NL-17-0236, License Amendment Request to Revise Technical Specification Surveillance Requirement 3.3.1.3
| ML17144A408 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/24/2017 |
| From: | Hutto J Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-17-0236 | |
| Download: ML17144A408 (18) | |
Text
.t. Southern Nuclear J. J. Hutto Regulatory Affairs Director 40 Inverness Center Parkway Post Office Box 1295 Birmingham, AL 35242 205 992 5872 tel 205 992 7601 fax jjhutto@southernco.com May 24,2017 Docket Nos.: 50-424 NL-17-0236 50-425 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 I Vogtle Electric Generating Plant- Units 1 and 2 License Amendment Request to Revise Technical Specification Surveillance Requirement 3.3.1.3 Ladies and Gentlemen:
Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC) hereby requests an amendment to Facility Operating License Nos. NPF-68 and to NPF-81 for the Vogtle Electric Generating Plant (VEGP) Units 1 and 2. This amendment request proposes to revise Technical Specification (TS) Surveillance Requirement (SR) 3.3.1.3.
The NOTE in SR 3.3.1.3 is proposed to be revised to: "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 POWER is 2:. 50% RTP." Appropriate TS Bases changes will also be made consistent with the SR change discussed. provides the basis for the proposed change to the VEGP TS, and the Significant Hazards Consideration and Environmental Consideration determination. provides the VEGP TS marked-up page showing the proposed changes. provides the VEGP TS clean-typed page showing the proposed changes. provides the VEGP TS Bases marked-up page showing the proposed changes for information only. The Bases will be revised under the Technical Specification Bases Control Program following NRC approval of the proposed Technical Specification changes.
As described in Enclosure 1, the proposed changes have been analyzed in accordance with 10 CFR 50.91 (a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that the changes involve no significant hazards consideration.
SNC requests approval of the proposed license amendment by May 30, 2018. The proposed changes will be implemented within 90 days of issuance of the amendments for VEGP.
In accordance with 10 CFR 50.91 (b)(1 ), "State Consultation," a copy of this application and its reasoned analysis about no significant hazards considerations is being provided to the designated Georgia officials.
U. S. Nuclear Regulatory Commission NL-17-0236 Page 2 This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369.
Mr. J. J. Hutto states he is Regulatory Affairs Director of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and to the best of his knowledge and belief, the facts set forth in this letter are true.
Respectfully submitted, 99*k::>
J. J. Hutto Regulatory Affairs Director JJH/GKMIGLS Sworn to and subscribed before me this . of _ __,M'---'---'-a'-lf'-----' 2017.
Z../*p-r.. day c::Jf) _y£ Notary Public
~ L--
U My commission expires: 1 2-0 I 8
Enclosures:
- 1. VEGP Basis for Proposed Change
- 2. VEGP Technical Specifications Marked-Up Page
- 3. VEGP Technical Specifications Clean-Typed Page
- 4. VEGP Technical Specifications Bases Marked-Up Page (Information Only) cc: Regional Administrator, Region II NRR Project Manager- Vogtle 1 & 2 Senior Resident Inspector- Vogtle 1 & 2 Director, Environmental Protection Division - State of Georgia RType: CVC7000
Vogtle Electric Generating Plant- Units 1 and 2 License Amendment Request to Revise Technical Specification Surveillance Requirement 3.3.1.3 Enclosure 1 VEGP Basis for Proposed Change to NL-17-0236 Basis for Proposed Change Table of Contents 1.0 Summary Description 2.0 Detailed Description 3.0 Technical Evaluation 4.0 Regulatory Evaluation 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration Determination 4.4 Conclusions 5.0 Environmental Consideration E1-1
Enclosure 1 to NL-17-0236 Basis for Proposed Change 1.0 Summary Description This evaluation supports a request to amend Operating Licenses NPF-68 and to NPF-81 for the Vogtle Electric Generating Plant Units 1 and 2 (VEGP). This amendment request proposes to revise VEGP Technical Specification (TS) Surveillance Requirement (SR) 3.3.1.3. This SR verifies the accuracy of the Axial Flux Difference (AFD) input to the reactor trip system as well as the indication of AFD for satisfying VEGP TS 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)
(Relaxed Axial Offset Control (RAOC) Methodology)". The NOTE in SR 3.3.1.3 states: "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 POWER is 2:. 15% RTP." During power asqension following a refueling, the SR is typically performed at about 30% RATED THERMAL rowER (RTP) where plant conditions allow for meaningful measurements. Once 15% RTP is reached, the remainder of the allowed time for performing the SR becomes challenging. It is proposed to change the power level in the NOTE from 15% RTP to 50% RTP. This will provide sufficient time during power ascension after refueling to perform SR 3.3.1.3. The power level of 50% RTP is consistent with the power level at which AFD control is required by the safety analyses and TS 3.2.3.
