NL-24-0259, License Amendment Request to Revise Technical Specification 3.4.14, Reactor Coolant System Pressure Isolation Valve Leakage, Surveillance Requirement 3.4.14.3 Acceptance Criteria and Miscellaneous Obsolete Changes
| ML24248A273 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 09/04/2024 |
| From: | Coleman J Southern Nuclear Operating Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| NL-24-0259 | |
| Download: ML24248A273 (1) | |
Text
A Southern Nuclear Regulatory Affairs 3535 Colonnade Park w ay Birmingham, AL 35243 205 992 5000
September 4, 2024 NL-24-0259 10 CFR 50.90
U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001
Farley Nuclear Plant Units 1 &2 Docket Nos.: 50-348 & 50-364
Subject:
License Amendment Request to Revise Technical Specification 3.4.14, Reactor Coolant System Pressure Isolation Valve Leakage, Surveillance Requirement 3.4.14.3 Acceptance Criteria and Miscellaneous Obsolete Changes
Ladies and Gentlemen:
In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR)
Section 50.90, Southern Nuclear Operating Company (SNC) requests an amendment to Renewed Facility Operating License Nos. NPF-2 and NPF-8 for the Joseph M. Farley Nuclear Plant (FNP) Units 1 and 2. This amendment request proposes to revise Technical Specification (TS) TS 3.4.14, Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage, Surveillance Requirement (SR) 3.4.14.3 Acceptance Criteria and remove other miscellaneous obsolete changes.
The Enclosure provides a description and assessment of the proposed changes including the technical analyses, regulatory analyses, and environmental considerations.
This letter contains no regulatory commitments. This letter has been reviewed and determined not to contain security-related information.
SNC requests approval of the proposed license amendment within 12 months of acceptance with the amendment being implemented within 90 days of issuance.
In accordance with 10 CFR 50.91 (b )(1 ), SNC is notifying the State of Alabama of this license amendment request by transmitting a copy of this application and the reasoned analysis about no significant hazards consideration to the designated State Official.
If you should have any questions about this submittal, please contact Ryan Joyce at 205.992.6468.
U.S. Nuclear Regulatory Commission NL-24-0259 Page 2
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 4th day of September 2024.
Respectfully submitted,
~Wvvuv~
Jamie Coleman Director, Regulatory Affairs Southern Nuclear Operating Company
Enclosure:
Basis for Proposed Change
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 for Changes to Technical Specification 3.4.14, Reactor Coolant System Pressure Isolation Valve Leakage, Surveillance Requirement 3.4.14.3 Acceptance Criteria and Miscellaneous Obsolete Changes
Enclosure
Basis for Proposed Change
- 1.
SUMMARY
DESCRIPTION
- 2. DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Technical Specifications Requirements 2.3 Reason for the Proposed Change 2.4 Description of the Proposed Change
- 3. TECHNICAL EVALUATION
- 4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 No Significant Hazards Consideration Determination Analysis 4.3 Conclusions
- 5. ENVIRONMENTAL CONSIDERATION
- 6. REFERENCES
ATTACHMENTS:
- 1. Technical Specification Page Markups
- 2. Retyped Technical Specification Pages
- 3. Bases Page Markups (for information only)
Enclosure to NL-24-0259 Basis for Proposed Change
1
SUMMARY
DESCRIPTION
Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC) hereby requests an amendment to Renewed Facility Operating License Nos. NPF-2 and NPF-8 for the Joseph M. Farley Nuclear Plant (FNP) Units 1 and 2. The proposed amendment involves changes to Technical Specifications (TS) 3.4.14, Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage, to:
(a) Delete Condition C Note, Condition C text "autoclosure or, " and SR 3.4.14.2 (including renumbering SR 3.4.14.3 to SR 3.4.14.2), which were no longer applicable after restart from Unit 1 Refueling Outage 1 R27 and restart from Unit 2 Refueling Outage 2R25;
(b) Revise Surveillance Requirement (SR) 3.4.14.3 (which is renumbered to SR 3.4.14.2)
Residual Heat Removal (RHR) System open permissive interlock (OPI) interlock acceptance criterion; and
(c) Revise the SR 3.4.14.3 Note reference to SR 3.4.12.3 to instead reference SR 3.4.12.4 and delete duplicative "valves. "
The proposed amendment also involves changes to TS 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation, since these provisions are no longer applicable after restart from Unit 1 outage 1 R28 and restart from Unit 2 outage 2R25. The changes include:
(d) Delete the Limiting Condition for Operation (LCO) Notes,
(e) Delete Conditions A and B Notes,
(f) Delete Table 3.3.5-1 and references to it,
(g) Delete Actions D, E, and F, and
(h) Renumber Table 3.3.5-2 and references to it to Table 3.3.5-1.
2 DETAILED DESCRIPTION
2.1 System Design and Operation
During normal and emergency conditions, the low pressure RHR System (design pressure of 600 psig) is isolated from the high pressure Reactor Coolant System (RCS)
(normal operating pressure of 2235 psig). Isolation is necessary to: 1) avoid RHR System over pressurization, and 2) minimize the potential for loss of integrity of the low pressure system and possible radioactive releases to the environment.
Two suction/isolation valves are provided on each inlet line from the RCS to the RHR System inside containment. These motor-operated gate valves are normally-closed to keep the low pressure RHR System isolated from the high pressure RCS, and are opened only when the RHR System is in operation or when required open to support low temperature overpressure protection (L TOP). The RHR suction isolation valves are
E-1 Enclosure to NL-24-0259 Basis for Proposed Change
interlocked with RCS pressure signals to prevent opening when the RCS pressure is greater than the Open Permissive Interlock (OPI) interlock setpoint. The setpoint (i.e., ::; 415 psig) takes into account instrument uncertainty and calibration tolerances.
This value also provides assurance that the RHR System suction relief valves setpoint will not be exceeded. Thus, the OPI prevents inadvertent opening of the RHR System isolation valves when the RCS pressure is above the valve opening interlock.
TS 3.4.12, "Low Temperature Overpressure Protection (L TOP) System ", SR 3.4.12.4 requires these RHR suction valves to be open when the RHR suction relief valves are required to be operable for overpressure mitigation (i.e., when the reactor coolant system is not vented with a ~ 2.85 sq in opening).
