View source for ULNRC-05297, Proposed Revision to Technical Specifications for Use of Beacon Power Distribution Monitoring System

{{Adams
| number = ML061570437
| issue date = 05/25/2006
| title = Proposed Revision to Technical Specifications for Use of Beacon Power Distribution Monitoring System
| author name = Young K
| author affiliation = AmerenUE
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000483
| license number = 
| contact person = 
| case reference number = OL-1268, ULNRC-05297
| document type = Final Safety Analysis Report (FSAR), Letter type:ULNRC, Technical Specification, Amendment
| page count = 69
}}

=Text=
{{#Wiki_filter:AmerenUE                                                                       PO Box 620 Cal/away Plant                                                                  Fulton, MO 65251 May 25, 2006 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-137 Washington, DC 20555-0001 Ladies and Gentlemen:                                           ULNRC-05297 DOCKET NUMBER 50-483 CALLAWAY PLANT UNION ELECTRIC COMPANY AISF/PROPOSED                                              REVISION TO TECHNICAL SPECIFICATIONS FOR USE OF BEACON POWER WAmeren                                   DISTRIBUTION MONITORING SYSTEM UiE                         (LICENSE AMENDMENT REQUEST OL-1268)
Pursuant to 10 CFR 50.90, AmerenUE (Union Electric) hereby requests amendment of the Facility Operating License (No. NPF-30) for the Callaway plant in order to incorporate proposed changes to the Callaway Technical Specifications.
Specifically, AmerenUE proposes to revise Technical Specification (TS) 3.1.7, "Rod Position Indication," TS 3.2.1, "Heat Flux Hot Channel Factor," TS 3.2.4, "Quadrant Power Tilt Ratio (QPTR)," and TS 3.3.1, "Reactor Trip System (RTS)
Instrumentation," for use of the BEACON power distribution monitoring system (PDMS) described in WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," to perform power distribution surveillances.
Essential information is provided in the attachments to this letter.
Attachment 1 provides a detailed description and technical evaluation of the proposed TS changes for the PDMS, including AmerenUE's determination that the proposed changes involve no significant hazards consideration. Attachment 2 provides a special evaluation for the PDMS as described in Attachment 1.
Attachment 3 provides the affected, existing TS pages marked-up to show the proposed changes. Attachment 4 provides a copy of the revised TS pages retyped with the proposed changes incorporated (if approved). Attachment 5 provides the affected existing TS Bases pages marked-up to show the associated proposed Bases changes (for information only). Lastly, Attachment 6 provides marked-up pages from Callaway's Final Safety Analysis Report (FSAR), indicating the changes to be made to that document, including changes to Chapter 16 to incorporate appropriate administrative controls and surveillance requirements for the BEACON PDMS. The proposed FSAR changes are provided for information only.
a subsidiaryof Ameren Corporation

ULNRC-05297 May 25, 2006 Page 2 The Callaway Plant Review Committee and a subcommittee of the Nuclear Safety Review Board have reviewed and approved this amendment application. In addition, it has been determined that this amendment application involves no significant hazards consideration as determined per 10 CFR 50.92, and that pursuant to 10 CFR 51.22(b) no environmental assessment should be required to be prepared in connection with the issuance of the requested amendment. It should also be noted that no regulatory commitments are made or identified in this amendment application.
With regard to when the requested amendment is needed, AmerenUE respectfully requests approval of the proposed license amendment by January 31, 2007, i.e., prior to the next scheduled refueling outage (Refuel 15). Approval of the requested license amendment in advance of the outage will support planning and completion of the necessary procedure changes and operator training for use of the system, as the intent is to begin using the PDMS in Cycle 16, i.e., upon restart from Refuel 15.
Pursuant to 10 CFR 50.91(b)(1), AmerenUE is providing the State of Missouri with a copy of this proposed amendment.
If you should have any questions on the above or attached, please contact Dave Shafer at (314) 554-3104 or Tom Elwood at (314) 554-4593.
I declare under penalty of perjury that the foregoing (and attached) is true and correct.
Very truly yours, Executed on: May 25, 2006 Manager, Regulatory Affairs TBEjdg Attachments: 1) Evaluation
: 2) Evaluation for Excluding PDMS Instrumentation Requirements from the Technical Specifications
: 3) Markup of Technical Specification Pages
: 4) Retyped Technical Specification Pages
: 5) Markup of Technical Specification Bases Pages (For information only)
: 6) Markup of FSAR pages

ULNRC-05297 May 25, 2006 Page 3 cc:    Mr. Bruce S. Mallett Regional Administrator U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Jack N. Donohew (2 copies)
Licensing Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop O-71I Washington, DC 20555-2738 Missouri Public Service Commission Governor Office Building 200 Madison Street P0 Box 360 Jefferson City, MO 65102-0360

ATTACHMENT 1 ULNRC-05297 EVALUATION

.01 Attachment I to ULNRC-05297 1.0   Introduction AmerenUE (Union Electric) is proposing to revise the Technical Specifications to allow use of a power distribution monitoring system (PDMS). The PDMS to be used at the Callaway plant utilizes the NRC-approved Westinghouse proprietary computer code, the Best Estimate Analyzer for Core Operations - Nuclear (BEACON), and plant data fed to the plant process computer from the incore thermocouples and excore nuclear instruments. Incore detector measurements are used to periodically calibrate the BEACON PDMS. The BEACON PDMS serves as a 3-D core monitor, operational analysis tool, and operational support package.
Westinghouse submitted a topical report for BEACON (WCAP-12472-P) to the NRC in the early 1990s. The NRC issued a Safety Evaluation Report (SER) thereby approving the topical report on February 16, 1994. In its SER, the NRC concluded that BEACON provides a greatly improved continuous online power distribution measurement and operation prediction information system for Westinghouse reactors.
AmerenUE proposes to use BEACON to augment the functional capability of the flux mapping system for the purpose of power distribution surveillances.
WCAP-12472-P-A discusses an application of BEACON in which the Technical Specifications and core power distribution limits are revised to take credit for continuous monitoring by plant operators. However, AmerenUE proposes to use a more conservative application of BEACON in which the core power distribution limits themselves remain unchanged. AmerenUE intends to use the BEACON PDMS as the primary method for performing power distribution measurements and surveillances, and to use the flux mapping system as an alternative for such purposes, when thermal power is greater than 25 percent rated thermal power (RTP). At thermal power levels less than or equal to 25 percent RTP, or when the PDMS is inoperable, the movable incore detector system will be used.
The Technical Specifications that are to be revised for implementation of the BEACON PDMS are Technical Specification (TS) TS 3.1.7, "Rod Position Indication," TS 3.2.1, "Heat Flux Hot Channel Factor," TS 3.2.4, "Quadrant Power Tilt Ratio," and TS 3.3.1, "Reactor Trip System (RTS) Instrumentation."
In addition, a section is to be added to the Core Operating Limits Report (COLR) for supporting application of BEACON. The new section defines the equations and constants used for determining the applicable measurement uncertainties that are applied to the core peaking factors when determined by either the PDMS or the flux mapping system. The constants found in this section of the COLR are used as coefficients in the uncertainty calculations and are determined using the methodology approved by the NRC in its review of the Westinghouse BEACON topical report. The constants may be revised periodically as appropriate to reflect cycle-specific variables.
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t Attachment 1 to ULNRC-05297 As noted in the cover letter, AmerenUE requests approval of the proposed amendment in advance of the next refueling outage (Spring 2007) based on planned implementation of BEACON during the outage and use of the BEACON PDMS upon restart from the outage. Movable incore detectors will continue to be used for the monthly flux map surveillances prior to issuance and implementation of the license amendment.
2.0 Description of Proposed TS Changes TS 3.1.7, "Rod Position Indication":
For conditions involving inoperable digital control rod position indicators, Required Actions A.1, B.3 and C.1 of TS 3.1.7 require plant operators to "verify the position of the rods with inoperable position indicators indirectly by using movable incore detectors." The Required Actions will be revised to state, "Verify the position of the rods with inoperable position indicators indirectly by using core power distribution measurement information." The generic phrase "core power distribution measurement information" would allow the use of an operable PDMS or the movable incore detectors for verifying the position of the rod with an inoperable digital rod position indicator.
TS 3.2.1, "Heat Flux Hot Channel Factor (Fc (Z) )(FQ Methodology)":
Surveillance Requirement (SR) 3.2.1.2, for verifying that F (Z) is within its limit, is modified by a Note. This Note currently states that if Fe (Z) measurements indicate that maximum over z                has increased since the previous evaluation of F c (Z), plant operators are required (per part "b" of this Note) to do the following:
: b. Repeat SR 3.2.1.2 once per 7 EFPD until two successive flux maps indicate maximum over z    -
has not increased.
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1&#xfd;-
Attachment 1 to ULNRC-05297 This requirement will be revised so that it is worded as follows::
: b. Repeat SR 3.2.1.2 once per 7 EFPD until 'two successive power distribution measurements indicate maximum over z    [-I--;]I has not increased.
This change would allow the surveillance to be performed using either the movable incore detectors or an operable PDMS.
TS 3.2.4, "Quadrant Power Tilt Ratio (QPTR)":
SR 3.2.4.2 currently states, "Verify QPTR is within limit using the movable incore detectors." It will be revised to state, "Verify QPTR is within limit using power distribution measurement information." This change would allow the surveillance to be performed using either the movable incore detectors or an operable PDMS.
TS 3.3.1, "Reactor Trip System (RTS) Instrumentation":
SR 3.3.1.3 requires comparing results from the incore system to the nuclear instrument system (NIS) channel output with respect to the indicated axial flux difference (AFD). Specifically, SR 3.3.1.3 currently states, "Compare results of the incore detector measurements to Nuclear Instrumentation System (NIS) AFD. Adjust NIS channel if absolute difference is > 2%."
SR 3.3.1.3 will be revised to state "Compare results of incore power distribution measurements to Nuclear Instrumentation System (NIS) AFD.
Adjust NIS channel if absolute difference is _2%."
SR 3.3.1.6 requires periodically calibrating the excore channels against the incore channels. Specifically, SR 3.3.1.6 currently states, "Calibrate excore channels to agree with incore detector measurements." It will be revised to state, "Calibrate excore channels to agree with incore power distribution measurements."
These changes would allow SR 3.3.1.3 and SR 3.3.1.6 to be performed using either the movable inccre detectors or an operable PDMS.
The above-described TS changes are shown as mark-ups to the current Technical Specifications on the pages provided in Attachment 3. The pages provided in Attachment 4 indicate how the affected TS pages would appear with the proposed changes incorporated.
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Attachment 1 to ULNRC-05297 The COLR and TS Bases will also be revised for consistency with the proposed TS changes. A markup of the TS Bases pages reflecting the needed changes is provided in Attachment 5 for information only. (Itimay be noted that in addition to the Bases for TS 3.1.7, TS 3.2.1, TS 3.2.4, and TS 3.3.1, the Bases for TS 3.1.4, "Rod Group Alignment Limits," and TS 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor F:=," are also to be revised due to references to "incore flux mapping," etc. in those Bases sections.) The COLR changes will be implemented in accordance with TS 5.6.5, "Core Operating Limits Report (COLR)," and the TS Bases changes will be implemented in accordance with TS 5.5.14, "Technical Specification (TS) Bases Control Program," as part of implementing the requested amendment (after NRC approval and issuance of the requested amendment).
With regard to maintenance, operability and control of the PDMS and its associated instrumentation, it has been determined that no TS changes are needed for this purpose since the PDMS does not meet the selection criteria set forth in 10 CFR 50.36(c)(2)(ii) for inclusion in the Technical Specifications. The evaluation for this determination is provided in Attachment 2.
In lieu of TS requirements, requirements for the PDMS and associated instrumentation will be contained in Chapter 16 of the Callaway FSAR. This section of the FSAR is dedicated to providing administrative controls and operational/testing requirements as appropriate for equipment not required to be included in the Technical Specifications. Changes to FSAR Chapter 16 are controlled in accordance with 10 CFR 50.59. The changes for incorporating PDMS instrumentation requirements and controls into FSAR Chapter 16, along with other proposed changes to the FSAR, are indicated in Attachment 6. The indicated changes are provided for information only.
In summary, the proposed license amendment would allow the use of the Westinghouse proprietary 3-D nodal code BEACON for performing power distribution surveillances provided that the PDMS instrumentation is operable.
The proposed amendment would also allow (or continue to allow) the use of the movable incore detector system for meeting power distribution surveillances and TS actions, in addition to their use for calibration of BEACON.

