ML100820517
| ML100820517 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 07/23/2010 |
| From: | Balwant Singal Plant Licensing Branch IV |
| To: | Matthew Sunseri Wolf Creek |
| Singal, Balwant, 415-3016, NRR/DORL/LPL4 | |
| References | |
| TAC ME2375 | |
| Download: ML100820517 (23) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 23, 2010 Mr. Matthew W. Sunseri President and Chief Executive Officer Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, KS 66839
SUBJECT:
WOLF CREEK GENERATING STATION -ISSUANCE OF AMENDMENT RE:
REVISION TO TECHNICAL SPECIFICATIONS FOR USE OF BEACON POWER DISTRIBUTION MONITORING SYSTEM (TAC NO. ME2375)
Dear Mr. Sunseri:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 188 to Renewed Facility Operating License No. NPF-42 for the Wolf Creek Generating Station. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated October 10, 2009, as supplemented by letter dated March 8, 2010.
The amendment revises TS 3.1.7, "Rod Position Indication," TS 3.2.1, "Heat Flux Hot Channel Factor (Fa{Z)) {Fa Methodology)," TS 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor {FN~H)'"
TS 3.2.4, "Quadrant Power Tilt Ratio {QPTR)," and TS 3.3.1, "Reactor Trip System (RTS)
Instrumentation," to permit use of the Westinghouse computer code, the Best Estimate Analyzer for Core Operations - Nuclear (BEACON) power distribution monitoring system (PDMS), as described in WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," to perform power distribution surveillances.
A copy of our related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely, B~~ k~se~r~nager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482
Enclosures:
- 1. Amendment No. 188 to NPF-42
- 2. Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 188 License No. NPF-42
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Wolf Creek Generating Station (the facility)
Renewed Facility Operating License No. NPF-42 filed by the Wolf Creek Nuclear Operating Corporation (the Corporation), dated October 10, 2009, as supplemented by letter dated March 8, 2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-42 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 188, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated in the license. The Corporation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
The license amendment is effective as of its date of issuance and shall be implemented by December 29, 2010.
FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: Jul y 23, 2010
ATTACHMENT TO LICENSE AMENDMENT NO. 188 RENEWED FACILITY OPERATING LICENSE NO. NPF-42 DOCKET NO. 50-482 Replace the following pages of the Renewed Facility Operating License No. NPF-42 and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
The corresponding overleaf pages are provided to maintain document completeness.
Renewed Facility Operating License REMOVE INSERT 4
4 Technical Specifications REMOVE INSERT 3.1-16 3.1-16 3.1-17 3.1-17 3.2-1 3.2-1 3.2-3 3.2-3 3.2-4 3.2-4 3.2-8 3.2-8 3.2-13 3.2-13 3.3-10 3.3-10 3.3-11 3.3-11
4 (5)
The Operating Corporation, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)
The Operating Corporation, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission, now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level The Operating Corporation is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100% power) in accordance with the conditions specified herein.
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 188, and the Environmental Protection Plan contained in Appendix S, both of which are attached hereto, are hereby incorporated in the license. The Corporation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Antitrust Conditions Kansas Gas & Electric Company and Kansas City Power & Light Company shall comply with the antitrust conditions delineated in Appendix C to this license.
(4)
Environmental Qualification (Section 3.11 SSER #4, Section 3.11 J
J SSER #5)*
Deleted per Amendment No. 141.
- The parenthetical notation following the title of many license conditions denotes the section of the supporting Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
Renewed License No. NPF-42 Amendment 1\\10. 188
3.1.7 Rod Position Indication 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.
CONDITION REQUIRED ACTION COMPLETION TIME A.
One DRPI per group inoperable for one or more groups.
A.1 OR A.2 Verify the position of the rods with inoperable position indicators indirectly by using core power distribution measurement information.
Reduce THERMAL POWER to ~ 50% RTP.
Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 8 hours (continued)
Wolf Creek - Unit 1 3.1-16 Amendment No. ~, 188
3.1.7 ACTIONS (continued)
CONDITION B. More than one DRPI per group inoperable for one or more groups.
C. One or more rods with inoperable DRPls have been moved in excess of 24 steps in one direction since the last determination of the rod's position.
Rod Position Indication REQUIRED ACTION B.1 Place the control rods under manual control.
AND B.2 Monitor and record RCS Tavg.
AND B.3 Verify the position of the rods with inoperable position indicators indirectly by using core power distribution measurement information.
AND B.4 Restore inoperable position indicators to OPERABLE status such that a maximum of one DRPI per group is inoperable.
C.1 Verify the position of the rods with inoperable position indicators indirectly by using core power distribution measurement information.
OR C.2 Reduce THERMAL POWER to s 50% RTP.
COMPLETION TIME Immediately Once per 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 24 hours 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 8 hours (continued) 3.1-17 Amendment No. ~, 188 Wolf Creek - Unit 1
3.2.1 Fa(Z) (Fa Methodology) 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Heat Flux Hot Channel Factor (Fa(Z)) (Fa Methodology)
LCO 3.2.1 Fa(Z), as approximated by Fac(Z) and Faw(Z), shall be within the limits specified in the COLR.
APPLICABI L1TY:
MODE 1.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Fac(Z) not within limit.
A.1 AND A.2 Reduce THERMAL POWER ~ 1% RTP for each 1% Fac(Z) exceeds limit.
Reduce Power Range Neutron Flux - High trip setpoints ~ 1% for each 1% Fac(Z) exceeds limit.
15 minutes after each Fac(Z) determination 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each Fac(Z) determination A.3 AND A.4 Reduce Overpower ~T trip setpoints ~ 1% for each 1% FQc(Z) exceeds limit.
Perform SR 3.2.1.1.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each Fac(Z) determination Prior to increasing THERMAL POWER above the limit of Required Action A.1 (continued)
Wolf Creek - Unit 1 3.2-1 Amendment No. 123, 159, 188
3.2.1 Fa(Z) (Fa Methodology)
SURVEILLANCE REQUIREMENTS
N()TE------------------------------------------------------------
During power escalation following shutdown, THERMAL P()WER may be increased until an equilibrium power level has been achieved, at which a power distribution measurement is obtained.
SURVEILLANCE SR 3.2.1.1 Verify Fac(Z) is within limit.
FREQUENCY Once after each refueling prior to THERMAL P()WER exceeding 75% RTP Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after achieving equilibrium conditions after exceeding, by 2: 10% RTP, the THERMAL P()WER at which Fac(Z) was last verified AND 31 EFPD thereafter (continued)
Wolf Creek - Unit 1 3.2-3 Amendment No. ~, 188
3.2.1 Fa(Z) (Fa Methodology)
SURVEILLANCE SR 3.2.1.2
N()lrE--------------------------------
If Fac(Z) measurements indicate maximum over z [
FaC (Z) ]
K(Z) has increased since the previous evaluation of Fac(Z):
- a.
Increase Faw(Z) by the appropriate factor specified in the C()LR and reverify Faw(Z) is within limits; or
- b.
Repeat SR 3.2.1.2 once per 7 EFPD until two successive power distribution measurements indicate FaC (Z) ]
maximum over z [
K(Z) has not increased.
Verify Faw(Z) is within limit.
FREQUENCY Once after each refueling prior to lrHERMAL P()WER exceeding 75% RTP Wolf Creek - Unit 1 3.2-4 Amendment No. ~, 188
3.2.2 FXH SURVEILLANCE REQUIREMENTS
N()TE--------------------------------------------------------------
During power escalation following shutdown, THERMAL P()WER may be increased until an equilibrium power level has been achieved, at which a power distribution measurement is obtained.
SURVEILLANCE SR 3.2.2.1 Verify FXH is within limits specified in the C()LR.