Appropriate Bases changes would also be made consistent with the TS change discussed above.
2.0 Detailed Description The NOTE in SR 3.3.1.3 is proposed to be revised to:
"Not required to be performed until24 hours after THERMAL POWER is 2:.50% RTP."
Specification 1.1, "Definitions", defines Axial Flux Difference (AFD) as: "AFD shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector."
There are four excore detectors (external to the core) used to measure AFD. The excore detectors monitor the flux and power in the corresponding four core quadrants. The normalized flux signals are used to provide control room indication of AFD and Quadrant Power Tilt Ratio (QPTR) to satisfy TS 3.2.3 and 3.2.4, respectively. In addition, AFD is an input to the Overtemperature 11T (OT11T) reactor trip function (Function 6 in Table 3.3.1-1 of TS 3.3.1 ). As described in Note 1 of Table 3.3.1-1, the term f1(AFD) modifies (reduces) the OT/1T setpoint when the AFD exceeds the values specified in Note 1 of Table 3.3.1-1.
Surveillance Requirement (SR) 3.3.1.3 is performed to ensure that the AFD input to the OT11T reactor trip function accurately reflects the power distribution in the core. The SR requires a comparison of the incore AFD to the indicated excore AFD and adjustment of the instrument channels if the acceptance criterion is not satisfied. Following a refueling outage, the first performance of this SR is not required until24 hours after THERMAL POWER is 2:.15% RTP.
The SR is typically performed at about 30% RTP where plant conditions allow for meaningful measurements. If adjustment of the instrument channels is required, the remainder of the 24-hour time limit is restrictive and poses challenges to completing the SR. The only way to avoid this is to remain below 15% RTP to perform this SR. However, plant conditions at such a low power level are not well suited for obtaining meaningful measurements.
E1-2 to NL-17-0236 Basis for Proposed Change The threshold of 15% RTP is a historical value based on the original (pre-ITS) VEGP Technical Specifications when the AFD specification (TS 3/4.2.1) applicability began at 15% RTP. When VEGP revised the AFD specification from Constant Axial Offset Control (CAOC) to Relaxed Axial Offset Control (RAOC), the applicability was revised to Mode 1 above 50% RTP. However, the threshold power level of 15% RTP for performing the surveillance to compare the incore AFD to the indicated excore AFD (pre-ITS specification Table 4.3-1, Functional Unit 2a, Notation 3) was not revised accordingly.
During the conversion to ITS, NUREG-1431 retained the 15% RTP threshold and added the time requirement of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in SA 3.3.1.3. In SA 3.3.1.3, the ITS does not differentiate between plants using RAOC AFD control versus plants using CAOC AFD control. During the conversion to ITS, the AFD specification (pre-ITS TS 3/4.2.1) was moved to TS 3.2.3.
In the case of RAOC AFD control (TS 3.2.3), AFD limits are not applicable below 50% RTP. A more appropriate requirement is to ensure that the AFD indication (TS 3.2.3) and AFD input to the OTf1 T reactor trip function satisfy the accuracy requirement of SA 3.3.1.3 when the AFD limits of TS 3.2.3 are applicable. Revising the requirement in the NOTE in SA 3.3.1.3 to "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 POWER is~ 50% RTP" is consistent with the power level at which AFD control is required by the safety analyses and TS 3.2.3.
3.0 Technical Evaluation 3.1 Proposed Change The current NOTE in SR 3.3.1.3 states:
"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 POWER is ~ 15%
RTP."
The NOTE in SA 3.3.1.3 is proposed to be revised to:
"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 POWER is~ 50%
RTP."
