 REFERENCES

1) BAW-10187, Statistical Core Design forB&W-Designed 177FA Plants, B&W Fuel Company, March 1994.

2) BAW-10179, Revision 5, Safety Criteria and Methodology for Acceptable Cycle Reload Analyses, Framatome ANP, December 2002.

3) BAW-10156, Revisionl, LYNXT - Core Transient Thermal-Hydraulic Program, B&W Fuel Company, August 1993.

4) Eggs, W.J., Statisticsfor Nuclear Engineers and Scientists, Part 1: Basic Statistical Inference, DOE Research and Development Report No. WAPD-TM-1292, February, 1981.

5) Calculation C-ICE-58.01-005, RPS High Temperature and Pressure-Temperature Field Trip Setpoints, Revision 4, November 25, 2002.

6) Instrument Society of America Recommended Practice 67.04 Part II, Committee Draft 10, August, 1992, Methodologiesfor the Determination of Setpointsfor Nuclear Safety, Related Instrumentation.

7) AREVA / Framatome ANP document 51-5057725-00, DB Low PressureReactor Trip Setpoint Evaluation, January 2005.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 15, 2005 - non-proprietary Page 1 of 10 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION DATED JUNE 15, 2005 REGARDING LICENSE AMENDMENT REQUEST 05-0002 FOR DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER I

1. (This question pertainedto the paginationofEnclosure I to License Amendment Request 05-0002 (Davis-Besse SerialNo. 3131). After discussion with the NRC staff it was agreed that the paginationis correct, and that there are no missingpages.

Therefore, this question was withdrawn.)

2. Pleasediscuss how instrnuzent uncertainty (including uncertainty in the estimation of core thernialpower) is accountedfor in the proposed limnits. Specifically, these address thefollowing:

a. RegardingProposedFigure2.1-1, confirnr that the "Acceptable Operation" region refers to measuredvalies ofpressure and temperature and that the "Safety Limit" line refers to actual values. If this is not the case, explain how thefigure is to be appliedin plant operation and explain how measurement uncertainty is accounted for.

RESPONSE 2.a Regarding proposed Figure 2.1-1, the limits of the "Acceptable Operation " region reflect the Allowable Values as depicted on Table 2.2-1, Reactor Protection System Instrumentation Trip Setpoints. The development of the Allowable Values is consistent with ISA Standard 67.04.01-2000, Setpointsfor Nuclear Safety-Related Instnumnentation and ISA Recommended Practice 67.04.02-2000, Methodologiesfor the Determninationof Setpointsfor Nuclear Safety-Related Instnimentation. The Allowable Values and trip setpoints are developed consistent with Method 1 of the Recommended Practice. The instrument uncertainties that are not tested during a normal periodic surveillance establish the region between the Safety Limit and the Allowable Value. The instrument uncertainties that are tested during a normal periodic surveillance establish the region between the Allowable Value and the field trip setpoint. The trip setpoint is not depicted on Figure 2.1-1 and therefore limits operation to values more conservative than that reflected on the Figure.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 15, 2005 - non-proprietary Page 2 of 10

b. Address the concerns and RAIpresented in the NRC letter to NEI, datedMarch 31, 2005, concerning trip setpoints and allowable values (ADAMS ML050870008).

RESPONSE 2.b Due to the number of items addressed in the March 31, 2005 letter, this response is contained in a separate enclosure. See Enclosure 6 for this response.

c. Demonstrate that,for operation with measured values at the limits of the "Acceptable Operation" region ofproposed Figure2.1-1, there would be adequate assurancethat the actual operatingpoint will be conservative relative to the indicatedSafety Limit line, given measurement errors consistent with the instrumeint channel uncertainties.

RESPONSE 2.c The Acceptable Operating region is defined by the pressure and temperature limits provided in Technical Specification Figure 2.1-1. The normal operating point for the Davis-Besse reactor coolant system (RCS) corresponds to a hot leg temperature of 606'F at a pressure of 2155 psig. The variation in the nominal RCS temperature and pressure is typically less than +20 F and +25 psi. Use of the Mark-BHTP fuel will have an almost negligible affect on the overall nominal plant conditions.