===3.0 Background===
As described in WCAP-12472-P-A, the Westinghouse BEACON PDMS was developed to provide operational support for pressurized water reactors (PWRs).
BEACON is an advanced core monitoring and support software package that utilizes existing plant instrumentation for providing incore thermocouple temperatures, reactor coolant system cold leg temperatures, control bank demand positions, power range detector output, and reactor power measurement data to the PDMS. These data are sent by the plant computer in the form of a file that Page 4 of 14

Attachment 1 to ULNRC-05297 BEACON can interpret to perform nodal power distribution prediction calculations.
The PDMS includes an on-line 3-D nodal model that is continuously updated to reflect the current plant operating conditions. The nodal solution method used by the PDMS is consistent with the NRC-approved Westinghouse Advanced Nodal Code (ANC) core design code. The core-exit thermocouple and excore neutron flux detector readings are used with the reference 3-D power distribution to determine the measured power distribution. By coupling the measured 3-D power distribution with an on-line evaluation, actual core margins can be better understood. The PDMS provides an understanding of operating and design margins to address strategic fuel cycle changes. The BEACON methodology improves the quality of the surveillance process since it uses a depleted model to match the actual operational profile. The PDMS continuously monitors the limiting FQ(Z) and F*.
As previously noted, the movable in-core detector system will remain available for use. The movable in-core detector system will also be used, however, to calibrate BEACON.
AmerenUE personnel at Callaway intend to utilize the BEACON PDMS to take advantage of its capability for continuous monitoring of the limiting core thermal peaking factors, FQ(Z) and F.N, without the need to obtain a full-core flux map.
The BEACON PDMS will provide operational support for TS compliance, and its continuous monitoring feature will permit instantaneous identification of core anomalies, as well as providing predictive capabilities for both operators and reactor engineers.
4.0 Technical Analysis As noted above, the requested license amendment will allow utilization of the BEACON PDMS at Callaway to provide operational support for the facility. The PDMS maintains an on-line 3-D nodal model that is continuously updated to reflect the current plant operating conditions. The following is a summary/excerpt of Brookhaven National Laboratory's (BNL) Technical Evaluation Report (TER) for WCAP-12472-P-A.
4.1      BEACON Core Monitoring Methodology The BEACON core monitoring system uses the NRC-approved Westinghouse SPNOVA nodal method for core power distribution measurements. The SPNOVA data libraries and core models are consistent with the NRC-approved Westinghouse PHOENIXIANC design Page 5 of 14

Attachment 1 to ULNRC-05297 models and have been benchmarked against operating reactor measurements.
The BEACON core monitoring process is carried out in three steps. In the first step, the SPNOVA model, individual thermocouples, and the excore axial offset are calibrated to the full-core incore flux measurement. In the second step, the SPNOVA model is updated based on the most recent operating history, and adjusted using the thermocouple and excore measurements. Continuous monitoring is performed for the third step, using the thermocouples and excores to update the BEACON model.
The BEACON power distribution calculation is updated using the thermocouple and excore detector measurements. The thermocouple measurements are interpolated/extrapolated radially using the spline fit.
The BEACON system provides both a full three-dimensional nodal power distribution calculation as well as a simplified, more approximate one-dimensional calculation. The BEACON on-line limits evaluation will be performed in three dimensions, and the one-dimensional calculation will only be used as a scoping tool in predictive analysis.
The continuous core monitoring of the current reactor statepoint (fuel burnup, xenon distribution, soluble boron concentration, etc.) provided by BEACON allows a more precise determination of the parameters used in the transient analyses, and therefore relaxes the requirement to limit the transient initial conditions via power distribution control. As part of the continuous monitoring, the fuel limits are calculated using the standard Westinghouse methods.
For the application of BEACON at Callaway, credit will not be taken for the continuous monitoring of the power distribution (as described above). Instead, BEACON will be used as a Technical Specification monitor for present peaking factor limits, and the transient initial condition limits for Callaway will not be relaxed.
Operability and calibration of the BEACON PDMS is dependent on the number and distribution of available core exit thermocouples. The criteria for the core-exit thermocouples, with BEACON operable, require at least 25% of the thermocouples, with at least two per quadrant, with the added requirement that the operable pattern normally covers all internal fuel assemblies within a chess "knight" move (an adjacent plus a diagonal square away) or otherwise a more frequent calibration is required. With optimum thermocouple coverage, calibration with the movable incore detectors is required every 180 effective full-power days. However, calibration is required every 30 days when the knight's move requirement is not satisfied. The accuracy of the power distribution information with decreased incore or thermocouple detector operability has been Page 6 of 14

Attachment 1 to ULNRC-05297 analyzed by Westinghouse and penalties are applied to the calculated peaking factors (refer to TER section 2.3). The analysis concluded that the minimum available incore and thermocouple detectors, when coupled with the increased uncertainty penalties, provide reasonable and acceptable power distribution information.
4.2      Model Calibration and Uncertainty BEACON uses the incore flux detector measurements, core-exit thermocouples, and excore detectors to perform the local calibration of the SPNOVA three-dimensional power distribution. The SPNOVA-predicted detector reaction rates are normalized to the incore measurements at the incore radial locations and over an axial mesh. The thermocouple adjustment is two-dimensional and is made by normalizing the SPNOVA radial power distribution to the assembly power inferred from the core-exit thermocouples. The thermocouple assembly power measurement is periodically calibrated to the incore-measured assembly power.
Since the incore detectors and core-exit thermocouples do not provide complete coverage of the core, BEACON employs a two-dimensional spline fit to interpolate/extrapolate measurements to the unmonitored assemblies. The spline fit includes a tolerance factor which controls the degree to which the fit is forced to match the individual measurements. If, for example, the measurements are believed to be extremely accurate (inaccurate) a low (high) tolerance factor is used and the SPNOVA solution is (not) forced to be in exact agreement with the measurements.
The BEACON axial power shape is adjusted to ensure agreement with the axial offset measured the by the excore detectors. This adjustment is made by adding a sinusoidal component to the SPNOVA-calculated axial power shape. The SPNOVA excore axial offset is determined by an appropriate weighting of the peripheral assembly powers. The excore detector axial offset is periodically calibrated to the incore power distribution measurements.
As an initial assessment of the power distribution calculation, Westinghouse performed detailed comparisons of BEACON to the predictions of the INCORE system presently used at Westinghouse plants. (INCORE is a data analysis code written to process information obtained by the movable incore detector system in Westinghouse pressurized water reactors. INCORE is presently used at Callaway for processing information obtained by the movable incore detector system and verifying Technical Specification surveillance requirements.) These comparisons were made for three plants over four cycles, and included a range of fuel bumup, core loadings, power levels and control rod insertion levels. The averages of the standard deviation between the BEACON results and the actual measured reaction rates were 1.5% for assemblies with power greater than the average (1.0) value and 2% for all measured assemblies (WCAP-12472-P-A section 4.1.1.) The Page 7 of 14

Attachment 1 to ULNRC-05297 averages of the standard deviation of the inferred assembly power between BEACON and INCORE were 1.10% for assemblies with power greater than the average (1.0) value and 1.37% for all assemblies. (See WCAP-12472-P-A Table 4-6.) From the results of this study, Westinghouse concluded that the BEACON processing of the incore flux map and the inferred assembly power distribution accuracy is statistically consistent with the INCORE computer code.
The uncertainties applied to the BEACON power distribution measurements are different than those applied to the traditional flux map systems because BEACON uses a more comprehensive scope of instrumentation. An uncertainty analysis of the BEACON power distribution measurement is reported in WCAP-12472-P-A.
Portions of the BNL TER for WCAP-12472-P-A relevant to the uncertainty analysis are summarized/excerpted as follows:
        "Due to the change in reactor statepoint, SPNOVA modeling approximations, and instrumentation error, a model calibration uncertainty is introduced into the BEACON predictions. Westinghouse has evaluated this uncertainty by comparing BEACON predicted and measured incore reaction rates over four cycles and a range of operating conditions, and has found that the model calibration uncertainty was very small and varied only slightly for these comparisons."
        "The thermocouple calibration uncertainty is due to the change in reactor statepoint and to instrument error. Westinghouse has evaluated this uncertainty by comparing the assembly powers inferred from the thermocouples to SPNOVA incore-corrected assembly powers.
Comparisons for three plants and a range of operating conditions indicate a difference of less than a few percent at full power. The observed calibration uncertainty increased at lower powers due to the reduced enthalpy rise and changes in cross-flow."
        "In order to determine the axial power distribution uncertainty, Westinghouse has compared SPNOVA incore-updated and SPNOVA excore-updated predictions of the axial power shape. These comparisons included a range of fuel bumups and rod insertions, and indicated a 95/95 upper tolerance limit of less than a few percent with a slight dependence on rod movement since calibration."
       "Based on an extensive set of calibration data, the model calibration uncertainty is observed to increase as the calibration interval (in units of fuel burnup) increases. Using the observed fuel burnup dependence, an additional assembly power uncertainty is determined to account for the effects of increased calibration interval."
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Attachment 1 to ULNRC-05297 "The failure of (incore and thermocouple) detectors (used by) the BEACON system results in a relaxation of the local calibration to measurement, and an increase in the power distribution uncertainty. The effect of random failures of the incore and thermocouple detectors on the assembly power was evaluated for failure rates of up to 75%. The assembly power uncertainty was found to increase linearly with incore detector failure and quadratically with the failure of thermocouples."
         "The BEACON calculation requires local power distribution factors for (1) the ratio of assembly power-to-detector response, (2) assembly local peaking factor, and (3) the grid power-depression factor (correction factor to the assembly axial power distribution to take the power depression due to the grid of the assembly into account). The BEACON uncertainty analysis employs previously approved upper tolerance values for the assembly power-to-detector response ratio and the local peaking factor.
The grid (power depression) factor uncertainty was determined by comparison to measured flux traces and is found to be relatively small."
        "The uncertainty in the BEACON power peaking resulting from errors in the SPNOVA model calibration and thermocouple calibration is determined using an analog Monte Carlo error propagation technique. In this analysis, the BEACON three-step calibration model update and power distribution update procedure are simulated. The SPNOVA model and thermocouple calibration factors are subjected to random variations (based on their uncertainties) and the resulting variations in the BEACON power distribution are used to determine the 95% probability upper tolerance limit on the assembly power for the twenty highest powered assemblies."
        "The analysis is performed for a range of operating conditions including off-normal power distributions and extended calibration intervals. A typical set of thermocouple uncertainties is used together with a relatively large tolerance factor which results in substantial smoothing of the thermocouple measurements. The upper tolerance limit on the assembly power peaking factor is calculated and found to increase as the square-root of the thermocouple uncertainty."
4.3     Acceptance Criteria/Conditions In the NRC Safety Evaluation Report for WCAP-12472-P-A, the NRC staff evaluated the BEACON methodology, the uncertainty analysis, and the operation of the overall system and concluded that the BEACON PDMS is acceptable for performing core monitoring and operations support functions for Westinghouse pressurized water reactors (PWR) but subject to certain conditions as specified in the BNL TER. These conditions are listed below. After each condition listed, a description of how the condition will be met at Callaway is provided.
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Attachment 1 to ULNRC-05297
: 1. In the cycle-specific applicationof BEACON, the powerpeaking uncertaintiesUAH and UQ must provide 95% probabilityupper tolerance limits at the 95% confidence level.
Although not specifically described in this submittal, cycle-specific BEACON calibrations performed before startup and at beginning-of-cycle conditions will ensure that power peaking uncertainties provide 95% probability upper tolerance limits at the 95% confidence level. These calibrations are to be performed using the Westinghouse approved methodology. Until these calibrations are complete, more conservative default uncertainties will be applied. The calibrations will be documented and retained as records.
: 2. In order to ensure that the assumptions made in the BEACON uncertainty analysis remain valid, the generic uncertainty components may require reevaluationwhen BEACON is appliedto plant or core designs that differ sufficiently to have a significant impact on the WCAP-12472-Pdatabase.
Callaway utilizes a Westinghouse 4-loop nuclear steam supply system (NSSS) with Westinghouse movable incore instrumentation. All fuel is presently of Westinghouse manufacture. Therefore, Callaway does not differ significantly from the plants that form the WCAP database, and no additional review of WCAP applicability to Callaway is necessary.
During the review of the Westinghouse topical report WCAP-12472-P, the NRC requested additional information on how BEACON treats core loadings with fuel designs from multiple fuel vendors, and the impact to the BEACON uncertainty analysis. Westinghouse responded that for all BEACON applications, the previous operating cycle is examined to establish reference uncertainties. This examination accounts for loading of fuel supplied by multiple vendors by comparing a BEACON model to actual operating data over the cycle. At the beginning of cycle, thermocouple data is verified and calibration/uncertainty components are updated as necessary. In addition, the initial flux mapping at the start of the cycle insures model calibration factors that reflect the actual fuel in the reactor before the BEACON system is declared operable.
: 3. The BEACON Technical Specifications should be revised to include the changes describedin Section 3 (of the BNL TER) concerningSpecifications 3.1.3.1 and 3.1.3.2 and the Core Operating Limits Report.
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Attachment 1 to ULNRC-05297 The WCAP describes an application of BEACON where the core operating limits are changed. As noted previously, AmerenUE is proposing only to use BEACON as a core Technical Specification monitor for conformance to Callaway's existing limits. The TS changes of concern per this question or condition are not applicable or of concern to the more limited changes being proposed by AmerenUE for the intended use of BEACON. Therefore, this condition does not apply to the amendment requested for Callaway.
5.0 Regulatory Analysis 5.1      No Significant Hazards Consideration AmerenUE has evaluated whether or not a significant hazards consideration is involved with the proposed Technical Specification changes for supporting use of the BEACON power distribution monitoring system (PDMS), by focusing on the three standards set forth in 10 CFR 50.92(c) as discussed below:
: 1. Does the proposed change involve a significant increasein the probability or consequences of an accidentpreviously evaluated?
Response: No The PDMS performs continuous core power distribution monitoring with data input from existing plant instrumentation. This system utilizes an NRC-approved Westinghouse proprietary computer code, i.e., Best Estimate Analyzer for Core Operations - Nuclear (BEACON), to provide data reduction for incore flux maps, core parameter analysis, load follow operation simulation, and core prediction. The PDMS does not provide any protection or control system function. Fission product barriers are not impacted by these proposed changes.
The proposed changes occurring with PDMS will not result in any additional challenges to plant equipment that could increase the probability of any previously evaluated accident. The changes associated with the PDMS do not affect plant systems such that their function in the control of radiological consequences is adversely affected. These proposed changes will therefore not affect the mitigation of the radiological consequences of any accident described in the Final Safety Analysis Report (FSAR).
Use of the PDMS supports maintaining the core power distribution within required limits. Further continuous on-line monitoring through the use of PDMS provides significantly more information about the power distributions present in the core than is currently available. This results in more time (i.e.,
earlier determination of an adverse condition developing) for operator action prior to having an adverse condition develop that could lead to an accident condition or to unfavorable initial conditions for an accident.
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Attachment 1 to ULNRC-05297 Therefore, the proposed change does not involve a significant increase in the probability or consequence of an accident previously evaluated.
: 2. Do the proposedchanges create the possibility of a new or different kind of accidentfrom any accidentpreviously evaluated?
Response: No Other than use of the PDMS to monitor core power distribution, implementation of the PDMS and associated Technical Specification changes has no impact on plant operations or safety, nor does it contribute in any way to the probability or consequences of an accident. No safety-related equipment, safety function, or plant operation will be altered as a result of this proposed change. The possibility for a new or different type of accident from any accident previously evaluated is not created since the changes associated with implementation of the PDMS do not result in a change to the design basis of any plant component or system. The evaluation of the effects of using the PDMS to monitor core power distribution parameters shows that all design standards and applicable safety criteria limits are met.
The proposed changes do not result in any event previously deemed incredible being made credible. Implementation of the PDMS will not result in any additional adverse condition and will not result in any increase in the challenges to safety systems. The cycle-specific variables required by the PDMS are calculated using NRC-approved methods. The Technical Specifications will continue to require operation within the required core operating limits, and appropriate actions will continue to be taken when or if limits are exceeded.
The proposed change, therefore, does not create the possibility of a new or different kind of accident from any previously evaluated.
: 3. Do the proposedchanges involve a significant reduction in a margin of safety?
Response: No No margin of safety is adversely affected by the implementation of the PDMS.
The margins of safety provided by current Technical Specification requirements and limits remain unchanged, as the Technical Specifications will continue to require operation within the core limits that are based on NRC-approved reload design methodologies. Appropriate measures exist to control the values of these cycle-specific limits, and appropriate actions will continue to be specified and taken for when limits are violated. Such actions remain unchanged.
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Attachment 1 to ULNRC-05297 Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above evaluations, AmerenUE concludes that the activities associated with the proposed amendment present no significant hazards consideration under the standards set forth in 10 CFR 50.92 and accordingly, a finding by the NRC of no significant hazards consideration is justified.
5.2      Applicable Regulatory Requirements/Criteria 10 CFR 50, Appendix A, General Design Criterion 13 states:
Criterion13 -- Instrumentationand control. Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.
Implementation of the PDMS at Callaway does not replace, eliminate, or modify existing plant instrumentation. The PDMS software runs on a workstation connected to the plant process computer. The PDMS combines inputs from currently installed plant instrumentation and design data generated for each fuel cycle. Together, this provides a means to continuously monitor the power distribution limits including limiting peaking factors and quadrant power tilt ratio.
Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
6.0 Environmental Consideration AmerenUE has evaluated the proposed change and has determined that the change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in the individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.
Page 13 of 14