2.2 Current Technical Specifications Requirements
FNP Amendment Nos. 201 and 197 (Unit 1 and 2, respectively) (Reference 1) provided for elimination of the TS and SR requirements for the RHR system suction valve autoclosure interlock (ACI) function, by including TS Notes that state when the current TS requirement would no longer be applicable for each unit with the following changes, which remain in current TS:
- TS 3.4.14 Condition C Note states: "Not applicable to the autoclosure interlock for Unit 1 after restart from 1 R27 and for Unit 2 after restart from 2R25, " and Condition C applies to inoperabilities of the ACI, as well as the OPI by stating:
"RHR System autoclosure or open permissive interlock function inoperable. "
- SR 3.4.14.2 provides the surveillance for verifying the RHR System ACI, while Note 2 states: " Not applicable to Unit 1 after restart from 1 R27 and not applicable to Unit 2 after restart from 2R25. "
Current SR 3.4.14.3 provides the surveillance to verify the OPI. The current acceptance criterion for the OPI interlock is stated as "~ 295 psig and ::; 415 psig. "
Current SR 3.4.14.3 also includes a Note stating: "Not required to be met when the RHR System valves valves are required open in accordance with SR 3.4.12.3. " The referenced SR 3.4.12.3 states: "Verify each accumulator is isolated, " while SR 3.4.12.4 states: "Verify RHR suction isolation valves are open for each required RHR suction relief valve. "
FNP Amendment Nos. 206 and 202 (Unit 1 and 2, respectively) (Reference 2) revised the setpoint requirements in TS 3.3.5, "Loss of Power Diesel Generator Start Instrumentation," with delayed implementation provisions that are currently expired.
2.3 Reason for the Proposed Change
For the three proposed changes to TS 3.4.14 outlined in the Summary
Description:
(a) Condition C Note, Condition C text "autoclosure or, " and SR 3.4.14.2 are proposed for deletion, due to no longer being applicable after restart from Unit 1 Refueling Outage 1 R27 and restart from Unit 2 Refueling Outage 2R25. FNP Unit 1 has
E-2 Enclosure to NL-24-0259 Basis for Proposed Change
completed Refueling Outage 1 R31 and Unit 2 has completed Refueling Outage 2R29. Therefore, the Action and surveillance associated with the ACI function are no longer applicable. Their deletion reflects an administrative change to eliminate extraneous verbiage.
(b) The RHR OPI interlock acceptance criterion for SR 3.4.14.3 (which is proposed to be renumbered to SR 3.4.14.2) should reflect the value that prevents opening valves which could lead to RHR System over pressurization. The pressure interlock functions at a single setpoint and not between a range as suggested by the current acceptance criterion of "~ 295 psig and ::; 415 psig. " The safety basis is to prevent opening the valves when above a specified pressure.
(c) FNP Amendment Nos. 193 and 189 (Units 1 and 2, respectively) (Reference 3) revised TS 3.4.12, "Low Temperature Overpressure Protection (L TOP) System, "
and in part split SR 3.4.12.1 into SR 3.4.12.1 and SR 3.4.12.2, thereby renumbering the remaining SRs. Prior to this Amendment SR 3.4.12.3 was "Verify RHR suction isolation valves are open for each required RHR suction relief valve, " which was referenced by SR 3.4.14.3. The Amendment renumbered this SR to SR 3.4.12.4, however, SR 3.4.14.3 was not updated to reflect the renumbering. As such, the current reference to SR 3.4.12.3 in the Note to SR 3.4.14.3 is proposed to be revised to SR 3.4.12.4. Additionally, the Note includes a repetitive "valves "
requiring an editorial deletion.
The proposed changes to TS 3.3.5 outlined in the Summary Description are proposed due to no longer being applicable after restart from Unit 1 outage 1 R28 and restart from Unit 2 outage 2R25. FNP Unit 1 has completed Refueling Outage 1 R31 and Unit 2 has completed Refueling Outage 2R29. Therefore, the provisions associated with the temporary TS 3.3.5 changes reflect an administrative change to eliminate extraneous verbiage.
2.4 Description of the Proposed Change
TS 3.4.14, RCS PIV Leakage, is proposed to be revised as follows:
(a) The following are proposed to be deleted:
- i. Condition C Note, ii. Condition C text "autoclosure or, " and iii. SR 3.4.14.2 (and concurrently renumbering SR 3.4.14.3 to SR 3.4.14.2)
(b) SR 3.4.14.3 (which is renumbered to SR 3.4.14.2) is proposed to state:
"Verify RHR System open permissive interlock prevents the valves from being opened with a simulated or actual RCS pressure signal ~ 415 psig "
(c) SR 3.4.14.3 (which is renumbered to SR 3.4.14.2) Note is proposed to be revised:
- i. To replace reference to SR 3.4.12.3 with reference to SR 3.4.12.4, and ii. Revise the Note to delete the repetitive "valves "
E-3 Enclosure to NL-24-0259 Basis for Proposed Change
TS 3.3.5, LOP DG Start Instrumentation, is proposed to be revised as follows:
(d) Delete the Limiting Condition for Operation (LCO) Notes,
(e) Delete Conditions A and B Notes,
(f) Delete Table 3.3.5-1 and references to it,
(g) Delete Actions D, E, and F, and
(h) Renumber Table 3.3.5-2 and references to it to Table 3.3.5-1.
Markups of the TS are provided in Attachment 1. Attachment 2 provides the clean typed copy of the proposed TS changes. Appropriate Bases changes consistent with the TS changes discussed above will be made during implementation of the approved Amendment. Attachment 3 provides a markup depicting conforming changes to the TS Bases for information only.
3 TECHNICAL EVALUATION
For the proposed changes to TS 3.4.14, RCS PIV Leakage:
(a) FNP Amendment Nos. 201 and 197 (Unit 1 and 2, respectively) approved the elimination of the TS and SR requirements for the RHR system suction valve ACI function after restart from Unit 1 Refueling Outage 1 R27 and restart from Unit 2 Refueling Outage 2R25. With the recent startup from Refueling Outage 1 R31 for Unit 1 and startup from Refueling Outage 2R29 for Unit 2, the requirements for the RHR system suction valve ACI function no longer apply. The obsolete requirements pose extraneous verbiage, which represents an unnecessary distraction. Removal of unnecessary distracting verbiage reflects a human-factors improvement in the presentation of TS requirements.