FREQUENCY Once after each refueling prior to THERMAL P()WER exceeding 75% RTP AND 31 EFPD thereafter Wolf Creek - Unit 1 3.2-8 Amendment No. 123, 131, 188
3.2.4 QPTR SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1
N()TES--------------------------------
- 1.
With input from one Power Range Neutron Flux channel inoperable and THERMAL P()WER
- 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
N()TE--------------------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after input from one Power Range Neutron Flux channel is inoperable with THERMAL P()WER > 75% RTP.
Verify QPTR is within limit using core power distribution measurement information.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Wolf Creek - Unit 1 3.2-13 Amendment No..:lU, 188
3.3.1 RTS Instrumentation SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.1.2
N()TES-------------------------------
Not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL P()WER is :?: 15% RTP.
Compare results of calorimetric heat balance 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.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SR 3.3.1.3
N()"rES-------------------------------
Not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL P()WER is :?: 50% RTP.
Compare results of the core power distribution measurements to Nuclear Instrumentation System (NIS) AFD. Adjust NIS channel if absolute difference is z 3%.
31 effective full power days (EFPD)
N()TE--------------------------------
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 TAD()T.
62 days on a STAGGERED TEST BASIS (continued)
Wolf Creek - Unit 1 3.3-10 Amendment No. 123, 148, 156, 188
3.3.1 RTS Instrumentation SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.1.5 Perform ACTUATION LOGIC TEST.
92 days on a STAGGERED TEST BASIS SR 3.3.1.6
NOTE--------------------------------
Not required to be performed until 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after achieving equilibrium conditions with THERMAL POWER ~ 75 % RTP.
Calibrate excore channels to agree with core power distribution measurements.
92 EFPD SR 3.3.1.7
NOTES------------------------------
- 1.
Not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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)
Wolf Creek - Unit 1 3.3-11 Amendment No. 123, 156, 188
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 188 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-42 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482
1.0 INTRODUCTION
By application dated October 10, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML092930130), as supplemented by letter dated March 8, 2010 (ADAMS Accession No. ML100750459), Wolf Creek Nuclear Operating Corporation (WCNOC, the licensee) requested changes to the Technical Specifications (TSs) for the Wolf Creek Generating Station (WCGS). The proposed amendment would revise TS 3.1.7, "Rod Position Indication," TS 3.2.1, "Heat Flux Hot Channel Factor (Fa(Z)) (Fa Methodology),"
TS 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FNLiH)," TS 3.2.4, "Quadrant Power Tilt Ratio (QPTR)," and TS 3.3.1, "Reactor Trip System (RTS) Instrumentation," to permit use of the Westinghouse computer code, the Best Estimate Analyzer for Core Operations - Nuclear (BEACON) power distribution monitoring system (PDMS), as described in WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System" (ADAMS Accession No. ML092050097)1 to perform power distribution surveillances.
The new BEACON PDMS would augment the functional capability of the neutron flux mapping system for the purposes of power distribution surveillances at WCGS. Certain required actions, for when a limiting condition for operation (LCO) is not met, and certain surveillance requirements (SRs) would be changed to refer to power distribution measurements or measurement information of the core.
The BEACON system was developed by Westinghouse to improve operational support for pressurized-water reactors (PWRs). It is a core monitoring and support package that uses Westinghouse standard instrumentation in conjunction with an analytical methodology for online generation of three-dimensional power distributions. The system provides core monitoring, core measurement reduction, core analysis, and core predictions. Since BEACON does not have any direct inputs to the RTS, BEACON will not affect any of the accident analyses in the WCGS licensing basis. Furthermore, as stated in its application dated October 10, 2009, the licensee will not use BEACON to relax the key safety parameter limits or levels of margin at WCGS.
1 This is a proprietary non-public document.
- 2 In Attachments V and VI to its application dated October 10,2009, the licensee identified changes to the TS Bases and to the Technical Requirements Manual (TRM), respectively.