3.2 Reason for Proposed Change For the CHANNEL CALIBRATION of the OTflT reactor trip function channels during a refueling outage, the AFD input to the OTf1 T reactor trip function is initially calibrated with projected full-power detector currents until plant conditions allow for more accurate measurement of detector currents and incore AFD.
Performance of SR 3.3.1.3 requires taking a flux map to determine the incore AFD to compare to the indicated AFD based on the nuclear instrumentation (excore AFD). It is not uncommon that indicated excore AFD in one or more channels and the incore AFD differ by more than the 3% criterion in SA 3.3.1.3. In such cases, the channels are adjusted as required by SA 3.3.1.3.
E1-3 to NL-17-0236 Basis for Proposed Change To obtain meaningful incore measurements, the flux map needs to be taken at a sufficiently high power level. Typically, this is at about 30% RTP following a refueling outage. The requirement in the NOTE in SR 3.3.1.3 is challenging during power ascension following a refueling outage due to the time required to raise power to about 30% RTP, place the generator on line, stabilize reactor power, perform a flux map, perform SR 3.3.1.3 to determine if the AFD channels meet the acceptance criterion, and prepare the information to perform the adjustments, all within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 15% RTP.
3.3 Original VEGP Pre-ITS Technical Specifications In the original VEGP Technical Specifications, AFD and Heat Flux Hot Channel Factor limits were contained in TS 3/4.2.1 and 3/4.2.2, respectively. Also in the original VEGP Technical Specifications, the Surveillance Requirement for performing the comparison of the incore AFD and indicated excore AFD was in Table 4.3-1, "Reactor Trip System Instrumentation Surveillance Requirements." The SR was a requirement of Functional Unit 2a, "Power Range Neutron Flux High Setpoint" with a frequency designated as "M (3, 4)", i.e., Monthly frequency with Table Notations 3 and 4.
TS 3/4.2.1 for AFD and the SR in Table 4.3-1 for the comparison of the incore AFD and indicated excore AFD were related by the power level of 15% RTP at which both became applicable. However, for the SR in Table 4.3-1 Notation 3, there was no time limit for performing the surveillance.
TS 3/4.2.1 for AFD limits and TS 3/4.2.2 for Heat Flux Hot Channel Factor limits were later amended. TS 3/4.2.1 for AFD was amended to adopt the Westinghouse Relaxed Axial Offset Control (RAOC) methodology. TS 3/4.2.2 for Heat Flux Hot Channel Factor limits was amended to adopt the Westinghouse FQ methodology. These changes were included in License Amendments 43 and 44 for VEGP Unit 1 and Amendments 23 and 24 for VEGP Unit 2 issued on September 19, 1991 (ML012290256).
In the amendment of TS 3/4.2.1 for AFD limits, the applicability was changed from MODE 1 above 15% RTP to MODE 1 above 50% RTP. However, the threshold of 15%
RTP for the SR in Table 4.3-1 for the comparison of the incore AFD and indicated excore AFD was not revised accordingly.
3.41TS Conversion- NUREG-1431 The original format of the Technical Specifications was changed as part of the Improved Technical Specifications program. NUREG-1431, "Standard Technical Specifications-Westinghouse Plants", became the new standard and is herein referred to as Standard Technical Specifications (STS).
During the development of STS, the SR in Table 4.3-1 (in the original Technical Specifications) for the comparison of the incore AFD and indicated excore AFD was moved to STS SR 3.3.1.3 which is associated with the OT~ T reactor trip function (Function 6 in STS Table 3.3.1-1, "Reactor Trip System Instrumentation"). In STS SR E1-4 to NL-17-0236 Basis for Proposed Change 3.3.1.3, the default power threshold of 15% RTP was retained and a default time limit of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> was added for performing the SR.
Also during the development of STS, TS 3/4.2.1 for AFD limits (in the original Technical Specifications) was moved to STS TS 3.2.3. The power level in the applicability was changed to "~50% RTP" in STS TS 3.2.38 (RAOC Methodology).