The Safety Limit Line specifically represents the temperature and pressure combinations that will result in a minimum departure from nucleate boiling ratio that is equal to, or slightly greater than, the thermal design limit for the fuel type being analyzed. This limit also accounts for 4- and 3-pump operation at the respective maximum allowed core power level as described in the response to Question 2.D in this enclosure. The uncertainties associated with the instrumentation string and process measurement described in the response to Question 2.D in this enclosure, are applied to the safety limit line to establish the Allowable Value for the variable low pressure trip function.

The actual reactor trip setpoint is more restrictive than the Allowable Value in that the trip setpoint also takes into consideration the ability to set a specific setpoint.

(See Enclosure I or Enclosure 3, Question 2.B.) Assuming the plant is operating at the trip setpoint and the uncertainty in setting the setpoint results in an actual measured temperature that is at the limits of Acceptable Operation, it can be shown that the trip setpoint calculated by the variable low pressure trip setpoint will occur prior to reaching the Safety Limit Line as shown below:

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 15, 2005 - non-proprietary Page 3 of 10 The equation that defines the variable low pressure trip setpoint is:

Low Pressure Trip Setpoint = 16.25 * (Thot) - 7899 psi For example, if we use a hot leg temperature of 612'F, the resulting trip setpoint would be 2046 psig. The total uncertainty for this trip function is [ ] psi. By subtracting the uncertainty from the calculated trip setpoint, a direct comparison can be made to the Safety Limit Line as shown on the proposed Technical Specification Figure 2.2-1, i.e., [ I psig vice a limit of 1979.8 psig. Similarly, if we assume that the hot leg temperature is 6181F, the resulting trip setpoint after subtracting the uncertainty will be l I psig. By interpolation of the data on the Safety Limit Line, the corresponding pressure at 618'F, is 2098.61 psig. As can be seen in these examples, the margin between the error-adjusted trip setpoint and the Safety Limit Line provides confidence that there is adequate conservatism in the proposed setpoint.

Furthermore, a review of the safety analyses that are included in the Davis-Besse FSAR was performed to ascertain which events credit the variable low pressure trip function for accident mitigation. This review is summarized in Table 2.C. The review concluded that the Reactor Protection System Variable Low Pressure Trip function is not credited for transient mitigation in any USAR analysis. The only specific credit for this trip function is for protection against minimum DNB relative to steady state and not transient operation.

In addition, ISA standard 67.04 Method 1 was used to develop the Allowable Value equation that defines the Acceptable Operation region in proposed Figure 2.1-1. As is explained in more detail in Enclosure 5, Method 1 ensures that the proposed Allowable Value equation meets the "95% / 95%" criteria endorsed by RG 1.105.

Therefore, all operating points within the Acceptable Operation region are conservative with respect to the indicated Safety Limit line.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 Enclosure 4 - RAI Questions Dated June 15, 2005 - non-proprietary Page4oflO Table 2.C - Summary of USAR Accident Analyses USAR Accident Analyses Event Classification RPS Trip Function Startup Accident Reactivity High Pressure/Hligh Flux Rod Withdrawal at Power Reactivity High Pressure/lHigh Flux Dropped Rod Reactivity Low RCS Pressure Moderator Dilution Accident Reactivity High Pressure Loss of Flow Overheating/DNB Power-to-Pump/RCP Monitor. Flux to Flow Startup of an Idle Loop Reactivity High Flux Turbine Trip Overheating High Pressure Loss of Feedwvater Overheating High Pressure Station Blackout / Loss of AC Power Overheating NA - trip due to loss of power Feedwater Malfunction Overcooling High Flux Excessive Load Increase Overcooling Bounded by small SLB Opening of a Pressurizer Safety Valve Loss of Coolant Bounded by SBLOCA Uncompensated Reactivity Changes Reactivity No automatic trip required Failure of Regulating Instrumentation Reactivity No automatic trip required Small Steam Line Break Overcooling SFRCS via ARTs/Low RCS Pressure/High Flux Inadvertent Loading of a Fuel Assembly in an Improper Position Reactivity No automatic trip required Waste Gas Decay Tank Rupture Offsite Dose NA - plant cooldowvn Steam Generator Tube Rupture Offsite Dose/Loss of Coolant Low RCS Pressure Control Rod Assembly Ejection Accident Reactivity High Flux Steam Line Break - mass and energy release Overcooling SFRCS via ARTs/Low RCS Pressure/High Flux Steam Line Break - core response Overcooling SFRCS via ARTs/Low RCS Pressure/High Flux Letdown Line Break Offsite Dose/Loss of Coolant Low RCS Pressure Large Break LOCA - core response Loss of Coolant Low RCS Pressure Large Break LOCA - mass and energy release Loss of Coolant Low RCS Pressure Small Break LOCA Loss of Coolant Low RCS Pressure Fuel Handling Accident Offsite Dose NA - refueling Toxic Material Release Control Room Habitability NA - historical non-RCS event Maximum Hypothetical Accident Offsite Dose NA non-mechanistic LBLOCA Fuel Cask Drop Offsite Dose NA - refueling Feedwater Line Break Overheating High Pressure Anticipated Transients Without Scram. ATWS Overheating DSS (alternate high pressure)

