ML041540457

License Amendment Request, Activation of the Trip Outputs of the Oscillation Power Range Monitor Portion of the Power Range Neutron Monitoring System

ML041540457
Person / Time
Site:	Limerick
Issue date:	05/20/2004
From:	Jury K Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML041540457 (198)

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Exelon Nuclear www.exeloncorp.com Nuclear 200 Exelon Way Kennett Square, PA 19348 10 CFR 50.90 May 20, 2004 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Limerick Generating Station, Units 1 and 2 Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353

Subject:

License Amendment Request Activation of the Trip Outputs of the Oscillation Power Range Monitor Portion of the Power Range Neutron Monitoring System

References:

(1)

Generic Letter 94-02, "Long-Term Solutions and Upgrade of Interim Operating Recommendations for Thermal Hydraulic Instabilities in Boiling Water Reactors," dated July 11, 1994 (2)

NEDC-3241OP-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option IlIl Stability Trip Function," dated October 1995 (3)

NEDC-3241OP-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option IlIl Stability Trip Function," Supplement 1, dated November 1997 (4)

Letter from Keith R. Jury (Exelon Generation Company, LLC) to U. S.

NRC, "Schedule for Completing Actions to Implement Long-Term Stability Solution," dated December 19, 2003 Pursuant to 10 CFR 50.90, "Application for amendment of license or construction permit,"

Exelon Generation Company, LLC (Exelon), hereby requests changes to the Technical Specifications (TS), Appendix A, of Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2, respectively.

The proposed changes to TS support the activation of the trip outputs of the previously-installed Oscillation Power Range Monitor (OPRM) portion of the Power Range Neutron Monitoring (PRNM) system. Specifically, these proposed changes will revise TS Sections 2.2.1, "Reactor Protection System Instrumentation Setpoints," 3/4.3.1, "Reactor Protection System Instrumentation," 3/4.3.6, "Control Rod Block Instrumentation," and 3/4.4.1, "Recirculation System" and their associated TS Bases and TS Section 6.9.1, "Routine Reports."

In addition, the proposed changes delete interim corrective action requirements from the Recirculation System TS.

LGS Activation of the Trip Outputs of the OPRM Portion of the PRNM System May 20, 2004 Page 2 The current plant design depends on manual operator action to avoid reactor operating regions where thermal-hydraulic instability may occur, to exit such operating regions when necessary, and to detect an actual thermal-hydraulic instability and take mitigating action by manual means.

The proposed changes replace these procedural actions with the long-term stability solution required by Generic Letter 94-02 (Reference 1) The PRNM hardware incorporates the Option IlIl detect and suppress solution (i.e., the OPRM Upscale Function) reviewed and approved by the NRC. The OPRM Upscale Function includes sophisticated algorithms that can automatically detect an instability condition and provide a reactor protection system trip input if the oscillation magnitude exceeds acceptable limits. The OPRM Upscale Function meets the requirements of 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 10, "Reactor design," and GDC 12, "Suppression of reactor power oscillations," by automatically detecting and suppressing design basis thermal-hydraulic oscillations prior to exceeding the Minimum Critical Power Ratio Safety Limit.

Exelon has been following the industry approach for implementation/activation of the OPRM Upscale Function in accordance with NRC approved Licensing Topical Reports (LTRs). In addition, Exelon has been reviewing and incorporating industry operating experience, as appropriate, regarding OPRM Upscale Function activations at other utilities. License amendments to support the activation of the trip outputs of the OPRM portion of the PRNM system have been approved by the NRC for Hatch Units 1 and 2, Browns Ferry Unit 2, Browns Ferry Unit 3, Nine Mile Point Unit 2, Fermi Unit 2 and Brunswick Units 1 and 2.

Information supporting this License Amendment Request (LAR) is contained in Attachment 1 to this letter. The proposed changes to the LGS Unit 1 and Unit 2 TS are contained in Attachment 2 (marked-up TS pages) and Attachment 3 (typed TS pages).

In addition, Attachment 4 provides plant-specific responses required by the generic NRC approved General Electric Nuclear Measurement Analysis and Control (NUMAC) PRNM LTR NEDC-3241 OP-A, including Supplement 1 (References 2 and 3). Attachment 4 also provides descriptions and justifications for each deviation from the NUMAC PRNM LTRs, and includes a discussion of changes not addressed in the LTRs.

Exelon has concluded that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c).

Exelon requests approval of the proposed amendments by May, 2005. Once approved, the amendments shall be implemented within 60 days. However, as indicated in the schedule provided in Reference 4, the OPRM portion of the PRNM system will be declared operational on both Units 1 and 2 following NRC approval of this license amendment request, but no earlier than 30 days following the end of the next refueling outage for LGS, Unit 2, currently scheduled to be completed by the end of March, 2005.

There are no commitments contained within this letter.

The proposed changes have been reviewed by the Plant Operations Review Committee and approved by the Nuclear Safety Review Board.

LGS Activation of the Trip Outputs of the OPRM Portion of the PRNM System May 20, 2004 Page 3 We are notifying the State of Pennsylvania of this application for changes to the TS by transmitting a copy of this letter and its attachments to the designated State Official.

If you have any questions or require additional information, please contact Glenn Stewart at 610-765-5529.

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

Respectfully, Executed on & i Lo, M L(S {.

V4 Keith Jury Director, Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: 1. Evaluation of the Proposed Changes

2. Technical Specifications and Bases Markup Pages

3. Technical Specifications and Bases Typed Pages

4. Plant-Specific Responses Required by Licensing Topical Report NEDC-3241 OP-A cc:

Regional Administrator - NRC Region I w/attachments NRC Senior Resident Inspector - Limerick Generating Station NRC Project Manager, NRR - Limerick Generating Station Director, Bureau of Radiation Protection - Pennsylvania Department of Environmental Protection

ATTACHMENT 1 LICENSE AMENDMENT REQUEST LIMERICK GENERATING STATION, UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 EVALUATION OF THE PROPOSED CHANGES

SUBJECT:

Activation of the Trip Outputs of the Oscillation Power Range Monitor Portion of the Power Range Neutron Monitoring System

2.0 PROPOSED CHANGE

S

7.0 REFERENCES

Evaluation of Proposed Changes LGS OPRM Trip Activation LAR Page 1 of 15

1.0 DESCRIPTION

In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company, LLC (Exelon), requests changes to the Technical Specifications (TS), Appendix A, of Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2, respectively.

The proposed changes to TS support the activation of the trip outputs of the previously-installed Oscillation Power Range Monitor (OPRM) portion of the Power Range Neutron Monitoring (PRNM) system. Concurrent with the OPRM TS changes, additional TS and TS Bases will be changed to address some issues that have arisen during the time since the installation of the PRNM system.

Exelon has been following the industry approach for implementation/activation of the OPRM Upscale Function in accordance with NRC approved Licensing Topical Reports. In addition, Exelon has been reviewing and incorporating industry operating experience, as appropriate, regarding OPRM Upscale Function activations at other utilities.

As indicated in the schedule provided in Reference 1, the OPRM portion of the PRNM system will be declared operational on both Units 1 and 2 following NRC approval of this license amendment request, but no earlier than 30 days following the end of the next refueling outage for LGS, Unit 2, currently scheduled to be completed by the end of March, 2005.

The proposed changes to the LGS Unit 1 and Unit 2 TS are contained in Attachment 2 (marked-up TS pages) and Attachment 3 (typed TS pages). In addition, Attachment 4 provides plant-specific responses required by the generic NRC approved General Electric (GE) Nuclear Measurement Analysis and Control (NUMAC) PRNM Licensing Topical Report (LTR) NEDC-3241 OP-A (including Supplement 1), "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," (References 2 and 3). Attachment 4 also provides descriptions and justifications for each deviation from the NUMAC PRNM LTRs, and includes a discussion of changes not addressed in the LTRs.

This Attachment 1 provides a discussion and description of the proposed TS changes, a technical analysis, a regulatory analysis, including information supporting a finding of No Significant Hazards Consideration, and information supporting an Environmental Assessment.

2.0 PROPOSED CHANGE

S The proposed changes will revise TS Sections 2.2.1, "Reactor Protection System Instrumentation Setpoints," 3/4.3.1, "Reactor Protection System Instrumentation," 3/4.3.6, "Control Rod Block Instrumentation," and 3/4.4.1, "Recirculation System" and their associated TS Bases and TS Section 6.9.1, "Routine Reports."

In addition, the proposed changes delete the Interim Corrective Action (ICA) requirements from the Recirculation System TS. A detailed description of the proposed TS changes is provided below.

Evaluation of Proposed Changes LGS OPRM Trip Activation LAR Page 2 of 15 Technical Specifications:

A.

Setpoints and Allowable Values, TS Section 2.2.1 There are no allowable values associated with the OPRM Upscale Function. The OPRM Upscale trip period based detection algorithm (PBDA) setpoints are determined based on the Option Ill licensing methodology developed by the Boiling Water Reactor Owner's Group (BWROG) and described in NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," (Reference

4) previously approved by the NRC. Note "****" will be added to Table 2.2.1-1 for the OPRM Upscale Function to state that the PBDA trip setpoints are defined in the Core Operating Limits Report (COLR).

There are also TS related setpoints for the auto-enable (not-bypassed) region, which are established as nominal setpoints only. These values are also included in the new Note

"****" to Table 2.2.1-1 and are further described in the TS Bases markup.

The minimum operable OPRM cells setpoint (23 cells per channel) is defined by GE analyses based on LGS's selection of the OPRM cell assignments and a minimum of 2 operable Local Power Range Monitors (LPRMs) per cell. The setpoint is established to conform to the licensing bases defined in NEDO-31960-A (including Supplement 1),

"BWR Owner's Group Long-Term Stability Solutions Licensing Methodology,"

(References 5 and 6) and NEDO-32465-A. This setpoint, along with the LGS selection of a minimum of 2 operable LPRMs per cell, is documented in a new Note l(p)" to Table 3.3.1-1 and in the TS Bases as part of the operability requirement for Function 2.f.

The PBDA includes several "tuning" parameters. These are established in accordance with LGS procedures as part of the system setup, and are not defined in TS.

Finally, there are also setpoints for the "defense-in-depth" algorithms discussed in the OPRM Upscale Function description in the TS Bases markup. These are treated as nominal setpoints and controlled by LGS procedures, as described in the TS Bases markup.

A minor non-OPRM related change is being made to Table 2.2.1-1 to show the Single Loop Operation (SLO) equation in the form 0.66(W-AW) + offset value, with the offset value the same for both SLO and Two Loop Operation (TLO). Currently, the equations are shown in the form 0.66W +offset value, with 5% difference in offset values for SLO vs. TLO. In the reformatted representation, AW equals zero for TLO and 7.6% for SLO (7.6% = 5%/0.66). The revised representation, while mathematically equivalent, states the equation in the same form that is actually implemented in the equipment. In addition, a notation has been added to Table 2.2.1-1 addressing the limits of application of the flow offset. No change related to SLO is required for the OPRM Upscale Function implementation. However, the form of this setpoint expression is being modified to address a TS concern identified during the system's installation at LGS. The concern involves the system's miscalculation of this equation when indicated recirculation drive flow (W) becomes less than AW. Although this flow condition is not operationally possible, proper description of the system's calculation of the single loop setpoint was deemed to be warranted.

Evaluation of Proposed Changes LGS OPRM Trip Activation LAR Page 3 of 15 B.

Reactor Protection System (RPS) Instrumentation, TS Section 3/4.3.1, APRM Related Functions B.1 Functions This modification has no impact on any of the existing PRNM functions. The OPRM monitoring function is currently installed and fully functional but is not connected to the associated Reactor Protection System (RPS) or trip annunciator circuitry. The only change in this modification is connecting the existing OPRM trip outputs in series with the Average Power Range Monitor (APRM) trip outputs. This series output configuration produces a logical "OR" relationship between the OPRM trip outputs and the existing APRM trip outputs to RPS.

A new OPRM Upscale Function 2.f will be added to Table 3.3.1-1.

B.2 Minimum Number of Operable OPRM Channels The required minimum number of operable OPRM channels will be three channels.

The OPRM Upscale Function will have operability requirements associated with OPRM cells of a minimum of 2 operable LPRMs per cell for a cell to be operable and a minimum of 23 OPRM cells per OPRM channel for channel operability. The specific numerical values for these two parameters are identified as "plant specific" in the NUMAC PRNM LTRs.

B.3 Applicable Operational Conditions The new OPRM Upscale Function is safety-related and will be required to be operable only with reactor power 2 25% of Rated Thermal Power (RTP).

B.4 Channel Check Surveillance Requirements The new OPRM Upscale Function will have a Channel Check requirement with a frequency of "D" (once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

B.5 Channel Functional Test Surveillance Requirements The new OPRM Upscale Function will have a Channel Functional Test requirement with a frequency of "SA" (184 days). The Channel Functional Test requirement includes both the OPRM channels and the 2-Out-Of-4 Voter channels plus the flow input function, excluding the flow transmitters. Note "e" will be referenced for Function 2.f in Table 4.3.1.1-1 to show that the flow input function, except for the flow transmitter, is also included in the Function 2.f Surveillance Requirement. The NUMAC PRNM LTR Supplement 1 includes this requirement only for the APRM Simulated Thermal Power function, but it has been included for the OPRM Upscale function for LGS since that function uses flow for the auto-enable function.

B.6 Channel Calibration Surveillance Requirements The new OPRM Upscale Function will have a Channel Calibration requirement with a frequency of "R" (24 months). A separate LPRM calibration requirement (Table 4.3.1.1-

Evaluation of Proposed Changes LGS OPRM Trip Activation LAR Page 4 of 15 1, Note f) with a frequency of every 1000 effective full power hours applied at the "Function 2" (APRM) level will also apply to the OPRM Upscale Function. Table 4.3.1.1-1, Note g, will also be applied to the OPRM Upscale Function to require Channel Calibration of the recirculation loop flow input function, the same as the current requirement for APRM Simulated Thermal Power - Upscale Function. The NUMAC PRNM LTR Supplement 1 shows Note g only against the APRM Simulated Thermal Power - Upscale Function, but since the flow is used in the OPRM trip auto-enable function, Note g should also apply to the OPRM Upscale Function. (See item B.9 below for related discussion of addition of Note c.)

B.7 Response Time Testing Surveillance Requirements The new OPRM Upscale Function will have no Response Time Testing Surveillance Requirement. However, the response time testing for the 2-Out-Of-4 Voter including the output relays to RPS must be modified to account for the OPRM Upscale Function outputs. This is accomplished by revising Note `"h to Table 3.3.1-2 to redefine "N", and by additional discussion in TS Bases 3/4.3.1 to clarify the sequence of testing. This testing is consistent with the sequencing described in the NUMAC PRNM LTR Supplement 1. Specifically, the testing procedure for the 2-Out-Of-4 Voter function will alternate testing of the voter OPRM output with the voter APRM output except the net testing rate for the components is twice the rate required by the LTR. This testing rate was submitted as part of the LGS PRNM modification license amendment request to the NRC in a letter dated February 11, 2000 (Reference 21), and approved by the NRC in a Safety Evaluation Report (SER) dated April 12, 2000 (Reference 22).

B.8 Logic System Functional Testing (LSFT) Surveillance Requirements The new OPRM Upscale Function will have no LSFT Surveillance Requirement.

However, the SR 4.3.1.2 requirements applicable to the 2-Out-Of-4 Voter, Function 2.e will be modified slightly to add "OPRM" to show that the simulated trip conditions must include the OPRM logic as well as the APRM Upscale/lnop logic. This clarification is required because the 2-Out-Of-4 Voter, Function 2.e, votes the OPRM Upscale trip independently from the APRM Upscale/lnop trip. The TS Bases 3/4.3.1 description for Function 2.e will be modified to document the independent voting of the OPRM and APRM trips. TS Bases description additions will clarify that the 2-Out-Of-4 Voter Function does not need to be declared inoperable if portions of the Two-Out-Of-Four Logic Module hardware that are not part of the 2-Out-Of-4 Voter are found to be inoperable. The proposed TS Bases wording for Function 2.e is somewhat different from, but consistent with the intent of, the NUMAC PRNM LTRs.

B.9 Verify OPRM Auto-Enable Setpoints The new OPRM Upscale Function will have a new Surveillance Requirement, Note c to Table 4.3.1.1-1, to confirm, with a frequency of "R" (24 months), that the OPRM auto-enable setpoints are correctly set. This addition is consistent with the NUMAC PRNM LTR Supplement 1 except for some minor rewording of the notation to improve clarity.

B.10 LCO Conditions and Actions LCO 3.3.1 Actions a and b apply to the OPRM Upscale Function 2.f the same as for the APRM Functions 2.a, 2.b., 2.c and 2.d. Action c does not apply to Function 2.f.

Evaluation of Proposed Changes LGS OPRM Trip Activation LAR Page 5 of 15 Therefore, the "Notes" for Actions a, b and c will be modified to add "2.f," consistent with the NUMAC PRNM LTR Supplement 1.

New Action Statement 10, which includes sub-actions 1 Qa and 1 Ob, will be defined for Table 3.3.1-1. The new Actions apply to the OPRM Upscale Function when the required completion times for LCO 3.3.1 Action a or b are not met, or when the Function is not available due to fewer than two operable OPRM channels. Action Statement 1 Oa applies when the Function is not available due to an unanticipated characteristic of the instability detection algorithm or equipment that would render all OPRM channels inoperable, and allows a completion time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to initiate alternate methods of detecting and suppressing instabilities, and a completion time of 120 days to restore the OPRM operability. Action Statement 1 Ob applies for conditions other than those specified in Action Statement 1 Oa, or when the allowable completion times for Action Statement 1 Oa are not met. Action Statement 1 Ob will allow 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to reduce thermal power to less than 25% RTP.