3.5 VEGP ITS Conversion During the conversion of the VEGP Technical Specifications to STS, VEGP adopted the default power level threshold of 15% RTP and the default time limit of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in STS SR 3.3.1 .3. In addition, VEGP adopted the power level of~ 50% RTP in the applicability of STS TS 3.2.38 (RAOC Methodology) because VEGP had previously converted to RAOC Methodology.
These changes were included in License Amendments 96 and 74 for VEGP Unit 1 and 2, respectively, issued on September 25, 1996 (ML012390239).
3.6 Basis for Proposed Change As discussed in Section 3.2 above, the requirement in the NOTE in VEGP SR 3.3.1.3 is challenging during power ascension following a refueling outage.
A review of Technical Specifications for the Westinghouse fleet was performed to determine if there was precedent for changing the 24-hour time limit. The review demonstrated that there is little consistency within the fleet. Time limits for SR 3.3.1.3 varied from no limit up to 7 days after exceeding a power level threshold. Also, there was little consistency in the power level threshold. Power level thresholds varied from 15%
VEGP TS 3.2.3 for AFD control is not applicable below 50% RTP. There are no limits on AFD below 50% RTP. At power levels~ 50% RTP, AFD must be maintained within specified limits. VEGP SR 3.3.1.3 ensures the accuracy of the indication of AFD used to comply with TS 3.2.3. In addition, the SR ensures the accuracy of the AFD input to the OT~T reactor trip function (Function 6 in Table 3.3.1-1, "Reactor Trip System Instrumentation"). Even though SR 3.3.1.3 is associated with the OT~T reactor trip function, it ensures the accuracy of the indication of AFD for both these Specifications.
As discussed previously in Section 3.3 above, there was a basis for the 15% RTP threshold in the SR for the comparison of the incore AFD and indicated excore AFD. For AFD control using the RAOC methodology as in VEGP TS 3.2.3, there is no basis for retaining the 15% RTP threshold in SR 3.3.1.3 for verifying the accuracy of the indication of AFD. VEGP TS 3.2.3 for AFD control is not applicable below 50% RTP. There are no limits on AFD below 50% RTP.
The OT~T reactor trip function is required to be operable in Modes 1 and 2. At the beginning of each fuel cycle, the AFD input to the OT~T reactor trip function , as well as E1-5 to NL-17-0236 Basis for Proposed Change the indication of AFD for satisfying VEGP TS 3.2.3, is initially calibrated with projected full-power detector currents until plant conditions allow for a more accurate measurement of detector currents and incore AFD. The initial calibration of the AFD input to the OTl\T trip function is sufficient to meet the requirements of the safety analyses until SR 3.3.1.3 is performed.
The proposed change revises the requirement in the NOTE in SR 3.3.1.3 to: "Not required to be performed until24 hours after THERMAL POWER is~ 50% RTP." The change from 15% RTP to 50% RTP will provide sufficient time to achieve suitable plant conditions for the first performance of SR 3.3.1.3 during power ascension after refueling.
This is consistent with the power level at which AFD control is required by the safety analyses and TS 3.2.3.
4.0 Regulatory Evaluation 4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36(c), "Technical specifications," requires Technical Specifications to be included for the following categories:
(1) Safety limits, limiting safety system settings, and limiting control settings, (2) Limiting conditions for operation, (3) Surveillance requirements, (4) Design features, and (5) Administrative controls.
10 CFR 50.36(c) (3) "Surveillance requirements," states:
"Surveillance requirements 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."
None of the TS categories are impacted by the proposed TS changes. The requirement to perform SR 3.3.1.3 is not being deleted. The NOTE in SR 3.3.1.3 is being modified to be consistent with the power level at which AFD control is required by the safety analyses and TS 3.2.3.
Therefore, the requirements of 10 CFR 50.36(c) continue to be met.
Appendix A to 10 CFR 50, General Design Criterion (GDC) 10, "Reactor Design" states:
The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences."
E1-6 to NL-17-0236 Basis for Proposed Change Appendix A to 10 CFR 50, General Design Criterion (GDC} 20, "Protection System Functions" states:
The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety."
Neither the reactor design nor the protection system functions are being modified by the proposed SR change therefore, GDC 10 and GDC 20 continue to be met.
Therefore, the proposed amendment does not impact the Regulatory Requirements discussed above.