Attachment 1 to ULNRC-05297 7.0     References
: 1. WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994 (NRC approved version with Safety Evaluation Report).
: 2. License Amendment No. 142 to Facility Operating License No. NPF-12 Regarding Best Estimate Analyzer for Core Operations - Nuclear (BEACON), Virgil C. Summer Nuclear Station, Unit No. 1.
: 3. License Amendment Nos. 237 and 218 to Facility Operating License Nos.
50-272 and 50-311 for Salem Nuclear Generating Station, Unit Nos. 1 and 2.
: 4. License Amendment Nos. 164 and 166 to Facility Operating License Nos.
DPR-80 and DPR-82 for the Diablo Canyon Power Plant, Unit Nos. 1 and 2.
7.1    Precedent The BEACON Technical Specification monitor was approved by the NRC for use at the Salem Nuclear Generating Station in License Amendment Nos. 237 (Unit 1) and 218 (Unit 2), on November 6, 2000, and for the Virgil C. Summer Nuclear Station in License Amendment No. 142 (Unit 1) on April 9, 1999. Similar changes were also approved for the Byron Station in License Amendment Nos. 116 (Units 1 and 2), on February 13, 2001, and for the Braidwood Station in License Amendment Nos. 110 (Units 1 and 2), on February 13, 2001. The application of BEACON to the Byron/Braidwood stations uses BEACON to take credit for the direct and continuous monitoring of Departure from Nuclear Boiling Ratio (DNBR), whereas the application of BEACON at Callaway is for power distribution surveillances.
More recently, use of the PDMS for power distribution measurements was approved for each of the Diablo Canyon units which are very similar in design to Callaway. Those license amendments were issued as amendment Nos. 164 and 166 (for Units 1 and 2, respectively) on March 31, 2004. The license amendment requested for Callaway is most similar to the amendment(s) requested and approved for Diablo Canyon.
Page 14 of 14

ATTACHMENT 2 ULNRC-05297 EVALUATION FOR EXCLUDING PDMS INSTRUMENTATION REQUIREMENTS FROM THE TECHNICAL SPECIFICATIONS

Attachment 2 to ULNRC-05297 Evaluation for Excluding PDMS Instrumentation Requirements from the Technical Specifications The justification for not including requirements for the PDMS and associated instrumentation in the Technical Specifications is explained per the evaluation provided below. The purpose of this evaluation is to demonstrate that the structures, systems, or components (i.e., instrumentation) that constitute the PDMS are not required to be contained in the Technical Specifications. This evaluation is done in accordance with the requirements contained in 10 CFR 50.36(c)(2)(ii).
Per 10 CFR 50.36(c)(2)(ii) a Technical Specification Limiting Condition for Operation must be established for each item meeting one or more of the following criteria:
(A) Installed Instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
PDMS instrumentation is not associated with monitoring of any aspect of the reactor coolant pressure boundary.
(B) A process variable, design feature, or operating restriction that is an initial condition of a design basis accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
The limits for the power distribution parameters FQ(Z) and FN are operating restrictions which ensure that the accident analyses and assumptions for all applicable, analyzed DBAs remain valid. These limits are included in the Technical Specifications. The PDMS supports the capability to monitor core power distribution for verifying conformance to such limits, but it does not control core power distribution and cannot itself cause or effect any condition assumed in the accident/transient analyses.
The PDMS provides the capability to monitor power distribution parameters at more frequent intervals than is currently required by the Technical Specifications. Additionally, these parameters or limits can be determined independent of the operability of PDMS. Therefore, the PDMS does not constitute a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
Page 1 of 2

Attachment 2 to ULNRC-05297 (C) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
The PDMS performs only a monitoring function and does not affect any of the key safety parameter limits or levels of margin considered in the DBA design-basis evaluations. The PDMS has no active/control functions or actuation capability and as such is not part of any primary success path for mitigation of a DBA or transient that either assumes the failure of, or presents the challenge to the integrity of a fission product barrier.
(D) A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
The PDMS and its associated instrumentation provide the capability to monitor power distribution parameters at more frequent intervals than is currently required by the Technical Specifications, but the PDMS has no active safety functions and its use has no impact on the results or consequences of anyiDBA or transient analysis. Further, the PDMS is an alternative means for performing core power distribution measurements and related surveillances, as the current means of performing such activities (by use of the movable incore detectors) will still be available. PDMS unavailability therefore is not significant relative to plant risk. Based on these considerations and facts, the PDMS is not a feature that is significant to public health and safety.
The evaluation completed above indicates that PDMS instrumentation does not meet any of the criteria for inclusion in the Technical Specifications. The PDMS requirements and controls to be incorporated into Chapter 16 of the Callaway FSAR are consistent with the recommendations in WCAP-12472-P-A and will suffice to provide the necessary operability and test requirements for the PDMS apart from the Technical Specifications.
Page 2 of 2

ATTACHMENT 3 ULNRC-05297 MARKUP OF TECHNICAL SPECIFICATION PAGES

LAR OL-1268 Rod Position Indication 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Rod Position Indication LCO 3.1.7             The Digital Rod Position Indication (DRPI) System and the Demand Position Indication System shall be OPERABLE.
APPLICABILITY:        MODES 1 and 2.
ACTIONS NOTE Separate Condition entry is allowed for each inoperable rod position indicator and each demand position indicator.
COMPLETION REQUIRED ACTION                           TIME CONDITION                                                                TIME A. One DRPI per group             A.1         Verify the position of the  Once per 8 hours inoperable for one or more                  rods with inoperable groups.                                     position indicators n indirectly by using core A.2         Reduce THERMAL              8 hour POWER to
* 50% RTP.
(continued)
CALLAWAY PLANT                                3.1-16                         Amendment No. 133

LAP, OL-1268 Rod Position Indication 3.1.7 ACTIONS (continued)
CONDITION                      REQUIRED ACTION                       COMPLETION TIME B. More than one DRPI per           B.1      Place the control rods         Immediately group inoperable for one or               under manual control.
more groups.
AND B.2      Monitor and record RCS         Once per 1 hour Tavg.
AND B.3     Verify the position of the      Once per 8 hours rods with inoperable position indicatorsv,.
n indirectly by using core             A" B.4     Restore inoperable              24 hours position indicators to OPERABLE status such that a maximum of one DRPI per group is inoperable.
C. One or more rods with             C.1     Verify the position of the      4 hours inoperable been movedDRPIs         haveof in excess                rods withindicators,-,
position   inoperable 24 steps in one direction                               statusing core since the last determination inoeraleDRPs hverodsrr     -one with inoprable         . iA of the rod's position.                                            of  n   4 C.2    Reduce THERMAL                   8 hours of tho rod's posin                             t (continued)
CALLAWAY PLANT                             3.1-17                              Amendment No. 133

LAR OL- IMG8 FQ(Z) (FQ Methodology) 3.2.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE                                     FREQUENCY SR 3.2.1.2                                 NO IL If Fc(Z) measurements indicate L      -
m axim um over z
_-_
has increased since the previous evaluation of Fc(Z)
: a. Increase Fw(Z) by the appropriate factor specified in the COLR and reverify Fw(Z) is within limits; or
: b. Repeat SR 3.2.1.2 On per 7 EFPD until two successive              indicate h                   Fic(Z)ed er zmaxim--
o--r has not increased.
Verify Fw(Z) is within limit                            Once after each refueling prior to THERMAL POWER exceeding 75%
RTP AoD (continued)
CALLAWAY PLANT                             3.2-4                      Amendment No. 133

LAR OL- 12C8 QPTR 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY NUlIlh SR 3.2.4.1     1. With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER
                       < 75% RTP, the remaining three power range channels can be used for calculating QPTR.
: 2.      SR 3.2.4.2 may be performed in lieu of this Surveillance.
Verify QPTR is within limit by calculation.            7 days SR 3.2.4.2 Not required to be performed until 12 hours after input from one or more Power Range Neutron Flux channels are Inoperable with THERMAL POWER
               > 75% RTR 12 hours
            .S CALLAWAY PLANT                          3.2-13                       Amendment No. 133