The alternate method for detection and suppression required by Action Statement 1 Oa is intended to be temporary re-establishment of the ICAs, but controlled by plant procedures rather than TS.

Action Statements 1 Oa and 1 Ob are consistent with the intent of the Actions as described in the Improved Technical Specifications Bases section of the NUMAC PRNM LTR.

B.1 1 To more clearly state the TS 3.3.1 Action requirements and application of the notations provisions, the existing Note (n) to Table 3.3.1-1 is being moved as a new footnote "***"

to the Action b. and Action c. for TS 3.3.1. The movement of the notation does not make any technical change to the TS requirements, but more clearly relates the note to the applicable Action statements. This change is not related to OPRM and is not covered in the NUMAC PRNM LTRs, but is being made concurrently to eliminate ambiguity in the intent and application of the note. No TS Bases changes are involved in this TS change.

C.

Control Rod Block Instrumentation, TS Section 3/4.3.6 C.1 A minor non-OPRM related change is being made to Table 3.3.6-2 to show the Single Loop Operation (SLO) equation in the form 0.66(W-AW) + offset value, with the offset value the same for both SLO and Two Loop Operation (TLO). Currently, the equations are shown in the form 0.66W + offset value, with 5% difference in offset values for SLO vs. TLO. In the reformatted representation, AW equals zero for TLO and 7.6% for SLO (7.6% = 5%/0.66). The revised representation, while mathematically equivalent, states the equation in the same form that is actually implemented in the equipment. In addition, a notation has been added to Table 3.3.6-2 addressing the limits of application of the flow offset. No change related to SLO is required for the OPRM Upscale Function implementation. However, the form of this setpoint expression is being modified to address a TS concern identified during the system's installation at LGS. The concern involves the system's miscalculation of this equation when indicated recirculation drive flow (W) becomes less than AW. Although this flow condition is not operationally possible, proper description of the system's calculation of the single loop setpoint was deemed to be warranted.

C.2 To make the Intermediate Range Monitor (IRM) rod block OPCON 5 operability requirements consistent with the current IRM RPS OPCON 5 operability requirements,

Evaluation of Proposed Changes LGS OPRM Trip Activation LAR Page 6 of 15 footnote "**" will be added for Functions 4.a, 4.b, 4.c and 4.d in Tables 3.3.6-1 and 4.3.6-

1. The existing footnote "**" for these two tables is the same, and is consistent with the equivalent Note "(i)" requirement for RPS for OPCON 5. This change is not related to OPRM and is not covered in the NUMAC PRNM LTRs, but is being made concurrently to eliminate an inconsistency between the RPS and rod block TS. No TS Bases changes are involved in this TS change.

D.

Recirculation System, TS Section 3/4.4.1 D.1 LCO Conditions and Actions LCO restrictions on operating region (references to 45% core flow and Figure 3.4.1.1-1) will be deleted from the LCO (the TS Index will also be revised to reflect the deletion of Figure 3.4.1.1-1). Actions c. and d. and Surveillances 4.4.1.1.3 and 4.4.1.1.4d, each associated with operation in the restricted zone and included previously as part of the ICAs, will be deleted. Similarly, Action b. will be modified to delete the 2-hour action associated with exiting the restricted zone. These changes, along with deletion of the related TS Bases discussions, effectively delete the TS requirements for the ICAs. The NUMAC PRNM LTRs do not address deletion of ICA related TS. Therefore, all TS 3/4.4.1 changes are beyond those covered by the NUMAC PRNM LTRs.

D.2 In addition to deletion of requirements to exit the restricted zone within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, Action b.

has been simplified to require only that the plant be in at least Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This change, which is beyond the change required to activate the OPRM Upscale Function, does not increase the total time allowed to reach Hot Shutdown, but removes the requirement to be in Startup within the first 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This change makes the LGS TS 3.4.1.1 allowed completion time for this Action consistent with a similar allowed Action completion time for LGS TS 3.4.1.2 get pumps inoperable) and the completion time for the equivalent Required Action in the Improved Standard Technical Specifications, NUREG-1433, "Standard Technical Specifications, General Electric Plants (BWR/4)," (Reference 7).

E.

Routine Reports, TS Section 6.9.1 The procedural method of controlling the limits used to establish the OPRM Upscale PBDA trip setpoints is not discussed in the NUMAC PRNM LTR, but a required utility action is to identify the method that will be used. The requirements for cycle specific confirmation or change of the limits is established in the BWROG LTRs NEDO-32465-A and NEDO-31960-A (including Supplement 1) but not the specific method of documentation. The required information will be included in the reload licensing report.

It has been determined that recording the OPRM Upscale PBDA trip setpoints in the COLR is the preferred method. This method is utilized at LGS for documenting similar cycle specific limits such as Rod Block Monitor limits, and has been utilized by other Licensees for the OPRM.

To document this requirement, a new item 6.9.1.9h will be added to note that the OPRM Upscale PBDA trip setpoints for TS 2.2.1 will be included in the COLR. Also, a new item 6.9.1.1 Ob will be added to identify the BWROG LTR NEDO-32465-A as the NRC approved documentation of the method for establishing the setpoints.

Evaluation of Proposed Changes LGS OPRM Trip Activation LAR Page 7 of 15 Technical Specifications Bases:

The TS Bases for the requested TS changes are consistent with Appendix H of the NUMAC PRNM LTRs (which have been approved by the NRC) with the following exceptions:

TS 2.2.1 - RPS Instrumentation Setpoints. APRM Functions Specifically, the TS Bases discussion for TS 2.2.1 has been limited somewhat to focus on setpoint related discussion. Some of the discussion has been expanded from the examples in the NUMAC PRNM LTRs, while some of the NUMAC PRNM LTR discussion has been moved to TS Bases 3.3.1.

Function 2.b APRM Simulated Thermal Power - Upscale The TS Bases will be expanded to include specific discussion of the 'W - 7.6%" term in the APRM Simulated Thermal Power - Upscale equation for single recirculation loop operation, and the limits of applicability of the required adjustment. This is being added to document the basis for the "offset" and to clarify that a hardware "clamp" limits the Allowable Value for flow values of W < 7.6%.

TS 3.3.1 - RPS Instrumentation, APRM Functions Function 2.e 2-Out-Of-4 Voter The TS Bases for Function 2.e will be modified slightly differently from that shown in the NUMAC PRNM LTR. The LTR discussion of "partial operability" related to the separate voting of the APRM High/Inop and the OPRM Upscale function will not be included.

Deletion of this discussion is conservative. In addition, discussion will be added to clarify that the "APRM Interface" part of the Two-Out-Of-Four Logic Module hardware is separate from the voter functions, and that inoperability of APRM Interface only hardware does not necessitate declaring the voter function inoperable. Specific examples are inoperable APRM Interface output modules, which might affect only annunciator functions or even rod block functions, but which do not affect any of the RPS functions and should not require entering an RPS LCO.

Function 2.f The specific number of LPRMs per OPRM cell, the minimum required number of OPERABLE LPRMs for an OPRM cell to be considered OPERABLE, and the minimum number of OPERABLE OPRM cells required for an OPRM channel to be considered OPERABLE are identified as plant specific values in the NUMAC PRNM LTRs with no specific criteria on selection or calculation of the values. The NUMAC PRNM LTR also does not discuss the specific assignment of LPRMs to OPRM cells or any criteria for those assignments. The NRC approved BWROG LTRs, NEDO-31960-A, including Supplement 1, and NEDO-32465-A, provide the criteria related to determination of those values.

Based on the criteria in the BWROG LTRs, LGS has selected an LPRM-to-OPRM Cell assignment pattern that includes either 3 or 4 LPRMs per OPRM cell, depending on where the cell is located in the core. This selection meets the criteria in the BWROG LTRs.

Similarly, LGS has selected 2 OPERABLE LPRMs as the minimum required per OPRM cell for OPRM cell operability. Based on these two LGS selected aspects of the OPRM

Evaluation of Proposed Changes LGS OPRM Trip Activation LAR Page 8 of 15 system, an analysis will be performed in accordance with the methodology defined in the BWROG LTRs to establish the PBDA trip setpoint limit criteria. The setpoint values will be documented in the COLR prior to OPRM Upscale Function activation. Also, based on the LGS selected cell assignments and minimum number of LPRMs per cell, an analyses was performed to establish a recommended minimum of 23 OPRM cells required for OPRM channel operability. The minimum number of LPRMs per cell and minimum required OPRM cells are included in the LGS specific TS Bases markups.

The TS Bases description of the OPRM power level for operability (25% RTP) and for "auto-enable" (30% APRM Simulated Thermal Power, 60% recirculation drive flow) has been modified somewhat from the NUMAC PRNM LTR version for clarity. There is no technical change from the intent. These values are identified as plant-specific in the NUMAC PRNM LTRs. Based on a BWROG letter to the NRC providing background and guidance (the letter is included as Reference 5 in proposed TS Bases Section 2.2.1), LGS has selected the above values. The values will be treated as nominal values with no additional margin added to determine the actual setpoints to be entered in the equipment.

Channel Calibration The NUMAC PRNM LTRs identify no specific changes to the Channel Calibration TS Bases. However, reviews associated with the OPRM Upscale function TS changes identified a concern that the TS Bases discussion of the flow input function calibration requirements previously added (to support the APRM Simulated Thermal Power - High function channel calibration) was not clear. In addition, the TS Bases should identify that the flow input function calibration also applies to the OPRM Upscale function (auto-enable). To address these issues and assure that the flow input function calibration requirements are correctly and completely understood, the TS Bases discussion related to flow input function calibration was expanded. The expanded discussion clarifies that channel calibration includes the once-per-cycle drive flow / core flow correlation adjustment.

Confirmation of OPRM Upscale enable setpoints The TS Bases description of this Surveillance Requirement has been reworded somewhat from that which is in the NUMAC PRNM LTR to clarify that the surveillance is only a confirmation of setpoints, and that the setpoints are considered "nominal" (reference to a BWROG letter supporting this position has been added as Reference 5 to proposed TS Bases Section 2.2.1). The APRM Simulated Thermal Power/THERMAL POWER and core flow/recirculation drive flow correlations are confirmed by weekly and refueling APRM channel calibrations, respectively. Some additional rewording has been done to clarify the intent of the Surveillance Requirement and to identify alternate actions available to satisfy the Surveillance Requirement.

Response Time Testing Surveillance Requirements Other than clarification that the OPRM Upscale Function has no specific response time testing requirement, the NUMAC PRNM LTRs show no added TS Bases discussion for the Response Time Testing SR 4.3.1.3. However, in defining the application of the Surveillance Requirement to the 2-Out-Of-4 Voter channels, and the required sequencing of tests between APRM and OPRM outputs, it was determined that some TS Bases discussion was required. Therefore, the TS Bases 3.3.1 have been expanded to address

Evaluation of Proposed Changes LGS OPRM Trip Activation LAR Page 9 of 15 application of the Surveillance Requirement to the 2-Out-Of-4 Voter channels, and to include the sequencing of tests.

TS 3.4.1 - Recirculation System Changes to the Recirculation System TS and associated TS Bases are not addressed in the NUMAC PRNM LTRs.

The changes are to delete TS Bases discussion of the Recirculation Loop requirements, Actions and Surveillance Requirements that were established only to address the ICAs, which are no longer required to be incorporated in TS with the implementation of the OPRM Upscale Function.

3.0 BACKGROUND

The NRC issued Generic Letter (GL) 94-02, "Long-Term Solutions and Upgrade of Interim Operating Recommendations for Thermal Hydraulic Instabilities in Boiling Water Reactors,"

(Reference 8) which required PECO Energy Company to develop and submit to the NRC a plan for long term stability corrective actions. In response to GL 94-02, by letter dated March 2, 1998, from G. D. Edwards, PECO Energy Company to the USNRC (Reference 9), PECO Energy (now Exelon Generation Company, LLC) committed to implement the long-term solution designated as Option IlIl in NEDO-31960-A (including Supplement 1).

The functionality required for the Option IlIl stability solution was implemented as part of the initial phase of PRNM system replacement modification P00224, which has been completed for both LGS, Units 1 and 2. That modification replaced the original Power Range Monitoring system, including the APRM system, the Rod Block Monitor system and the LPRM system, except for the detectors and signal cables, with the GE NUMAC PRNM system.

The NUMAC PRNM system utilizes the OPRM detect-and-suppress function to implement Option ll. The safety function of the OPRM function within the PRNM is to monitor its LPRM signals for signs of neutron flux oscillations. The OPRM also monitors simulated thermal power and recirculation drive flow conditions to automatically enable the OPRM Upscale Function when in a predefined region of the power to flow map. The OPRM Upscale Function initiates a trip whenever it detects an instability condition when in the predefined region of the power to flow map. Following installation of the new PRNM system, the OPRM Upscale Function has been fully operational except for the trip and associated trip alarm functions. The OPRM Upscale Function has been de-activated (not connected to the Reactor Protection System logic) in order to allow evaluation of the performance of the OPRM algorithms without the risk of spurious scrams. During this evaluation period, in 2001, GE initiated a report in accordance with 10 CFR Part 21, "Reporting of defects and noncompliances," concerning stability reload licensing calculations that support the development of setpoints for the OPRM Upscale Function. The OPRM Upscale Function was not armed pending resolution of this reportable condition. The reportable condition has now been resolved as described in the letter dated September 30, 2003, from the BWROG to the NRC (Reference 10). Consistent with NRC Bulletin 88-07, Supplement 1 (Reference 11), as committed to in the letter dated September 9, 1994, from G. A. Hunger, PECO Energy to USNRC (Reference 12), Exelon has continued to implement the Interim Corrective Actions (ICAs) to detect and suppress power oscillations.

During this time frame, the OPRM system has been tuned per recent GE criteria (Reference 13) to establish proper sensitivity. Performance of the system at LGS during this interim phase, as well as at other plants, has been reliable thus warranting activation of the trip outputs.

Evaluation of Proposed Changes LGS OPRM Trip Activation LAR Page 10 of 15 License amendments to support the activation of the trip outputs of the OPRM portion of the PRNM system have been approved by the NRC for Hatch Units 1 and 2 (Reference 14),

Browns Ferry Unit 2 (Reference 15), Browns Ferry Unit 3 (Reference 16), Nine Mile Point Unit 2 (Reference 17), Fermi Unit 2 (Reference 18) and Brunswick Units 1 and 2 (Reference 19).

The final phase of this modification will accomplish the following:

Activate the OPRM Upscale Function and annunciators; Add OPRM TS; Delete the Recirculation System TS requirements associated with the Interim Corrective Actions; Implement equipment modifications limited to minor wiring changes in the PRNM Panel 1(2)0C608, Annunciator Panel 1 (2)0C892 and annunciator window label changes; and Implement appropriate procedures and training to reflect the OPRM system.

4.0 TECHNICAL ANALYSIS

The technical bases for the requested TS changes are presented in Section 8.0 of the NRC approved NUMAC PRNM LTR NEDC-3241OP-A (including Supplement 1), and as described above in Section 2.0 of this attachment, including exceptions to the NUMAC PRNM LTRs, and as discussed below. In addition, Attachment 4 provides plant-specific responses required by the NUMAC PRNM LTRs, including a discussion of the justifications and deviations related to the Utility Required Actions specified in the LTRs.

Enabling the OPRM Upscale Function of the GE NUMAC PRNM system implements the long-term stability solution required by Generic Letter 94-02. The PRNM hardware incorporates the Option IlIl detect and suppress solution reviewed and approved by the NRC. The OPRM meets 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 10, "Reactor design," and GDC 12, "Suppression of reactor power oscillations," requirements by automatically detecting and suppressing design basis thermal-hydraulic oscillations prior to exceeding the fuel MCPR Safety Limit.

The current plant design utilizing Interim Corrective Actions (ICAs) depends on operator action to, if possible, avoid regions where instability may occur, to exit such regions when necessary, and to detect an actual instability and take mitigating action by manual means. The modification replaces procedural actions (i.e., the ICAs) with an NRC approved automatic detect and suppress function (i.e., the OPRM Upscale Function). The OPRM Upscale Function includes sophisticated algorithms that can automatically detect an instability condition and provide a RPS trip input if the oscillation magnitude exceeds acceptable limits.

The OPRM Upscale Function is capable of more quickly and reliably detecting a true reactor instability than was possible with the manual procedures. The OPRM also provides a reactor scram trip only if an actual instability is detected while the current ICAs require reactor shutdown if the plant is in a condition that may result in an instability, regardless of whether or not an instability occurs. Extensive analyses performed by the Boiling Water Reactor Owner's Group (BWROG) and reviewed and approved by the NRC demonstrate that the OPRM can detect reactor instabilities and initiate a scram trip before the MCPR Safety Limit is exceeded, thus maintaining the integrity of the fuel.

The only hardware impact for the proposed activity is disconnecting a few terminations to remove connections which "jumpered out" the OPRM trip outputs, the addition of jumpers to

Evaluation of Proposed Changes LGS OPRM Trip Activation LAR Page 11 of 15 connect the OPRM trip and alarm to the annunciator and plant computer inputs, and minor re-labeling of some annunciator windows. All jumper removals/additions are accomplished in Panels 1(2)0C608 and 1(2)0C892.

The OPRM trip actuation phase of LGS Modification P00224 and its associated TS changes will not adversely affect the ability of the RPS to perform its intended function. The significant change in this phase of the modification involves removing jumpers across the existing OPRM trip outputs to RPS. The Surveillance Requirements and their frequency of performance will assure reliability of the OPRM portion of the PRNM system. The modification replaces procedural actions (ICAs) with an NRC approved automatic detection and suppression function, which provides an RPS trip input if acceptable reactor operational limits are exceeded.