Appendix R-overcooling Overcooling Operator initiated Appendix R-loss of feedwater Overheating Operator initiated NA - Not Applicable

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 15, 2005 - non-proprietary Page 5 of 10

d. Specify the uncertaintiesin the pressure, temperature,and reactorflux measurements associatedwithi the proposed TS change, and show how those uncertaintiesare derived. Show that those uncertaintiesare consistent with the operationaland analytical limits addressedabove.

RESPONSE 2.d See Response 2.a above.

The Reactor Protection System variable low pressure trip setpoint defines a minimum allowed Reactor Coolant System pressure based on measurement of the hot leg temperature. The trip function is represented by a straight line in pressure-temperature space that is based on steady-state operation at the maximum allowed power level for the number of reactor coolant pumps that are operating. The trip setpoint is based on Reactor Coolant System pressure and Reactor Coolant System hot leg temperature. Core power is not a direct input to the trip. The core power and any uncertainty related to core power is factored into generation of the pressure/temperature limits.

The hot leg temperature is measured by a Resistive Temperature Detector (RTD).

The RTD signal is processed through a linear bridge to a signal converter. The signal converter converts the temperature measurement to an allowable Reactor Coolant System (RCS) pressure based on the equation defined as the Allowable Value in the technical specification. A bistable compares this allowable pressure to the measured RCS pressure to determine if a trip condition exists. The uncertainty of the RTD, the linear bridge, signal converter, pressure transmitter, buffer amplifier, and bistable are combined into a single error term by statistically (SRSS) combining the random terms and adding the correlated errors associated with the process measurement. The error terms for the pressure and temperature measurement are taken from the Bailey and Rosemount manufacturer's data for the individual components. The total uncertainty used to determine the Allowable Value is I I psi.

The variable low pressure trip setpoint provides steady-state protection for the pressure and temperature limits defined by operation at the maximum overpower condition for the number of Reactor Coolant Pumps (RCPs) in operation: 112%

Rated Thermal Power (RTP) for four RCPs and 91% RTP for three RCPs. The total power measurement error accounts for RCS flow measurement for either four-pump or three-pump operation and the core power measurement error as described in Section 7.4.2.3 of topical report BAW-10179P-A, Safety Criteriaand Methodologyfor Acceptable Cycle Reload Analyses.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 15, 2005 - non-proprietary Page 6 of 10

e. Demonstrate that the Allowable Values in proposed TS Table 2.1-1, and the limiting values in Note 1 of that table, provide adequate assurancethat the actuation values assumed in the plant safety analyses will not be violated. This should include considerationof theproposednewfitel configuration and considerationof instrumnent setpoint/measurement uncertainty. Forvalues for which no change is proposed,show that the analytical limits remain unchanged or show how the existing values adequatelyprotect the new analytical limits.

RESPONSE 2.e The only effect that the proposed fuel design has on the RCS or the reactor trip setpoints listed in Table 2.2-1 relates to the measurement of Departure from Nucleate Boiling (DNB). The trip setpoints that are defined by the calculation of DNB are the high flux and the reactor coolant pressure-temperature trip setpoint.

(Note that the flux - Aflux/flow setpoint is also affected by DNB; however, this setpoint is contained in the Core Operating Limits Report and is verified for each fuel cycle.) The calculations that support the reactor coolant pressure-temperature trip setpoint are based on operating at the maximum overpower condition for the number of RCPs that are in operation as described in Response 2.d. Since the pressure-temperature limits are defined by the maximum overpower level for the plant, the high flux trip setpoint is also validated because the fuel design does not impact the power measurement uncertainty. The high flux trip setpoint specified in Table 2.2-1 is simply the maximum overpower level for four-pump operation less the power measurement uncertainties described in Section 7.4.2.3 of BAW-10179P-A. The high flux trip setpoint with three pumps includes an additional uncertainty for partial flow.