These changes provide additional detail as discussed in the licensee's evaluation of the proposed amendment in Attachments I and II to its license amendment request. The U.S. Nuclear Regulatory Commission (NRC) staff reviewed this information as part of its review of the license amendment request. The changes to the TS Bases would be made by the licensee in accordance with TS 5.5.14, "Technical Specifications (TS) Bases Control Program,"
at the time the amendment is implemented. These changes were reviewed to determine whether the NRC staff had any disagreements with the identified changes based on the proposed amendment.
The supplemental letter dated March 8, 2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register on January 26, 2010 (75 FR 4120).
2.0 REGULATORY EVALUATION
Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The TSs ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the public. The NRC's regulatory requirements related to the content of the TSs are contained in Section 50.36, "Technical specifications," of Title 10 of the U.S. Code of Federal Regulations (10 CFR 50.36), which requires that the TSs include items in the following categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) SRs; (4) design features; and (5) administrative controls. However, the rule does not specify the particular requirements to be included in a plant's TSs. As stated in 10 CFR 50.36(c)(2)(i), LCOs are "the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications... " The regulations in 10 CFR 50.36(c)(3) state that SRs are "requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met."
As required by 10 CFR 50.36(c)(2)(ii), a TS LCO of a nuclear reactor must be established for each item meeting one or more of the following criteria:
Criterion 1:
Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
Criterion 2:
A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3 Criterion 3:
A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
Criterion 4:
A structure, system, or component, which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.
Those items that do not fall within or satisfy any of the above criteria do not need to be included in the LCOs of the TSs. The PDMS instrumentation does not meet any of the criteria of 10 CFR 50.36(c)(2)(ii) for inclusion in the TSs. Therefore, the licensee will include the PDMS requirements and controls in the TRM. The Technical Requirements are plant-specific administrative controls on equipment, similar to TS controls, but are maintained by the licensee in accordance with 10 CFR 50.59, "Changes, tests, and experiments."
There are no specific regulatory requirements on PDMSs, such as the BEACON system; however, the use of such systems by licensees in monitoring the power distribution in the reactor core during power operation must be consistent with the safe operation of the plant.
3.0 TECHNICAL EVALUATION
3.1 Proposed TS Changes
In its application dated October 10, 2009, the licensee proposed the following changes to TSs 3.1.7, 3.2.1,3.2.2, 3.2.4, and 3.3.1:
For TS 3.1.7, replace the phrase "moveable incore detectors" by the phrase "core power distribution measurement information" in Required Actions A.1, B.3, and C.1.
For TS 3.2.1, (1) replace the phrase "power distribution map" by the phrase "power distribution measurement" in the NOTE to the SRs; (2) replace the phrase "flux maps" by the phrase "power distribution measurements" in SR 3.2.1.2; and (3) LCO 3.2.1 from "FaC(Z), as approximated by FaC(Z) and FaW(Z), shall be within the limits specified in the COLR [core operating limit report]" to "Fa(Z), as approximated by FaC(Z) and FaW(Z), shall be within the limits specified in the COLR" to correct an editorial error.
For TS 3.2.2, replace the phrase "power distribution map" by the phrase "power distribution measurement" in the NOTE to the SRs.
For TS 3.2.4, replace the phrase "the movable incore detectors" by the phrase "core power distribution measurement information" in SR 3.2.4.2.
For TS 3.3.1, replace the phrase "incore detector" by the phrase "core power distribution" in SRs 3.3.1.3 and 3.3.1.6.
- 4 These changes would allow the surveillance to be performed using either the moveable incore detectors or an OPERABLE PDMS.
3.2
NRC Staff Evaluation
The licensee is proposing the changes to the TSs to allow the use of a PDMS at WCGS. The PDMS would be an enhancement to its core power distribution measurement and indication capability. The core power distribution information that is to be referred to in the proposed changes to the TSs 3.1.7, 3.2.1, 3.2.2, 3.2.4, and 3.3.1 would be information from either the existing incore detector system or the PDMS.