Therefore, the proposed modifications and associated TS changes will not adversely affect the health and safety of the public.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration Exelon Generation Company, LLC (Exelon) proposes changes to the Technical Specifications (TS) for Limerick Generating Station (LGS), Units 1 and 2. The proposed changes support the activation of the trip outputs of the previously-installed Oscillation Power Range Monitor (OPRM) portion of the Power Range Neutron Monitoring (PRNM) system. The OPRM system monitors neutron flux signals for signs of neutron flux oscillations and initiates a reactor trip whenever it detects an instability condition when in the predefined region of the power to flow map. Activation of the OPRM will replace manual methods for avoiding instabilities and for detecting and suppressing potential instabilities.

Exelon has evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No. This modification has no impact on any of the previously installed PRNM functions. Plant operation in portions of the former restricted zone may potentially cause a marginal increase in the probability of occurrence of an instability event. This potential increase in probability is acceptable because the OPRM Upscale Function will automatically detect the condition and initiate a reactor scram before the Minimum Critical Power Ratio (MCPR) Safety Limit is reached. Consequences of the potential instability event are reduced because of the more reliable automatic detection and suppression of an instability event, and elimination of dependence on the manual operator actions.

The change to align the operability requirements for the Intermediate Range Monitor (IRM) rod block function with those for the corresponding IRM Reactor Protection System (RPS) functions affects only the rod block function. The justification for the change to IRM RPS function (done with the original PRNM modification) concluded that the RPS change would not increase the probability of occurrence of an accident previously evaluated; therefore, changing the associated rod block to align with those requirements would not do so either.

Evaluation of Proposed Changes LGS OPRM Trip Activation LAR Page 12 of 15 Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No. The modification replaces procedural actions that were established to avoid operating conditions where reactor instabilities might occur with an NRC approved automatic detect and suppress function.

Potential failures in the OPRM Upscale Function could result in either failure to take the required mitigating action or an unintended reactor scram. These are the same potential effects of failure of the operator to take the appropriate action under the current procedural directions. The net effect of the modification changes the method by which an instability event is detected and by which mitigating action is initiated, but does not change the type of stability event that could occur. The effects of failure of the OPRM equipment are limited to reduced or failed mitigation, but such failure cannot cause an instability event or other type of accident.

The change to align the operability requirements for the IRM rod block function with those for the corresponding IRM RPS functions affects only the rod block function. The justification for the change to IRM RPS function (done with the original PRNM modification) concluded that the RPS change could not create the possibility of a new type of accident; therefore, changing the associated rod block to align with those requirements would not do so either.

Therefore, since no radiological barrier will be challenged as a result of activating the OPRM Upscale Function, it is concluded that the proposed changes do 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 current safety analysis assumes that the existing procedural actions are adequate to prevent an instability event. As a result, there is currently no quantitative or qualitative assessment of an instability event with respect to its impact on the MCPR Safety Limit.

The OPRM Upscale function is being implemented to automate the detection (via direct measurement of neutron flux) and subsequent suppression (via scram) of an instability event prior to exceeding the MCPR Safety Limit. The OPRM Upscale function provides a trip output of the same type as currently used for the Average Power Range Monitor (APRM). Its failure modes and types are identical to those for the present APRM output.

Currently, the MCPR Safety Limit is not challenged by an instability event since the event is "mitigated" by manual means via the procedural actions, which prevent plant operating conditions where an instability event is possible. In both methods of mitigation (manual and automated), the margin of safety associated with the MCPR Safety Limit is still maintained.

Evaluation of Proposed Changes LGS OPRM Trip Activation LAR Page 13 of 15 Therefore, based on the fact that the MCPR Safety Limit will still be enforced, implementation of the OPRM Upscale function in place of the existing manual actions does not reduce the margin of safety.

The IRM rod block function is not considered in any safety analysis. As a result, its failure will not affect the margin of safety.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, Exelon concludes that the proposed amendment 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.

5.2 Applicable Regulatory Requirements/Criteria 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 10, "Reactor design," requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. GDC 12, "Suppression of reactor power oscillations," requires assurance that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed.

Enabling the OPRM Upscale function implements the long-term stability solution required by Generic Letter 94-02. The PRNM hardware incorporates the Option IlIl detect and suppress solution reviewed and approved by the NRC in Licensing Topical Reports. The OPRM meets the GDC 10 and 12 requirements by automatically detecting and suppressing design basis thermal-hydraulic oscillations prior to exceeding the fuel Minimum Critical Power Ratio (MCPR) Safety Limit.

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.

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement, or environmental assessment need be prepared in connection with the proposed amendment.

Evaluation of Proposed Changes LGS OPRM Trip Activation LAR Page 14 of 15

6.0 REFERENCES

1. Letter from Keith R. Jury (Exelon Generation Company, LLC) to U. S. NRC, "Schedule for Completing Actions to Implement Long-Term Stability Solution," dated December 19, 2003.

2.

NEDC-3241OP-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option IlIl Stability Trip Function," dated October 1995.

3.

NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option IlIl Stability Trip Function," Supplement 1, dated November 1997.

4.

NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

5.

NEDO-31960-A, "BWR Owner's Group Long-Term Stability Solutions Licensing Methodology," dated November 1995.

6.

NEDO-31960-A, "BWR Owner's Group Long-Term Stability Solutions Licensing Methodology," Supplement 1, dated November 1995.

7.

NUREG-1433, "Standard Technical Specifications, General Electric Plants (BWR/4)."

8.

Generic Letter (GL) 94-02, "Long-Term Solutions and Upgrade of Interim Operating Recommendations for Thermal Hydraulic Instabilities in Boiling Water Reactors," dated July 11, 1994.

9.

Letter dated March 2,1998, from G. D. Edwards, PECO Energy Company to the USNRC.

10. Letter from K. S. Putnam (Boiling Water Reactor Owners' Group) to U. S. NRC, "Resolution of Reportable Condition for Stability Reload Licensing Calculations Using Generic Regional Mode DIVOM Curve," dated September 30, 2003.

11. NRC Bulletin 88-07, "Power Oscillations In Boiling Water Reactors (BWR)," Supplement 1, dated December 30, 1988.

12. Letter dated September 9, 1994, from G. A. Hunger, PECO Energy to USNRC.

13. Letter dated October 4, 2003, from J. S. Post, GE Nuclear Energy, to USNRC, "Part 21 Notification: Stability Option III Period Based Detection Algorithm Allowable Settings."

14. Letter dated August 20,1998, from L. N. Olshan, USNRC to H. L. Sumner, Hatch Plant, Southern Nuclear Operating Company, Inc., issuance of Amendments - Edwin I. Hatch Nuclear Plant, Units 1 and 2 (TAC Nos. M99066 and M99067)."

15. Letter dated March 5,1999, from L. Raghavan, USNRC, to J. A. Scalice, Tennessee Valley Authority, "Amendment No. 258 to Facility Operating License No. DPR-52: Oscillation Power Range Monitor Upscale Trip Function in the Average Power Range Monitor -

Technical Specification Change TS-354 (TAC No. MA3556)."

Evaluation of Proposed Changes LGS OPRM Trip Activation LAR Page 15 of 15

16. Letter dated September 27, 1999, from W. 0. Long, USNRC, to J. A. Scalice, Tennessee Valley Authority, "Browns Ferry Nuclear Plant, Unit 3 - Issuance of Amendment Regarding Oscillation Power Range Monitor (TAC No. MA5976)."

17. Letter dated March 2, 2000, from P. S. Tam, USNRC, to J. H. Mueller, Niagara Mohawk Power Corporation, Nine Mile Point Station, "Nine Mile Point Station, Unit No. 2 - Issuance of Amendment RE: Oscillation Power Range Monitoring System (TAC No. MA7119)."

18. Letter dated March 31, 2000, from A. J. Kugler, USNRC, to D. R. Gipson, Detroit Edison company, "Fermi 2-Issuance of Amendment RE: Enabling the Oscillation Power Range Monitor Upscale Trip Function (TAC No. MA6267)."

19. Letter dated March 8, 2002, from A. G. Hansen, USNRC, to J. S. Keenan, Brunswick Steam Electric Plant, Carolina Power & Light, "Brunswick Steam Electric Plant, Units 1 and 2 - Issuance of Amendment to Incorporate the General Electric Digital Power Range Neutron Monitoring System (TAC Nos. MB2321 and MB2322)."

20. Letter dated October 14, 1999, from J. A. Hutton, PECO Energy Company, to USNRC, "Limerick Generating Station, Units 1 and 2, Technical Specifications Change Request No.

99-05-0."

21. Letter dated February 11, 2000, from J. A. Hutton, PECO Energy Company, to USNRC, "Response to Request for Additional Information Related to Technical Specifications Change Request No. 99-05-0."

22. Letter dated April 12, 2000, from B. C. Buckley, USNRC, to J. A. Hutton, PECO Energy Company, "Limerick Generating Station, Unit 1 - Issuance of Amendment RE: Power Range Neutron Monitoring (TAC No. MA6965)."

23. Letter dated January 16, 2001, from J. W. Clifford, USNRC, to J. A. Hutton, PECO Energy Company, "Limerick Generating Station (LGS), Unit 2 - Issuance of Amendment RE: Power Range Neutron Monitoring (TAC No. MA6966)."

ATTACHMENT 2 LICENSE AMENDMENT REQUEST LIMERICK GENERATING STATION, UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 ACTIVATION OF THE TRIP OUTPUTS OF THE OSCILLATION POWER RANGE MONITOR PORTION OF THE POWER RANGE NEUTRON MONITORING SYSTEM MARKED-UP TECHNICAL SPECIFICATIONS AND BASES PAGES FOR PROPOSED CHANGES UNIT 1 xi 2-4 B2-6 B2-7 B2-9 3/4 3-1 3/43-1a 3/4 3-2 3/4 3-4 3/4 3-5 3/4 3-6 3/43-7 3/43-8 3/ 4 3-58 3/ 4 3-60 3/4 3-60a UNIT 1 3/ 4 3-61 3/4 4-1 3/44-la 3/ 4 4-2 3/4 4-3 B3/43-1 B3/4 3-1a B3/ 4 3-1b B3/ 4 3-1c B3/ 4 3-7 B3/ 4 4-1 B3/ 44-2 6-18a UNIT 2 xi 2-4 B2-6 B2-7 B2-9 3/4 3-1 3/43-1a 3/43-2 3/43-4 3/43-5 3/4 3-6 3/ 4 3-7 3/4 3-8 3/ 4 3-58 3/ 4 3-60 3/4 3-60a UNIT 2 3/ 4 3-61 3/4 4-1 3/44-la 3/44-2 3/4 4-3 B3/ 4 3-1 B3/4 3-1a B3/4 3-1b B3/ 4 3-1c B3/ 4 3-7 B3/ 4 4-1 B3/ 4 4-2 6-18a

TECH SPEC MARKUP for Stability (PRNM) OPRM Trip Activation MOD at LIMERICK UNIT 1

MINDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

_SECTI_

EA (VeJUee 0

G REACTOR COOLANT SYSTEM (Continued)

Figure 3.4.1.1-1 Jet Pumps.3/4 4-4 Recirculation Pumps............

3/4 4-5 Idle Recirculation Loop Startup.............................. 3/4 4-6 3/4.4.2 SAFETY/RELIEF VALVES....................

3/4 4-7 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems...................................... 3/4 4-8 Operational Leakage................

3/4 4-9 Table 3.4.3.2-1 Reactor Coolant System Pressure Isolation Valves.3/4 4-11 3/4.4.4 CHEMISTRY........

3/4 4-12 Table,3.4.4-1 Reactor Coolant System Chemistry Limits..........................

3/4 4-14 3/4.4.5 SPECIFIC ACTIVITY.............................

3/4 4-15 Table 4.4.5-1. Primary Coolant Specific Activity Sample and Analysis Program.3/4 4-17 3/4.4.6 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System.3/4 4-18 Figure 3.4.6.1-1 Minimum Reactor Pressure Vessel Metal Temperature Vs. Reactor Vessel Pressure.3/4 4-20 Table 4.4.6.1.3-1 Deleted.3/4 4-21 Reactor Steam Dome.............................

3/4 4-22 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES.3/4 4-23 3/4.4.8 STRUCTURAL INTEGRITY.3/4 4-24 K

LIMERICK -

UNIT 1 xi Amendment No. 167

TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS r-b.-4 rn

0 C-1

K I

,_4

__4 b.-

FUNCTIONAL UNIT

1.

Intermediate Range Monitor, Neutron Flux-High

2. Average Power Range Monitor:

y

a. Neutron Flux-Upscale (Setdown)

b. Simulated Thermal Power - Upscale:

Two Recirculation Loop Operation

- Single Recirculation Loop Operation

c. Neutron Flux -

Upscale TRIP SETPOINT s 120/125 divisions of full scale i 15.0% of RATED THERMAL POW1 K 0.66 W + 62.8% and s 116.6% of RATED THERMAL POWER THERMAL POWER 118.3% of RATED THERMAL POWER ALLOWABLE VALUES t 122/125 divisions of full scale

-R s 20.0% of RATED THERMAL POWER s 0.66 W + 63.3% and

& 117.0% of RATED 5 1170XofRAT ED THERMAL POWER 118.7% of RATED THERMAL POWER

d. Inoperative

e. 2-Out-Of-4 Voter

-3. Reactor Vessel Steam Dome Pressure - High

4.

Reactor Vessel Water Level - Low, Level 3

5. Main Steam Line Isolation Valve - Closure

6.

DELETED

7. Drywell Pressure -

High

8. Scram Discharge Volume Water Level - High

a.

Level Transmitter 1->

b.

Float Switch C

9. Turbine Stop-Valve - Closure

10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low

11. Reactor Mode:Switch Shutdown Position X

ma

12. Manual Scram N.A.

N. A.

N.A.

N.A.

& 1096 psig 2 12.5 inches above instrument zero*

i 8%..closed DELETED g 1.68 psig s 1103 psig 2 11.0 inches above instrument zero s 12% closed DELETED s 1.88 psig 5 261' 5 5/8" elevation g 261' 5 5/8" elevation

&7% closed t 260'.9 5/8" t 260' 9 5/8" t 5% closed elevation**

elevation**

2 500 psig N.A.

N.A.

2 465 psig N.A.

N.A.

b Iha

• • See Bases Figure B,3/4.3-1.

Equivalent to 25.45Igallons/scram discharge volume.

TECH SPEC MARKUP INSERT 1:

f.

OPRM Upscale N.A.

INSERT 2:

The 7.6% flow "offset" for Single Loop Operation (SLO) is applied for W 2 7.6%. For flows W < 7.6%, the (W-7.6%) term is set equal to zero.

See COLR for OPRM period based detection algorithm trip setpoints. OPRM Upscale trip output auto-enable (not bypassed) setpoints shall be APRM Simulated Thermal Power 2 30% and recirculation drive flow < 60%.

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System instrumentation setpoints specified in Table 2.2.1-1 are the values at which the reactor trips are set for each para-meter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design.basis anticipated operational occurrences and to assist in'miti'gating the consequences of accidents.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint. andthe Allowable Value is equal.,to or less than the drift allowance assumed for each trip in the safety analyses.

1.

Intermediate Range Monitor. Neutron Flux -

High The IRM system consists of 8 chambers,.4 in each of the reactor trip systems.

The IRM is a 5 decade 10 range instrument. The trip setpoint of 120 divisions of scale is active in each of the 10 ranges.

Thus as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up.

The.IRM instruments provide for overlap with' both the APRM and SRM systems.

The most significant source of reactivity changes during'the power increase is due to control rod withdrawal.

In order to ensure that the IRM provides the required protection,.a range of rod withdrawal accidents have been analyzed.

The results ofthese analyses :are!in Section.15!4;of'the" FSAR.

The most severe case involves' an initial condition in which THERMAL POWER is at approximately -1% of RATED'THERMALcPOWER. 'Additional conservatism was taken in this analysis ~by assuming the IRMIchannel closest to the control rod being withdrawn is bypassed. The results of this'analysis show that the reactor is shutdown.and peak power-is limited to 21%,of RATED THERMAL POWER with the peak fuel.enthalpy well belowtherfuel failure threshold of '170 cal/gm.

Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

2. Average Power Range Monitor The APRM system is divided into four APRH channels and four 2-Out-Of-4 Voter channels. The four voter channels are divided into two groups of two each, with each group of two providing inputs to one RPS trip system. All four voters will trip (full scram) when any tw APR channels exceed their trip setpoints.

For operation at low pran ow flow during STARTUP, the APRM Neutron L

'Flux-Upscale (Setdown) scram setting of 15% of RATED THERMAL POWER provides'adequate thermal margin between the setpoint and the Safety Limits.

The margin accommodates the anticipated maneuvers associated with power plant startup.

Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system. Tempera-ture coefficients are small and control rod patterns are constrained by the RWM.

Of all the possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power increase.

APR 12 2000 LIMERICK - UNIT I B 2-6 Amendment No.

17, 141

TECH SPEC MARKUP INSERT 3:

APRM trip Functions 2.a, 2.b, 2.c, and 2.d are voted independently from OPRM Upscale Function 2.f. Therefore, any Function 2.a, 2.b, 2.c, or 2.d trip from any two unbypassed APRM channels will result in a full trip in each of the four voter channels.

Similarly, a Function 2.f trip from any two unbypassed APRM channels will result in a full trip from each of the four voter channels.

LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Average Power Range Monitor (Continued)

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant amount, the rate of power rise is very slow. Generally the heat flux is in near equilibrium with the fission rate.

In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit.

The 15% Neutron Flux - Upscale (Setdown) trip remains active until the mode switch is placed in the Run position.

The APRM trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the system and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Neutron Flux - Upscale setpoint; i.e.,

for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel.