The values presented in Note 1 to TS Table 2.2-1 are not affected by the changes in the fuel design. The values identified in Note 1 are only to represent the bypass trip functions that facilitate a normal plant shutdown and/or heatup.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 15, 2005 - non-proprietary Page 7 of 10

3. Pleasedemonstrate that the increasedsignal range appliedto the computation unit that produces the variablelow-pressure setpoint does not compromise the operation of that unit or of any other unit. In addition,please discuss how the Uncertainties associatedwith the largersignalspans areproperly accountedforin the Uncertainty analyses.

RESPONSE 3 The computation unit is a signal converter unit in the Reactor Protection System. The revised values have been verified to be within the normal capabilities of the unit. This verification was performed by evaluating the new Allowable Values, adding additional conservatism to reflect a field trip value (the field trip setpoint calculation revisions have not yet been completed, so actual values were not used), and verifying it was within the normal operating band of the instrument. The largest contributor to the string error (by an order of magnitude) is the pressure transmitter itself. This error was determined by reference to the device range, and is therefore unaffected by a change to the setting span. The other error calculations are similarly unaffected by the proposed change. There is no impact on other computation units, as can be verified by reference to Figure 2.C-1 in Enclosure 1. It is anticipated that the field trip setpoint calculation revisions will be completed by December 30, 2005.

4. The 5' paragraphof Section 3 addressesonly accidents resulting in pressure reduction. Pleasediscuss howv reductions inflow or RCS heat removal and increasein fiuel temperature are considered.

RESPONSE 4 The discussion in the 5 th paragraph of Section 3 acknowledges the pressure reduction scenarios that are protected by the low pressure and the pressure-temperature trip functions. However, the protection of the core, particularly the DNB protection, is established using more than the trip functions in TS Figure 2.1-1. Sections 6, 7, and 8 of Reference 1 of Enclosure 1 provide a general discussion of the safety criteria and methodology used for the Davis-Besse core relating to the TS 2.0 and overall core protection.

All eight of the Reactor Protection System (RPS) trip functions work together to provide steady-state and transient protection for the Reactor Coolant System (RCS) and more specifically the core. These trip functions and the events/transients for which they provide protection are provided in Table 4.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 15, 2005 - non-proprietary Page 8 of 10 The Mark-B-HTP fuel design is a more robust design, but provides a slight increase in the hydraulic resistance and as a result there is a different CHF correlation that is applied to the fuel. As a result, there must be an assurance that fuel will not exceed the prescribed acceptance criteria. This assurance must not only address decreasing pressure conditions, but steady-state operation, the performance during loss of flow, events that evolve into an increase in the RCS/fuel temperature, as well as pressure and power increasing transients. An evaluation of all USAR Chapter 6 and Chapter 15 events was performed. It was concluded that, for most events, the fuel design has a negligible affect on the overall system response and only those events that specifically address the fuel response need to be reevaluated, i.e., the DNB-related transients such as LOCA, loss of flow, and steady-state operation. Consequently, the high- and low-RCS pressure setpoint, the high temperature setpoint, the high containment pressure and the high flux setpoints would not be affected by this change in fuel design.

Since new LOCA analyses are being performed for the Mark B-HTP fuel, only the loss of flow and steady-state conditions were assessed. The loss of flow transients (four RCP coast down, single RCP coast down, and locked rotor) are reanalyzed for each new reload using the appropriate CHF correlation. As noted above, the hydraulic resistance is higher for the Mark B-HTP fuel design; however, in comparison to the total RCS pressure drop this change is small. For the loss of flow events, the RCS flow is dominated by the inertia of the reactor coolant pumps such that the power-to-pump flow-related trip function is not affected as long as the minimum DNBR is greater than the thermal design limit (TDL). As part of the fuel reload analyses, it was confirmed that the TDL was not exceeded for the Mark-B-HTP fuel and therefore no change is required to this RPS setpoint.

