NL-19-0331, License Amendment Request: Revise Measurement Units Associated with Reactor Trip System Nuclear Instrument Comparison Surveillance and Trip Setpoint for Control Room Air Intake Radiation Monitors

License Amendment Request: Revise Measurement Units Associated with Reactor Trip System Nuclear Instrument Comparison Surveillance and Trip Setpoint for Control Room Air Intake Radiation Monitors

ML19346E959
Person / Time
Site:	Farley
Issue date:	12/12/2019
From:	Gayheart C Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-19-0331
Download: ML19346E959 (24)

Text

,~ Southern Nuclear Cheryl A. Gayheart Regulatory Affairs Director 3535 Colonnade Parkway Birmingham . AL 35243 205 992 5316 tel 205 992 7795 fax DEC 1 2 2019 cagayhea@southernco. com Docket Nos. : 50-348 NL-19-0331 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant- Units 1 and 2 License Amendment Request: Revise Measurement Units Associated with Reactor Trip System Nuclear Instrument Comparison Surveillance and Trip Setpoint for Control Room Air Intake Radiation Monitors Ladies and Gentlemen:

Pursuant to 10 CFR 50.90 , Southern Nuclear Operating Company (SNC) hereby requests a proposed amendment to the technical specifications (TS) for Joseph M. Farley Nuclear Plant (FNP), Unit 1 renewed facility operating license NPF-2 and Unit 2 renewed facility operating license NPF-8. The proposed amendment will:

1. Revise TS 3.3.1, "Reactor Trip System (RTS) Instrumentation," to delete the measurement unit "RTP" (rated thermal power) from the 3% absolute difference acceptance criterion specified in surveillance requirement (SR) 3.3.1.3. This change is needed to correct the acceptance criterion because the subject absolute difference is a percentage of deviation not expressed in RTP.

2. Revise TS 3.3.7, "Control Room Emergency Filtration/Pressurization System (CREFS)

Actuation Instrumentation," to change the unit of measure associated with the trip setpoint of TS Table 3.3.7-1, Function 3, "Control Room Radiation Control Room Air Intake (R-35A, B)," from "s; 800 cpm" to an equivalent value of "S: 1.0x1 0*5 ~Ci/cc" and add a footnote clarifying that the value represents radiation above background with no system flow. This change is needed to make the setpoint independent of the radiation monitor type and manufacturer and to be consistent with the units of the containment radiation gaseous and spent fuel pool room radiation gaseous monitors in TS 3.3.6, "Containment Purge and Exhaust Isolation Instrumentation," and TS 3.3.8 , "Penetration Room Filtration (PRF) System Actuation Instrumentation," respectively.

The enclosure provides a basis for the proposed changes, including a proposed no significant hazards considerations analysis. Attachments 1 and 2 contain marked-up TS pages and revised TS pages, respectively. Attachment 3 contains TS Bases pages marked to show the accompanying proposed changes for information only.

Approval of the proposed amendment is requested by December 31 , 2020. SNC will implement the amendments within 60 days of the NRC approval date.

U.S. Nuclear Regulatory Commission NL-19-0331 Page2 In accordance with 10 CFR 50.91, a copy of this letter and enclosure, including attachments, is being provided to the designated Alabama Official.

No regulatory commitments are made in this submittal.

If you should have any questions regarding this submittal , please contact Jamie Coleman at 205.992.6611.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the

~day of December 2019.

Respectfully submitted, a

Director, egu atory Affairs Southern Nuclear Operating Company CAGfTLE/scm

Enclosure:

Basis for Proposed Change Attachments 1. Technical Specification Marked-Up Pages

2. Revised (Clean) Technical Specification Pages

3. Technical Specification Bases Marked-up Pages (for information only) cc: Regional Administrator, Region II NRR Project Manager- Farley Senior Resident Inspector- Farley Director, Alabama Office of Radiation Control RType: CFA04.054

Joseph M. Farley Nuclear Plant - Units 1 and 2 License Amendment Request: Revise Measurement Units Associated with Reactor Trip System Nuclear Instrument Comparison Surveillance and Trip Setpoint for Control Room Air Intake Radiation Monitors Enclosure Basis for Proposed Change

Enclosure to NL-19-0331 Basis for Proposed Change

1.