4.2 Precedent A review of Technical Specifications for the Westinghouse fleet was performed to determine if there was precedent for changing the 24-hour time limit. The review demonstrated that there is little consistency within the fleet. Time limits for SR 3.3.1.3 varied from no limit up to 7 days after exceeding a power level threshold. Also, there was little consistency in the power level threshold. Power level thresholds varied from 15%
4.3 Significant Hazards Consideration This evaluation supports a request to amend Operating Licenses NPF-68 and to NPF-81 for the Vogtle Electric Generating Plant Units 1 and 2 (VEGP). This amendment request proposes to revise VEGP Technical Specification (TS) Surveillance Requirement (SR) 3.3.1.3. This SR verifies the accuracy of the Axial Flux Difference (AFD) input to the reactor trip system as well as the indication of AFD for satisfying VEGP TS 3.2.3, "AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC) Methodology)". The NOTE in SR 3.3.1.3 states: "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 POWER is?. 15% RTP." During power ascension following a refueling, the SR is typically performed at about 30% RATED THERMAL POWER (RTP) where plant conditions allow for meaningful measurements. Once 15% RTP is reached, the remainder of the allowed time for performing the SR becomes challenging. It is proposed to change the power level in the NOTE from 15% RTP to 50% RTP. This will provide sufficient time during power ascension after refueling to perform SR 3.3.1.3. The power level of 50% RTP is consistent with the power level at which AFD control is required by the safety analyses and TS 3.2.3.
As required by 10 CFR 50.91 (a)(1 ), Southern Nuclear Company (SNC) has evaluated the proposed amendment to the VEGP TS using the criteria in 10 CFR 50.92(c) and has determined that the proposed amendment does not involve a significant hazards consideration. An analysis of the issue of no significant hazards consideration is presented below:
E1-7 to NL-17-0236 Basis for Proposed Change
- 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed amendment to the TS does not affect the initiators of any analyzed accident. In addition, operation in accordance with the proposed amendment to the TS ensures that the previously evaluated accidents will continue to be mitigated as analyzed. The proposed amendment does not adversely affect the design function or operation of any structures, systems, and components important to safety.
The probability or consequences of accidents previously evaluated in the UFSAR are unaffected by this proposed amendment because there is no change to any equipment response or accident mitigation scenario. There are no new or additional challenges to fission product barrier integrity.
Therefore, it is concluded that the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No The proposed amendment does not involve a physical alteration of the plant (no new or different type of equipment will be installed). The proposed amendment does not create any new failure modes for existing equipment or any new limiting single failures. The proposed amendment does not involve a change in the methods governing normal plant operation and all safety functions will continue to perform as previously assumed in accident analyses. Thus, the proposed amendment does not adversely affect the design function or operation of any structures, systems, and components important to safety.
No new accident scenarios, failure mechanisms, or limiting single failures are introduced due to the proposed amendment. The proposed amendment does not challenge the performance or integrity of any safety-related system.
Therefore, it is concluded that the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.
- 3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No The margin of safety associated with the acceptance criteria of any accident is unchanged. The proposed amendment will have no affect on the availability, operability, or performance of the safety-related systems and components. No change is being made to the requirement to perform the surveillance. The NOTE E1-8 to NL-17-0236 Basis for Proposed Change in the surveillance is being changed to clarify when the initial surveillance after refueling is to be performed. The Technical Specification Limiting Condition for Operation (LCO) limits are not being changed.
The proposed amendment will not adversely affect the operation of plant equipment or the function of equipment assumed in the accident analysis.
Therefore, it is concluded that the proposed amendment does not involve a significant reduction in a margin of safety.
Based upon the above analysis, SNC concludes that the proposed amendment does not involve a significant hazards consideration, under the standards set forth in 10 CFR 50.92(c), "Issuance of Amendment," and accordingly, a finding of "no significant hazards consideration" is justified.
4.4 Conclusions In conclusion, 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.
5.0 Environmental Consideration A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
E1-9
Vogtle Electric Generating Plant- Units 1 and 2 License Amendment Request to Revise Technical Specification Surveillance Requirement 3.3.1.3 Enclosure 2 VEGP Technical Specifications Marked-Up Page to NL-17-0236 VEGP Technical Specification Marked Up Page RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS
NOTE----------------------------------
Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.
SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.1.2 ------------------NOTES-------------------
Not required to be performed until12 hours after THERMAL POWER is~ 15% RTP.
Compare results of calorimetric heat balance In accordance with calculation to power range channel output. Adjust the Surveillance power range channel output if calorimetric heat Frequency Control balance calculation results exceed power range Program channel output by more than +2% RTP.
SR 3.3.1 .3 -------------------NOTES---------------------
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 POWER is ~ +a% RTP .
so Compare results of the in core detector In accordance with measurements to Nuclear Instrumentation System the Surveillance (NIS) AFD. Adjust NIS channel if absolute Frequency Control difference is~ 3%. Program (continued)
Vogtle Units 1 and 2 3.3.1-9 Amendment No. 458 (Unit 1)
Amendment No. -14G (Unit 2)
Vogtle Electric Generating Plant- Units 1 and 2 License Amendment Request to Revise Technical Specification Surveillance Requirement 3.3.1.3 Enclosure 3 VEGP Technical Specifications Clean-Typed Page
RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS
N0 T E-----------------------------------------------------------
Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.
SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.1.2 ----------------------------N 0 TE S-----------------------------
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 THERMAL POWER is;:::: 15% RTP.
Compare results of calorimetric heat balance In accordance with calculation to power range channel output. Adjust the Surveillance power range channel output if calorimetric heat Frequency Control balance calculation results exceed power range Program channel output by more than +2% RTP.
SR 3.3.1.3 ---------------------------N0 T ES------------------------------
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 POWER is ;:::: 50% RTP.
Compare results of the incore detector In accordance with measurements to Nuclear Instrumentation System the Surveillance (NIS) AFD. Adjust NIS channel if absolute Frequency Control difference is ;:::: 3%. Program (continued)
Vogtle Units 1 and 2 3.3.1-9 Amendment No. (Unit 1)
Amendment No. (Unit 2)
Vogtle Electric Generating Plant- Units 1 and 2 License Amendment Request to Revise Technical Specification Surveillance Requirement 3.3.1.3 Enclosure 4 VEGP Technical Specifications Bases Marked-Up Page (Information Only) to NL-17-0236 VEGP Technical Specification Bases Marked Up Page (for information only)
RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.3 (continued)
REQUIREMENTS If the absolute difference is ~ 3%, the NIS channel is still OPERABLE, but must be readjusted. If the NIS channel cannot be properly readjusted, the channel is declared inoperable. This surveillance is primarily performed to verify the (AFD) input to the overtemperature ~T function.
SR 3.3.1 .3 compares the incore system to the NIS channel output.
If the absolute difference is ~ 3%, the NIS channel is still OPERABLE, but must be readjusted . If the NIS channel cannot be properly readjusted, the channel is declared inoperable. This surveillance is primarily performed to verify the f(AFD) input to the overtemperature ~T function .
The Note clarifies that the Surveillance is required only if reactor power is ~~% RTP and that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allowed for performing the first Surveillance after reaching ~% RTP.
Axial offset is the difference between the power in the top half of the core and the bottom half of the core expressed as a fraction (percent) of the total power being produced by the core.
Mathematically, it is expressed as:
A0=100 X
( Fluxr- Fluxe )
(Power ) ( Fluxr + Fluxs) where Fluxr =neutron flux at the top of the core, and Fluxs =neutron flux at the bottom of the core The relationship between AFD and axial offset is:
AFD =AO x ( Power ( % )/1 00)
AFD as displayed on the main control board and as determined by the plant computer use inputs from the power range NIS detectors which are located outside the reactor vessel. Axial offset is measured using incore detectors. For the performance of SR 3.3.1.3, WCAP-8648-A, EXCORE Detector Recalibration Using Quarter-Core Flux Maps, provides an acceptable method of measuring axial offset using incore detectors.
The surveillance assures that the AFD as displayed on the main control board and as determined by the plant computer is within 3% of the AFD as calculated from the axial offset equation.
Agreement is required so that the reactor is operated within the bounds of the safety analysis regarding axial power distribution.
(continued)
Vogtle Units 1 and 2 B 3.3.1-57 REVISION 3a