The Reactor Coolant Pressure-Temperature Trip defines the limit for the steady-state conditions at which Davis-Besse can operate at the respective maximum overpower condition for three or four RCPs operating with acceptable DNB performance. The safety limit that is presented in TS Figure 2.1-1 represents the combinations of reactor outlet temperatures and reactor system pressure that yield a minimum DNB equal to the DNB design criterion when the core is operating at the maximum overpower condition for the more limiting of three or four RCP operation, as described in Reference 2 of Enclosure 1. Typically, the limiting core safety limit is based on four RCP operation. The core safety limit represents the most severe pressure/temperature conditions the core can operate in a steady-state condition without violating the DNB criterion for the plant. For the Mark B-HTP fuel design, the existing technical specification trip setpoint must be revised as discussed in the LAR.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 15, 2005 - non-proprietary Page 9 of 10 The flux/Aflux/flow trip setpoint is included in the Core Operating Limits Report (COLR) and is subject to change for each new reload. The flow portion of this trip function is not affected as described above, however, the allowed imbalance (flux versus Aflux) may be affected because this trip function is also based on operation at the maximum overpower condition for the number of RCPs that are operating. The required setpoint for this trip function is not defined until all of the reload analyses for the cycle are completed. At this time, the COLR may need to be revised. Since this trip setpoint is in the COLR, it was not discussed in the LAR.

In summary, the trip functions that are required to provide a level of protection based on the overall system response to a postulated event are not affected by the implementation of the Mark B-HTP fuel design and only the trip functions that provide steady-state protection (Reactor Coolant Pressure-Temperature and the Flux/AFlux/Flow Trip functions) at the maximum overpower conditions are affected.

The change required for the RC pressure-temperature trip is provided in the LAR. The Flux/AFlux/Flow Trip is defined in the COLR and will be verified once all cycle-specific reload calculations are performed.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - RAI Questions Dated June 15, 2005 - non-proprietary Page 10 of 10 Table 4 - Summary of Reactor Protection Trip Functions Trip Function Events Credited System Response High Flux Rod withdrawal from zero power (high reactivity insertion rates) Positive reactivity insertion Rod withdrawal at power (high reactivity insertion rates) Positive reactivity insertion Steam line break (small area) Positive reactivity insertion Rod ejection (high rod worth) Positive reactivity insertion Startup of an idle loop Positive reactivity insertion Feedwater malfunction Positive reactivity insertion High RCS Pressure Rod withdrawal from zero power (low reactivity insertion rates) Positive reactivity insertion Rod withdrawal at power (low reactivity insertion rates) Positive reactivity insertion Rod ejection (low rod worth) Positive reactivity insertion Boron dilution at power Positive reactivity insertion Turbine trip Loss of heat transfer Loss of main feedwater Loss of heat transfer Feedwater line break Loss of heat transfer Low RCS Pressure Loss of coolant accident Loss of RCS inventory Letdown line rupture (offsite dose consequence) Offsite dose consequence SG tube rupture Offsite dose consequence Dropped rod assembly Core peaking Steam line break (large area) Positive reactivity insertion RC pressure-Temperaturc This trip function is not credited in the USAR Chapter 6 or DNB (variable low RCS pressure) Chapter 15 analyses, but provides a limit for steady-state DNB Centerline fuel temperature

_______________________protection._______________

High RCS This trip function is not credited in the USAR Chapter 6 or DNB Tempcrature Chapter 15 analyses, but provides a limit for steady-state DNB Centerline fuel temperature Temperature ~~protection.Cetriefltmpaur Flux/AFlux/Flow Loss of a single RCP Partial loss of RCS flow Locked RCP rotor/sheared shaft Rapid loss of RCS flow High Flux vs. No. RCPs Loss of all RCPs (4 pump coast down) Loss of all RCS forced flow (power-to-pump monitors) Loss of 2 RCPs in one loop Partial loss of RCS flow High Containment Pressure This trip function is not credited in the USAR Chapter 6 or Loss of RCS inventory Chapter 15 analyses. Loss of SG inventory

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - Response to March 31, 2005 Letter Page I of 5 RESPONSE TO CONCERNS AND RAI PRESENTED IN THE NRC LETTER TO NEI, DATED MARCH 31, 2005, CONCERNING TRIP SETPOINTS AND ALLOWABLE VALUES (ADAMS 1%IL050870008)

There are three items described on page 3 of the letter dated March 31, 2005 from Mr.