SUMMARY

DESCRIPTION The proposed amendment to the Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2 renewed facility operating licenses consists of two changes:

1. Revise surveillance requirement (SR) 3.3.1.3 of Technical Specification (TS) 3.3.1, "Reactor Trip System (RTS) Instrumentation," to delete the measurement unit "RTP" (rated thermal power) from the 3% absolute difference comparison between the incore thermal power detector measurements versus the excore nuclear instrumentation system (NIS) axial flux difference (AFD) ; and

2. Revise TS 3.3.7, "Control Room Emergency Filtration/Pressurization System (CREFS)

Actuation Instrumentation," to change the unit of measure associated with the trip setpoint of TS Table 3.3.7-1, Function 3, "Control Room Radiation Control Room Air Intake (R-35A, 8) ," from "S 800 cpm" to an equivalent value of "S 1.0x1 o-s IJCi/cc" and add a footnote clarifying that the value represents radiation above background with no system flow.

2. DETAILED DESCRIPTION 2.1 System Design and Operation SR 3.3.1.3 Acceptance Criterion Measurement Units The overtemperature !::. T reactor trip function is provided to ensure that the design limit departure from nucleate boiling ratio is met. The inputs to the overtemperature !::. T trip include pressure, coolant temperature, axial power distribution, and reactor power as indicated by loop !::. T assuming full reactor coolant flow. The overtemperature !::. T trip uses each loop's !::. T as a measure of reactor power and is compared with a setpoint that is automatically varied with reactor coolant average temperature, pressurizer pressure, and axial power distribution. AFD is defined as the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector. AFD is expressed in units of % of flux difference.

SR 3.3.1.3 compares the incore system to the NIS channel output. When the absolute difference is~ 3% between the incore and excore AFD, the NIS channel is adjusted.

CREFS Control Room Air Intake Radiation Monitors Trip Setpoint Measurement Units The FNP Units 1 and 2 common control room must be kept habitable for the operators stationed there during normal operation, anticipated transients, and design basis accidents (DBAs). The control room function is supported by CREFS , which provides an enclosed control room environment from which the units can be operated following an uncontrolled release of radioactivity. Upon receipt of an actuation signal, the CREFS initiates filtered ventilation and pressurization of the control room .

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Enclosure to NL-19-0331 Basis for l?roposed Change TS Table 3.3.7-1, Function 3, "Control Room Radiation Control Room Air Intake (R-35A, B)," provides a trip setpoint to isolate the control room air intake. A high radiation signal from one of these detectors isolates the normal unfiltered outside air supply from the control room. The setpoint is selected to limit the radionuclide concentration in the main control room to ensure that the radiological dose to control room occupants remains well below the 10 CFR 50.67(b)(2)(iii) limit, 10 CFR 50, Appendix A general design criterion (GDC) 19 limit, and the applicable 10 CFR 20 Appendix B limit following a DBA.

2.2 Current Technical Specifications Requirements SR 3.3.1.3 Acceptance Criterion Measurement Units SR 3.3.1.3 compares the results of the incore detector measurements to the NIS AFD.

SR 3.3.1.3 states, in part: "Adjust NIS channel if absolute difference is~ 3% RTP."

CREFS Control Room Air Intake Radiation Monitors Trip Setpoint Measurement Units Table 3.3.7-1, CREFS Actuation Instrumentation, establishes a trip setpoint of "S 800 cpm" for Function 3, "Control Room Radiation Control Room Air Intake (R-35A, B)."

2.3 Reason for the Proposed Change SR 3.3.1.3 Acceptance Criterion Measurement Units This change is needed to correct the acceptance criterion because the subject absolute difference is a percentage deviation not expressed in RTP. The proposed change will also align FNP TS to the equivalent surveillance acceptance criterion specified in SR 3.3.1.3 of NUREG-1431, Revision 4 (Reference 1).

CREFS Control Room Air Intake Radiation Monitors Trip Setpoint Measurement Units The current trip setpoint units of counts per minute (cpm) are specific to the installed Victoreen radiation monitors and would require a change to the TSs if or when the radiation monitors were replaced with a different type that uses different units of measure.

The change in units to microcuries per cubic centimeter (IJCilcc) is consistent with the units assumed in the safety analysis and would be independent of the unit of measure based on a specific radiation monitor type or manufacturer. This change also makes the trip setpoint units consistent with the setpoint units for Function 3 of TS Table 3.3.6-1, "Containment Radiation Gaseous (R-24A, B)," and Function 3 of TS Table 3.3.8-1, "Spent Fuel Pool Room Radiation Gaseous (R-25A, B)."