For the Simulated Thermal Power - Upscale setpoint, a time constant of 6 i 0.6 seconds is introduced into the flow-biased APRM in order to simulate the fuel thermal transient characteristics. A more conservative maximum value is used for the flow-biased setpoint as shown in Table 2.2.1-1.

The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unneces-

______.sary shutdown.

3.'

Reactor Vessel Steam Dome Pressure-High High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity.

The trip will quickly reduce the neutron flux, counteracting the pressure increase. *The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account-the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient.

This trip setpoint is effective at low power/flow conditions when the turbine stop valve and control fast closure trips are bypassed.

For a turbine trip or load rejection under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.

APR 12 2G9 LIMERICK - UNIT I B 2-7 Amendment No.

60, 141

TECH SPEC MARKUP INSERT 4:

A reduced Trip Setpoint and Allowable Value is provided for the Simulated Thermal Power - Upscale Function, applicable when the plant is operating in Single Loop Operation (SLO) per LCO 3.4.1.1. In SLO, the drive flow values (W) used in the Trip Setpoint and Allowable Value equations is reduced by 7.6%. The 7.6% value is established to conservatively bound the inaccuracy created in the core flow/drive flow correlation due to back flow in the jet pumps associated with the inactive recirculation loop. The Trip Setpoint and Allowable Value thus maintain thermal margins essentially unchanged from those for two-loop operation. The Trip Setpoint and Allowable Value equations for single loop operation are only valid for flows down to W = 7.6%. The Trip Setpoint and Allowable Value do not go below 62.8% and 63.3% RATED THERMAL POWER, respectively. This is acceptable because back flow in the inactive recirculation loop is only an issue with drive flows of approximately 40% or greater (Reference 1).

INSERT 5:

The APRM channels also include an Oscillation Power Range Monitor (OPRM)

Upscale Function. The OPRM Upscale Function provides compliance with GDC 10 and GDC 12, thereby providing protection from exceeding the fuel MCPR Safety Limit due to anticipated thermal-hydraulic power oscillations. The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into "cells" for evaluation by the OPRM algorithms.

References 2, 3 and 4 describe three algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. All three are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period based detection algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY for Technical Specification purposes is based only on the period based detection algorithm.

The OPRM Upscale trip output shall be automatically enabled (not bypassed) when APRM Simulated Thermal Power is Ž 30% and recirculation drive flow is < 60%

as indicated by APRM measured recirculation drive flow. (NOTE: 60% recirculation drive flow is the recirculation drive flow that corresponds to 60% of rated core flow.

Refer to TS Bases 3/4.3.1 for further discussion concerning the recirculation drive flow/core flow relationship.) This is the operating region where actual thermal-hydraulic instability and related neutron flux oscillations may occur. See Reference 5 for additional discussion of OPRM Upscale trip enable region limits. These setpoints, which are sometimes referred to as the "auto-bypass" setpoints, establish the boundaries of the OPRM Upscale trip enabled region. The APRM Simulated Thermal Power auto-enable setpoint has 1% deadband while the drive flow setpoint has a 2%

deadband. The deadband for these setpoints is established so that it increases the enabled region.

TECH SPEC MARKUP INSERT 5 (continued):

An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in a cell, with period confirmations and relative cell amplitude exceeding specified setpoints. One or more cells in a channel exceeding the trip conditions will result in a channel trip. An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithms detect oscillatory changes in the neutron flux for one or more cells in that channel.

There are four "sets" of OPRM related setpoints or adjustment parameters:

a) OPRM trip auto-enable setpoints for APRM Simulated Thermal Power (30%) and recirculation drive flow (60%); b) period based detection algorithm (PBDA) confirmation count and amplitude setpoints; c) period based detection algorithm tuning parameters; and d) growth rate algorithm (GRA) and amplitude based algorithm (ABA) setpoints.

The first set, the OPRM auto-enable region setpoints, are treated as nominal setpoints with no additional margins added as discussed in Reference 5. The settings, 30% APRM Simulated Thermal Power and 60% recirculation drive flow, are defined (limit values) in a note to Table 2.2.1-1. The second set, the OPRM PBDA trip setpoints, are established in accordance with methodologies defined in Reference 4, and are documented in the COLR. There are no allowable values for these setpoints.

The third set, the OPRM PBDA "tuning" parameters, are established or adjusted in accordance with and controlled by station procedures. The fourth set, the GRA and ABA setpoints, in accordance with References 2 and 3, are established as nominal values only, and controlled by station procedures.

LIMI~TING SA;:ZTy SYSTEVM SET7ING BASES REACTOR PROTECTION SYSTEH INSTRUMENTATION SETPOINTS (Continued)

8.

Scram Discharce Volume Water Level-Hich The scram discharge volume receives the water displaced.by the :motion of the control rod drive pistons during a reactor scram.

Should this volume fill-up to a point-where there is insufficient volume-to accept the displaced water

-at pressures below 65 psig, control rod insertion would be hindered..

The reactor is therefore tripped when the watedr level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough to accommodate the water from the movement of the rods at pressures below 65 psig when they are. tripped.

The trip setpoint for each scram discharge volume is equivalent to a contained volume of 25.45 gallons of water.

9.

Turbine -StoD Valve-Closure The turbine stop valve closure trip anticipates the pressure, neutron flux, and heat flux increases that would result from closure of the stop valves.

With. a trip setting of 15 of valve closure -from full open, the resultant increase in heat flux is such that adequate'thernal margins are maintained during the worst design basis transient.

10.

Turbine Control Valve Fast Closure, Trim Oil Pressure-L6w (1

C

,The turbine control' valve fast closure trip anticipates the pressure, neutron

.flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection with or without coincident fail-ure of. the turbine bypass valves.

The Reactor Protection System initiates a trip when fast clqsure of the control valves is initiated by the fast acting solenoid valves and in less than 30 milliseconds after the start of control valve fast closure.

This is achieved by the action of the fast acting solenoid valves in rapidly reducing hydraulic trip oil pressure at the main turbine control valve actuator disc dump valves.

This loss of pressure is sensed by pressure switches whose contacts form

.tne one-out-Of-tWO-twiCe logic input to the Reactor Protection System.

This trip setting,-a faster closure time, and-a different valve characteristic From that of the turbine stop valve, combine to produce transients which are very similar to t.nat for the stop valve.

Relevant transient analyses are discussed in Section 13.2.2 of the Final Safety Analysis Report.

21.

Reactor Hode Switch Shutdown Position The reactor mode switch Shutdown position is a redundant channel to the automatic protective instrumentation channels and provides additional manual reactor trip capability.

12.

Manual Scramr The Manual Scram is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

LIMERICK -

UNIT 2 B 2-9 6

E1g

TECH SPEC MARKUP INSERT 6:

REFERENCES:

1.

NEDC-31300, "Single-Loop Operation Analysis for Limerick Generating Station, Unit 1", August 1986.

2.

NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

3.

NEDO-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

4.

NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

5.

BWROG Letter 96113, K. P. Donovan (BWROG) to L. E. Phillips (NRC),

"Guidelines for Stability Option IlIl 'Enable Region' (TAC M92882),"

September 17,1996.

3/4.3 INSTRUMENTATION 4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

APPLICABILITY:

As shown in Table 3.3.1-1.

ACTION:

Note:

Separate condition entry is allowed for each channel.

a.

With the number of OPERABLE channels in either trip system for one or more Functional Units less than the Minimum OPERABLE Channels per Trip System required by Table 3.3.1-1, within one hour for each affected functional unit either verify that at least one* channel in each trip system is OPERABLE or tripped or that the trip system is tripped, or place either the affected trip system or at least one inoperable channel in the affected trip system in the tripped condition.

b.

With the number of OPERABLE channels in either trip system less than the Minimum OPERABLE Channels per Trip System required by Table 3.3.1-1, place either the inoperable channel(s) or te affected trip system** in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> rs**'

c.

With the number of OPERABLE channels in both trip systems for one or more Functional Units less than the Minimum OPERABLE Channels per Trip System required by Table 3.3.1-1, place either the inoperable channel(s) in one trip system or one trip system in the tripped condition within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s**i

d.

If within the allowable time allocated b ions a, b or c, it is not desired to place the inoperable channel o trip system in trip (e.g., full scram would occur), Then no later than expiration of that allowable time initiate the action identified in Table 3.3.1-1 for the applicable Functional Unit.

• For Functional Units 2.a, 2.b, 2.c, 2.

a east two channels shall be OPERABLE or tripped.

For Functional Unit 5, both trip systems shall have each channel associated with the MSIVs in three main steam lines (not necessarily the same main steam lines for both trip systems) OPERABLE or tripped.

For Function 9, at least three channels per trip system shall be OPERABLE or tripped.

• • For Functional Units 2.a, 2.b, 2.c, 2.d,~irioerable channels shall be placed in the tripped condition to comply with Action b. Action c does not apply for these Functional Units.

A APR 1 2 Z.3 LIMERICK - UNIT 1 3/4 3-1 Amendment No. %%, 71, 141

TECH SPEC MARKUP INSERT 6A:

A channel or trip system which has been placed in the tripped condition to satisfy Action b. or c. may be returned to the untripped condition under administrative control for up to two hours solely to perform testing required to demonstrate its operability or the operability of other equipment provided Action a. continues to be satisfied.

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1 4.3.1.2 LOGIC SYSTEM FUNCTIONAL ESTS and simulated automatic operation of all channels shall be performed at least once per 24 months, except Table 4.3.1.1-1 Functions 2.a, 2.b, 2.c, 2.d, C 2.e.

Functions 2.a, 2.b,'2.c,40b2.d require separate LOGIC SYSTEM FUNCTIO AL TESTS.

For Function 2.e, tests shall be performed at least once per 24 months.

LOGIC SYSTEM FUNCTIONAL TEST for Function 2.e includes simulating APRM rp conditions at the APRM channel inputs to the voter channel to check all combina ion of two tripped inputs to the 2-Out-Of-4 voter logic in the voter channels.

4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its limit at least once per 24 months.

Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 24 months where N is the total number of redundant channels in a specific reactor trip system.

I APR 1 2 200 LIMERICK - UNIT 1 I

3/4 3-la Amendment -No.

141

FUNCTIONAL UNIT

1.

Intermediate Range Monitors"':

a.

Neutron Flux - High

b.

Inoperative

2.

Average Power Range Monitories:

a.

Neutron Flux - Upscale (Se

b.

Simulated Thermal Power -

c.

Neutron Flux - Upscale

d.

Inoperative Qe.

2-Out-Of-4 Voter

3.

Reactor Vessel Steam Dome Pressure - High

4.

Reactor Vessel Water Level -

Low Level 3

5.

Main Steam Line Isolation Valve-Closure TABLE 3.3.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION APPLICABLE MINIMUM OPERATIONAL OPERABLE CHANNELS CONDITIONS PER TRIP SYSTEM (a) 2 3

3(i), 4(i) 3 5(i) 3(d) 2 3

3(i), 4(i) 3 5(i) 3(d) tdown) 2 3(m)

Upscale 1

3(m) 1 3(m) 1, 2 3(m) 1, 2 2

1, 2(f) 2 ACTION 1

2 3

1 2

3 I.

1 4

4 1

1 1

1, 2 2

1 1(g) 1/valve 4

LIMERICK - UNIT I 3/4 3-2 Amendment No. X8, 44, 44-1, 149

TECH SPEC MARKUP INSERT 7:

f.

OPRM Upscale 1 (o)(p) 3(m) 10

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATJ.PP' ACTION 'STATEMENTS ACTION I Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 2.

Verify all insertable control rods to bb inserted in the core and lock the'reactor mode switch in the SHUTDOWN position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 3 Suspend all operations involving CORE ALTERATIONS and insert all insertablea control rods within l hour.'

ACTION 4 Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 5 Be in STARTUP with the main steam line isolation valves closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 6 Initiate a reduction-in THERMAL POWER within ]5 minutes and reduce turbine first stage pressure until the function is

'automatically bypassed, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

ACTION 7 Verify all insertable control rods to be inserted, within I hour.

ACTION 8 L6ck the reactor mode switch in the -Shittdown position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 9 Suspend all operations involving CORE ALTERATIONS, and' insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

A.

LiME2IC-UN1.........

.3-14. 3.4

.. Am'cndm'ent No.

1-ai, 149

TECH SPEC MARKUP INSERT 8:

ACTION 10 a.

If the condition exists due to a common-mode OPRM deficiency*, then initiate alternate method to detect and suppress thermal-hydraulic instability oscillations within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND restore required channels to OPERABLE status within 120 days, OR

b.

Reduce THERMAL POWER to < 25% RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Unanticipated characteristic of the instability detection algorithm

• or equipment that renders all OPRM channels inoperable at once.

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a)

A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

(b) This function shall be automatically bypassed when the reactor mode switch is in the Run position.

(c) DELETED (d) The noncoincident NMS reactor trip function logic is such that all channels go to both trip systems. Therefore, when the "shorting links' are removed, the Minimum OPERABLE Channels Per Trip System is 6 IRMs.

(e)

An APRM channel is inoperable if there are less than 3 LPRM inputs per level or less than 20 LPRM inputs to an APRM channel, or if more than 9 LPRM inputs to the APRM channel have been bypassed since the last APRM calibration (weekly gain calibration).

(f) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

(g) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.

(h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(i) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(j)

This function shall be automatically bypassed when turbine first stage pressure is equivalent to a THERMAL POWER of less than 30% of RATED THERMAL POWER.

(k) Also actuates the EOC-RPT system.

(1)

DELETED (m) Each APRM channel provides inputs to both trip systems.

I I

TECH SPEC MARKUP INSERT 9:

(n)

DELETED (o)

With THERMAL POWER 2 25% RATED THERMAL POWER. The OPRM Upscale trip output shall be automatically enabled (not bypassed) when APRM Simulated Thermal Power is 2 30% and recirculation drive flow is < 60%. The OPRM trip output may be automatically bypassed when APRM Simulated Thermal Power is < 30% or recirculation drive flow is 2 60%.

(p)

A minimum of 23 cells, each with a minimum of 2 OPERABLE LPRMs, must be OPERABLE for an OPRM channel to be OPERABLE.

TABLE 3.3,1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES FUNCTIONAL UNIT I; Intermediate Range Monitors:

• a.

Neutron Flux - High

a. b.

Inoperative RESPONSE TIME (Seconds)

N.A.

H.A.

Aly-f

2. Average Power Range Monitor*:

a.

Neutron Flux - Upscale (Setdown)

b.

Simulated Thermal Power - Upscale

c.

Neutron Flux - Upscale

d.

Inoperative

e.

2-Out-Of-4 Voter N.A.

N.A.

N.A.

N.A.

s0.05*

3.

Reactor Vessel Steam Dome Pressure - High

&O.55

4. Reactor Vessel Water Level - Low, Level 3 g1.05#

5. Main Steam Line Isolation Valve - Closure A0.06 w

W an

('3

6. DELETED DELETED

7. Drywell Pressure - High

8.

Scram Discharge Volume Water Level - High

a.

Level Transmitter

b.

Float Switch N.A.

N.A.

N.A.

0 I.-

a

9. Turbine Stop Valve - Closure

10.

Turbine Control Valve Fast Closure, Trip Oil Pressure - Low

&O.06 K0.08**

N.A.

11.

Reactor Mode Switch Shutdown Position

12.

Manual Scram N.A.

r-a

• Neutron detectors, APRM channel and 2-Out-Of-4 Voter channel digital electronics are exempt from response time testing. Response time shall be measured from activation of the 2-Out-Of-4 Voter output relay.

For application of Specification 4.3.1.3, the redundant outputs from each 2-Out-Of-4 Voter channel are considered part of the same channel, I

• • Measured from start of turbine control valve fast closu

1. Sensor is eliminated from response time testing for the RPS circuits.

Response time testing and conformance to the administrative limits for the remaining channel including trip unit and relay logic are required.

TECH SPEC MARKUP INSERT 9A:

f.

OPRM Upscale N.A.

INSERT 9B:

but the OPRM and APRM outputs are considered to be separate channels, so N = 8. Testing of OPRM and APRM outputs shall alternate.

TABLE 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL FUNCTIONAL CHANNEL CHECK TEST CALIBRATION(z OPERATIONAL CONDITIONS FOR WHICH LI SURVEILLANCE REQUIRED I...

FUNCTIONAL UNIT

1. Intermediate Range Moni

a.

Neutron Flux -

tors:

High

b.

Inoperative

2.

Average Power Range Monitor(f):

a.

Neutron Flux - Upscale (Setdown)

b.

Simulated Thermal Power - Upscale

c.

Neutron Flux - Upscale X I

d.

Inoperative e

2-Out-Of-4 Voter

3.

Reactor Vessel Steam Dome Pressure - High

4.

Reactor Vessel Water Level-Low, Level 3

5.

Main Steam Line Isolation Valve - Closure

6.

DELETED

7.

Drywell Pressure - High

8.

Scram Discharge Volume Water Level - High

a. Level Transmitter

b. Float Switch S(b)

S N.A.

D(b)

D D

N.A.

D S

W W(j)

W(j)

SA(l)

SA(e)

SA SA SA a

R R

N.A.

R W(d),

W(d),

N.A.

N.A.

2 3(i), 4(i), 5(i) 2, 3(i), 4(i), 5(i),

R(g)

R 2

1 1

1, 2 1, 2 R

R 1, 2(h) 1, 2 S

N.A.

DELETED S

S.

N.A.

Q DELETED Q

R DELETED R

1 DELETED 1, 2 Q

Q R

R 1, 2, 5(i) 1, 2, 5(i)

LIMERICK UNIT I 3/4 3-7 Amendment No. 44, i3, 89, 149

TECH SPEC MARKUP INSERT 10:

f.