Lyons to Alex Marion of the Nuclear Energy Institute. The following is taken from that letter:

Duringthe public meeting held on March 11, 2005, at the NRC Headquarters Office, you requested that the NRC withdraw the RAI question and continue its review of existing licensing actionspendingfinalizationof the TSTF technical specification change to addressthis issue. You reiteratedthis request in your March 18, 2005 letter. As notedpreviously, the NRC staff is continuing the review of existing in-house LARs. To moveforward with LARs that are currently under NRC staff review, an interim approachhas been developed that addresses the NRC staffs concerns until the TSTF technicalspecification change is reviewed and approved. As part of these licensee specific reviews, the revised RAI question (see Enclosure) will remain outstanding, with the understanding.

that, in addition to a briefdiscussion of the licensee's methodologyfor establishingLSSS, the licensee 's response to the question needs to contain the following in orderforthe staff to complete its review:

1. An explicit regulatory commitment to adopt thefinal TSTF technical specification change to come into conformance with the existing understandingof the requirements of 10 CFR 50.36.

2. An explicit regulatory conmzitment to assess the operabilityoftested instrumentationbased on the previous as-left instrument setting and accountingforthe uncertaintiesassociatedwith the test or calibration.

3. A revision to the technicalspecificationsfor the LSSS being changed by the LAR to incorporateafootnote that states:

Tile as-left instrumnent setting shall be returned to a setting within the tolerance band of the trip setpoint establishedto protect the safety limit.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - Response to March 31, 2005 Letter Page 2 of 5 RESPONSE 1-3 Davis-Besse uses Method 1 from the ISA Recommended Practice 67.04.02 for the Reactor Coolant Pressure-Temperature Functional Unit trip. Using this method, the Allowable Value is calculated from the Analytical Limit, by subtracting instrument uncertainties that are not tested during periodic surveillances. The trip setpoint is then calculated from the Allowable Value by subtracting instrument uncertainties that are tested during periodic surveillances.

As part of the uncertainty calculation, a calibration tolerance is established. The periodic surveillance tests include acceptance criteria for the trip setpoint, including the calibration tolerance. If these acceptance criteria are not met, the instrument is re-calibrated within the calibration tolerance value before returning it to service.

For item 1, FENOC commits to evaluate the final TSTF technical specification change recommendations after NRC approval of the associated TSTF traveler.

However, FENOC can not commit to submital of a license amendment request proposing adoption of these recommendations before having a chance to evaluate them.

For item 2, FENOC does not commit to assessing the operability of the tested instrumentation based on the previous as-left instrument setting and accounting for the uncertainties associated with the test or calibration. As previously described, the instruments are evaluated for operability based on the as-found setting being consistent with the calculated trip setpoint and the calibration tolerance. This provides assurance that the instrument is performing within the criteria established in the calculations. Evaluation of the difference between the previous as-left setting and the current as-found setting would create an additional administrative burden while providing no additional assurance that the instrument is performing within the calculation limits. As industry evaluation of this issue concludes with the anticipated issuance of the TSTF, additional consideration will be given to this methodology under the evaluation committed to above.

For item 3, FENOC will submit a separate letter to propose an additional TS change for LAR 05-0002 to add a footnote to TS Table 4.3-1, "Reactor Protection System Instrumentation Surveillance Requirements," applicable to Functional Unit 7, "RC pressure-temperature." The new footnote will read as follows, "The as-left instrument setting shall be returned to a setting within the tolerance band of the trip setpoint established to protect the safety limit." This proposed footnote is consistent with the existing practice of resetting the trip setpoint within the established calibration tolerance prior to returning this equipment to service.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - Response to March 31, 2005 Letter Page 3 of 5 In addition to the items discussed above, the Enclosure to the March 31, 2005 letter requests additional information:

REVISED METHOD 3 REOUEST FOR ADDITIONAL INFORAIATION The [insertplant nameI technical specifications define Limiting Safety System Settings (LSSS) as an allowable value (A V). During reviews ofproposed license amendments that contain changes to LSSS setpoints, the NRC staff identified concerns regardingthe method used by some licensees to deternine the allowable value (A V) identified in thle technical specifications (TS). A Vs are identified in the TS as LSSS to provide acceptance criteria for determination of instrument channel operabilityduringperiodicsurveillance testing.