2.4 Description of the Proposed Change SR 3.3.1.3 Acceptance Criterion Measurement Units The measurement unit "RTP" is proposed to be deleted from the 3% acceptance criterion in the last sentence of SR 3.3.1 .3.

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Enclosure to NL-19-0331 Basis for Proposed Change CREFS Control Room Air Intake Radiation Monitors Trip Setpoint Measurement Units The trip setpoint for Function 3, "Control Room Radiation Control Room Air Intake (R-35A, B) ," in TS Table 3.3.7-1 is proposed to be modified from s 800 cpm to s 1.0 X 10-5 1-JCi/cc.

In addition, proposed footnote (c) is referenced for the trip setpoint. Proposed footnote (c) states: "Above background with no flow."

3. TECHNICAL EVALUATION SR 3.3.1.3 Acceptance Criterion Measurement Units The current 3% acceptance criterion for the overtemperature 6.T instrument comparison between incore detector measurements and the NIS AFD in% RTP is not correct because the definition of AFD is % of difference from the normalized flux signals. This change does not alter the intent or meaning of the requirement but rather corrects the error introduced in FNP license amendments 203 and 199, issued on August 3, 2016 for Units 1 and 2, respectively (Reference 2) . The surveillance requirement will continue to require the overtemperature 6. T NIS channels to be adjusted when the comparison results between the incore detector measurements and the NIS AFD reaches or exceeds the 3%

acceptance criterion. Therefore, this change is considered administrative.

CREFS Control Room Air Intake Radiation Monitors Trip Setpoint Measurement Units 10 CFR 20, Appendix D defines the derived air concentration (DAC) as the airborne concentration of a given radionuclide, which, if breathed by the reference man for a working year of 2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, would result in a committed effective dose equivalent of 5 rem or a committed dose equivalent of 50 rem to any individual organ or tissue. DAC values for occupational dose are listed in Table 1, Column 3, of Appendix B to 10 CFR 20.

The predominant inhalation radionuclides released during a fuel handling accident (FHA) are Xenon (Xe) -133 and Krypton (Kr) -85 and the DAC value for each of these isotopes is 1.0E-04 1-JCi/ml, or 1.0E-04 1-JCi/cc.

The current beta scintillation detector setpoint of 800 cpm is based on the detector's Kr-85 sensitivity and is equivalent to one-tenth of the Kr-85 DAC value; i.e., 1.0E-05 1-JCi/cc.

During an FHA within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown , the inhalation radionuclide mixture released is predominantly Xe-133 and Kr-85 . During an FHA at 30 days or more following shutdown, the inhalation radio nuclide mixture released is 100% Kr-85 due to the decay of Xe-133. SNC has determined that with a control room intake isolation setpoint sensitive to Kr-85 and a release mixture of 10:1 Xe-133 to Kr-85, inhalation concentrations remain well below the associated DAC limits. Therefore, the current instrument setpoint value ensures main control room ventilation intake isolation occurs well before the control room intake concentration reaches the 10 CFR 20 occupational limit. The proposed change from 800 cpm to 1.0 X 1o-s 1-JCi/cc results in equivalent control room ventilation isolation protection and is considered administrative.

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Enclosure to NL-19-0331 Basis for Proposed Change The addition of proposed footnote (c) represents a clarification that the trip setpoint value is that value above the normal background radiation level with no adjustment for flow in the system. The addition of the clarifying footnote is consistent with the footnote associated with the trip setpoint for Function 3 of TS Tables 3.3.6-1 and 3.3.8-1 and is considered administrative.

The proposed change continues to ensure the main control room ventilation intake isolation occurs well before the intake concentration reaches the applicable 10 CFR 20 occupational limit, which is consistent with the GDC 19 and 10 CFR 50.67(b)(2)(iii) dose limit for control room personnel.

4. REGULATORY EVALUATION Following implementation of the proposed changes, FNP Units 1 and 2 will remain in compliance with applicable regulations and design criteria as described in the FNP UFSAR and as previously licensed and approved by the NRC.

4.1 Applicable Regulatorv Requirements/Criteria The RTS and CREFS radiation monitor designs satisfy 10 CFR 50.36 , "Technical specifications," paragraph (c)(2)(ii) , Criterion 3. A reactor trip from overtemperature Ll T and control room ventilation isolation from inlet radiation monitors are considered primary success paths to mitigate certain accidents and transients as described in Chapters 7 and 15 of the FNP UFSAR.