LAP, OL-j246 RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS Refer to Table 3.3.1 1 to determine which SRsNOTE apply for each RTS Function.
SURVEILLANCE                                      FREQUENCY SR 3.3.1.1          Perform CHANNEL CHECK.                                  12 hours SR 3.3.1.2                                    NOTE Not required to be performed until 24 hours after THERMAL POWER is -2:15%RTP.
Compare results of calorimetric heat balance             24 hours calculation to power range channel output. Adjust power range channel output if calorimetric heat balance calculation results exceed power range channel output by more than +2% RTP.
SR 3.3.1.3                                    NOTE Not required to be performed until 24 hours after THERMAL POWER is _>       50% RTR Compare results ofeincore(* .L l                         31 effective full measurements to Nuclear Instrumentation System           power days (NIS) AFD. Adjust NIS channel if absolute difference     (EFPD) is > 2%.
(continued)
CALLAWAY PLANT                                3.3-12                       Amendment No. 148

LAR OL-1Z26.
RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE                                      FREQUENCY SR 3.3.1.4                                NOTE This Surveillance must be performed on the reactor trip bypass breaker for the local manual shunt trip only prior to placing the bypass breaker in service.
Perform TADOT.                                            62 days on a STAGGERED TEST BASIS SR 3.3.1.5    Perform ACTUATION LOGIC TEST.                             92 days on a STAGGERED TEST BASIS SR 3.3.1.6 Not required to be performed until 72 hours after achieving equilibrium conditions with THERMAL POWER z 75 % RTP.
channels to agree with incore          92 EFPD ements.
SR 3.3.1.7                                NOTE
: 1.      Not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 4 hours after entry into MODE 3.
: 2.      Source range instrumentation shall include verification that interlocks P-6 and P-10 are in their required state for existing unit conditions.
Perform COT.                                               184 days (continued)
CALLAWAY PLANT                            3.3-13                         Amendment No. 165

ATTACHMENT 4 ULNRC-05297 RETYPED TECHNICAL SPECIFICATION PAGES (TO BE PROVIDED LATER)

ATTACHMENT 5 ULNRC-05297 MARKUP OF TECHNICAL SPECIFICATION BASES PAGES (FOR INFORMATION ONLY)

TS B CN 04-008 Rod Group Alignment Limits B 3.1.4 BASES APPLICABLE              1.       specified acceptable fuel design limits, or SAFETY ANALYSES                2.       Reactor Coolant System (RCS) pressure boundary (continued)                     integrity; and
: b.      The core remains subcritical after accident transients.
Two types of misalignment are distinguished. During movement of a rod group, one rod may stop moving, while the other rods in the group continue. This condition may cause excessive power peaking. The second type of misalignment occurs if one rod fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition requires an evaluation to determine that sufficient reactivity worth is held in the control and shutdown rods to meet the SDM requirement, with the maximum worth rod stuck fully withdrawn.
Two types of analysis are performed in regard to static rod misalignment (Ref. 3). With control and shutdown banks at their insertion limits, one type of analysis considers the case when any one rod is completely inserted into the core. The second type of analysis considers the case of a completely withdrawn single rod from control bank D, inserted to its insertion limit. Satisfying limits on departure from nucleate boiling ratio in both of these cases bounds the situation when a rod is misaligned from its group by 12 steps.
Another type of misalignment occurs if one RCCA fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition is assumed in the evaluation to determine that the required SDM is met with the maximum worth RCCA also fully withdrawn (Ref. 4).
The Required Actions in this LCO ensure that either deviations from the alignment limits will be corrected or that THERMAL POWER will be adjusted so that excessive local linear heat rates (LHRs) will not occur, and that the requirements on SDM and ejected rod worth are preserved.
Continued operation of the reactor with a misaligned rod is allowed if the heat flux hot channel factor (Fo(Z)) and the nuclear enthalpy rise hot channel factor (FYH) are verified to be within their limits in the COLR and the safety analysis is verified to remain valid. When a rod is misaligned, the assumptions that are used to determine the rod insertion limits, AFD limits, and quadrant power tilt limits are not preserved. Therefore, the limits may not preserve the design peaking factors, and Fo(Z) and F-N must be verified directly bySection                               3.2 (Power crses Section 3.2(Powrtin (continued)
CALLAWAY PLANT                           B 3.1.4-3                                  Revision 0

Rod Group Alignment Limits B 3.1.4
                                                    .TSo CN 0(0-oo8 BASES ACTIONS         B.2.2, B.2.3, B.2.4, B.2.5, and B.2.6 (continued) verification of SDM is required. A Frequency of 12 hours is sufficient to ensure this requirement continues to be met.
Verifying that FQ(Z), as approximated by FQc(Z) and FQw(Z), and F)H are within the required limits ensures that current operation at 75% RTP with a rod misaligned is not resulting in power distributions that may invalidate safety analysis assumptions at full power. The Completion Time of 72 hours allows sufficient time to obtain flux map of the core power distribution and FM. r- usn,                              -
and to calculate FQ(Z) a "Hcore power s'ribItou me surev              nj"
                                                                          ;vlforrEIsfionl Once current conditions have been verified acceptable, time is available to perform evaluations of accident analysis to determine that core limits will not be exceeded during a Design Basis Event for the duration of operation under these conditions. The accident analyses presented in FSAR Chapter 15 (Ref. 5) that may be adversely affected will be evaluated to ensure that the analyses results remain valid for the duration of continued operation under these conditions. A Completion Time of 5 days is sufficient time to obtain the required input data and to perform the analysis.
C.1 When Required Actions cannot be completed within their Completion Time, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours, which obviates concerns about the development of undesirable xenon or power distributions. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging the plant systems.
D.1.1 and D.1.2 More than one control rod becoming misaligned from its group demand position is not expected, and has the potential to reduce SDM. Therefore, SDM must be evaluated. One hour allows the operator adequate time to determine SDM. Restoration of the required SDM, if necessary, requires increasing the RCS boron concentration to provide negative reactivity, as described in the Bases or LCO 3.1.1. The required Completion Time of 1 hour for initiating boration is reasonable, based on the time required for (continued)
CALLAWAY PLANT                         B 3.1.4-7                                   Revision 0

Rod Position Indication B 3.1.7
.4.
TS       CN 046-OO8 BASES LCO            b.       The Bank Demand Indication System has been calibrated either in (continued)            the fully inserted position or to the DRPI System.
The 12 step agreement limit between the Bank Demand Position Indication System and the DRPI System indicates that the Bank Demand Position Indication System is adequately calibrated, and can be used for indication of the measurement of rod bank position.
A deviation of less than the allowable limit, given in LCO 3.1.4, in position indication for a single rod, ensures high confidence that the position uncertainty of the corresponding rod group is within the assumed values used in the analysis (that specified rod group insertion limits).
These requirements ensure that rod position indication during power operation and PHYSICS TESTS is accurate, and that design assumptions are not challenged. OPERABILITY of the position indicator channels ensures that inoperable, misaligned, or mispositioned rods can be detected. Therefore, power peaking, ejected rod worth, and SDM can be controlled within acceptable limits.
APPLICABILITY  The requirements on the DRPI and step counters are only applicable in MODES 1 and 2 (consistent with LCO 3.1.4, LCO3.1.5, and LCO 3.1.6),
because these are the only MODES in which power is generated, and the OPERABILITY and alignment of rods have the potential to affect the safety of the plant. In the shutdown MODES, the OPERABILITY of the shutdown and control banks has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the Reactor Coolant System.
ACTIONS        The ACTIONS table is modified by a Note indicating that a separate Condition entry is allowed for each inoperable rod position indicator and each demand position indicator. This is acceptable because the Required Actions for each Condition provide appropriate compensatory actions for each inoperable position indicator.
A.1                                      INSERT /REPLACE WITH ATTACHED.
When one DRPI per group fails, the position of the rod ma still be determined indirectly by use of ..            .... f4   e'The
_......a, Required Action may also be satisfied by ensuring at least once per 8 hours that F0 (Z) satisfies LCO 3.2.1, FN satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided (continued)
CALLAWAY PLANT                          B 3.1.7-3                                  Revision 0

TS" CN 0(-0o08 TEXT INSERT for BASES for TS 3.1.7 core power distribution measurement information. Core power distribution measurement information can be obtained from flux maps using the movable incore detectors, or from an OPERABLE power distribution monitoring system (PDMS) (Reference 5).

Rod Position Indication B 3.1.7 TSB Chi O6-O0 BASES                                                             On ACTIONS        A.1 (continued)                                 mowcure d         nsorabnton the nonindicating rods have not been moved. The alternate use of peaking factor and SDM verification is limited to those rodded core locations where rod position can not be determined by, These locations are either.ot instrumented (or has an out of service incore thimble) or not fage-Valjacent to instrumented assembly locations.
Based on experience, no          I power operation does not require excessive movement of banks. If a bank has been significantly moved, the Required Action of C.1 or C.2 below is required. Therefore, verification of RCCA position within the Completion Time of 8 hours is adequate for allowing continued full power operation, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small. While in Condition A, the performance of Required Action A.1 can be used with the Bank Demand Position Indication System (group step counters) to verify alignment limits are met for SR 3.1.4.1.
A.2 Reduction of THERMAL POWER to
* 50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factors (Ref. 3).
The allowed Completion Time of 8 hours is reasonable, based on operating experience, for reducing power to *50% RTP from full power conditions without challenging plant systems and allowing for rod position determination by Required Action A. 1 above.
B.1, B.2, B.3, and B.4 When more than one DRPI per group fail, additional actions are necessary to ensure that acceptable power distribution limits are maintained, minimum SDM is maintained, and the potential effects of rod misalignment on associated accident analyses are limited. Placing the Rod Control System in manual assures unplanned rod motion will not power A;Wi6Adsan   occur. TO ether with the idirect position determination available via CAsaigrenment                       njtrmailo-        his will minimize the potential for rod misalignment.
The Immediate Completion Time for placing the Rod Control System in manual reflects the urgency with which unplanned rod motion must be prevented while in this Condition. Monitoring and recording reactor coolant system Tavg help to assure that significant changes in power (continued)
CALLAWAY PLANT                         B3 3-1.7-4                                    Revision 1.

Rod Position Indication B 3.1.7 TSB CN      OC-o00 BASES ACTIONS                   B.1, B.2, B.3, and B.4 (continued) distribution and SDM are avoided. The once per hour Completion Time is acceptable because only minor fluctuations in RCS temperature are power Aisiribttdion               expected at steady state plant operating conditions.
4.. 1        V I-- Af IIa
       &#xfd;&sect; A
F"Clue
                         *4 9-6   wt.The position of the rods may be determined indirectly by use ofer
(.r,         ,r         t    . he Required Action may also be satisfied by ensuring at least once per 8 hours that Fo(Z) satisfies LCO 3.2.1, 15H satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the nonindicating rods have not been moved. The alternate use of peaking factor and SDM verification is limited to o e rodded core Ioc tions where rod position can not be e ermined by                       hese locations are either not V instrumented (or has an out of service incore thimble) or not fa ge-acjacent to instrumented assembly locations. Verification of RCCA           tionce per 8 hours is adequate for allowing continued full power operation for a limited, 24 hour period, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small. While in Condition B, the performance of Required Action B.3 can be used with the Bank Demand Position Indication System (group step counters) to verify alignment limits are met for SR 3.1.4.1. The 24 hour Completion Time provides sufficient time to troubleshoot and restore the DRPI system to operation while avoiding the plant challenges associated with a shutdown without full rod position indication (Ref. 4).
Based on operating experience, normal power operation does not require excessive rod movement. If one or more rods has been significantly moved, the Required Action of C.1 or C.2 below is required.
C.1 and C.2 These Required Actions clarify that when one or more rods with inoperable DRPIs have been moved in excess of 24 steps in one direction, since the position was last determined, the Required Actions of A.1 and A.2 or B.1 are still appropriate but must be initiated promptly under Required Action C. 1 to begin indirectly verifying that these rods are still properly positioned, relative to their group positions.
If, within 4 hours, the rod positions have not been determined, THERMAL POWER must be reduced to : 50% RTP within 8 hours to avoid undesirable power distributions that could result from continued operation at > 50% RTP, if one or more rods are misaligned by more than 24 steps.
             .%st.R'T/ADt            The allowed Completion Time of 4 hours provides an acceptable period of ttime to verify the rod      it            I          "nboedeet~oro
                                                                                 -0     e(continued)
CALLAWAY PLANT                                   B 3.1.7-5                                   Revision 1