(b) The RHR OPI interlock acceptance criterion for SR 3.4.14.3 (which is proposed to be renumbered to SR 3.4.14.2) should reflect the value that prevents opening valves which could lead to RHR System over pressurization. The pressure interlock precludes opening at and above a specified pressure, and not between a range of pressures as suggested by the current acceptance criterion of "~ 295 psig and ::; 415 psig. " Although the format of the SR requirement is changed, the intent in meeting the SR is unchanged. For the revised SR to be considered met, calibration of the interlock setting is performed ::; 415 psig with appropriate uncertainty margin, which will ensure that the valves will not open with a simulated or actual RCS pressure signal ~ 415 psig. The safety basis to prevent opening the valves when above a specified pressure should be reflected in the surveillance as a greater than or equal to pressure. There is no safety basis to prevent opening at lower pressures (e.g., below 415 psig, including pressures as low as 295 psig). Preventing RHR suction valve opening at lower pressures nonconservatively precludes operation of RHR for performing the safety-related function of a decay heat removal function. As such, that proposed change revises the range to instead to be the pressure above which the interlock is required to be in effect,
E-4 Enclosure to NL-24-0259 Basis for Proposed Change
i.e.,~ 415 psig, which provides adequate margin to the low pressure RHR System design pressure of 600 psig.
The Standard Technical Specifications (STS), Westinghouse Plants, NUREG-1431, similarly prescribes the RHR System autoclosure interlock (i.e., FNP "open permissive interlock ") setting as a single pressure above which the interlock is required to be in effect (i.e., "~ [425] psig " ). During the FNP proposed amendment to convert to the NUREG-1431 Standard TS, FNP added this SR with an FNP-specific RCS pressure "range " for the open permissive interlock in place of the single value in the STS (Reference 4 ). Regarding demonstration of the specified safety function (preventing RHR system overpressurization), the change proposed ensures that the valves will not open ~ 415 psig and restores consistency with the STS for this SR acceptance criterion and is therefore considered an administrative change.
(c) FNP Amendment Nos. 193 and 189 (Units 1 and 2, respectively) (Reference 3) revised TS 3.4.12, "Low Temperature Overpressure Protection (L TOP) System, " in part to split SR 3.4.12.1 into SR 3.4.12.1 and SR 3.4.12.2, thereby renumbering the remaining TS 3.4.12 SRs. Prior to this Amendment SR 3.4.12.3 was "Verify RHR suction isolation valves are open for each required RHR suction relief valve, " which was referenced by SR 3.4.14.3 Note, which states "Not required to be met when the RHR System valves are required open in accordance with SR 3.4.12.3. " The Amendment renumbered the referenced SR 3.4.12.3 to SR 3.4.12.4, however, SR 3.4.14.3 Note was not updated to reflect the renumbering. The intent of the currently referenced "SR 3.4.12.3 " is to align with what is now SR 3.4.12.4. As such, the current reference to SR 3.4.12.3 in the Note to SR 3.4.14.3 is proposed to be revised to SR 3.4.12.4. Additionally, the Note includes a repetitive "valves " requiring an editorial deletion.
For the proposed changes to TS 3.3.5, LOP DG Start Instrumentation, the provisions are no longer applicable after restart from Unit 1 outage 1 R28 and restart from Unit 2 outage 2R25.
FNP Unit 1 has completed Refueling Outage 1 R31 and Unit 2 has completed Refueling Outage 2R29. Therefore, the provisions associated with the temporary TS 3.3.5 changes reflect an administrative change to eliminate extraneous verbiage.
4 REGULATORY ANALYSIS
4.1 Applicable Regulatory Requirements/Criteria
General Design Criterion (GDC) 14 - Reactor coolant pressure boundary. The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.
GDC 30 - Quality of reactor coolant pressure boundary. Components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.
E-5 Enclosure to NL-24-0259 Basis for Proposed Change
The leakage from the RHR System suction isolation valves will continue to be tested and verified in the same manner as before the proposed change. Thus, the proposed change will not affect the leakage limit required by TS 3.4.14.
Therefore, GDC 14 and 30 will continue to be met.
GDC 20 - Protection system functions. 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.
The RHR System OPI is not a protection system that is required to ensure that the specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences, nor is it required to respond to accident conditions and initiate the operation of systems and components important to safety. Revising the OPI interlock pressure acceptance criterion will not adversely affect the ability of the plant instrumentation and systems to assure that the specified acceptable fuel design limits are not exceeded and to respond to accident conditions and initiate the operation of systems and components important to safety.
Therefore, GDC 20 will continue to be met.
GDC 34 - Residual heat removal. A system to remove residual heat shall be provided.
The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.
Revising the RHR System OPI interlock pressure acceptance criterion does not adversely affect the capability of the RHR System to perform its intended safety function. The revision provides adequate margin to the low pressure RHR System design pressure of 600 psig. Thus, the RHR System integrity will not be affected by the revised acceptance criterion.
Therefore, GDC 34 continues to be met.
4.2 No Significant Hazards Consideration Determination
Southern Nuclear Operating Company (SNC) has evaluated the proposed changes to the Joseph M. Farley Nuclear Plant (FNP) Units 1 and 2 Technical Specifications (TS) 3.4.14, "Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage, " and TS 3.3.5, "Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation, " using the criteria in Section 50.92 to Title 10 of the Code of Federal Regulations (10 CFR) and has determined that the proposed changes do not involve a significant hazards consideration.
E-6 Enclosure to NL-24-0259 Basis for Proposed Change
The proposed amendment would revise TS requirements to:
(a) Delete TS 3.4.14 Condition C Note, Condition C text "autoclosure or, " and SR 3.4.14.2 (including renumbering SR 3.4.14.3 to SR 3.4.14.2);
(b) Revise Surveillance Requirement (SR) 3.4.14.3 (which is renumbered to SR 3.4.14.2) Residual Heat Removal System (RHR) open permissive interlock (OPI) interlock acceptance criterion; and
(c) Revise the SR 3.4.14.3 Note reference to SR 3.4.12.3 to instead reference SR 3.4.12.4.
(d) Delete TS 3.3.5 Limiting Condition for Operation (LCO) Notes,
(e) Delete TS 3.3.5 Conditions A and B Notes,
(f) Delete TS 3.3.5 Table 3.3.5-1 and references to it,
(g) Delete TS 3.3.5 Actions D, E, and F, and
(h) Renumber TS 3.3.5 Table 3.3.5-2 and references to it as Table 3.3.5-1.
As required by 10 CFR 50.91 (a), the SNC analysis of the issue of no significant hazards consideration is presented below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No
The two motor-operated gate valves located in each RHR System suction line are normally-closed to maintain the low pressure RHR System (design pressure of 600 psig) isolated from the high pressure RCS (normal operating pressure of 2235 psig). The RHR OPI function prevents inadvertent opening of the RHR System isolation valves when the RCS pressure is above the valve opening interlock, thereby protecting the RHR System from overpressure. The proposed change to the OPI interlock acceptance criterion continues to provide this protective function.