The NRC staff's concern relates to one of the three methodsfor determining the A V as describedin the Instrument Society ofnAmerica (ISA) recommendedpracticeISA-RP67.04-1994, PartII, "Methodologyfor Determinationof Setpointsfor NuclearSafety-Related Instrumentation."

The NRC staff has determined that to ensure a plant wvill operate in accordancewith the assumptions upon which the plant safety analyses have been based, additionalinformation is requiredregardlessof the methodology used to establish LSSS values in technical specifications. Details about tihe NRC staff's concerls are availableoi the NRC's public website under ADAMSAccession Numbers ML041690604, ML041810346, and ML050670025.

In Orderfor the NRC staff to assess the acceptabilityof your license amendment request relatedto this issue, the NRC staffrequests thefollowing additionalinformation:

1. Discuss the setpoint methodology used at [insertplant name] to establish A Vs associatedwith LSSS setpoints.

RESPONSE 1 Please refer to page 7 of Enclosure 1 of the May 2, 2005 license amendment application. ISA Recommended Practice 67.04.02, Method 1 was used to develop the Allowable Value and trip setpoints for the Reactor Coolant Pressure-Temperature Functional Unit. Using this method, the Allowable Value was calculated from the Analytical Limit, by subtracting instrument uncertainties that are not tested during periodic surveillances. The trip setpoint is then calculated from the Allowable Value by subtracting instrument uncertainties that are tested during periodic surveillances.

2. Regardless of the methodology used, the NRC staffhas thefollowing questions regardingthe use of the methodology at [insertplant name]:

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - Response to March 31, 2005 Letter Page 4 of 5

a. Discuss how the methodology and controlsyou have in place ensure the analytical limit (AL) associatedwith an LSSS will not be exceeded (the AL is a surrogatethat ensures the safety limits will not be exceeded). Include in your discussion infornation on the controlsyou employ to ensure the trip setpoint established after completingperiodicsurveillances satisfiesyourmnethodology. If the controls are located in a document other than the TS, discuss how those controls satisfy the requirements of 10 CFR 50.36.

RESPONSE 2.a As stated above, the Allowable Value for the Reactor Coolant Pressure-Temperature Functional Unit was calculated using Method 1 of ISA Recommended Practice 67.04.02. Due to the availability of references supporting the use of Method 2, all three methods are considered in the discussion that follows, with Method 1 being shown to be acceptable in part by comparison to Method 2. Using either Method 1 or Method 2 of ISA RP 67.04.02 would result in the same Allowable Value.

Because this Allowable Value is calculated from the Analytical Limit, it is conservative with respect to the Allowable Value calculated using Method 3.

Method 2 calculates the Trip Setpoint by subtracting the channel uncertainty (plus margin) from the Analytical Limit. Method I calculates the Trip Setpoint by subtracting drift, calibration uncertainty, and uncertainties during normal operation from the Allowable value. Uncertainties that are random, normally distributed, and independent may be combined using the Square-Root-Sum-of-the-Squares (SRSS) method. Method I and Method 2 both account for the same uncertainties, but Method 2 may apply SRSS in one step, so that the resulting setpoint is likely to be slightly more conservative using Method 1, depending on the extent to which SRSS is used.

Regulatory Guide (RG) 1.105, Revision 3, December 1999, states that conformance with Part 1 of ISA-S67.01-1994, "Setpoints for Nuclear Safety-Related Instrumentation," with the listed clarifications and exceptions, provides a method acceptable to the NRC staff for satisfying the NRC's regulations for ensuring that setpoints for safety-related instrumentation are established and maintained within the technical specification limits. Regulatory Position #I of Regulatory Guide 1.105, Revision 3 provides the following position with regard to the degree of certainty with which a licensee must ensure the analytical limit (AL) associated with an LSSS will not be exceeded:

Section 4 ofISA-S67. 04-1994 specifles the methods, but not the criterion,for combining uncertaintiesin determining a trip setpoint and its allowable values.