Rod Position Indication B 3.1.7 TSB CNOG-O09 BASES SURVEILLANCE   SR 3.1.7.1 (continued)
REQUIREMENTS This surveillance is performed prior to reactor criticality after each removal of the reactor head, since there is potential for unnecessary plant transients if the SR were performed with the reactor at power.
REFERENCES     1.      10 CFR 50, Appendix A, GDC 13.
: 2.      FSAR, Chapter 15.
: 3.      WCAP-10216-P-A, Rev. 1A, "Relaxation of Constant Axial Offset Control and Fa Surveillance Technical Specification," February 1994.
: 4.      Amendment 61 to Callaway Plant Facility Operating License NPF-30, February 1, 1991.
: 5. WCAP- 124712 -P-A, BEACON core Movi loran3 anJ Operafloris Surrort :Sy34CJ,' Aiigist ISS-.
CALLAWAY PLANT                       B 3.1.7-7                                     Revision 1

TSB Co06,-008 FJ(Z) (F,, Methodology)
B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Heat Flux Hot Channel Factor (Fo(Z)) (F, Methodology)
BASES BACKGROUND           The purpose of the limits on the values of Fo(Z) is to limit the local (i.e., pellet) peak power density. The value of F,(Z) varies along the axial height (Z) of the core.
Fo(Z) is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions. Therefore, F(Z) is a measure of the peak fuel pellet power within the reactor core.
During power operation, the global power distribution is limited by LCO 3.2.3, "Axial Flux Difference (AFD)," and LCO 3.2.4, "Quadrant Tilt Power Ratio (QPTR)," which are directly and continuously measured process variables. These LCOs, along with LCO 3.1.4, "Rod Group Alignment Limits," LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits," maintain the core limits on power distributions on a continuous basis.
F3(Z) varies with fuel loading patterns, bank insertion, fuel burnup, and changes in axial power *bution.
F       isnot directly measura e but is inferred from a power distribution obtained with the movable incore detector s sterk The results of e ree- imensional power is r utiio            a an-ayzed to derive a otr Jro_4 a4n PERABLE power                           measured value for F.(Z). Thes  mea   uremen   s are generally taken with disiet'=*fion         the core at or near qiiru           o t CI,Re(-rCvnc' 5 )o
* However, because this value re                n qui ibrium condition, it does not include the variations in the value of Fo(Z) that are present during nonequilibrium situations, such as load following. To account for these possible variations, the steady state value of FE,(Z) is adjusted by an elevation dependent factor that accounts for the calculated worst case transient conditions.
Core monitoring and control under nonsteady state conditions are accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, and control and shutdown bank insertion.
(continued)
CALLAWAY PLANT                               B 3.2.1 -1                                  Revision 0

T S B CN O-OO&
Fo(Z) (F. Methodology)
B 3.2.1 BASES LCO                 where: CFQ = FRTP is the F,(Z) limit at RTP provided in the COLR, (continued)
K(Z) is the normalized Fr(Z) as a function of core height provided in the COLR, and p= THERMAL POWER RTP The actual values of CFQ and K(Z) are given in the COLR.
For Relaxed Axial Offset Control operation, F(Z) is approximated by F. (Z) and Fw (Z). Thus, both F&(Z) and Fw Zrmust meet the preceding limits on F,(Z).,         1a wer        s1Iis iio, rneasurer'ent
()w-b,.io*( ... An F (Z) evaluation requires obtaining                       --din MODE 1.
aSigrfe~nt          From th         -e.je         esults             the measured value (Fa"(Z))
of F0(Z) Then, ~        Land FS (Z) FmZ.
wher (13eis a factor that            nts for fuel manufacturing tolerances and measurement uncertainty.
FS (Z) is an excellent approximation for Fo(Z) when the reactor is at the steady state power at which the                        was taken.
The expression for Fw (Z) is:                         ric Fw (Z) = F8 (Z) W(Z) where W(Z) is a cycle dependent function that accounts for power distribution transients encountered during normal operation. W(Z) is included in the COLR.
The F.(Z) limits define limiting values for core power peaking that precludes peak cladding temperatures above 2200&deg;F during either a large or small break LOCA.
This LCO requires operation within the bounds assumed in the safety analyses. Calculations are performed in the core design process to confirm that the core can be controlled in such a manner during operation (continued)
CALLAWAY PLANT                              B 3.2.1-3                                     Revision 0

TSS CN OG-008 FJ(Z) (F. Methodology)
B 3.2.1 BASES LCO            that it can stay within the LOCA F(Z) limits. If Fo(Z) cannot be maintained (continued)   within the LCO limits, reduction of the core power is required.
Ifthe power distribution measurements are performed at a power level less than 100% RTP, then the F8 (Z) and F*' (Z) values that would result from measurements if the core was at 100% RTP should be inferred from the available information. A comparison of these inferred values with F RTP assures compliance with the LCO at all power levels.
Violating the LCO limits for Fo(Z) may produce unacceptable consequences if a design basis event occurs while Fo(Z) is outside its specified limits.
APPLICABILITY  The FI(Z) limits must be maintained in MODE 1 to prevent core power distributions from exceeding the limits assumed in the safety analyses.
Applicability in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require a limit on the distribution of core power.
ACTIONS        A.1 Reducing THERMAL POWER by 2t 1% RTP for each 1% by which FS (Z) exceeds its limit, ma      n ,i acceptable        olute power density. F. (Z) is F" (Z) multiplied          rc which accofr        manufacturing tolerances and measurement un' ertainti-s. F,-(Z) isse measured value of Fo(Z)
The Completion Time of 15 minutes provides an acceptable time to reduce power in an orderly manner and without allowing the plant to remain in an unacceptable condition for an extended period of time.
The maximum allowable power level initially determined by Required Action A.1 may be affected by subsequent determinations of F. (Z) and would require power reductions within 15 minutes of the F8 (Z) determination, if necessary to comply with the decreased maximum allowable power level. Decreases in F8 (Z)) would allow increasing the maximum allowable power level and increasing power up to this revised limit.
(continued)
CALLAWAY PLANT                        B 3.2.1-4                                    Revision 0

TSB CN 04-0OE6 F,(Z) (F. Methodology)
B 3.2.1 BASES SURVEILLANCE       THERMAL POWER at which they were last verified to be within specified REQUIREMENTS       limits. Because F8 (Z) and Fw (Z) could not have previously been (continued)       measured in a reload core, there is a second Frequency condition, applicable only for reload cores, that requires determination of these parameters before exceeding 75% RTP. This ensures that some determination of F8 (Z) and Fw (z) are made at a lower power level at which adequate margin is available before going to 100% RTP. Also, this Frequency condition, together with the Frequency condition requiring verification of F8 (Z) and Fw (Z) following a power increase of more than 10%, ensures that they are verified within 24 hours from when equilibrium conditions at RTP (or any other power level for extended operation) are achieved. Equilibrium conditions are achieved when the core is sufficiently stable such that the uncertainty allowances associated with the measurement are valid. In the absence of these Frequency conditions, it is possible to increase power to RTP and operate for 31 days without verification of F88 (Z) and Fw. (Z). The Frequency condition is not intended to require verification of these parameters after every 10% increase in power level above the last verification. It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which F, was last measured.
SR 3.2.1.1 Verification that F. (Z) is within its specified limits involves increasing Fm (Z) to allow for manufacturing tolerance and measurement power core V.           uncertainties in order to obtain Fa (Z). Specifically, Fmo (Z) is the
                 &#xfd;L!
Q111STrIVIA     On measurer4ent       measured value of F.(Z) obtained                    m. __     results and G E                   8 (Z) = F.m (Z)t  . W     (Ref.
I    4).I (Z) is then compared to its specified
                            #
The vIloC           UK The limit with which FO(Z) is compared varies inversely with power above ir.                     50% RTP and directly with a function called K(Z) provided in the COLR.
,r*VJd o ihe CO.        Performing this Surveillance in MODE 1 prior to exceeding 75% RTP, or at a reduced power level at any other time, and meeting the 100% RTP F,(Z) limit, provides assurance that the F80 (Z) limit is met when RTP is achieved, because peaking factors generally decrease as power level is increased.
(continued)
CALLAWAY PLANT                             B 3.2.1-8                                      Revision 0

F,(Z) (F, Methodology)
B 3.2.1 BASES SURVEILLANCE   SR 3.2.1.1 (continued)
REQUIREMENTS IfTHERMAL POWER has been increased by > 10% RTP since the last determination of FS (Z), another evaluation.of this factor is required within 24 hours after achieving equilibrium conditions (to ensure that Fc (Z) values are being reduced sufficiently with power increase to stay within the LCO limits).
The Frequency, of 31 EFPD is adequate to monitor the change of power distribution with core bumup because such changes are slow and well controlled when the plant is operated in accordance with the Technical Specifications (TS).
SR 3.2.1.2                                                  wer Aistrbujoa vWorn The nuclear design process includes calculations performed to determine e46er at      that the core can be operated Within the F4(z) limits. Becaus or iieir     tah        oeiBc are take       &#xfd;equilibrium conditions, the vari'ations in po r distribution resulting from normal operational maneuvers are not resent in t e ux map data. These variations are, however, conservatively calculated b considering a wide range of unit maneuvers in normal operation.
The maximum peaking factor increase over steady state values, calculated as a function of core elevation, Z, is called W(Z). Multiplying the measured total peaking factor, FS (Z), by W(Z) gives the maximum F(Z) calculated to occur in normal operation, Fw(Z).
The limit with which Fw (Z) is compared varies inversely with power and directly with the function K(Z) provided in the COLR.
The W(Z) curveis prqvided in the COLR for discrete core elevations.
Flux map data are typically taken for 30 to 75 core elevations. Fw WZ) evaluations are not applicable for the following axial core regions, measured in percent of core height:
: a.      Lower core region, from 0 to 15% inclusive; and
: b.      Upper core region, from 85 to 100% inclusive.
The top and bottom 15% of the core are excluded from the evaluation because of the low probability that these regions would be more limiting in the safety analyses and because of the difficulty of making a precise measurement in these regions.
(continued)
CALLAWAY PLANT                          -B 3.2.1-9                                    Revision 2

F (Z) (F, Methodology)
B 3.2.1 BASES REFERENCES     1. 10 CFR 50.46, 1974.
: 2. FSAR, Section 15.4.8.
: 3. 10 CFR 50, Appendix A, GDC 26.
: 4. WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties," June 1988.
S. WCAP-12472-P-A, i BEACONt Ccr          viioi    5   wdOrt~i Suort* Sys4 eM41           ,,us 1594.7 (continued)
CALLAWAY PLANT                     B 3.2.1 -11                            Revision 0

TSB CN 06-009                        FN*
B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F*)
BASES BACKGROUND           The purpose of this LCO is to establish limits on the power density at any point in the core so that the fuel design criteria are not exceeded and the accident analysis assumptions remain valid. The design limits on local (pellet) and integrated fuel rod peak power density are expressed in terms of hot channel factors. Control of the core power distribution with respect to these factors ensures that local conditions in the fuel rods and coolant channels do not challenge core integrity at any location during either normal operation or a postulated accident analyzed in the safety analyses.
FN is defined as the ratio of the integral of the linear power along the fuel rod with the highest integrated power to the average integrated fuel rod power. Therefore, F,"H is a measure of the maximum total power produced in a fuel rod. FN. is sensitive to fuel loading patterns, bank insertion, and fuel bur fnFNsur                FN  is not directl meas        e but is inferred from a power distribution obtained wit the movable incore detector syste          Specifically, the results of the three dimensional power distributio              ore analyzed to The Cdetermine Fpo. This factor is calculated at     ensureveryta EFPD.
However, during power operation, the global power distribution is ea3u r=C:rvenso   monitored by LCO 3.2.3, "Axial Flux Difference (AFD)," and LCO 3.2.4, "k'L'*.-)**'*    Quadrant Power Tilt Ratio (QPTR)," which address directly and continuously measured process variables. Compliance with these LCOs.
along with the LCOs governing shutdown and control rod insertion and alignment, maintains the core limits on power distribution on a continuous basis.
The COLR provides peaking factor limits that ensure that the design basi.
value of the departure from nucleate boiling (DNB) is met for normal operation, operational transients, and any transient condition arising from events of moderate frequency. All DN13 limited transient events are assumed to begin with an F.N value that satisfies the LCO requirements.
rv~oyisorih3 sysin (PriS) (Rejerotinued).
(continued)
CALLAWAY PLANT                               B 3.2.2-1                                     Revision 0