OPRM Upscale D

SA(e)

R(c)(g) 1(m)

TABLE 4.3.1.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH

'<FUNCTIONAL UNIT CHECK TEST CALIBRATION"'}

SURVEILLANCE REQUIRED c 9. Turbine Stop Valve - Closure N.A.

Q R

1

'10.

Turbine Control Valve Fast Closure, Trip Oil Pressure - Low N.A.

Q R

1

11.

Reactor Mode Switch Shutdown Position N.A.

R N.A.

1, 2, 3, 4, 5

12.

Manual Scram N.A.

W N.A.

1, 2, 3, 4, 5 (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

b 'The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after

• entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for a least 1/2 decades during each ntrolled shutdown, if not performed within the previous 7 days.

sd ThTf callbrtn consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER k25% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of-RATED THERMAL POWER.

(e) CHANNEL FUNCTIONAL TEST shall include the flow input function, excluding the flow transmitter.

(f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH).

(g) Calibration includes the flow input function.

• (h) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

(i) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(j) If the RPS shorting links are required to be removed per S ecification 3.9.2, they may be reinstalled for up to 2 a

r?

hours for required surveillance. During this time, CORE ALTERATIONS shall be suspended, and no control rod shall

- !4 be moved from its existing position.

(k) DELETED

>(l)

Not required to be performed when entering OPERATIONAL CONDITION 2 from OPERATIONAL CONDITION I until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering OPERATIONAL CONDITION 2.

A

'b=

D

TECH SPEC MARKUP INSERT 11:

Calibration includes verification that the OPRM Upscale trip auto-enable (not-bypass) setpoint for APRM Simulated Thermal Power is 2 30% and for recirculation drive flow is

< 60%.

INSERT 12:

(m)

With THERMAL POWER Ž 25% of RATED THERMAL POWER.

TABLE 3.3.6-1 CONTROL ROD BLOCK INSTRUMENTATION X3 MINIMUM APPLICABLE g

OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION PER TRIP FUNCTION CONDITIONS ACTION

1.

ROD BLOCK MONITOR<')

a.

Upscale 2

1*

60

b.

Inoperative 2

1*

60 C.

Downscale 2

1*

60

2.

APRM

a.

Simulated Thermal Power - Upscale 3

1 61

b.

Inoperative 3

1, 2 61

c.

Neutron Flux - Downscale 3

1 61

d.

Simulated Thermal Power - Upscale (Setdown) 3 2

61

e.

Recirculation Flow - Upscale 3

1 61

f.

LPRM Low Count 3

1, 2 61

3.

SOURCE RANGE MONITORS ***

X

a.

Detector not full in (b) 3 2

61 2

5 61

b.

Upscalefc) 3 2

61 2

5 61

c.

Inoperative"c) 3 2

61 2

5 61

d.

Downscale(d) 3 2

61 2

5 61

4.

INTERMEDIATE RANGE MONITORS 2^

6

a.

Detector not full in 6

2, '

61

b.

Upscale 6

2,

• 61

c.

Inoperative 6

2, 2,1*

61

d.

Downscale"(

6 2, 5 s 61

5.

SCRAM DISCHARGE VOLUME So

a.

Water Level-High 2

1, 2, 5**

62

6.

DELETED DELETED DELETED DELETED

7.

RATR'OESITC;HTONPSTO

,46 7 7.

REACTOR MODE SWITCH SHUTDOWN POSITION 2

3, 4 63

TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP SETPOINT 6-4 I-..

TRIP FUNCTION

1. ROD BLOCK MONITOR l.Upscale

• .~ 1) Low Trip Setpoint (LTSP)

2) Intermediate Trip Setpoint (ITSP)

3) High Trip Setpoint (HTSP)

b. Inoperative

c. Downscale (DTSP)

d. Power Range Setpointzb)

1) Low Power Setpoint (LPSP)

2) Intermediate Power Setpoint (IPSP) 3 High Power Setpoint (11PSP)

2. APRM

a. Simulated Thermal Power - Upscale:

Two Recirculation Loop Operation Single Recirculation Loop Operatio N/A 28.1% RATED THERMAL POWER 63.1% RATED THERMAL POWER 83.1% RATED THERMAL POWER s 0.66 W + 55.2% and i 108.0% of RATED THERMAl POWER eU &C.

r. Lto -a c& + SA.ED 7.

ALLOWABLE VALUE N/A 28.4X RATED THERMAL POWER 63.4% RATED THERMAL POWER 63.4% RATED THERMAL POWER W

(nJ al s 0.66 W + 55.7% and 5 108.4% of RATED TpJTML-M LF-r-

-' o.rcto-GX)+

S EMA.4L Of WR THERMAL POWER

-0 (b

Fin

1-0

'-P

-E 7%t 0

b. Inoperative

c. Neutron Flux - Downscale

d. Simulated Thermal Power -

Upscale (Setdown)

e. Recirculation Flow - Upscale

f. LPRM Low Count

3. SOURCE RANGE MONITORS

a. Detector not full in

b. Upscale

c. Inoperative

d. Downscale N.A.

2 3.2% of RATED THERMAL POWER s 12.0% of RATED THERMAL POWER

< 20 per channel 3 per axial level N.A.

s 1 x 105 cps N.A.

k 3 cps**

i 2.8% of RATED THERMAL POWER N.A.

g 13.0% of RATED THERMAL POWER

< 20 per channel

< 3 per axial level N.A.

S 1.6 x 10' cps N.A.

2 1.8 cps**

TABLE 33.6-2 (continued)

CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT

4.

INTERMEDIATE RANGE MONITORS

a. Detector not full in

b. Upscale c.

d.

Inoperative Downscale N.A.

! 108/125 divisions of full scale N.A.

2 5/125 divisions of full scale s 257' 5 9/16" elevation***

ALLOWABLE VALUE N.A.

S 110/125 divisions of full scale N.A.

2 3l25 divisions of full sca e 5~ 257' 7 9/16" elevation

5. SCRAM DISCHARGE VOLUME

a. Water Level-High

a. Float Switch

6. DELETED DELETED DELETED

7. REACTOR MODE SWITCH SHUTDOWN POSITION N.A.

N.A.

W

.Pk lal Cn Refer to the COLR for these setpoints.

May be reduced provided the Source Range above the curve shown in Figure 3.3.6-1.

Monitor has an observed count rate and signal-to-noise ratio on or Equivalent to 13 gallons/scram discharge volume.

(a)

There are three upscale trip levels. Each is applicable only over its specified operating core thermal power range.

All RBM trips are automatically bypassed below the low power setpoint (LPSP). The upscale LTSP is applied between the low power setpoint (LPSP) and the intermediate power setpoint (IPSP). The upscale ITSP is applied between the intermediate power setpoint and the high power setpoint (HPSP).

The HTSP is applied above the high power setpoint.

la 10 1:o (b)

Power range setpoints control enforcement of appropriate upscale trips over the proper core thermal power ranges. The power signal to the RBM is provided by the APRM.

TECH SPEC MARKUP INSERT 13:

The 7.6% flow "offset" for Single Loop Operation (SLO) is applied for W 2 7.6%.

For flows W < 7.6%, the (W-7.6%) term is set equal to zero.

'1 TA00E 4.3.6-1 CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL FUNCTIONAL CHANNEL CHECK TEST CALIBRATION'&

OPERATIONAL CONDITIONS FOR WHICH SURVEILLANCE REOUIRED r-m V-4

-4 TRIP FUNCTION

1. ROD BLOCK MONITOR a.

b.

C.

Upscale Inoperative Downscale N.A.

N.A.

N.A.

Q(C)

QCc)

Q(C)

R N.A.

R 1*

1*

1*

b-

2. APM

a. Simulated Thermal Power -

Upscale

b. Inoperative

c. Neutron Flux - Downscale

d. Simulated Thermal Power -

Upscale (Setdown)

e. Recirculation Flow - Upscale

f. LPRM Low Count N.

A.

N.A.

N.A.

N.

A.

N.

A.

N.A.

SA SA SA R

N.A.

R I

1, 2 1

2 1, 2 SA SA SA RR R

-N W

(/3

3. SOURCE RANGE MONITORS a.

b.

C.

d.

Detector not full in Upscale Inoperative Downscale N. A.

N.A.

N.A.

N.A.

N.A.

R N.A.

R 2,

2, 2,

2,

4. INTERMEDIATE RANGE MONITORS 5

55 S

• 5 1 I 5't sk V a.

b.

C.

d.

Detector not full in Upscale Inoperative Downscale N.A.

N.A.

N.A.

N.A.

14 14 14 14 N.A.

R N.A.

R 2,

2 2,9 2,9

5. SCRAM DISCHARGE VOLUME

-u

a. Water Level - High N.A.

Q DELETED R

DELETED 1, 2, 5**

DELETED

6. DELETED r',

=3

7. REACTOR MODE SWITCH SHUTDOWN C

POSITION DELETED I

M.

A.

N.

A.

3, 4

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation APPLICABILITY:

OPERATIONAL CONDITIONS 1* and 2*.

ACTION:

a.

With one reactor coolant system recirculation loop not in operation:

1.

Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

a.

Place the recirculation flow control system in the Local Manual mode, and

b.

Reduce THERMAL POWER to < 76.2% of RATED THERMAL POWER, and,

c.

Limit the speed of the operating recirculation pump to less than or equal to 90% of rated pump speed, and

d.

Verify that the differential temperature requirements of Surveillance Requirement 4.4.1.1.5 are met if THERMAL POWER is < 30% of RATED THERMAL POWER or the recirculation loop flow in the operating loop is S 50% of rated loop flow, or suspend the THERMAL POWER or recirculation loop flow increase.

I

• See Special Test Exception 3.10.4.

LIMERICK -

UNIT 1 3/4 4-1 Amendment No.

30,00,106 FEB 1 2 1996

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

(Continued)

2.

Within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s:

Reduce the Average Power Range Monitor (APRM) Simulated Thermal Power -

Upscale Scram and Rod Block Trip Setpoints and Allowable Values, to those applicable for single recirculation loop operation per Specifications 2.2.1 and 3.3.6, or declare the associated channel(s) inoperable and take the actions required by the referenced specificatios M

3.

Otherwise be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ieration, initiate measures to

)HOT SHUTDOWN within LIMERICK - UNIT I 3/4 4-la Amendment No. 3a, 6s, 4-4, 169

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.1.1 DELETED 4.4.1.1.2 DELETED 4.4.1.1.3l e

M a L

ntr fl isea thy T SbreionZ og c/ Xo gnXsr Xui

?D fi e Zis 3

, X agpr~i0s pe M rbe re"/ionX i et s

i 4.4.1.1.4 With one reactor coolant system recirculation loop not-in operation, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verify that:

a.

Reactor.THERMAL POWER is s 76.2% of RATED THERMAL POWER,

b.

The recirculation flow control system is in the Local Manual mode, and

c.

The speed of the operating recirculation pump is

• 90% of rated pump speed.

i: tac ponef e 3,f.fg-K/(

V

/( w G o' 4.4.1.1.5 With one reactor coolant system recirculation loop not in operation, within 15 minutes prior to either THERMAL POWER increase or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is s 30% of RATED THERMAL POWER or the recirculation loop flow in the operating recirculation loop is s 50% of rated loop flow.

a.

s 1450F between reactor vessel steam space coolant and bottom head drain line coolant,

b.

s 50'F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and C.

5 50'F between the reactor coolant within the loop not in operation and the operating loop.

The differential temperature requirements of Specification 4.4.1.1.5b. and c.

do not apply when the loop not in operation is isolated from the reactor pressure vessel.

LIMERICK - UNIT I 3/4 4-2 Amendment No. 3,39,'4,76,77,406,142

I-rn 7) z LkI

-'Irye

-aco K

no a'

3/4.3 INSTRUMENTATION t..,

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

a.

Preserve the integrity of the fuel cladding.

b.

Preserve the integrity of the reactor coolant system.

c.

Minimize the energy which must be adsorbed following a loss-of-coolant accident, and

d.

Prevent inadvertent criticality.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.

The reactor protection system is made up of two independent trip systems.

There are usually four channels to monitor each parameter with two channels in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram.

The APRM system is divided into four APRM channels and four 2-Out-Of-4 Voter channels.

Each APRM channel provides inputs to each of the four voter channels.

The four voter channels are divided into two groups of two each, with each group of two providing inputs to one RPS trip system. The system is designed to allow one APRM channel, but no voter channels, to be bypassed.

The system meets the intent of IEEE-279 for nuclear power plant protection systems. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System" and NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function."

The bases for the trip settings of the RPS are discussed in the bases for Specification 2.2.1.

6~Acctions a, b and c define th(

(s) required when RPS channels are discovered to be inoperable.

For thsE. 4~k ions, separate entry condition is allowed for each inp ble RPS channel.

Separ e entry means that the allowable time clock(s) fq¶ Options a, b or c start upon discovery of inoperability for that specific ch'T Fel.

Restoration of an inoperable RPS channel satisfies only the action statements for that particular channel.

Action statement(s) for remaining inoperable channel(s) must be met according to their original entry time.

Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> has been shown to be acceptable (NEDC-30851P-A and NEDC-32410P-A) to permit restoration of any inoperable channel to OPERABLE status. However, this out of service time is only acceptable provided that the associated Function's (identified as a "Functional Unit" in Table 3.3.1-1) inoperable channel is in one trip system and the Function still maintains RPS trip capability.

V..

LIMERICK - UNIT 1 B 3/4 3-1 Amendment No.

$0, Z3Z 141 APR t 2 2GCJ

TECH SPEC MARKUP INSERT 14:

The APRM Functions include five Functions accomplished by the four APRM channels (Functions 2.a, 2.b, 2.c, 2.d, and 2.f) and one accomplished by the four 2-Out-Of-4 Voter channels (Function 2.e). Two of the five Functions accomplished by the APRM channels are based on neutron flux only (Functions 2.a and 2.c), one Function is based on neutron flux and recirculation drive flow (Function 2.b) and one is based on equipment status (Function 2.d). The fifth Function accomplished by the APRM channels is the Oscillation Power Range Monitor (OPRM) Upscale trip Function 2.f, which is based on detecting oscillatory characteristics in the neutron flux. The OPRM Upscale Function is also dependent on average neutron flux (Simulated Thermal Power) and recirculation drive flow, which are used to automatically enable the output trip.

The Two-Out-Of-Four Logic Module includes 2-Out-Of-4 Voter hardware and the APRM Interface hardware. The 2-Out-Of-4 Voter Function 2.e votes APRM Functions 2.a, 2.b, 2.c, and 2.d independently of Function 2.f. This voting is accomplished by the 2-Out-Of-4 Voter hardware in the Two-Out-Of-Four Logic Module. The voter includes separate outputs to RPS for the two independently voted sets of Functions, each of which is redundant (four total outputs). The analysis in Reference 2 took credit for this redundancy in the justification of the 12-hour allowed out-of-service time for Action b, so the voter Function 2.e must be declared inoperable if any of its functionality is inoperable. The voter Function 2.e does not need to be declared inoperable due to any failure affecting only the APRM Interface hardware portion of the Two-Out-Of-Four Logic Module.

Three of the four APRM channels and all four of the voter channels are required to be OPERABLE to ensure that no single failure will preclude a scram on a valid signal. To provide adequate coverage of the entire core, consistent with the design bases for the APRM Functions 2.a, 2.b, and 2.c, at least 20 LPRM inputs, with at least three LPRM inputs from each of the four axial levels at which the LPRMs are located, must be OPERABLE for each APRM channel. In addition, no more than 9 LPRMs may be bypassed between APRM calibrations (weekly gain adjustments). For the OPRM Upscale Function 2.f, LPRMs are assigned to 'cells" of 3 or 4 detectors. A minimum of 23 cells (Reference 9), each with a minimum of 2 OPERABLE LPRMs, must be OPERABLE for each APRM channel for the OPRM Upscale Function 2.f to be OPERABLE in that channel.

TECH SPEC MARKUP INSERT 14 (continued):

References 4, 5 and 6 describe three algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. All three are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period based detection algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY for Technical Specification purposes is based only on the period based detection algorithm.

An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in any cell, with period confirmations and relative cell amplitude exceeding specified setpoints. One or more cells in a channel exceeding the trip conditions will result in a channel trip. An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithms detect growing oscillatory changes in the neutron flux for one or more cells in that channel.

The OPRM Upscale Function is required to be OPERABLE when the plant is at >

25% RATED THERMAL POWER. The 25% RATED THERMAL POWER level is.

selected to provide margin in the unlikely event that a reactor power increase transient occurring while the plant is operating below 30% RATED THERMAL POWER causes a power increase to or beyond the 30% RATED THERMAL POWER OPRM Upscale trip auto-enable point without operator action. This OPERABILITY requirement assures that the OPRM Upscale trip automatic-enable function will be OPERABLE when required.

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

The requirements of Action a are intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same trip system for the same Function result in the Function not maintaining RPS trip capability. A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal.

For the typical Function with one-out-of-two taken twice logic, including the IRM Functions and APRM Function 2.e (trip capability associated with APRM Functions 2.a, 2.b, 2.c,<g3)2.dYe discussed below), this would require both trip systems to have one channel OP in trip (or the associated trip system in trip).

For Function 5 (Main Steam Isolation Valve--Closure), this would require both trip systems to have each channel associated with the MSIVs in three main steam lines (not necessarily the same main steam lines for both trip systems) OPERABLE or in trip (or the associated trip system in trip).

For Function 9 (Turbine Stop Valve-Closure), this would require both trip systems to have three channels, each OPERABLE or in trip (or the associated trip system in trip).

The completion time to satisfy the requirements of Action a is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.