The proposed amendment does not delete requirements associated with the RTS or CREFS isolation instrumentation. Limiting condition for operation (LCO) 3.3 .1 and LCO 3.3.7 continue to maintain requirements associated with structures, systems, and components that are part of the primary success path and actuate to mitigate the related design basis accidents and transients. The proposed amendment does not alter the remedial actions or shutdown requirements required by 10 CFR 50.36(c)(2)(i) for RTS or CREFS isolation instrumentation.

The proposed change to the CREFS control room air intake radiation monitors isolation trip setpoint to an equivalent setpoint in units of IJCi/cc instead of cpm with the addition of a clarifying note continues to ensure adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent for the duration of the accident; thus, maintaining compliance with 10 CFR 50 .67(b)(2)(iii) .

The RTS and CREFS control room isolation instrumentation designs comply with the requirements of 10 CFR 50.55a(h)(2) , Protection and Safety Systems, and meet the requirements of the applicable Institute of Electrical and Electronics Engineers standard pursuant to 10 CFR 50.55a(2) , as previously licensed and approved by the NRC. The proposed amendment does not alter the design of any protection or safety system, E-4

Enclosure to NL-19-0331 Basis for Proposed Change including the RTS and CREFS isolation instrumentation. Therefore, the protection and safety system design continues to meet the requirements of 10 CFR 50.55a.

In addition , the following 10 CFR Part 50, Appendix A General Design Criteria are related to the instrumentation design:

General Design Criterion (GDC 10): Reactor design. The proposed amendment does not alter the design of the reactor core and associated coolant, control, and protection systems, including the RTS. The change deletes the unit "RTP" from the 3% deviation adjustment criterion since the correct unit is% of absolute deviation, not% of RTP.

GDC 13: Instrumentation and control. The proposed amendment does not alter the design of the instrumentation that is 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. The change deletes the unit "RTP from the 3% deviation adjustment criterion since the correct unit is % of absolute deviation, not% of RTP. The change also revises the trip setpoint in units of cpm to an equivalent setpoint in units of !JCi/cc with a clarifying note that the value is that above background level with no flow.

GDC 15: Reactor coolant system design. The proposed amendment does not alter the design of the control and protection instrumentation that is provided to assure, with sufficient margin, that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.

GDC 19: Control room . The proposed amendment does not alter the design or impact the capability of the control room to provide a habitable location from where actions can be taken to operate the FNP units safely under normal conditions and to maintain them in a safe condition under accident conditions, including loss-of-coolant accidents. The change revises the CREFS control room ventilation isolation instrumentation trip setpoint in units of cpm to an equivalent setpoint in units of !JCi/cc with a clarifying note that the value is that above background level with no flow. Because this change is an equivalent change in units, adequate radiation protection continues to be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident to the extent previously licensed and approved by the NRC.

GDC 20: Protection system functions. The proposed amendment does not alter the design of reactivity control protection systems or instrumentation that sense accident conditions to initiate systems or components important to safety. The change deletes the unit "RTP" from the 3% deviation adjustment criterion since the correct unit is % of absolute deviation, not% of RTP. The change also revises the CREFS control room ventilation isolation instrumentation trip setpoint in units of cpm to an equivalent setpoint in units of !JCi/cc with a clarifying note that the value is that above background level with E-5

Enclosure to NL-19-0331 Basis for Proposed Change no flow. The overtemperature!::.T and control room ventilation isolation instrumentation continue to initiate a reactor trip and control room ventilation isolation, respectively; both systems are important to safety.

GDC 21: Protection system reliability and testability. The proposed amendment does not alter the design of any protection system, including the RTS and control room ventilation isolation instrumentation. Therefore, the protection system design continues to provide high functional reliability and inservice testability commensurate with the safety functions to be performed and continues to be sufficient to assure that (1) no single failure results in loss of the protection function, and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy.

This change does not impact the RTS or control room ventilation isolation instrumentation design; thus, the design of these instrumentation systems continues to permit periodic testing of their funtionality when the reactor is in operation as previously licensed and approved by the NRC.

GDC 29: Protection against anticipated operational occurrences. The proposed amendment does not alter the design of any protection or reactivity control system, including the RTS and control room ventilation isolation instrumentation. Therefore, the protection and reactivity control system design continues to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences to the extent previously licensed and approved by the NRC.