TSB CN 0-0o0B B 3.2.2 BASES ACTIONS        A.1.1 (continued) requirements. In addition, Required Action A.2 is performed if power ascension is delayed past 24 hours.
A.1.2.1 and A.1.2.2 If the value of F. is not restored to within its specified limit either by adjusting a misaligned rod or by reducing THERMAL POWER, the alternative option is to reduce THERMAL POWER to < 50% RTP in accordance with Required Action A. 1.2.1 and reduce the Power Range Neutron Flux - High to *55% RTP in accordance with Required Action A.1.2.2. Reducing power to < 50% RTP increases the DNB margin and does not likely cause the DNBR limit to be violated in steady state operation. The reduction in trip setpoints ensures that continuing operation remains at an acceptable low power level with adequate DNBR margin. The allowed Completion Time of 4 hours for Required Action A.1.2.1 is consistent with those allowed for in Required Action A. 1.1 and provides an acceptable time to reach the required power level from full power operation without allowing the plant to remain in an unacceptable condition for an extended period of time. The Completion Times of 4 hours for Required Actions A.1.1 and A.1.2.1 are not additive.
The allowed Completion Time of 72 hours to reset the trip setpoints per Required Action A.1.2.2 recognizes that, once power is reduced, the safety analysis assumptions are satisfied and there is no urgent need to reduce the trip setpoints; however, for extended operations at the reduced power level, the reduced trip setpoints are required to protect against transients involving positive reactivity excursions. This is a sensitive operation that may inadvertently trip the Reactor Protection System .a                               po e       M b + a me s r            en A.2 Once actions have been taken to restore F,, to within its limits per Required Action A.1.1, or the power leve as been reducetto
               < 50% RTP per Required Action A.1.2.1, b, in                ... *e R 3.2.2.1) must be obtained and the measured value of F," verified not to exceed the allowed limit at the lower power level. The unit is provided 20 additional hours to perform this task over and above the 4 hours allowed by either Action A.1.1 or Action A.1.2.1. The Completion Time of 24 hours is acceptable because of the increase in the DNB margin, which (continued)
CALLAWAY PLANT                        B 3.2.2-5                                     Revision 0

TSB CN 06B-005                             FH B 3.2.2 BASES ACTIONS         A.2 (continued) is obtained at lower power levels, and the low probability of having a DNB limiting event within this 24 hour period. Additionally, operating experience has indicated that this Completion Time is sufficient to obtain (cjn          core flux map, perform the required calculations, and evaluate FH A.3 Verification that FN is within its specified limits after an out of limit occurrence ensures that the cause that led to the F"H exceeding its limit is identified, to the extent necessary, and corrected, and that subsequent operation proceeds within the LCO limit. This Action demonstrates that the F", limit is within the LCO limits prior to exceeding 50% RTP, again prior to exceeding 75% RTP, and within 24 hours after THERMAL POWER is > 95% RTP. SR 3.2.2.1 must be satisfied prior to increasing power above the allowable power level or restoration of any reduced Reactor Trip System setpoints. When F. is measured at reduced power levels, the allowable power level is determined by evaluating FN,, for higher power levels.
This Required Action is modified by a Note that states that THERMAL POWER does not have to be reduced prior to performing this Action.
B. I When Required Actions A.1.1 through A.3 cannot be completed within their required Completion Times, the plant must be placed in a mode in which the LCO requirements are not applicable. This is done by placing the plant in at least MODE 2 within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience regarding the time required to reach MODE 2 from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE   SR 3.2.2.1 REQUIREMENTS SR 3.2.2.1 is modified by a Note. The Note applies during power ascensions following a plant shutdown (leaving Mode 1). The note allows for power ascensions if the surveillances are not current. It states that THERMAL POWER may be increased until an equilibrium power level has been achieved at which a power distribution map can be obtained.
Equilibrium conditions are achieved when the core is sufficiently stable (continued)
CALLAWAY PLANT                         B 3.2.2-6                                      Revision 0

7S5 C?4 OC-006 B 3.2.2 BASES SURVEILLANCE      SR 3.2.2.1 (continued)
REQUIREMENTS such that the uncertainty allowances associated with the measurement are valid.
or  &#xfd;rv               eThe      value of FIH is determined by     i           vable incore detector 4nsouiwJ-eo system to obtain a flux distribution ma A data reduction computer provide4 by an OPEKABLE    program then calculates the maximum value of FNH from the measured PD.                ,flu                        The measured value of FN does not require a correction for measurement uncertainty before making comparisons to the FNH limit since a measurement uncertainty of 40 has been included in dis'ribu'+'on tvp,      the FN limit.
After each refueling, FNH must be determined in MODE 1 prior to exceeding 75% RTP. This requirement ensures that F" limits are met at the beginning of each fuel cycle. Performing this Surveillance in Mode 1 fl-Al                        prior to exceeding 75% RTP, or at a reduced power level at any other time, and verifying the inferred results for 100% RTP meet the 100% RTP or PDMS-Aerived      FiH values (Relcrence+)]         FH limit, provides assurance that F,,H limit will be met when RTP is Cappi"Icable    e4er       achieved, because peaking factors generally decrease as power level is W                      increased.
The 31 EFPD Frequency is acceptable because the power distribution changes relatively slowly over this amount of fuel burnup. Accordingly, this Frequency is short enough that the FNH limit cannot be exceeded for any significant period of operation.
REFERENCES       1.       FSAR, Section 15.4.8.
: 2.       10 CFR 50, Appendix A, GDC 26.
: 3.       10 CFR 50.46.
: 4. WCAP- 124-7Z-P-A[,,BEACON4 Core 14o,4Ior~n3 and Oj.rafionS Suvrrt Syj5teM1, Au~ust 1151C CALLAWAY PLANT                         B 3.2.2-7                                   Revision 0

TSB Co0(-008 QPTR B 3.2.4 BASES ACTIONS               A.1 (continued)
QPTR would allow raising the maximum allowable THERMAL POWER level and increasing THERMAL POWER up to this revised limit.
A.2 After completion of Required Action A. 1, the QPTR may still exceed its limits. As such, any additional changes in the QPTR are detected by requiring a check of the QPTR once per 12 hours thereafter. If the QPTR continues to increase, THERMAL POWER has to be reduced accordingly. A 12 hour Completion Time is sufficient because any additional change in QPTR would be relatively slow.
A.3 The peaking factors F" and FJ(Z), as approximated by Fl (Z) and oIkialni%5'.a power                Fw (z); are of primary importance in ensuring that the power distribution
  "*1sibu*.      measurement.,       remains consistent with the initial conditions used in the safety analyses.
Power aiswiebaoa ,nformva4o;o       Performing SRs on FN, and F,(Z) within the Completion Time of 24 hours can be . obizined esther by         after achieving equilibrium conditions from a THERMAL POWER tt.eisi +Ue moyable incare          reduction per Required Action A.1 ensures that these primary indicators
( t=-"orf or f        an   OPEfRABLE of power distribution are within their respective limits. Equilibrium power diSV-bji'on mnov'iforb'a       conditions are achieved when the cor sUfficiently stable at the i        .       in ended operating conditions to suppo                     A Completion
      -- (PDMS)(I.teference sysie-M                    ..... Time of 24 hours after achieving equilibrium conditions from a THERMAL
  -      i       -*lPOWER                       reduction per Required Action A. 1 takes into consideration the rate at which peaking factors are likely to change, and the time required to stabilize the plant and perfor        ux ma If these peaking factors are not within  their limits, the   equ red Actions associated with these power 4,1c*r'11%flonb           Surveillances provide an appropriate response for the abnormal vv."
       *ca~u~revv      . ,-condition.             Ifthe QPTR remains above its specified limit, the peaking factor surveillances are required each 7 days thereafter to evaluate F' and Fo(Z) with changes in power distribution. Relatively small changes are expected due to either burnup and xenon redistribution or correction of the cause for exceeding the QPTR limit.
(continued)
CALLAWAY PLANT                                  B 3.2.4-3                                 Revision 0

TSB CN OG-OoB QPTR B 3.2.4 BASES ACTIONS        A.4 (continued)
Although F" and F.(Z) are of primary importance as initial conditions in the safety analyses, other changes in the power distribution may occur as the QPTR limit is exceeded and may have an impact on the validity of the safety analysis. A change in the power distribution can affect such reactor parameters as bank worths and peaking factors for rod malfunction accidents. When the QPTR exceeds its limit, it does not necessarily mean a safety concern exists. It does mean that there is an indication of a change in the gross radial power distribution that requires an investigation and evaluation that is accomplished by examining the incore power distribution. Specifically, the core peaking factors and the quadrant tilt must be evaluated because they are the factors that best characterize the core power distribution. This re-evaluation is required to ensure that, before increasing THERMAL POWER above the limit of Required Action A. 1, the reactor core conditions are consistent with the assumptions in the safety analyses.
A.5 If the QPTR remains above the 1.02 limit and a re-evaluation of the safety analysis is completed and shows that safety requirements are met, the excore detectors are normalized to restore QPTR to within limit prior to increasing THERMAL POWER to above the limit of Required Action A. 1.
This is done to detect any subsequent significant changes in QPTR.
Normalization is accomplished in such a manner that the indicated QPTR following normalization is near 1.00.
Required Action A.5 is modified by two Notes. Note 1 states that excore detectors are not normalized to restore QPTR to within limit until after the re-evaluation of the safety analysis has determined that core conditions at RTP are within the safety analysis assumptions (i.e., Required Action A.4). Note 2 states that if Required Action A.5 is performed, then Required Action A.6 shall be performed. Required Action A.5 normalizes the excore detectors to restore QPTR to within limit, which restores
   ~ornpliance                witJ* 03.2.4. Th s, Note 2 prevents exiting the Actions rnIIsAreMnt)     prior to completino                  verify peaking factors per Required Action A.6. These Notes are intended to prevent any ambiguity about the required sequence of actions.
(continued)
CALLAWAY PLANT                       B 3.2.4-4                                   Revision 0

TSB      4cN o-OOB                       QPTR B 3.2.4 BASES ACTIONS          A.6 (continued)
Once the excore detectors are normalized to restore QPTR to within limit (i.e., Required Action A.5 is performed), it is acceptable to return to full power operation. However, as an added check that the core power distribution is consistent with the safety analysis assumptions, Required Action A.6 requires verification that F(Z), as approximated by F8 (Z) and F* (Z), and FNH are within their specified limits within 24 hours of achieving equilibrium conditions at RTP. Equilibrium conditions at RTP A6 aini      a power are achieved when t e core is sufficiently stable at the intended operating dconditions Mi Sir,'6fit.                   to suppor,-
does not reach                .,...,s      an added precaution, if the core equilibrium conditions at RTP within 24 hours, but power is increased slowly, then the peaking factor surveillances must be performed within 48 hours after increasing THERMAL POWER above the limit of Required Action A. 1. These Completion Times are intended to allow adequate time to increase THERMAL POWER to above the limit of Required Action A. 1, while not permitting the core to remain with unconfirmed power distributions for extended periods of time.
Required Action A.6 is modified by a Note that states that the peaking factor surveillances must be completed when the excore detectors have been normalized to restore QPTR to within limit (i.e., Required Action A.5). The intent of this Note is to have the peaking factor surveillances performed at operating power levels, which can only be accomplished after the excore detectors are normalized to restore QPTR to within limit.
B.1 If Required Actions A. 1 through A.6 are not completed within their associated Completion Times, the unit must be brought to a MODE or condition in which the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to < 50% RTP within 4 hours. The allowed Completion Time of 4 hours is reasonable, based on operating experience regarding the amount of time required to reach the reduced power level without challenging plant systems.
SURVEILLANCE     SR 3.2.4.1 REQUIREMENTS SR 3.2.4.1 is modified by two Notes. Note 1 allows QPTR to be calculated with three power range channels if THERMAL POWER is
                     <75% RTP and the input from one Power Range Neutron Flux channel is inoperable. Note 2 allows performance of SR 3.2.4.2 in lieu of SR 3.2.4.1.
(continued)
CALLAWAY PLANT                          B 3.2.4-5                                   Revision 0

TSB CN OG-00                        QPTR B 3.2.4 BASES SURVEILLANCE   SR 3.2.4.2 (continued)
REQUIREMENTS change in tilt has actuall occurred, which might cause the QPTR limit to
 "    E   s    beae     eded, the incor result may be compared against previousd- W9 either using the symmetric thimbles as described above or a complete flux map. Nominally, quadrant tilt from the Surv illance should be within 2% of the tilt shown by the most recent            data.
REFERENCES     1.      10 CFR 50.46.                p,*,,,,. ,i,*rbi+',i ,,,,-ir         t
: 2.      FSAR, Section 15.4.8.
: 3.      10 CFR 50, Appendix A, GDC 26.
: 4. Westinghouse Recommendations on Monitoring QPTR with One S       Power Range Channel Out of Service, (Proprietary).
: 5. WCAP- 1Z472.-P-A. a BEACON Core Movilorin xv4 O~erarictns Support SyS+4M," Au~ust 14)r CALLAWAY PLANT                       B 3.2.4-7                                    Revision 0