The PDMS to be used is the BEACON system, which was developed by Westinghouse to improve the monitoring support by Westinghouse-designed PWRs, such as WCGS. The BEACON PDMS is a core monitoring and support package, which uses Westinghouse standard instrumentation in conjunction with an analytical methodology for online generation of three dimensional power distributions to provide core monitoring, core measurement reduction, core analysis, and core predictions. The BEACON PDMS is calibrated by the existing incore detector system.
The BEACON system is described in the topical report WCAP-12472-P, "BEACON: Core Monitoring and Operations Support System," which was approved as WCAP-12472-P-A by the NRC staff for Westinghouse reactors in its letter dated February 16, 1994, which transmitted the NRC staff's safety evaluation report (SER) endorsing November 18, 1993 Brookhaven National Laboratory Technical Evaluation Report (TER), "Technical Evaluation of the BEACON Core Monitoring and Operations Support System Topical Report WCAP-12472-P," The topical report (ADAMS Accession No. ML092050097)2, which also contains the copy of the NRC staff's letter dated February 16, 1994, is subject to conditions and limitations delineated in Section 4.0 of the TER. These conditions are the following:
- 1.
In the cycle-specific applications of BEACON, the power peaking uncertainties U~h and Ua must provide 95% probability upper tolerance limits at the 95% confidence level (Section 3.3 [of the TER]),
- 2.
In order to ensure that the assumptions made in the BEACON uncertainty analysis remain valid, the generic uncertainty components may require reevaluation when BEACON is applied to the plant or core designs that differ sufficiently to have a significant impact on the WCAP-12472-P data base (Section 3.4 [of the TER]), and
- 3.
The BEACON Technical Specifications should be revised to include the changes described in Section 3 [of the TER] concerning Specifications 3.1.3.1 and 3.1.3.2, and the Core Operating Limits Report (Section 3.6 [of the TER]).
In addressing condition 1, in Attachment I to its application dated October 10, 2009, the licensee stated that:
2This is a proprietary non-public document.
- 5 Although not specifically described in [WCNOC's] 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 methodology. Until these calibrations are complete, more conservative, default uncertainties will be applied. The calibrations will be documented and retained as records.
In addressing condition 2 in Attachment I to its application dated October 10, 2009, the licensee stated that:
WCGS utilizes a Westinghouse 4-loop nuclear steam supply system (NSSS) with Westinghouse movable incore instrumentation. All fuel is presently of Westinghouse manufacture. Therefore, WCGS does not differ significantly from the plants that form the WCAP database, and no additional review of WCAP applicability to WCGS 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.
In addressing condition 3 in Attachment I to its application dated October 10, 2009, the licensee stated that:
The WCAP describes an application of BEACON where the core operating limits are changed. As noted previously, WCNOC is proposing only to use BEACON as a core TS monitor for conformance to WCGS'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 WCNOC for the intended use of BEACON. Therefore, this condition does not apply to the amendment requested for WCGS.
The NRC staff reviewed the licensee's responses to the three conditions and concludes that the responses are acceptable. The licensee has not proposed changes to the COLR or the core safety limits for WCGS, and the proposed TS changes are to allow the core power distribution to be determined at WCGS by either the existing movable incore detector system or the BEACON system; the proposed core power distribution measurement language will cover both surveillance approaches. Also, as stated by the licensee, and identified in the TRM, the BEACON PDMS is not required to be operable below 25 percent rated thermal power (RTP),
but is required to be operable at all other times in MODE 1 when used for the surveillances
- 6 described in the requested TS changes. Furthermore, the TRM Bases state that, below 25 percent RTP, the PDMS is inoperable, and may not be used. This is because the accuracy of the calculated core power distribution may not be bounded by the uncertainties in WCAP-12472-P-A at these reduced power levels. Based on this technical requirement, and on the discussion provided above, the NRC staff concludes that the licensee has provided an acceptable disposition of the WCAP-12472-P-A TER conditions, and it is acceptable for the licensee to use the BEACON system described in WCAP-12472-P-A at WCGS.