4.2 Precedent None identified.

4.3 No Significant Hazards Consideration Determination Analysis Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC) hereby requests an amendment to Joseph M. Farley Nuclear Plant (FNP) Unit 1 renewed facility operating license NPF-2 and Unit 2 renewed facility operating license NPF-8. The proposed amendment revises technical specification (TS) 3.3.1 , "Reactor Trip System (RTS)

Instrumentation," by deleting the unit "RTP" (rated thermal power) from the 3% deviation adjustment criterion associated with the overtemperature !::. T trip since the correct unit is

%of absolute deviation, not% of RTP. The proposed amendment also revises TS 3.3.7, "Control Room Emergency Filtration/Pressurization System (CREFS) Actuation Instrumentation," to change the control room ventilation radiation isolation trip setpoint in units of counts per minute (cpm) to an equivalent setpoint in units of microcuries per cubic centimeter (IJCi/cc) with a clarifying note that the value is that above background radiation level with no flow. The changes are considered administrative in nature with no technical impact.

SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92 ,

"Issuance of amendment," as discussed below:

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Enclosure to NL-19-0331 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 does not affect accident initiators or precursors nor adversely alter the design assumptions, conditions, and configuration of the facility.

The proposed amendment does not alter any plant equipment or operating practices with respect to such initiators or precursors in a manner that the probability of an accident is increased.

The proposed amendment represents an administrative change and does not involve a physical change to the RTS or the CREFS actuation instrumentation, nor does it change the safety function of the instrumentation or the equipment supported by the instrumentation. The change deletes an incorrect unit of measure from the adjustment criterion associated with the overtemperature 1:::.T reactor trip and revises the measurement units of CREFS control room ventilation high radiation isolation trip setpoint with an equivalent value and a clarifying note that the value is that above background level with no flow. As a result, the proposed amendment does not alter assumptions relative to the mitigation of an accident or transient event, and therefore, the consequences of an accident are not changed.

Therefore, the proposed change 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 With respect to a new or different kind of accident, there are no proposed design changes to the RTS; nor are there any changes in the method by which safety related plant structures, systems, and components (SSCs) perform their specified safety functions. The proposed amendment will not affect the normal method of plant operation or revise any operating parameters. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of this proposed change and the failure modes and effects analyses of SSCs important to safety are not altered as a result of this proposed administrative change. The change deletes an incorrect unit of measure from the adjustment criterion associated with the overtemperature 1:::. T reactor trip and revises the measurement units of CREFS control room ventilation high radiation isolation trip setpoint with an equivalent value and a clarifying note that the value is that above background level with no flow.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated .

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Enclosure to NL-19-0331 Basis for Proposed Change (3) Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The margin of safety is related to the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding , the reactor coolant system, and the containment. The performance of these fission product barriers is not affected by the proposed change.

Instrumentation safety margin is established by ensuring the limiting safety system settings automatically actuate the applicable design function to correct an abnormal situation before a safety limit is exceeded. Safety analysis limits are established for RTS instrumentation functions related to those variables having significant safety functions.

The overtemperature 11T reactor trip function and the control room ventilation high radiation isolation function are not altered as a result of the proposed administrative change. The change deletes an incorrect unit of measure from the adjustment criterion associated with the overtemperature 11T reactor trip and revises the measurement units of the control room ventilation high radiation isolation trip setpoint with an equivalent value and a clarifying note that the value is that above background level with no flow.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusion 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. ENVIRONMENTAL CONSIDERATION SNC has determined that the proposed change does not modify requirements with respect to installation or use of facility components located within the restricted area, as defined in 10 CFR 20, and does not change an inspection or surveillance requirement. The proposed change 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 E-8

Enclosure to NL-19-0331 Basis for Proposed Change 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}, no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6. REFERENCES

1. NRC NUREG-1431 , "Standard Technical Specifications- Westinghouse Plants, Volume 1 Specifications," Revision 4.0 (NRC Agencywide Documents Access and Management System (ADAMS) Accession No. ML12100A222).

2. Letter from S. A. Williams (NRC) to C. R. Pierce (SNC} , "Joseph M. Farley Nuclear Plant, Units 1 and 2- Issuance of Amendments Adopting 21 Previously NRC-approved TSTF Travelers and One Request Not Associated with TSTF Travelers (CAC Nos.

MF5317 and MF5318)," dated August 3, 2016 (NRC ADAMS Accession No. ML15233A448).