With trip capability maintained, i.e., Action a satisfied, Actions b and c as applicable must still be satisfied. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Action b requires that the channel or the associated trip system must be placed in the tripped condition.

Placing the inoperable channel in trip (or the associated trip system in trip) would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to co s,

Li or2 As noted, placing the trip system ns not applicable to satisfy Action b for APRM Functions 2.a, 2.b, 2:c,,)2.(. Inoperability of one required APRM channel affects both trip systems.

For that condition, the Action b requirements can only be satisfied by placing the inoperable APRM channel in trip.

Restoring OPERABILITY or placing the inoperable APRM channel in trip are the only actions that will restore capability to accommodate a single APRM channel failure.

Inoperability of more than one required APRM channel of the same trip function results in loss of trip capability and the requirement to satisfy Action a.

The requirements of Action c must be satisfied when, for any one or more Functions, at least one required channel is inoperable in each trip system.

In this condition, provided at least one channel per trip system is OPERABLE, normally the RPS still maintains trip capability for that Function, but cannot accommodate a single failure in either trip system (see additional bases discussion above related to loss of trip capability and the requirements of Action a, and special cases for Functions 2.a, 2.b, 2.c, 2.d, and 9).

APR 12 2ODS LIMERICK - UNIT 1 B 3/4 3-la Amendment No.

141 1

314.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

The requirements of Action c limit the time the RPS scram logic, for any Function, would not accommodate single failure in both trip systems (e.g., one-out-of-one and one-out-of-one arrangement for a typical four channel Function).

The reduced reliability of this logic arrangement was not evaluated in NEDC-30851P-A for the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time. Within the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance, the associated Function must have all required channels OPERABLE or in trip (or any combination) in one trip system.

Completing the actions required by Action c restores RPS to a reliability level equivalent to that evaluated in NEDC-30851P-A, which justified a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowable out of service time as allowed by Action b. To satisfy the requirements of Action c, the trip system in the more degraded state should be placed in trip or, alternatively, all the inoperable channels in that trip system should be placed in trip (e.g., a trip system with two inoperable channels could be in a more degraded state than a trip system with four inoperable channels if the two inoperable channels are in the same Function while the four inoperable channels are all in different Functions).

The decision of which trip system is in the more degraded state should be based on prudent judgment and take into account current plant conditions (i.e.,

what OPERATIONAL CONDITION the plant is in).

If this action would result in a scram or RPT, it is permissible to place the other trip system or its inoperable channels in trip.

The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowable out of service time is judged acceptable based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram.

As noted, Action c is not applicable for APRM Functions 2.a, 2.b, 2.c, 2

Inoperability of an APRM channel affects both trip systems and is not associated with a specific trip system as are the APRM 2-Out-Of-4 voter and other non-APRM channels for which Action c applies.

For an inoperable APRM channel, the requirements of Action b can only be satisfied by tripping the inoperable APRM channel. Restoring OPERABILITY or placing the inoperable APRM channel in trip are the only actions that will restore capability to accommodate a single APRM channel failure.

If it is not desired to place the channel (or trip system) in trip to satisfy the requirements of Action a, Action b or Action c (e.g., as in the case where placing the inoperable channel in trip would result in a full scram), Action d requires that the Action defined by Table 3.3.1-1 for the applicable Function be initiated immediately upon expiration of the allowable out of service time.

7o-t- -F r ogic Mo le i clu es out f-4 ot r h rd re nd RN fce ar ar.

evter u tio 2.e s com ish b th 2-ut-f-4 otF d r wh' h clu es du an out ts./ Th ana sis n DC 24 OP-to c di r Ifs r dun anc in }Me stica ion or Ie 1hou al owa e ut f sIvi fi eso theote Fu tio 2. mus be ecl ed opeabl if nyof e

0out/of4 t hrdw e'sfun ion it is 0nopabl I

vo er unco 02.e doe/not/n d o e /eclred nap abl d to ny ilur affctigo y e PRNInt fac /

rd{are forti nog the/wout-f-Fr Lo icMdul APR 12 2000 LIMERICK - UNIT I B 3/4 3-lb Amendment No. 141 1

TECH SPEC MARKUP INSERT 15:

Table 3.3.1-1, Function 2.f, references Action 10, which defines the action required if OPRM Upscale trip capability is not maintained. Action 1 Ob is required to address identified equipment failures. Action 1 Oa is to address common mode vendor/industry identified issues that render all four OPRM channels inoperable at once. For this condition, References 2 and 3 justified use of alternate methods to detect and suppress oscillations for a limited period of time, up to 120 days. The alternate methods are procedurally established consistent with the guidelines identified in Reference 7 requiring manual operator action to scram the plant if certain predefined events occur. The 12-hour allowed completion time to implement the alternate methods is based on engineering judgment to allow orderly transition to the alternate methods while limiting the period of time during which no automatic or alternate detect and suppress trip capability is formally in place. The 120-day period during which use of alternate methods is allowed is intended to be an outside limit to allow for the case where design changes or extensive analysis might be required to understand or correct some unanticipated characteristic of the instability detection algorithms or equipment.

The evaluation of the use of alternate methods concluded, based on engineering judgment, that the likelihood of an instability event that could not be adequately handled by the alternate methods during the 120-day period was negligibly small. Plant startup may continue while operating within the allowed completion time of Action 10a. The primary purpose of this is to allow an orderly completion, without undue impact on plant operation, of design and verification activities in the event of a required design change to the OPRM Upscale function. This exception is not intended as an alternative to restoring inoperable equipment to OPERABLE status in a timely manner.

Action 1 Oa is not intended and was not evaluated as a routine alternative to returning failed or inoperable equipment to OPERABLE status. Correction of routine equipment failure or inoperability is expected to be accomplished within the completion times allowed for LCO 3.3.1 Action a or Action b, as applicable. Action 1 Ob applies when routine equipment OPERABILTY cannot be restored within the allowed completion times of LCO 3.3.1 Actions a or b, or if a common mode OPRM deficiency cannot be corrected and OPERABILTY of the OPRM Upscale Function restored within the 120-day allowed completion time of Action 1Oa.

TECH SPEC MARKUP INSERT 15 (continued):

The OPRM Upscale trip output shall be automatically enabled (not-bypassed) when APRM Simulated Thermal Power is 2 30% and recirculation drive flow is < 60%

as indicated by APRM measured recirculation drive flow. NOTE: 60% recirculation drive flow is the recirculation drive flow that corresponds to 60% of rated core flow. This is the operating region where actual thermal-hydraulic instability and related neutron flux oscillations may occur. As noted in Table 4.3.1.1-1, Note c, CHANNEL CALIBRATION for the OPRM Upscale trip Function 2.f includes confirming that the auto-enable (not-bypassed) setpoints are correct. Other surveillances ensure that the APRM Simulated Thermal Power properly correlates with THERMAL POWER (Table 4.3.1.1-1, Note d) and that recirculation drive flow properly correlates with core flow (Table 4.3.1.1-1, Note g).

If any OPRM Upscale trip auto-enable setpoint is exceeded and the OPRM Upscale trip is not enabled, i.e., the OPRM Upscale trip is bypassed when APRM Simulated Thermal Power is 2 30% and recirculation drive flow is < 60%, then the affected channel is considered inoperable for the OPRM Upscale Function.

Alternatively, the OPRM Upscale trip auto-enable setpoint(s) may be adjusted to place the channel in the enabled condition (not-bypassed). If the OPRM Upscale trip is placed in the enabled condition, the surveillance requirement is met and the channel is considered OPERABLE.

As noted in Table 4.3.1.1-1, Note g, CHANNEL CALIBRATION for the APRM Simulated Thermal Power - Upscale Function 2.b and the OPRM Upscale Function 2.f, includes the recirculation drive flow input function. The APRM Simulated Thermal Power - Upscale Function and the OPRM Upscale Function both require a valid drive flow signal. The APRM Simulated Thermal Power-Upscale Function uses drive flow to vary the trip setpoint. The OPRM Upscale Function uses drive flow to automatically enable or bypass the OPRM Upscale trip output to RPS. A CHANNEL CALIBRATION of the APRM recirculation drive flow input function requires both calibrating the drive flow transmitters and establishing a valid drive flow / core flow relationship. The drive flow / core flow relationship is established once per refuel cycle, while operating within 10% of rated core flow and within 10% of RATED THERMAL POWER. Plant operational experience has shown that this flow correlation methodology is consistent with the guidance and intent in Reference 8. Changes throughout the cycle in the drive flow / core flow relationship due to the changing thermal hydraulic operating conditions of the core are accounted for in the margins included in the bases or analyses used to establish the setpoints for the APRM Simulated Thermal Power - Upscale Function and the OPRM Upscale Function.

TECH SPEC MARKUP INSERT 15 (continued):

As noted in Table 3.3.1-2, Note "*", the redundant outputs from the 2-Out-Of-4 Voter channel are considered part of the same channel, but the OPRM and APRM outputs are considered to be separate channels, so N = 8 to determine the interval between tests for application of Specification 4.3.1.3 (REACTOR PROTECTION SYSTEM RESPONSE TIME). The note further requires that testing of OPRM and APRM outputs shall be alternated.

Each test of an OPRM or APRM output tests each of the redundant outputs from the 2-Out-Of-4 Voter channel for that function, and each of the corresponding relays in the RPS. Consequently, each of the RPS relays is tested every fourth cycle. This testing frequency is twice the frequency justified by References 2 and 3.

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

Automatic reactor trip upon receipt of a high-high radiation signal from the Main Steam Line Radiation Monitoring System was removed as the result of an analysis performed by General Electric in NEDO-31400A. The NRC approved the results of this analysis as documented in the SER (letter to George J. Beck, BWR Owner's Group from A.C. Thadani, NRC, dated May 15, 1991).

The measurement of respopse lome at the specified frequencies provides assurance that the protective unctions associated with each channel are completed within the time lim assumed in the safety analyses.

No credit was taken for those channels wit response times indicated as not applicable except for APRM Simulated Thermal Power U scale and Neutron Flux - Upscale trip functions (Table 3.3.1-2, Items 2.b Response time may be demonstrated by any series of sequential, overlapping or otal channel test measurement, provided such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times.

Response time testing for the sensors as noted in Table 3.3.1-2 is not required based on the analysis in NEDO-32291-A. Response time testing for the remaining channel components is required as noted.

For the digital electronic portions of the APRM functions, performance characteristics that determine response time are checked by a combination of automatic self-test, calibration activities, and response time tests of the 2-Out-Of-4 Voter (Table 3.3.1-2, Item 2.e).

• ~g.A R 2 o LIMERICK - UNIT I B 3/4 3-1c Amendment No. 141 1

INSTRUMENTATION BASES 3/4.3.7.10 (Deleted)

I 3/4.3.7.11 (Deleted) -

INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.

3/4.3.7.12 OFFGAS MONITORING INSTRUMENTATION This instrumentation includes provisions for monitoring the concentrations of potentially explosive gas mixtures and noble gases in the off-gas system.

3/4.3.8 (Deleted) -

INFORMATION FROM THIS SECTION RELOCATED TO THE UFSAR.

3/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION The feedwater/main turbine trip system actuation instrumentation is provided to initiate action of the feedwater system/main turbine trip system in the event of failure of feedwater controller under maximum demand.

LIMERICK - UNIT 1 B 3/4 3-7 Amendment No. 33, 48, 04, 4-0, 1-04, 153

TECH SPEC MARKUP INSERT 16:

REFERENCES:

1.

NEDC-30851 P-A, 'Technical Specification Improvement Analyses for BWR Reactor Protection System", March 1988.

2.

NEDC-3241OP-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option IlIl Stability Trip Function", October 1995.

3.

NEDC-3241 OP-A, Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option IlIl Stability Trip Function", November 1997.

4.

NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

5.

NEDO-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

6.

NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

7.

Letter, L. A. England (BWROG) to M. J. Virgilio, "BWR Owners' Group Guidelines for Stability Interim Corrective Action", June 6,1994.

8.

GE Service Information Letter No. 516, "Core Flow Measurement - GE BWR/3, 4, 5 and 6 Plants", July 26,1990.

9.

GE Letter NSA 00-433, Alan Chung (GE) to Sujit Chakraborty (GE),

"Minimum Number of Operable OPRM Cells for Option IlIl Stability at Limerick 1 & 2," May 02, 2001.

,J/4.4.tKLQIUK LUULANI SbYItM BASES 3/4.4.1 RECIRCULATION SYSTEM The impact of single recirculation loop operation upon plant safety is assessed and shows that single-loop operation is permitted if the MCPR fuel cladding safety limit is increased as noted by Specification 2.1.2, APRM scram and control. rod block setpoints are adjusted as noted in Tables 2.2.1-1 and 3.3.6-2, respectively.

Additionally, surveillance on the pump speed of the operating recirculation loop is imposed to exclude the possibility of excessive internals vibration.

The surveillance on differential temperatures below 30% RATED THERMAL POWER or 50% rated recirculation loop flow is to mitigate the undue thermal stress on vessel nozzles, recirculation pump and vessel bottom head during the extended operation of the single recirculation loop mode.

An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable. Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation.

Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteria for two recirculation loop operation. The limits will ensure an adequate core flow coastdown from either recirculation loop.

following a LOCA. In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single recir-culation loop.mode.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50F of each other prior to startup of an idle loop. The loop temperature must also be within 50F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Sudden equalization of a temperature difference > 145-F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.

Th obVctiv of E BW plani and el sig is t pr vide stabl opera-t n w h rgin ver he rmal peratng mai How er at e W h wer/low rlow orr of(he era ng do in, sma pr abiV y li t cyae utro fl os la ons xist depe ing co inatons o op ati g coitiins /

.g. rod att n, p er s pe).

o p ovid assur nc tha neut n f ux I it ycl os la ons e det ted ds pres d,A L

netro flux no' e 1 eelsm'houl be itor whi op atm in is egio St ility ests t op atin WRs were vi ed de rmi a g eri re ion the ower/ ow p in hich urve&la e ofaneut n f x no se 1 vels oul be p forme.

A onser tive decay rat' of.6 w s ch sen the base or eter f ing e ge nc r gion or s vei anc to a coun for he ant o pI t var abili of ecay tio th c re a d fu des gns. Thi gen ic gio h bee dete ned o cor espon to core/flow/of 1 s t n or kqualto

% o ated ore f w an a THE MAL P ER eate tha tha spec fied n Fi ure.4..1-1 LIMERICK - UNIT 1 B 3/4 4-1 Amendment No. ;0 66 FEB 10 1994

REACTOR COOLANT SYSTEM BASES f ECRC(ATM0 SYSTJM(Cosnuedd, t

ange sl ofic gcu ons tio per o d tgbdete an cadle re i rcmul neuion1epoperion.

eut f

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in korin-to ie reneutn wfx s

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lngle ond al rercul on th ope

n. eura fl wnoisea vels wifich Siga lin plyi pun.

om t s an d era thing rmants echanite aesit rof GE B atlel Id orefwwl st pr ae 50%. In easeiti etabil ux shou bev tbnnate m imumr lineahiht ajr ofpetn ietut Howeatin Rs hb e da ed tta when o litn eiated 1 eutron r pro oamipt cycleveillion Ult in kt-ak neut flux a

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propof onal t5tepowe ve a a gie core 'fow.

314. 4.2 SAFETY /REL IEF VALVES The safety valve function of the safety/relief valves operates to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code. A total of 12 OPERABLE safety/

relief valves is required to limit reactor pressure to within ASME III allow-able values for the worst case upset transient.

Demonstration of the safety/relief valve lift settings will occur only during shutdown. The safety/relief valves will be removed and either set pressure tested or replaced with spares which have been previously set pres-sure tested and stored in accordance with manufacturers recommendations in the specified frequency.

Corrected by Ltr.-dated 3/l0t00 LIMERICK - UNIT I B 3/4 4-2 Amendment No. 34, 137

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.9.1.9 Core Operating Limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and.shall be documented in the CORE OPERATING LIMITS REPORT for the following:

a.

The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1,

b.

MAPFAC(P) and MAPFAC(F) factors for Specification 3.2.1,

c.

The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.3,

d.

The MCPR(P) and MCPR(F) adjustment factors for specification 3.2.3,

e.

The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4,

f.

The power biased Rod Block Monitor setpoints and the Rod Block Monitor MCPR OPERABILITY limits of Specification 3.3.6,

9.

The Reactor Coolant System Recirculation Flow upscale trip setpoint and allowable value for Specification 3:3.6, 6.9.1.10 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document-lo

a.

NEDE-24011-P-A "General Electric Standard Application for Reactor Fuel" (Latest approved revisio 7!*

6.9.1.11 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.

6.9.1.12 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator'and Resident Inspector.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

• For Cycle 8, specific documents were approved in the Safety Evaluation dated (5/4/98) to support License Amendment No. (127).

MAY o a 20,o LIMERICK - UNIT 1 6-18a Amendment No.127, 142

TECH SPEC MARKUP INSERT 17:

h.

The Oscillation Power Range Monitor (OPRM) period based detection algorithm (PBDA) setpoints for Specification 2.2.1.

INSERT 18:

b.

NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

TECH SPEC MARKUP for Stability (PRNM) OPRM Trip Activation MOD at LIMERICK UNIT 2

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE REACTOR COOLANT SYSTEM (Continued)

Figure 3.4.1.1

-1/e/F 3/4 4-3 Jet Pumps................................................... 3/4 4-4 Recirculation Pumps................

......................... 3/4 4-5 Idle Recirculation Loop Startup........

..................... 3/4 4-6 3/4.4.2 SAFETY/RELIEF VALVES.

......................................... 3/4 4.-7 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems...........

......................... 3/4 4-8 Operational Leakage................

......................... 3/4 4-9 Table 3.4.3.2-1 Reactor Coolant System Pressure Isolation Valves......