The 95/95 tolerance limit is an acceptable criterionfor uncertainties. That is, there is a 95% probabilitythat the constructedlimits contain 95% ofthe population of interestfor the surveillance intervalselected.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 - Response to March 31, 2005 Letter Page 5 of 5 Therefore, RG 1.105 accepts "95/95" as satisfying the requirements of 10 CFR 50.36. This conclusion is supported by the NRC document, "Setpoint Allowable Values for Instrument Channels in Safety-Related Service," ADAMS accession number ML041810346, June 23, 2004, at page 27. Review of ISA Recommended Practice ISA-RP67.04.02-2000, "Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation," has led FENOC to conclude that both Method 1 and Method 2 meet the 95/95 criterion and therefore the calculation described in FENOC's May 2, 2005 application (Serial 3131, LAR 05-0002) meets the requirements of 10 CFR 50.36.

With respect to the trip setpoint, this was discussed in FENOC's response (above) to items 1, 2, and 3 on page 3 of the March 31, 2005 letter from Mr. Lyons of the NRC to Mr. Marion of the NEL. As part of the uncertainty calculation, a calibration tolerance is established. The periodic surveillance tests include acceptance criteria for the trip setpoint being within the established tolerance. If these acceptance criteria are not met, the surveillance test has not satisfied the Technical Specification periodic surveillance requirement. Therefore, the instrument is recalibrated within the calibration tolerance value prior to returning the equipment to service.

b. Discuss how the TS surveillances ensure the operability of the instrument channel. This should include a discussion on how the surveillance test results relate to the technical specificationA V and describe how these are used to determine the operabilityof the instnrment channel. If the requirementsfor determining operabilityof the LSSS instrument being tested are in a document other than the TS (e.g., plant test procedure), discuss how this meets the requirements of 10 CFR 50.36.

RESPONSE 2.b

b. As discussed above, during the TS surveillances the trip setpoints are verified to be within the calibration tolerance.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 Enclosure 6 - Affidavit Page l of3 AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

CITY OF LYNCHBURG )

1. My name is Gayle F. Elliott. I am Manager, Product Licensing in Regulatory Affairs, for Framatome ANP ("FANP"), and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by FANP to determine whether certain FANP information is proprietary. I am familiar with the policies established by FANP to ensure the proper application of these criteria.

3. I am familiar with Engineering Information Record 51-5069986, Revision 1,

'RAI Response Information for Davis Besse Cycle 15 TS Change LAR," dated August 25, 2005 and referred to herein as "Document." Information contained in this Document has been classified by FANP as proprietary in accordance with the policies established by FANP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by FANP and not made available to the public.

Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure.

6. The following criteria are customarily applied by FANP to determine whether Information should be classified as proprietary:

Docket Number 50-346 License Number NPF-3 Serial Number 3166 Enclosure 6 - Affidavit Page 2 of 3 (a) The information reveals details of FANP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for FANP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for FANP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by FANP, would be helpful to competitors to FANP, and would likely cause substantial harm to the competitive position of FANP.

7. In accordance with FANP's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside FANP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. FANP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

Docket Number 50-346 License Number NPF-3 Serial Number 3166 Enclosure 6 - Affidavit Page 3 of 3

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this__c___

day of an A .2005.

Wanda L. Wade NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES:

WANDA L. WADE NOTARY PUBLIC COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIIS AUGUIT 31) 2009 l

Docket Number 50-346 License Number NPF-3 Serial Number 3166 Enclosure 7 Page 1 of 1 COMMITMENT LIST The following list identifies those actions committed to by the Davis-Besse Nuclear Power Station, Unit Number 1, (DBNPS) in this document. Any other actions discussed in the submittal represent intended or planned actions by the DBNPS. They are described only for information and are not regulatory commitments. Please notify Henry L. Hegrat, Supervisor

- Licensing (330-315-6944) of any questions regarding this document or associated regulatory commitments.

COMMITMENTS DUE DATE A letter will be submitted to revise LAR 05-0002 to Letter will be submitted as change Technical Specifications for the Variable Reactor soon as reasonably Coolant Pressure-Temperature Trip to add a footnote that achievable, with a target of states "The as-left instrument setting shall be returned to a 15 days following submittal setting within the tolerance band of the trip setpoint of this letter.

established to protect the safety limit."

FENOC commits to evaluate the final TSTF technical Evaluation will be completed specification change recommendations after NRC within 120 days of receipt of approval of the associated TSTF traveler. the NRC-approved TSTF traveler.

As industry evaluation of this issue concludes with the anticipated issuance of the TSTF, additional consideration will be given to assessment of tested instrumentation based on the previous as-left setting, as required by the evaluation committed to above.