The other proposed changes reflect administrative and editorial changes that do not affect the performance of or analyses crediting any plant systems, subsystems, or components.
As such, the there is no adverse effect on the integrity of fission product barriers, plant configuration, and operating procedures as described in the FSAR by the proposed changes. Therefore, the consequences of previously analyzed accidents will not increase by implementing these changes. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
E-7 Enclosure to NL-24-0259 Basis for Proposed Change
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No
The proposed changes do not change a safety related function. Changes to surveillances will not result in the RHR system being operated in any unanalyzed modes, either during normal or accident conditions. The RHR system will continue to be maintained and surveilled as it is currently. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. The proposed change does not challenge the performance or integrity of any safety-related system. Therefore, operation of the facility in accordance with this proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed changes do not affect equipment margins, which will continue to be maintained in accordance with the plant-specific design bases. The proposed changes do not alter or prevent any plant response such that the margin of safety to any applicable acceptance criteria is decreased. The proposed changes will not adversely affect operation of plant equipment. The actuation of safety-related components and the response of plant systems to accident scenarios are not affected, and thus will remain as assumed in the safety analysis..
TS changes maintain the assumed level of equipment performance. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment does not involve a 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.
4.3 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.
E-8 Enclosure to NL-24-0259 Basis for Proposed Change
5 ENVIRONMENTAL CONSIDERATION
The proposed TS changes 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, and would change an inspection or surveillance requirement. However, 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 effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed TS change meets the eligibility criteria 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 needs to be prepared in connection with the proposed TS change.
6 REFERENCES
- 1. Joseph M. Farley Nuclear Plant, Units 1 and 2 - Issuance of Amendments to Revise Technical Specification 3.4.14 (CAC Nos. MF6687 and MF6688), dated May 17, 2016 (ML16083A265).
- 2. Joseph M. Farley Nuclear Plant, Units 1 and 2 - Issuance of Amendments Related to Technical Specification 3.3.5 (CAC Nos. MF7106 AND MF7107) (ML16196A161 ).
- 3. Joseph M. Farley Nuclear Plant, Units 1 And 2, Issuance Of Amendments Regarding Technical Specifications Revisions Associated With The Low Temperature Overpressure Protection System And The Pressure And Temperature Limits Report (TAC Nos.
ME9256 and ME9257) (NL-12-0868), dated October 2, 2013 (ML13249A386).
- 4. Joseph M. Farley Nuclear Plant, Response to Request for Additional Information Related to the Conversion to the Improved Technical Specifications - Chapter 3.4 (ML20203A758) (refer to response to NRC Question 27).
E-9 Joseph M. Farley Nuclear Plant - Units 1 and 2
License Amendment Request for Changes to Technical Specification 3.4.14, Reactor Coolant System Pressure Isolation Valve Leakage, Surveillance Requirement 3.4.14.3 Acceptance Criteria and Miscellaneous Obsolete Changes
Attachment 1
Technical Specification Marked Up Pages LOP DG Start Instrumentation 3.3.5
3.3 INSTRUMENTATION
3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation
LCO 3.3.5 The LOP instrumentation for each Function in Table 3.3.5-1 -aoo Table 3.3.5 2 shall be OPERABLE.
APPLICABILITY: According to Table 3.3.5-1 and Table 3.3.5 2.
- 1. For Unit 1, use Table 3.3.5 1 until Mode 4 entry folloi.ving the spring 2018 outage (1 R28);
thereafter use Table 3.3.5 2.
- 2. For Unit 2, use Table 3.3.5 1 until Mode 4 entry following the fall 2017 outage (2R25);
thereafter use Table 3.3.5 2.
ACTIONS
NOTE--------------------------------------------------------
Separate Condition entry is allowed for each Function.
CONDITION REQUIRED ACTION COMPLETION TIME
A. t:>JGTE A.1 -------------NOTE------------
Qnly appliGable to The inoperable channel FunGtions 1 and 2. may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance One or more functions testing of other channels.
with one channel per ----------------------------------
train inoperable. Place channel in trip. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
B. t:>JGTE B.1 Restore all but one 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Qnly appliGable to channel per train to FunGtions 1 and 2. OPERABLE status.
One or more Functions with two or more channels per train inoperable.
Farley Units 1 and 2 3.3.5-1 Amendment No. - ~ (Unit 1)
Amendment No. - ~ (Unit 2)
LOP DG Start Instrumentation 3.3.5
ACTIONS
cor:.mlTIOt:>J REQl:JIREQ ACTIOt:>J COMPbETIOt:>J TIME
C. Required Action and C.1 Enter applicable Immediately associated Condition(s) and Required Completion Time of Action(s) for the Condition A or B not associated DG made met. inoperable by LOP DG start instrumentation.
g_ t:>JOTE g_1 VeFify velta§e eA OAGe peF 4 A91:1FS ORiy appliGable te asseGiated b1:1s is ~ 3850 Fl:IAGtieA 3. -veUs-:
OAe AlaFFA Fl:IAGtieA GA8AAel iAepeFable 9A 9Ae 9F FA9Fe tFaiAS.
E. ReEjl:liFed,A,GtieA aAd E.1 ResteFe b1:1s velta§e te ~ 1 A91:1F asseGiated CeFApletieA 3850 1.ielts.
TiFAe ef CeAditieA g Aet FAet.
i;:_ ReEjl:liFed,A,GtieA aAd F.1 Be iA MOQE 3. 6 A91:1FS asseGiated CeFApletieA TiFAe ef CeAditieA E A-N9 Aet FAet.
i;:_~ Be iA MOQE 5. 36 A91:1FS
Farley Units 1 and 2 3.3.5-2 Amendment No. - ~ (Unit 1)
Amendment No. - ~ (Unit 2)
LOP DG Start Instrumentation 3.3.5
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
SR 3.3.5.1 ---------------------------NOTES--------------------------------
- 1. TADOT shall exclude actuation of the final trip actuation relay for LOP Functions 1 and 2.
- 2. Setpoint verification not required.
Perform TADOT. In accordance with the Surveillance Frequency Control Program
SR 3.3.5.2 -------- --- -------- --- ------- --NOTE --- ------ --- ------- ------ ----
C HANN EL CALIBRATION shall exclude actuation of the final trip actuation relay for Functions 1 and 2.
Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program
SR 3.3.5.3 -------- --- -------- --- ------- --Note- -- -------- --- ------- ------ ----
Response time testing shall include actuation of the final trip actuation relay.
Verify ESF RESPONSE TIME within limit. In accordance with the Surveillance Frequency Control Program
Farley Units 1 and 2 3.3.5-3 Amendment No. - ~ (Unit 1)
Amendment No. - ~ (Unit 2)
LOP DG Start Instrumentation 3.3.5
Table 3.3.5 1 (page 1 of 1)
Loss of Po>.ver Diesel Generator Start Instrumentation
i;LJ~JCTlmJ APPl::IC,11,Bl::e ReQldlReg SldRVell::l::A~JCe Al::l::GWABl::e T-Rl-P MQges GR CHA~J~Jel::S ReQldlReMe~JTS VAWe SeTPGl~JT GTHeR PeR TRAl~J SPeCIFleg cmmITImJs
- 1. 4.16 kV emergency B1,1s 1,2,3,4, (a) SR 3.3.5. 1 2 3222 V 2 3255 V boss of Voltage gG Start SR 3.3.5.2 aAa SR 3.3.5.3,_; 3418 V
2. 4.16 kV emergency B1,1s 1,2,3,4, (a) SR 3.3.5. 1 > 3638 V > 3675 V gegraded Grid Voltage SR 3.3.5.2 aAa Act1,1ation SR 3.3.5.3,_; 3749 V
3. 4.16 kV emergency B1,1s ~ 4 SR 3.3.5. 1 gegraged Grid Voltage SR 3.3.5.2 2 3835 V 2 3850 V AlaFm
(a) When associated gG is req1,1ired to be GPeRABl::e by l::CG 3.8.2, "AC So1,1rces Shbitdown."
Farley Units 1 and 2 3.3.5-4 Amendment No. _ 4-a (Unit 1)
Amendment No. - ~ (Unit 2)
LOP DG Start Instrumentation 3.3.5
Table 3.3.5-1 i (page 1 of 1)
Loss of Power Diesel Generator Start Instrumentation
FUNCTION APPLICABLE REQUIRED SURVEILLANCE ALLOWABLE DELAY MODES OR CHANNELS REQUIREMENTS VALUE TIME OTHER PER TRAIN SPECIFIED CONDITIONS
- 1. 4.16 kV Emergency Bus 1,2,3,4, (a) 3 SR 3.3.5.1 2'. 3222 V NA Loss of Voltage DG Start SR 3.3.5.2 and SR 3.3.5.3 :o:; 3418V
- 2. 4.16 kV Emergency Bus 1,2,3,4, (a) 3 SR 3.3.5.1 Bus 1F: 2'. 3761 V :,:; 11.4 sec Degraded Grid Voltage SR 3.3.5.2 Bus 1G: 2'. 3752 V :,:; 11.4 sec Actuation SR 3.3.5.3 Bus 2F: 2'. 3757 V :,:; 9.9 sec Bus 2G: 2'. 3778 V :,:; 9.9 sec
(a) When associated DG is required to be OPERABLE by LCO 3.8.2, "AC Sources - Shutdown."
Farley Units 1 and 2 3.3.5-5 Amendment No. - ~ (Unit 1)
Amendment No. - ~ (Unit 2)
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
A. ( continued) A.1 Isolate the high pressure 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> portion of the affected system from the low pressure portion by use of one closed manual, deactivated automatic, or check valve.
AND
A.2 Isolate the high pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> portion of the affected system from the low pressure portion by use of a second closed manual, deactivated automatic, or check valve.
B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time for Condition A not AND met.
B.2 ------------NOTE-----------
LCO 3.0.4.a is not applicable when entering MODE 4.
Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
t:>JG+E t:>Jet af}f}liGaele te tl=le a1:1teGles1:1Fe iAteFleGk feF ldAit
~ afteF FestaFt fFeFA ~ R~7 8A9 feF ldAit ~ afteF FestaFt fF9FA
~
C. RHR System C.1 Place the affected valve(s) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> a1:1teGles1:1Fe eF open in the closed position and permissive interlock maintain closed under function inoperable. administrative control.
Farley Units 1 and 2 3.4.14-2 Amendment No. - ~ (Unit 1)
Amendment No. _ 4W (Unit 2)
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
SR 3.4.14.1 -----------------------------NOTES-----------------------------
- 1. Not required to be performed in MODES 3 and 4.
- 2. Not required to be performed on the RCS PIVs located in the RHR flow path when in the shutdown cooling mode of operation.
- 3. RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.
Verify leakage from each RCS PIV is equivalent to 18 months, prior
- 0.5 gpm per nominal inch of valve size up to a to entering maximum of 5 gpm at an RCS pressure ~ 2215 psig MODE2 and
- :::; 2255 psig.
Following valve actuation due to automatic or manual action or flow through the valve
( except for RCS PIVs located in the RHR flow path)
- 1. Not required to be met 1Nhen the RHR System valves are required open in accordance 1Nith SR 3.4.12.3.
- 2. Not applicable to Unit 1 after restart from 1 R27 and not applicable to Unit 2 after restart from 2R25.
Verify RHR System autoclosure interlock In accordance 1Nith causes the valves to close automatically the Surveillance 1Nith a simulated or actual RCS pressure Frequency Control signal ~ 700 psig and :::; 750 psig. Program
Farley Units 1 and 2 3.4.14-3 Amendment No. - ~ (Unit 1)
Amendment No. _ 97-(Unit 2)
SURVEILLANCE REQUIREMENTS
SR 3.4.14. ~.f. -------------------------- NOTE----------------------------------
Not required to be met when the RHR System valves valves are required open in accordance with SR 3.4.12. J1.
Verify RHR System open permissive interlock In accordance with prevents the valves from being opened with a the Surveillance simulated or actual RCS pressure signal Frequency Control
~ 295 psig and :'.:o ~ 415 psig. Program
Farley Units 1 and 2 3.4.14-4 Amendment No. _ 4-Sa (Unit 1)
Amendment No. _ 4-00 (Unit 2)
Joseph M. Farley Nuclear Plant - Units 1 and 2
License Amendment Request for Changes to Technical Specification 3.4.14, Reactor Coolant System Pressure Isolation Valve Leakage, Surveillance Requirement 3.4.14.3 Acceptance Criteria and Miscellaneous Obsolete Changes
Attachment 2
Technical Specification Revised Pages LOP DG Start Instrumentation 3.3.5
3.3 INSTRUMENTATION
3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation
LCO 3.3.5 The LOP instrumentation for each Function in Table 3.3.5-1 shall be OPERABLE.