................ 3/4 4-11 3/4.4.4 CHEMISTRY................................................... 3/4 4-12 Table 3.4.4-1 Reactor Coolant System Chemistry Limits..........

........................ 3/4 4-14 3/4.4.5 SPECIFIC ACTIVITY........................................... 3/4 4-15 Table 4.4.5-1 Primary Coolant Specific Activity Sample and Analysis Program............. 3/4 4-17 3/4.4.6 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System............

.......................... 3/4 4-18 Figure 3.4.6.1-1 Minimum Reactor Pressure Vessel Metal Temperature Vs. Reactor Vessel Pressure......

................ 3/4 4-20 Table 4.4.6.1.3-1 Deleted........

..................... 3/4 4-21 Reactor Steam Dome...............

3/4 4-22 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES............................. 3/4 4-23 3/4.4.8 STRUCTURAL INTEGRITY..........

................... 3/4 4-24 LIMERICK - UNIT 2 xi Amendment No. 130

r-

'_4 C-)

TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS c

FUNCTIONAL UNIT TRIP SETPQIN

1. Intermediate Range Monitor, Neutron Flux-High

2. Average Power Range Monitor:

a.

Neutron Flux-Upscale (Setdown)

b.

Simulated Thermal Power - Upscale:

Two Recirculation Loop Operation Single Recirculation Loop OperatioA

c. Neutron Flux - Upscale

d.

Inoperative re. 2-Out-Of-4 Voter

3. Reactor Vessel Steam Dome Pressure - High

4.

Reactor Vessel Water Level - Low, Level 3 120/125 divisions of full scale

• 15.0% of RATED THERMAL POWER 0.66 W + 62.8% and 116.6% of RATED THERMAL POWER anu s 116.6% of RATED THERMAL POWER 118.3% of RATED THERMAL POWER N.A.

N.A.

s 1096 psig 2 12.5 inches above instrume zero*

s 8% closed DELETED s 1.68 psig 5 261' 1 1/4" elevation**

s 261' 1 1/4" elevation**

ALLOWABLE VALUES s 122/125 divisions of full scale s 20.0% of RATED THERMAL POWER s 0.66 W +: 63.3% and

• 117.0% of RATED THE L

,6=~~

o

. c o~v two) 633o

(-3<:5hb~dEtr*-3Eand

~ 1.0% of RAED THERMAL POWER 118.7% of RATED THERMAL POWER N.A.

N.A.

S 1103 psig nt k 11.0 inches above instrument zero s 12% closed DELETED s 1.88 psig S 261' 9 1/4". elevation s 261' 9 1/4" elevation 5.

6.

7.

8.

Main Steam Line Isolation Valve - Closure DELETED Drywell Pressure - High Scram Discharge Volume Water Level - High

a.

Level Transmitter

b.

Float Switch rrI 0-1 n

. -1 n

t )

I

• See Bases Figure B 3/4.3-1.

Hi quivalent to 25.58 gallons/scram discharge voluti

TECH SPEC MARKUP INSERT 1:

f.

OPRM Upscale N.A.

INSERT 2:

The 7.6% flow "offset" for Single Loop Operation (SLO) is applied for W 2 7.6%. For flows W < 7.6%, the (W-7.6%) term is set equal to zero.

See COLR for OPRM period based detection algorithm trip setpoints. OPRM Upscale trip output auto-enable (not bypassed) setpoints shall be APRM Simulated Thermal Power 2 30% and recirculation drive flow < 60%.

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System instrumentation setpoints specified in Table 2.2.1-1 are the values at which the reactor trips are set for each para-meter.

The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist in mitigating the consequences of accidents. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

1. Intermediate Range Monitor. Neutron Flux -

High The IRM system consists of 8 chambers, 4 in each of the reactor trip systems.

The IRM is a 5 decade 10 range instrument. The trip setpoint of 120 divisions of scale is active in each of the 10 ranges. Thus as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up.. The IRM instruments provide for overlap with both the APRM and SRM systems.

The most significant source of reactivity changes during the power increase is due to control rod withdrawal.

In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed.

The results of these analyses are in Section 15.4 of the FSAR.

The most severe case involves an initial condition in which THERMAL POWER is at approximately 1% of RATED THERMAL POWER. Additional conservatism was taken in this analysis by assuming the IRM channel closest to the control rod being withdrawn is bypassed. The results of this analysis show that the reactor is shutdown and peak power is limited to 21% of RATED THERMAL POWER with the peak fuel entha py well below the fuel failure threshold of 170 cal/gm.

Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

2.

Average Power Range Monitor The APRM system is divided into four APRM channels and four 2-Out-Of-4 Voter channels. The four voter channels are divided into two groups of two each, with each group of two providing inputs to one RPS trip system.

All four voters will trip (full scram) when any twoYAPRM chann exceed their trip setpoints.

For operation at low pressure an ow Tow during STARTUP, the APRM Neutron Flux-Upscale (Setdown) scram setting of 15% of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits.

The margin accommodates the anticipated maneuvers associated with power plant startup.

Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system. Teippera-ture coefficients are small and control rod patterns are constrained by the RWM. Of all the possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power increase.

LIMERICK - UNIT 2 B 2'-6 AI~ht Mb. 109

TECH SPEC MARKUP INSERT 3:

APRM trip Functions 2.a, 2.b, 2.c, and 2.d are voted independently from OPRM Upscale Function 2.f. Therefore, any Function 2.a, 2.b, 2.c, or 2.d trip from any two unbypassed APRM channels will result in a full trip in each of the four voter channels.

Similarly, a Function 2.f trip from any two unbypassed APRM channels will result in a full trip from each of the four voter channels.

LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Average Power Range Monitor (Continued)

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant amount, the rate of power rise is very slow.

Generally the heat flux is in near equilibrium with the fission rate.

In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit.

The 15% Neutron Flux - Upscale (Setdown) trip remains active until the mode switch is placed in the Run position.

The APRM trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the system and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Neutron Flux - Upscale setpoint; i.e.,

for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel.

For the Simulated Thermal Power -

Upscale setpoint, a time constant of 6 t 0.6 seconds is introduced into the flow-biased APRM in order to simulate the fuel thermal transient characteristics. A more conservative maximum value is used for the flow-biased setpoint as shown in Table 2.2.1-1.

The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unneces-sary shutdown.

3.

Reactor Vessel Steam Dome Pressure-High High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient. This trip setpoint is effective at low power/flow conditions when the turbine stop valve and control fast closure trips are bypassed.

For a turbine trip or load rejection under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.

LIMERICK - UNIT 2 B 2-7 A~nayhmt Nj.- 48-7 109)

TECH SPEC MARKUP INSERT 4:

A reduced Trip Setpoint and Allowable Value is provided for the Simulated Thermal Power - Upscale Function, applicable when the plant is operating in Single Loop Operation (SLO) per LCO 3.4.1.1. In SLO, the drive flow values (W) used in the Trip Setpoint and Allowable Value equations is reduced by 7.6%. The 7.6% value is established to conservatively bound the inaccuracy created in the core flow/drive flow correlation due to back flow in the jet pumps associated with the inactive recirculation loop. The Trip Setpoint and Allowable Value thus maintain thermal margins essentially unchanged from those for two-loop operation. The Trip Setpoint and Allowable Value equations for single loop operation are only valid for flows down to W = 7.6%. The Trip Setpoint and Allowable Value do not go below 62.8% and 63.3% RATED THERMAL POWER, respectively. This is acceptable because back flow in the inactive recirculation loop is only an issue with drive flows of approximately 40% or greater (Reference 1).

INSERT 5:

The APRM channels also include an Oscillation Power Range Monitor (OPRM)

Upscale Function. The OPRM Upscale Function provides compliance with GDC 10 and GDC 12, thereby providing protection from exceeding the fuel MCPR Safety Limit due to anticipated thermal-hydraulic power oscillations. The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into "cells" for evaluation by the OPRM algorithms.

References 2, 3 and 4 describe three algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. All three are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period based detection algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY for Technical Specification purposes is based only on the period based detection algorithm.

The OPRM Upscale trip output shall be automatically enabled (not bypassed) when APRM Simulated Thermal Power is > 30% and recirculation drive flow is < 60%

as indicated by APRM measured recirculation drive flow. (NOTE: 60% recirculation drive flow is the recirculation drive flow that corresponds to 60% of rated core flow.

Refer to TS Bases 3/4.3.1 for further discussion concerning the recirculation drive flow/core flow relationship.) This is the operating region where actual thermal-hydraulic instability and related neutron flux oscillations may occur. See Reference 5 for additional discussion of OPRM Upscale trip enable region limits. These setpoints, which are sometimes referred to as the "auto-bypass" setpoints, establish the boundaries of the OPRM Upscale trip enabled region. The APRM Simulated Thermal Power auto-enable setpoint has 1 % deadband while the drive flow setpoint has a 2%

deadband. The deadband for these setpoints is established so that it increases the enabled region.

TECH SPEC MARKUP INSERT 5 (continued):

An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in a cell, with period confirmations and relative cell amplitude exceeding specified setpoints. One or more cells in a channel exceeding the trip conditions will result in a channel trip. An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithms detect oscillatory changes in the neutron flux for one or more cells in that channel.

There are four "sets" of OPRM related setpoints or adjustment parameters:

a) OPRM trip auto-enable setpoints for APRM Simulated Thermal Power (30%) and recirculation drive flow (60%); b) period based detection algorithm (PBDA) confirmation count and amplitude setpoints; c) period based detection algorithm tuning parameters; and d) growth rate algorithm (GRA) and amplitude based algorithm (ABA) setpoints.

The first set, the OPRM auto-enable region setpoints, are treated as nominal setpoints with no additional margins added as discussed in Reference 5. The settings, 30% APRM Simulated Thermal Power and 60% recirculation drive flow, are defined (limit values) in a note to Table 2.2.1-1. The second set, the OPRM PBDA trip setpoints, are established in accordance with methodologies defined in Reference 4, and are documented in the COLR. There are no allowable values for these setpoints.

The third set, the OPRM PBDA "tuning" parameters, are established or adjusted in accordance with and controlled by station procedures. The fourth set, the GRA and ABA setpoints, in accordance with References 2 and 3, are established as nominal values only, and controlled by station procedures.

LIMITING SAFETY SYSTEM SETTING BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

8. Scram Discharge Volume Water Level-High The scram discharge volume receives the water displaced by the motion of the control rod drive pistons during a reactor scram.

Should this volume fill up to a point where there is insufficient volume to accept the displaced water at pressures below 65 psig, control rod insertion would be hindered.

The reactor is therefore tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough to accommodate the water from the movement of the rods at pressures below 65 psig when they are tripped.

The trip setpoint for each scram discharge volume is equivalent to a contained volume of 25.58 gallons of water.

9. Turbine Stop Valve-Closure The turbine stop valve closure trip anticipates the pressure, neutron flux, and heat flux increases that would result from closure of the stop valves. With a trip setting of 5% of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained during the worst design basis transient.

10.

Turbine Control Valve Fast Closure, Trip Oil Pressure-Low The turbine control valve fast closure trip anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of-the turbine control valves due to load rejection with or without coincident failure of the turbine bypass valves. The Reactor Protection System initiates a trip when fast closure of the control valves is initiated by the fast acting solenoid valves and in less than 30 milliseconds after the start of control.valve fast closure. *This is achieved by the action of the fast acting solenoid valves in rapidly reducing hydraulic trip oil pressure at the main turbine control valve actuator disc dump valves.

This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the Reactor Protection System.

This trip setting, a faster closure time, and a different valve characteristic from that of the turbine stop valve, combine to produce transients which are very similar to that for the stop valve.- Relevant transient analyses are discussed in Section 15.2.2 of the Final Safety Analysis Report.

11. Reactor Mode Switch Shutdown Position The reactor mode switch Shutdown position is a redundant channel to the automatic protective instrumentation channels and provides additional manual reactor trip capability.

12.

Manual Scram The Manual Scram is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

LIMERICK - UNIT 2 B 2-9 AU6 2 5 1989

TECH SPEC MARKUP INSERT 6:

REFERENCES:

1.

NEDC-31300, "Single-Loop Operation Analysis for Limerick Generating Station, Unit 1", August 1986.

2.

NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

3.

NEDO-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology,' November 1995.

4.

NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

5.

BWROG Letter 96113, K. P. Donovan (BWROG) to L. E. Phillips (NRC),

"Guidelines for Stability Option IlIl 'Enable Region' (TAC M92882),"

September 17,1996.

3/4.3 INSTRUMENTATION A/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

APPLICABILITY: As shown in Table 3.3.1-1.

ACTION:

Note:

Separate condition entry is allowed for each channel.

a.

With the number of OPERABLE channels in either trip system for one or more Functional Units less than the Minimum OPERABLE Channels. per Trip System required by Table 3.3.1-1, within one hour for each affected functional unit either verify that at least one* channel in each trip system is OPERABLE.or tripped or that the trip system is tripped, or place either the affected trip system or at least one inoperable channel in the affected trip system in the tripped condition.

b.

With the number of OPERABLE channels in either trip system less than the Minimum OPERABLE Channels per Trip System required by Table 3.3.1-1, place either the inoperable channel(s) or-the affected trip system** in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> OMD

c.

With the number of OPERABLE channels in both trip systems for one or more Functional Units less than the Minimum OPERABLE Channels per Trip System required by Table 3.3.1-1, place either the inoperable channel(s) in one trip system or one trip system in the trip ed condition within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s**.Ai

d.

If within the allowable time allocated by ( tions a, b or c, it is not desired to place the inoperable channel or rip system in trip (e.g., full scram would occur), Then no later than expiration of that allowable time initiate the action identified in Table 3.3.1-1 for the applicable Functional Unit.

• For Functional Units 2.a, 2.b, 2.c, 2.d, at least two channels shall be OPERABLE or tripped.

For Functional Unit 5, both trip systems shall have each channel associated with the MSIVs in three main steam lines (not necessarily the same main steam lines for both trip systems) OPERABLE or tripped.

For Function 9, at least three channels per trip system shall be OPERABLE or tripped.

• • For Functional Units 2.a, 2.b, 2.c, 2.d, inoperable channels shall be placed in the tripped condition to comply with Action b. Action c does not apply for these Functional Units.

(iii' an 2.

LIMERICK - UNIT3 3/4 3-1 Anminmt lb. 17, 34-, 109

TECH SPEC MARKUP INSERT 6A:

A channel or trip system which has been placed in the tripped condition to satisfy Action b. or c. may be returned to the untripped condition under administrative control for up to two hours solely to perform testing required to demonstrate its operability or the operability of other equipment provided Action a. continues to be satisfied.

3/4.3 INSTRUMENTATION 3/4.3.] REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE.by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1 4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation o all channels shall be performed at leas once per 24 months, except Table 4.3.1.1- _,

Functions 2.a, 2.b, 2.c, 2

.d<EjP2.

Functions 2.a, 2.b, 2.c,(?2.d no require separate LOGIC SYSTEM FUNCTI AL TESTS.

For Function 2.e, tests shall be performed at least once per 24 months.

LOGIC SYSTEM FUNCTIONAL TEST for Function 2.e includes simulating APRM4trip conditions at the APRM channel inputs to the voter channel to check all combination of two tripped inputs to the 2-Out-Of-4 voter logic in the voter channels.

Nina o 4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE -TIME of each reactor trip functional unit shown in Table 3.3.1-2 shall be demonstrated to be. within its limit at least once per 24 months.

Each test shall include at least one channel per trip system such that all channels are tested.at least once every N times 24 months where N is the total number of redundant channels in a specific reactor trip system.

LIMERICK - UNIT 2 3/4 3-la Ln oiwnt lb. 109 I

TABLE i 1l REACTOR PROTECTION SYSTEM INSTRUMENTATION APPLICABLE MINIMUM OPERATIONAL OPERABLE CHANNELS CONDITIONS PER TRIP SYSTEM (a)

FUNCTIONAL UNIT

1.

Intermediate Range MonitorsWb:

a.

Neutron Flux - High

b.

Inoperative ACTION I.

2 3(i), 4(i) 5(i) 2 3(i), 4(i) 5(i) 3

.3 3(d) 3 3

3(d) 1 2

3 1

2 3

2.

Average Power Range Monitor`):

a.

Neutron Flux - Upscale (Setdown)

b.

Simulated Thermal Power - Upscale

c.

Neutron Flux - Upscale d,

Inoperative 2-Out-Of-4 Voter

3.

Reactor Vessel Steam Dome Pressure - High

4.

Reactor Vessel Water Level - Low, Level 3

5.

Main Steam Line Isolation Valve-Closure 2

1 1

1, 2 1, 2 1, 2(f) 3(m) 3(m) 3 ()

3(m) 2 1

4 4

1 1

2 2

1 1, 2 1

1(g) 1/valve 4

LIMERICK - UNIT 2 3/4 3-2 Amendment No.

.7, 4-9, 112

TECH SPEC MARKUP INSERT 7:

f.

OPRM Upscale 1(o)(p) 3(m) 10

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION STATEMENTS ACTION 1 Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 2 Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the SHUTDOWN position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 3 Suspend all operations involving CORE ALTERATIONS and insert all insertable control rods within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 4 Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 5 Be in STARTUP with the main steam line isolation valves closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 6 Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure until the function is automatically bypassed, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

ACTION 7 Verify all insertable control rods to be inserted within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 8 Lock the reactor mode switch in the Shutdown position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 9 Suspend all operations involving CORE ALTERATIONS, and insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

LIMERICK - UNIT 2 3/4 3 -4 Amendment No. 109, 112

TECH SPEC MARKUP INSERT 8:

ACTION 10 a.