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Joseph M. Farley Nuclear Plant- Units 1 and 2 License Amendment Request: Revise Measurement Units Associated with Reactor Trip System Nuclear Instrument Comparison Surveillance and Trip Setpoint for Control Room Air Intake Radiation Monitors Attachment 1 Technical Specification Marked-up Pages

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 ---------------------- ----NOTE--------------------------

Not required to be performed for source range instrumentation until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after THERMAL POWER is< P-6 .

Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.1.2 - ---------- --------------NOTE------------ -------------

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.

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-----------------------------

1. Not required to be performed until7 days after THERMAL POWER is~ 50% RTP.

2. Performance of SR 3.3.1.9 satisfies this SR.

Compare results of the in core detector In accordance with measurements to Nuclear Instrumentation System the Surveillance (NIS) AFD. Adjust NIS channel if absolute difference Frequency Control is ~ 3%--R-+P. Program Farley Units 1 and 2 3.3.1-9 Amendment No. 2-W (Unit 1)

Amendment No. 4-W (Unit 2)

CREFS Actuation Instrumentation 3.3.7 Table 3.3.7-1 (page 1 of 1)

CREFS Actuation Instrumentation FUNCTION APPLICABLE REQUIRED SURVEILLANCE TRIP SETPOINT MODES OR OTHER CHANNELS REQUIREMENTS SPECIFIED CONDITIONS

1. Manual Initiation 1,2,3,4, (a), (b) 2trains SR 3.3.7.6 NA

2. Automatic Actuation Logic 1,2,3,4 2trains SR 3.3.7.3 NA and Actuation Relays SR 3.3.7.4 SR 3.3.7.5

3. Control Room Radiation 1,2,3,4 SR 3.3.7.1 s ~ 1 . ox1o* 5 Control Room Air Intake (a) , (b) 2 SR 3.3.7.2 uCilcc (c)

(R-35A, B) SR 3.3.7.7

4. Containment Isolation - Refer to LCO 3.3.2, "ESFAS Instrumentation," Function 3.a., for all initiation functions and Phase A requirements.

(a) During CORE ALTERATIONS.

(b) During movement of irradiated fuel assemblies.

(c) Above background with no flow.

Farley Units 1 and 2 3.3.7-4 Amendment No. 449 (Unit 1)

Amendment No. J+. (Unit 2)

Joseph M. Farley Nuclear Plant - Units 1 and 2 License Amendment Request: Revise Measurement Units Associated with Reactor Trip System Nuclear Instrument Comparison Surveillance and Trip Setpoint for Control Room Air Intake Radiation Monitors Attachment 2 Revised (Clean) Technical Specification Pages

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 -----------------------NOTE-------------------------

Not required to be performed for source range instrumentation until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after THERMAL POWER is< P-6.

Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.1.2 -------------------------N 0TE--------------------------

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.

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 TE S----------------------------

1. Not required to be performed until 7 days after THERMAL POWER is ;::: 50% RTP.

2. Performance of SR 3.3.1.9 satisfies this SR.

Compare results of the in core detector In accordance with measurements to Nuclear Instrumentation System the Surveillance (NIS) AFD. Adjust NIS channel if absolute difference Frequency Control is~ 3%. Program Farley Units 1 and 2 3.3.1-9 Amendment No. (Unit 1)

Amendment No. (Unit 2)

CREFS Actuation Instrumentation 3.3.7 Table 3.3 .7-1 (page 1 of 1)

CREFS Actuation Instrumentation FUNCTION APPLICABLE REQUIRED SURVEILLANCE TRIP SETPOINT MODES OR OTHER CHANNELS REQUIREMENTS SPECIFIED CONDITIONS

1. Manual Initiation 1,2,3,4, (a), (b) 2 trains SR 3.3.7.6 NA

2. Automatic Actuation Logic 1,2,3,4 2 trains SR 3.3.7.3 NA and Actuation Relays SR 3.3.7.4 SR 3.3.7.5

1.0 X 1o* 1-JCi/cc (c) 5

3. Control Room Radiation 1,2,3,4 1 SR 3.3.7.1 Control Room Air Intake (a), (b) 2 SR 3.3.7.2 (R-35A, B) SR 3.3.7.7

4. Containment Isolation- Refer to LCO 3.3.2, "ESFAS Instrumentation," Function 3.a., for all initiation functions and Phase A requirements.

(a) During CORE ALTERATIONS.

(b) During movement of irradiated fuel assemblies.