APPLICABILITY: According to Table 3.3.5-1.
ACTIONS
NOTE--------------------------------------------------------
Separate Condition entry is allowed for each Function.
CONDITION REQUIRED ACTION COMPLETION TIME
A. One or more functions A.1 -------------NOTE------------
with one channel per The inoperable channel train inoperable. may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels.
Place channel in trip. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
B. One or more Functions B.1 Restore all but one 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with two or more channel per train to channels per train OPERABLE status.
C. Required Action and C.1 Enter applicable Immediately associated Condition(s) and Required Completion Time of Action(s) for the Condition A or B not associated DG made met. inoperable by LOP DG start instrumentation.
Farley Units 1 and 2 3.3.5-1 Amendment No. _ (Unit 1)
Amendment No. _ (Unit 2)
LOP DG Start Instrumentation 3.3.5
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
SR 3.3.5.1 ---------------------------NOTES--------------------------------
- 1. TADOT shall exclude actuation of the final trip actuation relay for LOP Functions 1 and 2.
- 2. Setpoint verification not required.
Perform TADOT. In accordance with the Surveillance Frequency Control Program
SR 3.3.5.2 -------- --- -------- --- ------- --NOTE --- ------ --- ------- ------ ----
C HANN EL CALIBRATION shall exclude actuation of the final trip actuation relay for Functions 1 and 2.
Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program
SR 3.3.5.3 -------- --- -------- --- ------- --Note- -- -------- --- ------- ------ ----
Response time testing shall include actuation of the final trip actuation relay.
Verify ESF RESPONSE TIME within limit. In accordance with the Surveillance Frequency Control Program
Farley Units 1 and 2 3.3.5-2 Amendment No. _ (Unit 1)
Amendment No. _ (Unit 2)
LOP DG Start Instrumentation 3.3.5
Table 3.3.5-1 (page 1 of 1)
Loss of Power Diesel Generator Start Instrumentation
FUNCTION APPLICABLE REQUIRED SURVEILLANCE ALLOWABLE DELAY MODES OR CHANNELS REQUIREMENTS VALUE TIME OTHER PER TRAIN SPECIFIED CONDITIONS
- 1. 4.16 kV Emergency Bus 1,2,3,4, (a) 3 SR 3.3.5.1 2'. 3222 V NA Loss of Voltage DG Start SR 3.3.5.2 and SR 3.3.5.3 :o:; 3418V
- 2. 4.16 kV Emergency Bus 1,2,3,4, (a) 3 SR 3.3.5.1 Bus 1F: 2'. 3761 V :,:; 11.4 sec Degraded Grid Voltage SR 3.3.5.2 Bus 1G: 2'. 3752 V :,:; 11.4 sec Actuation SR 3.3.5.3 Bus 2F: 2'. 3757 V :,:; 9.9 sec Bus 2G: 2'. 3778 V :,:; 9.9 sec
(a) When associated DG is required to be OPERABLE by LCO 3.8.2, "AC Sources - Shutdown."
Farley Units 1 and 2 3.3.5-3 Amendment No. _ (Unit 1)
Amendment No. _ (Unit 2)
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
A. ( continued) A.1 Isolate the high pressure 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> portion of the affected system from the low pressure portion by use of one closed manual, deactivated automatic, or check valve.
AND
A.2 Isolate the high pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> portion of the affected system from the low pressure portion by use of a second closed manual, deactivated automatic, or check valve.
B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time for Condition A not AND met.
B.2 ------------NOTE-----------
LCO 3.0.4.a is not applicable when entering MODE 4.
Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
C. RHR System open C.1 Place the affected valve(s) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> permissive interlock in the closed position and function inoperable. maintain closed under administrative control.
Farley Units 1 and 2 3.4.14-2 Amendment No. (Unit 1)
Amendment No. (Unit 2)
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
SR 3.4.14.1 -----------------------------NOTES-----------------------------
- 1. Not required to be performed in MODES 3 and 4.
- 2. Not required to be performed on the RCS PIVs located in the RHR flow path when in the shutdown cooling mode of operation.
- 3. RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.
Verify leakage from each RCS PIV is equivalent to 18 months, prior
- 0.5 gpm per nominal inch of valve size up to a to entering maximum of 5 gpm at an RCS pressure ~ 2215 psig MODE2 and
- :::; 2255 psig.
Following valve actuation due to automatic or manual action or flow through the valve
( except for RCS PIVs located in the RHR flow path)
SR 3.4.14.2 -------------------------- NOTE----------------------------------
Not required to be met when the RHR System valves are required open in accordance with SR 3.4.12.4.
Verify RHR System open permissive interlock In accordance with prevents the valves from being opened with a the Surveillance simulated or actual RCS pressure signal ~ 415 psig. Frequency Control Program
Farley Units 1 and 2 3.4.14-3 Amendment No. (Unit 1)
Amendment No. (Unit 2)
Joseph M. Farley Nuclear Plant - Units 1 and 2
License Amendment Request for Changes to Technical Specification 3.4.14, Reactor Coolant System Pressure Isolation Valve Leakage, Surveillance Requirement 3.4.14.3 Acceptance Criteria and Miscellaneous Obsolete Changes
Attachment 3
Technical Specification Bases Marked Up Pages (For Information Only)
LOP DG Start Instrumentation B 3.3.5
BASES
APPLICABLE actual DG start has historically been associated with the ESFAS SAFETY ANALYSES actuation. The DG loading is included in the delay time associated
( continued) with each safety system component requiring DG supplied power following a loss of offsite power.
Monitoring by the offsite power system grid operators and the first level LOP instrumentation (alarm) provide the primary protection for a degraded grid event. The degraded grid voltage alarm provides notification to control room operators that an abnormally low voltage condition exists on a 4.16 kV emergency bus. For slow acting transient conditions, the alarm setpoint allows for the initiation of manual actions by the offsite power system operator to restore normal bus voltage and protect required ESF LOCA loads from the low voltage condition without initiating an unnecessary automatic disconnect from the preferred offsite power source.
For the 4.16 kV emergency buses to 1.vhich Technical Specification Table 3.3.5 1 is applicable, an administrative limit is established at a 1o<oltage le 1o<el between the degraded grid 1o<oltage alarm allowable value (3835V) and the automatic degraded grid voltage actuation upper allmvable value (3749V). Calculations verify that no ESF components require a 4.16kV bus voltage higher than the administrative limit to perform their safety functions. In the voltage range between the administrative limit and the degraded grid voltage actuation trip setpoint, a few ESF components may not have automatic protection from inadequate voltage. The manual actions provide the primary means of protecting these few ESF components from a sustained, slightly low voltage condition and all components from unnecessary automatic disconnection from the preferred offsite power source.