TSB CH 06-OO8 RTS Instrumentation B 3.3.1 BASES ACTIONS          D.1.1. D.1.2, and D.2 (continued)
Condition D applies to the Power Range Neutron Flux - High trip Function.
With one of the NIS power range detectors inoperable, 1/4 of the radial power distribution monitoring capability is lost. Therefore, SR 3.2.4.2 must be performed (Required Action D.11.1) within 12 hours of THERMAL POWER exceeding 75% RTP and once per 12 hours thereafter. If reactor power decreases to < 75% RTP, the measurement of both Completion Times for Required Action D.1.1 stops and SR 3.2.4.2 is no longer required. Completion Time tracking recommences upon reactor power exceeding 75% RTP. Calculating QPTR every 12 hours compensates for the lost monitoring capability due to the inoperable NIS power range channel and allows continued plant operation at power levels > 75% RTP.
At power levels 5 75% RTP, operation of the core with radial power distributions beyond the design limits, at a power level where DNB conditions may exist, is prevented. The 12 hour Completion Time is consistent with the SR 3.2.4.2 Frequency in LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)."
Required Action D.1.1 has been modified by a Note which only requires SR 3.2.4.2 to be performed if the Power Range Neutron Flux input to QPTR becomes inoperable. Failure of a component in the Power Range Neutron Flux Channel which renders the High Flux Trip Function inoperable may not affect the capability to monitor QPTR. As such, determining QPTR using*honce                     .                per 12 hours Smay not be necessary.,,
The NIS power range detectors provide input to the Rod Control System Jis4ri6t+'i'an     and the SG Water Level Control System and, therefore, have a two-out-of-four trip logic. A known inoperable channel must be placed in 0% r r^'A4 on     the tripped condition. This results in a partial trip condition requiring only W  core power        one-out-of-three logic for actuation. The 72 hours allowed to place the inoperable channel in the tripped condition is justified in Reference 17.
As an alternative to the above Actions, the plant must be placed in a MODE where this Function is no longer required OPERABLE.
Seventy-eight (78) hours are allowed to place the plant in MODE 3. The 78-hour Completion Time includes 72 hours for channel corrective maintenance, and an additional 6 hours for the MODE reduction as required by Required Action D.2. This is a reasonable time, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging plant systems. If Required Actions cannot be completed within their allowed Completion Times, LCO 3.0.3 must be entered.
(continued)
CALLAWAY PLANT                         B 3.3.1-37                                  Revision 5e

TSB CN OG-oo RTS Instrumentation B 3.3.1 BASES SURVEILLANCE     SR 3.3.1.4 (continued)
REQUIREMENTS capability is provided for performing such a test at power. The independent test for bypass breakers is included in SR 3.3.1.14. The bypass breaker test shall include a local manual shunt trip only. A Note has been added to indicate that this test must be performed on the bypass breaker prior to placing it in service.
The Frequency of every 62 days on a STAGGERED TEST BASIS is justified in Reference 18.
SR 3.3.1.5 SR 3.3.1.5 is the performance of an ACTUATION LOGIC TEST. The SSPS is tested every 92 days on a STAGGERED TEST BASIS, using the semiautomatic tester. The train being tested is placed in the bypassed condition, thus preventing inadvertent actuation. Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function, including operation of the P-7 permissive which is a logic function only. The Frequency of every 92 days on a STAGGERED TEST BASIS is justified in Reference 18.
The incore power             SR3..16_owr____6A~o distr~bufoon vvieasurc,&#xb6;ens SR33..
can be obtalied
          ,,,in              SR 3.3.1.6 is a calibration of the excore channels to the incore channels.
 +b wnovable incore          If the measurements do not agree, the excore channels are not declared deicctors or n OPERlIABLE   inoperable t must be calibrated to agree with the incore           __
power dis+6riAb-ieo         measurements. Ifthe excore channels cannot be adjusted, the channels r=____;n" S.        t .....
(P-_"   are declared inoperable. This Surveillance is performed to verify the f(AI)
(RMef" erecP)e               input to the Overtemperature AT Function. Determination of the loop-(Rc.*=rence, aa), *specific            vessel AT and Tag values should be made when performing this calibration, under steady state conditions (ATo and T' [T" for Overpower AT] when at 100% RTP).
A Note modifies SR 3.3.1.6. The Note states that this Surveillance is required only if reactor power is _>75% RTP and that 72 hours after achieving equilibrium conditions with THERMAL POWER _75% RTP is oL'ain a power        allowed for performing the first surveillance. Equilibrium conditions are disiridlon            achieved when the core is sufficiently stable at intended operating The SR is deferred until a scheduled testing plateau above 75% RTP is attained during a power ascension. During a typical power ascension, it (continued)
CALLAWAY PLANT                          B 3.3.1-53                                Revision 5e

TS5 CN 06-008 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE   SR 3.3.1.6 (continued)                      power dAis+yrlib$        wasuremeut REQUIREMENTS is usually necessary to control the axia ux difference at lower power levels through control rod insertion. Afte qulibdur conditions are achieved at the specified power plateau, a                ust be taken and the required data collected. The data is typically analyzed and the appropriate excore calibrations completed within 48 hours after achieving equilibrium conditions. An additional time allowance of 24 hours is provided during which the effects of equipment failures may be remedied and any required re-testing may be performed.
The allowance of 72 hours after equilibrium conditions are attained at the testing plateau provides sufficient time to allow power ascensions and associated testing to be conducted in a controlled and orderly manner at conditions that provide acceptable results and without introducing the potential for extended operation at high power levels with instrumentation that has not been verified to be OPERABLE for subsequent use.
The Frequency of 92 EFPD is adequate. Itis based on industry operating experience, considering instrument reliability and operating history data for instrument drift.
SR 3.3.1.7 SR 3.3.1.7 is the performance of a COT every 184 days.
A COT is performed on each required channel to ensure the channel will perform the intended Function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL OPERATIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
Setpoints must be within the Allowable Values specified in Table 3.3.1-1.
SR 3.3.1.7 is modified by two Notes. Note I provides a 4 hour delay in the requirement to perform this Surveillance for source range instrumentation when entering MODE 3 from MODE 2. This Note allows a normal shutdown to proceed without a delay for testing in MODE 2 and for a short time in MODE 3 until the Applicability is exited and SR 3.3.1.7 is no longer required to be performed. Ifthe unit is to be in MODE 3 with the Rod Control System capable of rod withdrawal of one or more rods (continued)
CALLAWAY PLANT                       B 3.3.1-54                                   Revision 5e

TS5 CN 06-006 RTS Instrumentation B 3.3.1 BASES REFERENCES     10. FSAR Table 15.0-4.
(continued)
: 11. WCAP-9226-P-A, "Reactor Core Response to Excessive Secondary Steam Releases," Revision 1, February 1998.
: 12. Deleted.
: 13. FSAR Section 15.1.1.
: 14. RFR- 18637A.
: 15. WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," October 1998.
: 16. FSAR Section 15.4.6.
: 17. WCAP-1 4333-P-A, Revision 1, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times," October 1998.
: 18. WCAP-1 5376-P-A, Revision 1, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," March 2003.
: 19. Westinghouse letter SCP-04-90 dated August 27, 2004.
: 20. ULNRC-03748 dated February 27, 1998.
: 21. IDP-ZZ-0017.
CALLAWAY PLANT                  B 3.3.1-63                              Revision 5e

ATTACHMENT 6 ULNRC-05297 MARKUP OF FSAR PAGES

FSAR CN O-017 CALLAWAY - SP            No cmys io 411             f 7.7.1.8.2       Plant Trip Steam Dump Controller                proYIded -r     covikx-/c Following a reactor trip, the load rejection steam dump controller is defeated, and the plant trip steam dump controller becomes active. Since control rods are not available in this situation, the demand signal is the error signal between the lead-lag compensated auctioneered Tavg and the no-load reference Tavg. When the error signal exceeds a predetermined setpoint, the dump valves are tripped open in a prescribed sequence. As the error signal reduces in magnitude, indicating that the RCS Tavg is being reduced toward the reference no-load value, the dump valves are modulated by the plant trip controller to regulate the rate of removal of decay heat and thus gradually establish the equilibrium hot shutdown condition.
7.7.1.8.3       Steam Header Pressure Controller Residual heat removal at operating temperature is maintained by the steam generator pressure controller (manually selected) which controls the amount of steam flow to the condensers. This controller operates a portion of the same steam dump valves to the condensers which are used during the initial transient following turbine or reactor trip on load rejection.
7.7.1.9         Incore Instrumentation The incore instrumentation system consists of chromel-alumel thermocouples (described in Section 18.2.13.2) at fixed core outlet positions and movable miniature neutron detectors which can be positioned at the center of selected fuel assemblies, anywhere along the length of the fuel assembly vertical axis. The basic system for insertion of these detectors is shown in Figure 7.7-9.
7.7.1.9.1       Thermocouples Chromel-alumel thermocouples are threaded into guide tubes that penetrate the reactor vessel head through seal assemblies, and terminate at the exit flow end of the fuel assemblies. The thermocouples are provided with two primary seals-a conoseal and swage-type seal from conduit to head. Thermocouple readings are monitored by the computer.
7.7.1.9.2       Movable Neutron Flux Detector Drive System Miniature fission chamber detectors can be remotely positioned in retractable guide thimbles to provide flux-mapping of the core. The stainless steel detector shell is welded to the leading end of the helical wrap drive cable and to stainless steel sheathed coaxial cable. The retractable thimbles, into which the miniature detectors are driven, are pushed into the reactor core through conduits which extend from the bottom of the reactor vessel down through the concrete shield area and then up to a thimble seal table.
Their distribution over the core is nearly uniform with about the same number of thimbles located in each quadrant.
7.7-19                             Rev. OL-14c 04/05

FSAR CN O-O17 SP         o chan~esro+o Nproviaed           t1'M ls pa e -
CALLWAY                           con~xt/coui+rv; The thimbles are closed at the leading ends, are dry inside, and serve as the pressure barrier between the reactor water pressure and the atmosphere. For each thimble, an in-line magnetic isolation ball check valve, located above the seal table, provides a second barrier between the reactor water pressure and the atmosphere. Mechanical seals between the retractable thimbles and the conduits are provided at the seal table.
During reactor operation, the retractable thimbles are stationary. They are extracted downward from the core during refueling to avoid interference within the core. A space above the seal table is provided for the retraction operation.
The drive system for the insertion of the miniature detectors consists basically of drive assemblies, 6-path transfer assemblies, and 15-path transfer assemblies, as shown in Figure 7.7-9. The drive system pushes hollow helical wrap drive cables into the core with the miniature detectors attached to the leading ends of the cables and small diameter sheathed coaxial cables threaded through the hollow centers back to the ends of the drive cables. Each drive assembly consists of a gear motor which pushes a helical wrap drive cable and a detector through a selective thimble path by means of a special drive box and includes a storage device that accommodates the total drive cable length.
Each detector has access to all thimble locations via the 6- and 15-path rotary assemblies.
7.7.1.9.3       Control and Readout Description The control and readout system provides means for inserting the miniature neutron detectors into the reactor core and withdrawing the detectors while plotting neutron flux versus detector position. The control system is located in the control room. Limit switches in each transfer device provide feedback of path selection operation. Each gear box drives a resolver for position feedback. One 6-path transfer selector is provided for each drive unit to insert the detector in one of six functional modes of operation. One 15-path transfer is also provided for each drive unit that is then used to route a detector into any one of up to 15 selectable paths. A common path is provided to permit cross calibration of the detectors.
The control room contains the necessary equipment for control, position indication, and flux recording for each detector.
A "flux-mapping" consists, briefly, of selecting flux thimbles in given fuel assemblies at various core quadrant locations. The detectors are driven to the top of the core and stopped automatically. An x-y plot (position versus flux level) is initiated with the slow withdrawal of the detectors through the core from top to a point below the bottom. In a similar manner, other core locations are selected and plotted. Each detector provides axial flux distribution data along the center of a fuel assembly.
Various radial positions of detectors are then compared to obtain a flux map for a region of the core.
7.7-20                              Rev. OL-14c 04/05

FSAR CN 06-017 tp,CALLAWAY                                                       - SP The number and location of these thimbles have been chosen to permit measurement of local-to-,average peaking factors to an accuracy of +/-5 percent (95-percent confidence).
Measured nuclear peaking factors will be increased by 5 percent to allow for this accuracy. Ifthe measured power peaking is larger than acceptable, reduced power capability will be indicated.
Operating plant experience has demonstrated the adequacy of the incore instrumentation in meeting the design bases stated.
7.7.1.10        Boron Concentration Monitoring System The boron concentration monitoring system utilizes a sampler assembly unit which contains a neutron source and neutron detector located in a shield tank. A thermal neutron absorption technique is used. Piping within the shield tank is arranged to provide coolant sample flow between the neutron source and the neutron detector.
Neutrons originating at the source are thermalized in the sample and the surrounding moderator. These neutrons then pass through the sample and impinge upon the detector. The number of neutrons which survive the transit from the source to the detector is inversely proportional to the boron concentration in the sample. The boron concentration is calculated by monitoring the neutron countrate in conjunction with the proper transfer function. The neutron cross-section of the boron in the sample is also a function of the neutron energy and, subsequently, the sample temperature. Therefore, the sample temperature is also monitored and the transfer function from the neutron countrate to boron concentration modified to compensate for the variance of temperature.
The processor assembly is used to convert the neutron countrate and temperature data from the sampler assembly to parts per million (ppm) of boron, and to prepare the data for local and remote display. The system characteristics are listed in Table 7.7-2.
: a. Sampler assembly The sampler assembly consists of a polyethylene cylinder encased in a iN s E F"
* stainless steel liner (see Figure 7.7-10). The polyethylene serves as a ATTACHED    ,neutron              moderator and shield. A cavity (source tube) is located in the ATTACK..                 center of the shield into which Is inserted a neutron source on the end of a N W SOMSECTION polyethylene              rod (source plug). Immediately adjacent to the source tube is a second larger cavity into which an annulus assembly and a top plug 7,"7* 94.              assembly are inserted. Details of these two assemblies are shown in
                            .Figure 7.7-11.
The annulus assembly consists of two concentric tubes with top and bottom plates. A neutron detector is positioned inside the smaller tube.
The coolant sample is circulated between the concentric tubes. The sample is brought into and taken out of the annular region via tubes 7.7-21                             Rev. OL-14c 04/05

FSAR, CN 06-017i 7.7.1.9.4     Power Distribution Monitoring System As an enhancement to power distribution measurement and indication capability, the power distribution monitoring system (PDMS) is provided, which consists of a set of coupled but independent computer software programs that execute on one or more workstations to generate an on-line, three-dimensional indication of the core power distribution. The PDMS uses the flux map together with a three-dimensional analytical model to yield the continuously measured three-dimensional power distribution. The movable incore neutron detectors are used to calibrate the PDMS.