(c) Above background with no flow.

Farley Units 1 and 2 3.3.7-4 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Joseph M. Farley Nuclear Plant - Units 1 and 2 License Amendment Request: Revise Measurement Units Associated with Reactor Trip System Nuclear Instrument Comparison Surveillance and Trip Setpoint for Control Room Air Intake Radiation Monitors Attachment 3 Technical Specification Bases Marked-up Pages (for information only)

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.2 (continued)

REQUIREMENTS SR 3.3.1.2 is modified by a Note. This Note clarifies that this Surveillance is required only if reactor power is ~ 15% RTP and that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are allowed for performing the first Surveillance after reaching 15% RTP. A power level of 15% RTP is chosen based on plant stability, i.e., automatic rod control capability and turbine generator synchronized to the grid.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.1.3 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 it must be adjusted. The excore NIS channel shall be adjusted if the absolute difference between the incore and excore AFD is~3%~ .

If the NIS channel cannot be properly adjusted, the channel is declared inoperable. This Surveillance is performed to periodically verify the f(ill) input to the overtemperature .llT Function.

Two Notes modify SR 3.3.1.3. Note 1 clarifies that the Surveillance is required only if reactor power is ~ 50% RTP and that 7 days are allowed for performing the Surveillance and channel adjustment, if necessary, after reaching 50% RTP. A power level of~ 50% RTP is consistent with the requirements of SR 3.3.1.9. Note 2 allows SR 3.3.1.9 to be performed in lieu of SR 3.3.1.3, since SR 3.3.1.9 calibrates (i.e.,

requires channel adjustment) the excore channels to the incore channels, it envelopes the performance of SR 3.3.1.3.

For each operating cycle, the initial channel normalization is performed under SR 3.3.1.9. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.1.4 SR 3.3.1.4 is the performance of a TADOT. This test shall verify OPERABILITY by actuation of the end devices.

The RTB test shall include separate verification of the undervoltage trip via the Reactor Protection System and the local manual shunt trip (continued)

Farley Units 1 and 2 B 3.3.1-52 Revision ++

CREFS Actuation Instrumentation B 3.3.7 BASES LCO 2. Automatic Actuation Logic and Actuation Relays (continued)

The LCO requires two trains of Actuation Logic and Relays OPERABLE to ensure that no single random failure can prevent automatic actuation.

Automatic Actuation Logic and Actuation Relays consist of the same features and operate in the same manner as described for ESFAS Function 3.a.2, Containment Isolation-Phase A, in LCO 3.3.2. The Actions Conditions for the CREFS portion of these Functions are different and less restrictive than those specified for their Phase A Isolation roles. If one or more of the Phase A Isolation Functions becomes inoperable in such a manner that only the CREFS Function is affected, the Conditions applicable to their Phase A Isolation Function need not be entered . The less restrictive Actions specified for inoperability of the CREFS Functions specify sufficient compensatory measures for this case.

3. Control Room Radiation The LCO specifies one required Control Room Air Intake Radiation Monitor in MODES 1-4 to ensure that the radiation monitoring instrumentation necessary to provide a backup initiation of control room isolation remains OPERABLE. The LCO requires two air intake radiation monitor channels OPERABLE during CORE ALTERATIONS and during movement of irradiated fuel assemblies when the radiation monitor channels provide the primary control room protection function.

For sampling systems, channel OPERABILITY involves more than OPERABILITY of channel electronics. OPERABILITY also requires correct valve lineups and sample pump operation , as well as detector OPERABILITY.The setpoint is based upon a release during a fuel handling accident in which Krvpton-85 and Xenon-133 are the predominant radionuclides. Derived air concentration (DAC) limits are listed in Table 1, Column 3 of Appendix B to 10 CFR 20 (Reference 1). The trip setpoint ensures control room ventilation intake isolation occurs well before the intake concentration reaches the 10 CFR 20 occupational limit, which is consistent with the GDC 19 and 10 CFR 50 .67(b)(2){iii) dose limit for control room personnel (References 2 and 3).

Footnote (c) clarifies that the trip setpoint value is that value above the normal background radiation level with no adjustment for flow in the system.

(continued)

Farley Units 1 and 2 B 3.3.7-3 Revision G

CREFS Actuation Instrumentation B 3.3.7 BASES LCO 4. Containment Isolation-Phase A (continued)

Refer to LCO 3.3.2, Function 3.a, for all initiating Functions and requirements except as described above in item 2, "Automatic Actuation Logic and Actuation Relays."