For the 4.16 kV emergency buse£ to which Technical Specification Table 3.3.5 2 is applicable, an analytical limit is established for each bus at a voltage level below the automatic degraded grid voltage actuation allowable value shown in Table 3.3.5-1 ~ - Calculations verify that no ESF components require a 4.16 kV bus voltage higher than the analytical limit to perform their safety functions.
The required channels of LOP DG start instrumentation, in conjunction with the ESF systems powered from the DGs, provide unit protection in the event of any of the analyzed accidents discussed in FSAR, Section 15 (Ref. 2), in which a loss of offsite power is assumed.
(continued)
Farley Units 1 and 2 B 3.3.5-3 Revision - ~
LOP DG Start Instrumentation B 3.3.5
BASES
APPLICABLE The delay times assumed in the safety analysis for the ESF SAFETY ANALYSES equipment bound the 12 second DG start delay and include the
( continued) appropriate sequencing delay, if applicable. The response times for ESFAS actuated equipment in LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation," include the appropriate DG loading and sequencing delay.
The LOP DG start instrumentation channels satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO The LCO for LOP DG start instrumentation requires that three channels per train of both the loss of voltage and degraded grid voltage actuation Functions shall be OPERABLE in MODES 1, 2, 3, and 4 when the LOP DG start instrumentation supports safety systems associated with the ESFAS. In MODES 5 and 6, the three channels must be OPERABLE whenever the associated DG is required to be OPERABLE to ensure that the automatic start of the DG is available when needed. Loss of the LOP DG Start Instrumentation Function could result in the delay of safety systems initiation when required. This could lead to unacceptable consequences during accidents. During the loss of offsite power the DG powers the motor driven auxiliary feedwater pumps. Failure of these pumps to start would leave only one turbine driven pump, as well as an increased potential for a loss of decay heat removal through the secondary system.
In addition, the LCO requires one channel of the degraded grid alarm function per train of 4.16 kV emergency buses to be OPERABLE in MODES 1, 2, 3, and 4. The required alarm channels include the Digital Voltmeter Relay Contacts (LO 27V) on buses P: and G and the associated alarm annunciators WE2, VE2 (Unit 1) and YE2, ZE2 (Unit 2). The alarm channels provide assurance that manual actions are taken to restore bus voltage and protect the required ESP: LOCA loads from a degraded grid voltage condition.
APPLICABILITY The LOP DG Start Instrumentation Functions are required in MODES 1, 2, 3, and 4 because ESF Functions are designed to provide protection in these MODES. Actuation in MODE 5 or 6 is required whenever the required DG must be OPERABLE so that it can perform its function on an LOP or degraded power to the vital bus.
(continued)
Farley Units 1 and 2 B 3.3.5-4 Revision - ~
LOP DG Start Instrumentation B 3.3.5
BASES
/\\PPLIC/\\BILITY P:or the 4.16 kV emergency buses to which Technical Specification (continued) Table 3.3.5 1 is applicable, the degraded grid alarm function is required OPERABLE in MODES 1, 2, 3, and 4 to support the voltage requirements of the ESP: loads required OPERABLE to mitigate a design basis LOCA. In MODES 5 and 6, the degraded grid alarm function is not required OPERABLE as no design basis LOCA is assumed to occur in these MODES and most of the ESP: loads required to mitigate a design basis LOCA are not required OPERABLE.
P:or the 4.16 kV emergency buses to 1.vhich Technical Specification Table 3.3.5 2 is applicable, the degraded grid alarm is not a function included in the Technical Specifications.
ACTIONS In the event a channel's Alarm or Trip Setpoint is found nonconservative with respect to the Allowable Value, or the channel is found inoperable, then the function that channel provides must be declared inoperable and the LCO Condition entered for the particular protection function affected.
Because the required channels are specified on a per train basis, the Condition may be entered separately for each train as appropriate.
A Note has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in the LCO. The Completion Time(s) of the inoperable channel(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.
A.1
Condition A applies to the LOP DG start Functions (Functions 1 and
- 2) with one loss of voltage or degraded grid voltage channel per train inoperable.
If one channel is inoperable, Required Action A.1 requires that channel to be placed in trip within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. With a channel in trip, the remaining LOP DG start instrumentation channels will provide a one-out-of-two logic to initiate a trip of the incoming offsite power.
A Note is added to Condition A indicating that it is only applicable to P:unctions 1 and 2.
(continued)
Farley Units 1 and 2 B 3.3.5-5 Revision - ~
LOP DG Start Instrumentation B 3.3.5
BASES
ACTIONS A.1 (continued)
A Note is added to allow bypassing an inoperable channel for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels. This allowance is made where bypassing the channel does not cause an actuation and where at least two other channels are monitoring that parameter.
The specified Completion Time and time allowed for bypassing one channel are reasonable considering the Function remains fully OPERABLE on each train and the low probability of an event occurring during these intervals.
B.1
Condition B applies to LOP Functions 1 and 2 when two or more loss of voltage or degraded voltage channels on a single train are inoperable.
A ~Jote is added to Condition B indicating that it is only applicable to Functions 1 and 2.
Required Action B.1 requires restoring all but one channel on a train to OPERABLE status. With a single inoperable channel remaining on a train, Condition A is applicable. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time should allow ample time to repair most failures and takes into account the low probability of an event requiring an LOP start occurring during this interval.
C.1
Condition C applies to each of the LOP DG start Functions when the Required Action and associated Completion Time for Condition A or B are not met.
In these circumstances the Conditions specified in LCO 3.8.1, "AC Sources - Operating," or LCO 3.8.2, "AC Sources - Shutdown," for the DG made inoperable by failure of the LOP DG start instrumentation are required to be entered immediately. The actions of those LCOs provide for adequate compensatory actions to assure unit safety.
(continued)
Farley Units 1 and 2 B 3.3.5-6 Revision - ~
LOP DG Start Instrumentation B 3.3.5
BASES
/\\CTIONS D.1 (continued)
Condition D applies i.vhen the required degraded grid voltage alarm function is inoperable on one or both trains of emergency buses. The affected bus voltage associated with each inoperable alarm function must be verified ~ 3850 volts every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Frequent bus voltage 1,erifications in lieu of an OPERABU: alarm effectively accomplish the same function as the alarm and allow operation to continue without the required alarm(s). A Note is added to Condition D indicating that it is only applicable to Function 3.