In its application dated October 10, 2009, the licensee stated that it intended to use BEACON to augment the functional capability of its core flux mapping system for the purpose of power distribution surveillances. Although WCAP-12472-P-A discusses an application of BEACON in which there is continuous flux monitoring by control room operators, the licensee is proposing a more conservative application of BEACON in which the core power distribution limits themselves remain unchanged. The licensee 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 the reactor power is greater than 25 percent RTP. At less than or equal to 25 percent RTP, or when the PDMS is inoperable, the existing movable incore detector system would be used.
In Attachment II to its application dated October 10, 2009, the licensee addressed whether the PDMS needed to have an LCO added to the TSs to state that the system is required to be operable. The licensee concluded that an LCO for the PDMS was not required, because it did not meet the criteria set forth in 10 CFR 50.36(c)(2)(ii).
The PDMS instrumentation provides the capability to monitor core parameters at more frequent intervals than is currently required by the TSs. The PDMS combines inputs from currently installed plant instrumentation and design data for each fuel cycle, and does not modify or eliminate existing plant instrumentation. It provides a means to continuously monitor the power distribution limits including limiting peaking factors and quadrant power tilt ratio. The PDMS is used for periodic measurement of the core power distribution to confirm operation within design limits, and for periodic calibration of the ex-core detectors, and it does not initiate any automatic protection action. The PDMS instrumentation does not change any of the key safety parameter limits or levels of margin as considered in the reference design basis evaluations. These limits are not revised by this license amendment, and can be determined independently of the operability of the PDMS. Based on these considerations, the NRC staff also concludes that the PDMS itself does not meet any of the 10 CFR 50.36(c)(2)(ii) selection criteria for inclusion in the TSs. Therefore, the NRC staff concludes that 10 CFR 50.36 does not require the PDMS to have an LCO in the TSs.
Based on the schedules for software and procedural implementation established by the licensee to implement the proposed TS changes, the licensee has requested an implementation date of no later than December 29, 2010, in its letter dated March 8, 2010, which is acceptable.
Based on its review of the proposed changes to TSs 3.1.7, 3.2.1,3.2.2, 3.2.4, and 3.3.1, identified in Section 3.1 of this SER, the NRC staff concludes that replacing the current TS references to incore detectors and neutron flux maps with the proposed references to core power distribution measurements or measurement information (from either the movable incore detector system or the BEACON system) is consistent with the technical requirements of the
- 7 NRC-approved WCAP-12472-P-A, and, therefore, the proposed changes are acceptable. The proposed change to LCO 3.2.1 is to correct an inadvertent change, is editorial in nature, and is consistent with the latest approved TSs for WCGS. Based on this conclusion, the NRC staff further concludes that the proposed changes are adhere to 10 CFR 50.36, and, therefore, the proposed amendment is acceptable.
4.0 REGULATORY COMMITMENT The licensee has made the following regulatory commitment in its letter dated October 10, 2009, to meet the acceptance criteria condition for power peaking uncertainties described in WCAP 12472-P:
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 methodology. Until these calibrations are complete, more conservative default uncertainties will be applied. The calibrations will be documented and retained as records.
The licensee stated that the regulatory commitment will be effective upon implementation of the amendment. The NRC staff reviewed the licensee's regulatory commitment and found it acceptable.
5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Kansas State official was notified of the proposed issuance of the amendment. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding published in the Federal Register on January 26,2010 (75 FR 4120). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
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7.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (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.
Principal Contributor: B. Parks, SRXBIDSS/NRR Date: July 23, 2010
- SE memo dated OFFICE NRR/LPL4/PM NRR/LPL4/LA DIRS/ITSB/BC DSS/SRXB/BC OGC NRR/LPL4/BC NRR/LPLR/PM NAME BSingal JBurkhardt REliiolt GCranston*
SUltal MMarkley (MThadani)
BSingal DATE 6/28/10 6/28/10 7/1/10 212110 7/12/10 7/23/10 7/23/10