FSAR. C14 oC-017 CALLAWAY - SP 16.2          POWER DISTRIBUTION LIMITS 16.2.2        QUADRANT POWER TILT RATIO ALARM 16.2.2.1      LIMITING CONDITION FOR OPERATION The QUADRANT POWER TILT RATIO (QPTR) Alarm shall be OPERABLE.
APPLICABILITY: MODE 1 with THERMAL POWER a 50% RTP NOTE: With input from one Power Range Neutron Flux channel inoperable and THERMAL POWERR. 75% RTP, the remaining three power range channels can be used for calculating QPTR.                                   cort power dsbrblahmfon ACTION:                                       utmIeof imijormiaf ion With The QPTR Alarm inoperable, verify QPTR is within limit by calculation or by using Ahs-io rovblc~sea ir.3oro 'i~ ~lo   nce per 12 hours during steady-state operation.
16.2.2.1.1     SURVEILLANCE REQUIREMENTS Not Applicable 16.2.2.1.2     BASES QPTR is required to be verified within limits on a nominal basis of once per 7 days per Technical Specification SR 3.2.4.1, with more frequent verifications required if the QPTR Alarm is inoperable. The more frequent verifications are sufficient to ensure QPTR to be within assumed limits.
The completion time for calculating QPTR when the QPTR Alarm is inoperable is adequate to detect any relatively slow changes in QPTR, because for those causes of QPT that occur quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt.
With one excore detector inoperable, the remaining three detectors shall be used for computing the average.
16.2-2                               Rev. OL-14 12/04

FSAK CMJ 04-017 CALLAWAY - SP 16.3.3         MONITORING INSTRUMENTATION 16.3.3.1       MOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION The Movable Incore Detection System shall be OPERABLE with:
: a.      At least 75% of the detector thimbles,
: b.      A minimum of two detector thimbles per core quadrant, and
: c.      Sufficient movable detectors, drive, and readout equipment to map these thimbles.
APPLICABILITY: When the Movable Incore Detection System is used for:
: a.      Recalibration of the Excore Neutron Flux Detection System, or
: b.      Monitoring the QUADRANT POWER TILT RATIO, or N
: c.      Measurement of FAH .and FQ(Z)
ACTION:
: a.      With the Movable Incore Detection System inoperable, do not use the system for the above applicable monitoring or calibration functions.
: b.      The provisions of Sections 16.0.1.3 and 16.0.1.4 are not applicable.
16.3.3.1.1     SURVEILLANCE REQUIREMENTS The Movable Incore Detection System shall be demonstrated OPERABLE at least once per 24 hours by normalizing each detector output when required for:
: a.      Recalibration of the Excore Neutron Flux Detection System, or
: b.      Monitoring the QUADRANT POWER TILT RATIO, or N
: c.      Measurement of FAH .and FQ(Z).
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16.3-3                              Rev. OL-14h 3/06

FSAR CN OG-017 CALLAWAY - SP 16.3.3.1.2     BASES The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the core. The OPERABILITY of this system is demonstrated by irradiating each detector used and cjT *determining the acceptability of its voltage curve. For the purpose of measuring FQ(Z) or FAH  a full incore flux map is used. Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the Excore Neutron Flux Detection System, and full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one or more Power Range Neutron Flux Channels are inoperable.
INSERT ADDITIONAL TEXT (ATTACKED) 16.3-4                               Rev. OL-14h 3/06

FSAK CH4 06-0I1 JV TEXT INSERT for FSAR Section 16.3.3.1.2 (BASES for FSAR Section 16.3.3.1)
Power distribution measurement is generally performed, however, via the power distribution monitoring system (PDMS) when thermal power is greater than 25 percent rated thermal power (RTP). At thermal power levels less than 25 percent RTP, or when the PDMS is inoperable, the movable incore detector system is used. Since the movable incore detector system is employed by the PDMS, it shall be operable before it is used to calibrate the PDMS.

CALLAWAY - SP 16.3.3.8         POWER DISTRIBUTION MONITORING SYSTEM LIMITING CONDITION FOR OPERATIONt The power distribution monitoring system (PDMS) and associated instrumentation shall be OPERABLE with the minimum required channels from the plant computer as specified in Table 16.3-9, whenever the PDMS is used to perform core power distribution monitoring and related surveillances required by the Technical Specifications.
APPLICABILITY:             When the PDMS is used to perform core power distribution monitoring and related surveillances required by the Technical Specifications during MODE 1 with reactor THERMAL POWER greater than or equal to 25% of RATED THERMAL POWER (RTP).
ACTION:
: a.        With one or more Functions with one or more required channels inoperable, declare the PDMS inoperable.
: b.        The provisions of Sections 16.0.1.4 are not applicable.
16.3.3.8.1        SURVEILLANCE REQUIREMENTS The PDMS shall be verified OPERABLE:
: a.        By performance of a CHANNEL CHECK of the PDMS input instrumentation specified in Table 16.3-9, prior to use of the PDMS for core power distribution measurement purposes;
: b.        By performance of a calibration of the PDMS following each refueling outage, and thereafter:
: 1)     Every 31 EFPD with only minimum thermocouple coverage, or
: 2)     Every 180 EFPD with optimum thermocouple coverage;
: c.        By verifying that a CHANNEL CALIBRATION has been performed for each required instrumentation channel listed in Table 16.3-9 in accordance with the instrument's applicable Technical Specification Surveillance Requirement(s),
prior to initial use of the PDMS for core power distribution measurement purposes following each refueling outage.
16.3-

.6 CALLAWAY - SP 16.3.3.8.2      BASES The power distribution monitoring system (PDMS) is used for periodic measurement of the core power distribution to confirm operation within design limits, and for periodic calibration of the excore detectors. The system does not initiate any automatic protection action.
Specifically, the PDMS generates a continuous measurement of the core power distribution and requires information on current plant and core conditions in order to determine the core power distribution using the core peaking factor measurement and measurement uncertainty methodology described in WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994. The core and plant condition information, which includes reactor power level, average reactor vessel inlet temperature, control bank positions, the Power Range Detector calibrated voltage values, and input from the Core Exit Thermocouples, is used as input to the continuous core power distribution measurement software (i.e., BEACON) that continuously and automatically determines the current core peaking factor values, including the most limiting factors, FAN and FQ(Z). The measured peaking factor values are provided at nominal one-minute intervals to allow operators to confirm that the core peaking factors are within design limits. The peaking factor limit margins include measurement uncertainty which bounds the actual measurement uncertainty of an OPERABLE PDMS.
Operability of the PDMS is dependent on the specified number of channels from the plant computer for each Function listed in Table 6.3-9. The PDMS is OPERABLE when the required channels are available, when it has been calibrated with an incore flux map within the required frequency, and when the reactor power level is at least 25% of RTP.
The PDMS must be calibrated above 25% RTP to assure the accuracy of the calibration data set which can be generated from an incore flux map, core exit thermocouples, and the other instrumentation channels. Below 25% RTP, the PDMS is inoperable since the calculated power distribution is of reduced accuracy and may not be bounded by the uncertainties documented in WCAP-12472.
With one or more required channels from the plant computer inoperable or unavailable as input to the PDMS (or if the PDMS is inoperable for any other reason), the PDMS must be declared inoperable and shall not be used to perform a core power distribution measurement.
16.3-

CALLAWAY - SP Surveillance Requirements are specified for the PDMS to support or provide assurance of its OPERABILITY, as follows:
: a.      Performance of a CHANNEL CHECK for the PDMS input instrumentation prior to using the PDMS to obtain a core power distribution measurement ensures that no gross instrumentation failure has occurred, thus providing added assurance that the required inputs to the PDMS are available. A featuire of the PDMS is its capability to automatically check its required inputs and provide a status indication that confirms all required inputs are available and working. The CHANNEL CHECK requirement can thus be satisfied by checking the BEACON status indicator on the BEACON PDMS monitor screen prior to use of the PDMS for core power distribution measurement purposes.
: b.        Upon initial plant startup following refueling, the PDMS uses a calibration data set calculated by the core designer for the new core. An accurate incore flux map for PDMS calibration may be obtained upon exceeding 25% RTP. The initial calibration data set generated in each operating cycle must utilize incore flux measurements from at least 75% of the incore thimbles, with at least two incore thimbles in each core quadrant. The incore flux measurements in combination with inputs from the Table 16.3-9 channels are used to generate the updated calibration data set, including the nodal calibration factors and the thermocouple mixing factors.
After completion of the first PDMS calibration following refueling, the PDMS must continue to be periodically calibrated at least once per 31 EFPD when only the minimum core exit thermocouple coverage is available for the PDMS calibration (which requires at least 13 thermocouples available with a minimum of two per core quadrant), or at least once per 180 EFPD when the optimum core exit thermocouple coverage is available for the PDMS calibration. Optimum thermocouple coverage also requires at least 13 thermocouples available with a minimum of two per core quadrant, but it also includes coverage of all interior fuel assemblies such that they are all within a knight's move from an operable thermocouple. (Coverage of fuel assemblies with a face along the baffle is not required.).
With regard to thermocouple coverage, the PDMS software automatically analyzes the available thermocouple coverage, consistent with the above criteria, and determines the next surveillance interval for calibration (i.e., 31 or 180 EFPD). Typically, the required interval is reported in the surveillance report that is generated upon completion of the periodic core power distribution measurement.
16.3-

CALLAWAY - SP
: c. Verification that a CHANNEL CALIBRATION ha 'been performed for each of the instrument channels that provide input to the PDMS provides assurance of channel operability and accurate input to the PDMS prior to use of the PDMS for core power distribution measurement purposes following each refueling outage.
The intent is that each instrument channel has been subject to a CHANNEL CALIBRATION at the frequency or interval required by the instrument's associated Technical Specification Surveillance Requirement(s). This can be verified by confirming that the instruments' surveillances are "current" with respect to their specified test intervals, based on their last performance and satisfactory completion (as scheduled per the plant surveillance program). A CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The CHANNEL CALIBRATION for the power range neutron detectors consists of a normalization of the detectors based on a power calorimetric and flux map performed above 15% RTP.
16.3-

TABLE 16.3-9 POWER DISTRIBUTION MONITORING SYSTEM INSTRUMENTATION FUNCTION                                   REOUIRED CHANNELS
: 1. Control Bank Position                                            4(a)
: 2. RCS Cold Leg Temperature, TCOed                                   2
: 3. Reactor Power Level                                              1cb)
: 4. NIS Power Range Excore Detector                                  6 (c)
Section Signals
: 5. Core Exit Thermocouple Temperatures                               13 with > 2 per core quadrant (a)  Control Bank position inputs may be bank positions from either valid Demand Position indications of the average of all valid individual RCCA positions in the bank determined from Rod Position Indication (RPI) System values for each Control Bank. A maximum of one rod position indicator per group may be inoperable when RCCA position indications are being used as input to the PDMS.
(b)  Reactor Power Level input may be reactor thermal power derived from either a valid secondary calorimetric measurement, the average Power Range Neutron Flux Power, or the average RCS Loop AT.
(c)   The total must consist of three pairs of corresponding upper and lower detector sections.}}