Condition E is applicable when the Required Action and associated Completion Time of Condition D is not met. If the voltage being verified per Required Action D.1 is< 3850 volts, action must be taken to restore the voltage to ~ 3850 volts within one hour. The Completion Time of one hour is reasonable to ensure prompt action is taken to restore adequate voltage to the affected emergency bus(es).
F.1 and F.2
Condition F becomes applicable when the Required Action and associated Completion Time of Condition E is not met. If the emergency bus 11oltage cannot be restored to ~ 3850 11olts within the Completion Time of Condition E, action must be taken to place the unit in a MODE where the LCO requirement for the Alarm function is not applicable. 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 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.3.5.1 REQUIREMENTS SR 3.3.5.1 is the performance of a TADOT. The test checks trip devices that provide actuation signals directly, bypassing the analog process control equipment.
The TADOT surveillance is modified by two Notes. The first Note excludes the actuation of the final trip actuation relay for LOP Functions 1 and 2 from this TADOT. The actuation of this relay would
(continued)
Farley Units 1 and 2 B 3.3.5-7 Revision - ~
BASES
ACTIONS B.1 and B.2 (continued)
Required Action B.2 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met.
However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit._ The allowed Completion Times are reasonable based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
C.1
The inoperability of the RHR autoclosure interlock renders the associated RHR suction isolation valves incapable of isolating in response to a high pressure condition. The inoperability of the RHR open permissive interlock renders the associated RHR suction isolation valves incapable of preventing inadvertent opening of the valves at RCS pressures in excess of the RHR systems design pressure. If the RHR autoclosure or open permissive interlocks are inoperable, operation may continue as long as the affected RHR suction valves are closed and administrative controls are in place in the control room to maintain them closed (e.g., tags on the main control board handswitches, etc.) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This Action accomplishes the purpose of the autoclosure or open permissive interlock function.
Required Action C.1 is modified by a Note that states the Required Action for the autoclosure interlock is not applicable to Unit 1 after 1 R27 and not applicable to Unit 2 after 2R25. The Required Action for the autoclosure interlock is no longer applicable after these refueling outages because the autoclosure interlock 1Nill be removed during the outages and 1Nill no longer be required OPERABLE.
Note to Operators: After 1 R27 (Unit 1) and 2R25 (Unit 2) 1.yhen the RHR autoclosure interlocks are removed from each unit, The administrative controls for the RHR suction isolation valves will be requires them to be closed with power removed from the valves in MODES 1, 2, and 3. The requirement to isolate the valves with power removed in these MODES is necessary to satisfy the conditions for removal of the RHR autoclosure interlock made in Amendment Nos.
201 (Unit 1) and 197 (Unit 2).
(continued)
Farley Units 1 and 2 B 3.4.14-5 Revision _ ~
BASES
ACTIONS C.1 (continued)
If the open permissive interlock becomes inoperable after the removal of the autoclosure interlock, the Required Action to ensure the valves are closed using the administrative controls described above would only be applicable in MODE 4. In MODES 1, 2, and 3, if an open permissive interlock becomes inoperable, the Required Action to close and maintain close the valves by administrative controls would be met by the administrative controls in place to ensure the valves are closed with power removed (as required for the removal of the autoclosure interlock).
SURVEILLANCE SR 3.4.14.1 REQUIREMENTS Performance of leakage testing on each RCS PIV or isolation valve used to satisfy Required Action A.1 and Required Action A.2 is required to verify that leakage is below the specified limit and to identify each leaking valve. However, the valves used to isolate the flow path to satisfy Required Actions A.1 and A.2 (which are not PIVs) do not have to be pre-qualified by periodic testing. When Required Action A is entered and the flow path isolated, the valves will be verified at that time to meet the leakage requirements of SR 3.4.14.1.
This is accomplished using the methodology of SR 3.4.13.1 (RCS water inventory balance) with the leakage limits of SR 3.4.14.1 applied. The leakage limit of 0.5 gpm per inch of nominal valve diameter up to a 3 or 5 gpm maximum applies to each valve.
Leakage testing requires a stable pressure condition.
For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost.
Testing is to be performed every 18 months, a typical refueling cycle, on all PIVs listed in the TRM. The 18 month Frequency is consistent with 10 CFR 50.55a(g) (Ref. 8) as contained in the INSERVICE TESTING PROGRAM, is within frequency allowed by the American Society of Mechanical Engineers (ASME) OM Code (Ref. 7), and is based on the need to perform such surveillances under the conditions that apply during an outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
(continued)
Farley Units 1 and 2 B 3.4.14-6 Revision _ 84 RCS PIV Leakage B 3.4.14
BASES
SURVEILLANCE SR 3.4.14.2 REQUIREMENTS
( continued) Verifying that the RHR autoclosure interlock is OPERABLE ensures that RCS pressure 1Nill not pressurize the RHR system beyond 125%
of its design pressure of 600 psig. The autoclosure interlock isolates the RHR System from the RCS 1.yhen the interlock setpoint is reached.
The setpoint ensures the RHR design pressure 1Nill not be exceeded.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
The SR is modified by ti.vo Notes. Note 1 provides an exception to the requirement to perform this surveillance when using the RHR System suction relief valves for cold overpressure protection in accordance with SR 3.4.12.3.
~Jote 2 states the Surveillance is not applicable to Unit 1 after 1 R27 and not applicable to Unit 2 after 2R25. The Surveillance is no longer applicable after these refueling outages because the autoclosure interlock will be remo 1,ed during the outages and will no longer be required OPERABLE.
Verifying that the RHR open permissive interlock is OPERABLE ensures that the RCS will not pressurize the RHR system beyond design of 600 psig. The open permissive interlock prevents opening the RHR System suction valves from the RCS when the RCS pressure is above the setpoint. The setpoint upper interlock value ensures the RHR System design pressure will not be exceeded at the RHR pump discharge and was chosen taking into account instrument uncertainty and calibration tolerances. This value also provides assurance that the RHR System suction relief valves setpoint will not be exceeded.
The minimum value of the setpoint range is chosen based upon operational considerations (differential pressure) for the RCP seals and thus does not have a safety related function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
The SR is modified by a Note that provides an exception to the requirement to perform this surveillance when using the RHR System suction relief valves for cold overpressure protection in accordance with SR 3.4.12.~J.
(continued)
Farley Units 1 and 2 B 3.4.14-8 Revision - ~