APPLICABILITY The CREFS Functions must be OPERABLE in MODES 1, 2, 3, 4, and the radiation monitor and manual initiation Functions must also be OPERABLE during CORE ALTERATIONS and movement of irradiated fuel assemblies to ensure a habitable environment for the control room operators. The Applicability for the CREFS actuation on the ESFAS Containment Isolation-Phase A Functions are specified in LCO 3.3.2. Refer to the Bases for LCO 3.3.2 for discussion of the Containment Isolation-Phase A Function Applicability.

ACTIONS The most common cause of channel inoperability is outright failure or drift of the bistable or process module sufficient to exceed the tolerance allowed by the unit specific calibration procedures.

Typically, the drift is found to be small and results in a delay of actuation rather than a total loss of function. This determination is generally made during the performance of a COT, when the process instrumentation is set up for adjustment to bring it within specification.

If the Trip Setpoint is less conservative than the tolerance specified by the calibration procedure, the channel must be declared inoperable immediately and the appropriate Condition entered .

A Note has been added to the ACTIONS indicating that separate Condition entry is allowed for each Function. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.7-1 in the accompanying LCO. The Completion Time(s) of the inoperable channel(s)/train(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.

Condition A applies to the actuation logic train Function of the CREFS, the radiation monitor channel Functions, and the manual channel Functions.

If one train is inoperable, or one required radiation monitor channel is inoperable in one or more Functions, 7 days are permitted to restore it to OPERABLE status. The 7 day Completion Time is the same as is allowed if one train of the mechanical portion of the system is inoperable. The basis for this Completion Time is the same as provided in LCO 3.7.10. If the channel/train cannot be restored to (continued)

Farley Units 1 and 2 B 3.3.7-4 Revision 4-8

CREFS Actuation Instrumentation B 3.3.7 BASES ACTIONS A.1 (continued)

OPERABLE status, one CREFS train must be placed in the emergency recirculation mode of operation. This accomplishes the actuation instrumentation Function and places the unit in a conservative mode of operation.

B.1.1 . B.1.2. and B.2 Condition B applies to the failure of two CREFS actuation trains, two required radiation monitor channels, or two manual initiation trains.

The first Required Action is to place one CREFS train in the emergency recirculation mode of operation immediately. This accomplishes the actuation instrumentation Function that may have been lost and places the unit in a conservative mode of operation.

The applicable Conditions and Required Actions of LCO 3.7.10 must also be entered for the CREFS train made inoperable by the inoperable actuation instrumentation. In the case of inoperable radiation monitors, one train of CREFS must be declared inoperable and the applicable Condition of LCO 3.7.10 entered . This ensures appropriate limits are placed upon train inoperability as discussed in the Bases for LCO 3.7.10.

Alternatively, both trains may be placed in the emergency recirculation mode. This ensures the CREFS function is performed even in the presence of a single failure.

C.1 and C.2 Condition C applies when the Required Action and associated Completion Time for Condition A or B have not been met and the unit is in MODE 1, 2, 3, or 4. Condition C is only applicable to those CREFS functions in Table 3.3.7-1 required OPERABLE in MODES 1-4. The unit must be brought to a MODE in which overall plant risk is reduced. To achieve this status, the unit must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Remaining within the applicability of the LCO is acceptable to accomplish short duration repairs to restore inoperable equipment because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 1). In MODE 4 the Steam Generators and Residual Heat Removal System are available to remove decay heat, which provides diversity and defense in depth. As stated in Reference ~. the steam turbine driven Auxiliary Feedwater Pump must be available to remain in MODE 4.

(continued)

Farley Units 1 and 2 B 3.3.7-5 Revision ~

CREFS Actuation Instrumentation B 3.3.7 BASES SURVEILLANCE SR 3.3.7.7 REQUIREMENTS (continued) The CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 20. Appendix B - Annual Limits on Intake (Alls) and Derived Air Concentrations <DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations: Concentrations for Release to Sewerage.

2. 10 CFR 50. Appendix A - General Design Criteria for Nuclear Power Plants.

3. 10 CFR 50.67, Accident source term.

4._

_ WCAP-16294-NP-A, Rev. 1, "Risk-Informed Evaluation of Changes to Technical Specification Required Action Endstates for Westinghouse NSSS PWRs," June 2010.

Farley Units 1 and 2 B 3.3.7-9 Revision ~