If the condition exists due to a common-mode OPRM deficiency*, then initiate alternate method to detect and suppress thermal-hydraulic instability oscillations within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND restore required channels to OPERABLE status within 120 days, OR

b.

Reduce THERMAL POWER to < 25% RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

• Unanticipated characteristic of the instability detection algorithm or equipment that renders all OPRM channels inoperable at once.

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a)

A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped conditic provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

(b) This function shall automatically be bypassed when the reactor mode switch is in the Run position.

(c) DELETED (d) The noncoincident NMS reactor trip function logic is such that all channels gs to both trip systems. Therefore, when the "shorting links" are removed, the Minimum OPERABLE Channels Per Trip System is 6 IRMs.

(e) An APRM channel is inoperable if there are less than 3 LPRM inputs per level or less than 20 LPRM inputs to an APRM channel, or if more than 9 LPRM inputs to the APRM channel have been bypassed since the last APRM calibration (weekly gain calibration).

(f) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

(g) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.

(h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(i) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(j) This function shall be automatically bypassed when turbine first stage pressure is equivalent to a THERMAL POWER of less than 30% of RATED THERMAL POWER.

(k) Also actuates the EOC-RPT system.

(1) DELETED (m) Each APRM channel provides inputs to both trip systems.

TECH SPEC MARKUP INSERT 9:

(n)

DELETED (o)

With THERMAL POWER Ž 25% RATED THERMAL POWER. The OPRM Upscale trip output shall be automatically enabled (not bypassed) when APRM Simulated Thermal Power is Ž 30% and recirculation drive flow is < 60%. The OPRM trip output may be automatically bypassed when APRM Simulated Thermal Power is < 30% or recirculation drive flow is 2 60%.

(p)

A minimum of 23 cells, each with a minimum of 2 OPERABLE LPRMs, must be OPERABLE for an OPRM channel to be OPERABLE.

TABLE 3.3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES RESPONSE TIME FUNCTIONAL UNIT (Seconds}

1. Intermediate Range Monitors:

a.

Neutron Flux - High N.A.

b.

Inoperative N.A.

2. Average Power Range Monitor*:

a.

Neutron Flux - Upscale (Setdown)

N.A.

b.

Simulated Thermal Power - Upscale N.A.

c.

Neutron Flux - Upscale N.A.

7A

d.

Inoperative N.A.

e.

2-Out-Of-4 Voter N0.05*

3.

Reactor Vessel Steam Dome Pressure - High

• 0.55

4.

Reactor Vessel Water Level - Low, Level 3 91.05#

5.

Main Steam Line Isolation Valve - Closure

• 0.06

6. DELETED DELETED

7. Drywell Pressure - High N.A.

8. Scram Discharge Volume Water Level - High

a.

Level Transmitter N.A.

b.

Float Switch N.A.

9. Turbine Stop Valve - Closure

• 0.06

10.

Turbine Control Valve Fast Closure, Trip Oil Pressure - Low

• O.08**

11.

Reactor Mode 'Switch Shutdown Position N.A.

12.

Manual Scram N.A.

W

• Neutron detectors, APRM channel and 2-Out-Of-4 Voter channel digital electronics are exempt from response time testing. Response time shall be measured from activation of the 2-Out-Of-4 Voter output relay.

For application of Specification 4.3.1.3, 3the r dundant outputs from each 2-Out-Of-4 Voter channel are considered part of the same channel,

,°

• Measured from start of turbine control valve fast closure -

k-4^_

cnsor is eliminated from response time testing fo:L i he RPS circuits.

Response time testing conformance to the administrative limits for oJ[. :emaining channel including trip unit

.,d relay loqic are reouired.

TECH SPEC MARKUP INSERT 9A:

f.

OPRM Upscale N.A.

INSERT 9B:

but the OPRM and APRM outputs are considered to be separate channels, so N = 8. Testing of OPRM and APRM outputs shall alternate.

TABLE 43.1.1 -1 REACTOR PROTECTION S TEMSINSTRU ETATION SYtVjdiLANCE REOUIREMENTS C fNNEL-FUtTNCTIONAL'.:

CHA 'EL', '

'CON91T INS;

'"WHICH '

' tEx.

TEST CALIBRATION-a,.SURVEILANCEREOU1RED Initors:-

J.

-High S(b)

Wi....

R

.2'..

N-.A.

)

?.A.

2, 3(i), 4(i), 5(i)

.,Mo r,

A..

FUNCTIONAL UNIT

1. Intermediate Range.Mo

a.

NeutronFlux -

b.

Inoperative

2.

Average Power,. Range

a.

Neutron"Flux

b.

Simulated The monitor tT) :.:

-.Upscale (S'e rmal.Power.-

c.

Neutron Flux - Upscale

d.

Inoperative P.

2-Out-Of-4 Voter

3.

Reactor Vessel Steam Dome '

Pressure -'

High itdown)

O(b)

SA(l)

R

.Upscale, SA(e)

W(d),R(g) 8 ;S _S

a.

\\4... '.

DA,,;

.O..,.>

.:2>. SA;

..,,,.....,W'd,.-R,.....

• ,:N.A'

,SA-NA.....-

0- *

.. :A.

.. S.'Q.

R' S

R N.A.

Q R

DELETED DELETED DELETEd

,S R

2..

1 1.2 1,.Z.

1

",'2 (h)5.

4.

Reactor Vessel Water Level-Low, Level 3

5.

Main. Steam Line Isol tion Valve Closure

6.

DELETED

7.

Drywell Pressure - High

8.

Scram Discharge Volume Water Level -.-High'

a. Level Transmitter b' Float Switch I,1,2 6.

DELETED'

. I

.' S I ": ' 'N.A.

"a:'Q :'

.Q.1..

,medmeit N.

2.

112 LIMERICK -

UNIT 2 3/4.3-7

TECH SPEC MARKUP INSERT 10:

f.

OPRM Upscale D

SA(e)

R(c)(g) 1(m)

TABLE 4.3.1.-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL FUNCTIONAL UNIT CHECK CHANNEL FUNCTIONAL TEST.

CHANNEL CALIBRATION(a)

OPERATIONAL CONDITIONS FOR WHICH SURVEILLANCE REQUIRED

9.

Turbine Stop Valve - Closure

10.

Turbine Control Valve Fast Closure, Trip Oil Pressure - Low N.A.

Q R

1-1 N.A.

R

11.

Reactor Mode Switch Shutdown Position

12.

Manual Scram N.A.

N.A.

R N.A.

N.A.

1, 2, 3, 4, 5 1, 2, 3, 4, 5 (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for a least 1/2 each cntrolled shutdown, if not performed within the previous 7 days.

(c) £ (d) This calibration shal consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER 225% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER.

(e) CHANNEL FUNCTIONAL TEST shall include the flow input function, excluding the flow transmitter.

(f) The LPRMs shall be calibrated at least once per 1000 effective'full power hours (EFPH).

(g) Calibration includes the flow input function.

(h) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

(i) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(j) If the RPS shorting links are required to be removed per Specification 3.9.2, they may be reinstalled for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance. During this time, CORE ALTERATIONS shall be suspended, and no control rod shall be moved from its existing position.

(k) DELETED (1) Not required to be performed when entering OPERATIONAL CONDITION 2 from OPERATIONAL CONDITION 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering OPERATIONAL CONDITION 2.

LIMERICK - UNIT 2 f#e7 3/4 3-8 Amendment No. 7,4,48,7.,79,9.,1O9 I

I 11

TECH SPEC MARKUP INSERT 11:

Calibration includes verification that the OPRM Upscale trip auto-enable (not-bypass) setpoint for APRM Simulated Thermal Power is Ž 30% and for recirculation drive flow is

< 60%.

INSERT 12:

(m)

With THERMAL POWER Ž 25% of RATED THERMAL POWER.

C,)

73 TABLE 3.3.6-1 CONTROL ROD BLOCK INSTRUMENTATION rN, MINIMUM OPERABLE CHANNELS PER TRIP FUNCTION TRIP FUNCTION

1.

ROD BLOCK MONITORY$

a.

Upscale

b.

Inoperative

c.

Downscale APPLICABLE OPERATIONAL CONDITIONS 1*

1*

1*

22 2

ACTION 60 60 60

2.

APRM a.

b.

C.

d.

e.

f.

Simulated Thermal Power - Upscale Inoperative Neutron Flux - Downscale Simulated Thermal Power - Upscale (Setdown)

Recirculation Flow - Upscale LPRM Low Count 3

3 3

3 3

3

3.

SOURCE RANGE MONITORS ***

a.

Detector not full in tb)

I 1, 2 1

2 1

1,2 2

2 25 2

5 2

5 61 61 61 61 61 61

coa,

b.

Upscale(c)

c.

Inoperative(c)

d.

Downscale(")

4.

INTERMEDIATE RANGE MONITORS 3

2 3

23 2

3 2

61 61 61 61 61 61 61 61 flTI a.

b.

C.

d.

Detector not full in Upscale Inoperative Downscale' 6

6 6

6 2,

2, 2,

2, Ebl 61 61 61 61

5.

SCRAM DISCHARGE VOLUME

a.

Water Level-High 2

1, 2, 5**

62

6.

DELETED DELETED DELETED DELETED I

7.

REACTOR MODE SWITCH SHUTDOWN POSITION 2

3, 4 63

TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SEPOINTS TRIP SETPOINT ALLOWABLE VALUE TRIP FUNCTION

1.

ROD BLOCK MONITOR

a. Upscale"'

1) Low Trip Setpoint (LTSP)

2) Intermediate Trip Setpoint (ITSP)

3) High Trip Setpoint (HTSP)

b.

Inoperative C. Downscale (DTSP)

d. Power Range Setpoint(b)

1) Low Power Setpoint (LPSP)

2) Intermediate Power Setpoint (IPSP)

3) High Power Setpoint (HPSP)

N/A 28.1% RATED THERMAL POWER

• 63.1% RATED THERMAL POWER 83.1% RATED THERMAL POWER Z. APRM

a. Simulated Thermal Power - Upscale:

Two Recirculation Loop Operation s 0.66 W + 55.2% and 5 108.0% of RATED TH£m nF Single Recirculation Loop Operatio atiD 06 W 5f and sTHERA of RATED THERMAL POWER N/A 28.4% RATED THERMAL POWER 63.4% RATED THERMAL POWER l 83.4% RATED THERMAL POWERl i 0.66 W + 55.7% and

& 108.4% of RATED 4 -108.4%-of RATEO THERMAL POWER N.A.

i 2.8% of RATED THERMAL POWER W

S-14.

W M

C

b. Inoperative

c. Neutron Flux - Downscale N.A.

2. 3.2% of RATED THERMAL POWER

d. Simulated Thermal Power - Upscale (Setdown)

e. Recirculation Flow - Upscale

f. LPRM Low Count s 12.0% of RATED THERMAL POWER 20 per channel

< 3 per axial level s 13.0% of RATED THERMAL POWER 20 per channel 3 per axial level

3. SOURCE RANGE MONITORS

a. Detector not full in

b. Upscale

c.

Inoperative

d. Downscale N.A.

£ 1 x 105 cps N.A.

i 3 cps**

01ti N.A.

• 1.6 x lO cps N.A.

Z 1.8 cps**

r-

-1

-4.I TABLE 3.3.6-2 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in

b. Upscale c.

d.

Inoperative Downscale TRIP SETPOINT N.A.

s 108/125 divisions of full scale N.A.

2 5/125 divisions of full scale K257' 7 3/8" elevation***

DELETED N.A.

ALLOWABLE VALUE N.A.

s 110/125 divisions ol full scale N.A.

2 3/125 divisions of scale 1 257' 9 3/8"' elevati DELETED N.A.

5. SCRAM DISCHARGE VOLUME

a. Water Level-High

a. Float Switch

6. DELETED

7. REACTOR MODE SWITCH SHUTDOWN POSITION F

ful l on on ae CA) rY Refer to the COLR for these setpoints.

lMay be reduced, provided the Source Range Monitor has an observed count rate and signal-to-noise ratio or above the curve shown in Figure 3.3.6-1.

Equivalent to 13.56 gallons/scram discharge volume.

(a)

There are three upscale trip levels. Each is applicable only over its specified operating core thermal power range.

All RBM trips are automatically bypassed below the low power'setpoint (LPSP).

The upscale LTSP is applied between the low power setpoint (LPSP) and the intermediate power setpoint (IPSP). The upscale ITSP is applied between the intermediate power setpoint and the high power setpoint (HPSP).

The HTSP is applied above the high power setpoint.

(b)

Power power range setpoints control enforcement of appropriate upscale trips over the proper core thermal ranges. The power signal to the RBM is provided by the APRM.

TECH SPEC MARKUP INSERT 13:

The 7.6% flow "offset" for Single Loop Operation (SLO) is applied for W 2 7.6%.

For flows W < 7.6%, the (W-7.6 %) term is set equal to zero.

TABLE A3.6-1 CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS

-CHANNEL

,-CHECK TRIP-FUNCTION

1. ROD BLOCK MONITOR

a. Upscale

b. Inoperative

c. Downscale CHANNEL FUNCTIONAL-TEST Q(c) 0(C) cc)

'CHANNEL '

CALIBRATIONI')

R N.A.

R-OPERATIONAL.

• .CONDITIONS FOR WHICH SURVEILLANCE REQUIRED N.A.

N.A.

N.A.

1*

1*.

• 1*

I.

2. APRM

a. Simulated Thermal Power -

.Upscale N.A.

b. Inoperative N.A.

c. Neutron. Flux - Downscale

N-.A.

d. Simulated Thermal Power -'

Upscale (Setdown)

N.A.

e. Recirculation Flow - Upscale.

N.A.

f. LPRM Low Count N.A.

SA SA SA

-SA SA SA R

• . N.A'.

R R

R

>1

• 1 2.

1 1,

2.

2

3. SOURCE RANGE MONITORS' a.

b.

c.

d.

Detector not full in Upscale Inoperative Downscale N.A.

N.A.

.N.A.

N.A.

N.A.

R N.A.

R n 2,

2, 2,

2, 5 -

5 5

5 1

4. INTERMEDIATE RANGE MONLTORS a.

b.

c.

d.

Detector not full in

• Upscale -

Inoperative Downscale N.A.

N.A.

N.A.

N.A.

N.

N N

N N.A.

I. N.A.

R R

DEE E

.,R -

DELETED..

2, 5

• 2, 5 f*

2, i*

2, 5 **

5. SCRAM DISCHARGE VOLUME

a. Water Level - High N.A.

Q DELETED 1, 2, 5**-

DELETED

6. DELETED DELETED

7.

REACTOR MODE SWITCH SHUTDOWN POSITION N.A.

'N.A.

3, 4 LIMERICK - UNIT 2 3/4 3-61 Amendment No. ;, 4-,

48, Corrected by letter dated May 63, 4-28, 2002

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS I TMTTTNG CONDITION FOR OPERATIION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operatio G

ACTION:

a.

With one reactor coolant system recirculation loop not in operation:

1.

Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

a.

Place the recirculation flow control system in the Local Manual mode, and

b.

Reduce THERMAL POWER to < 76.2% of RATED THERMAL POWER, and, I

c.

Limit the speed of the operating recirculation pump to less than or equal to 90% of rated pump speed, and

d.

Verify that the differential temperature requirements of Surveillance Requirement 4.4.1.1.5 are met if THERMAL POWER is 5 30% of RATED THERMAL POWER -or the recirculation loop flow in the operating loop is c 50% of rated loop flow, or suspend the THERMAL POWER or recirculation loop flow increase.

• See Special Test Exception 3.10.4.

LIMERICK - UNIT 2 3/4 4-1 Amendment No. 4(9, 51 FEB 1 6 1995.

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

(Continued)

2.

Within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s:

Reduce the Average Power Range Monitor (APRM) Simulated Thermal Power

- Upscale Scram and Rod Block Trip Setpoints and Allowable Values, to those applicable for single recirculation loop operation per Specifications 2.2.1 and 3.3.6, or declare the associated channel(s) inoperable and take the actions required by the referenced speci fi cati oR Dm

3.

Otherwise be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

With no 'reactor coolant system recirculation1

/-5iFiri~e y-sd e

itp~t vnte eye ngbLJO

{/Tir~ct/'wth~~h< fi£ifi~k<*5Nz he'uSidn"19i2:

RfofirN

§{3frf5Ri~iasu es to place Tteuit-in wN firt~yhf~f tyrRVHOT SHUTDOWN within the nex-LIMERICK - UNIT 2 3/4 4-la Amendment No. 48, 409. 132

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.1.1 DELETED 4.4.1.1.2 DELETED 4.4.1.13 bl. h basQ ineAPRM nd LAM**

ut fl noie va wi n

sor 1?i ec fuat 3r s q X

Z.

.1, 1 0 Op~r,'6rso Xnepfn to effnJ whit mHio ig isv equj ulnSs bflrn~

pd dviosl b~npro~di~hfgiof sinc 'heoatJEu/n 4.4.1.1.4 With one reactor coolant system recirculation loop not in operation, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verify that:

a.

Reactor THERMAL POWER is s 76.2% of RATED THERMAL POWER,

b.

The recirculation flow control system is in the Local Manual mode, and

c.

The speed of the operating recirculation pump is s 90% of rated pump speed.

t< /

~

X eX w at t

39 T W A

'C reS rictXd ine pftFj~ure /4/1 o

4.4.1.1.5 With one reactor coolant system recirculation loop not in operation, within 15 minutes prior to either THERMAL POWER increase or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is s 30% of RATED THERMAL POWER or the recirculation loop flow in the operating recirculation loop is s 50% of rated loop flow.

a.

s 1450F between reactor vessel steam space coolant and bottom head drain line coolant,

b.

s 50'F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and

c.

s 50'F between the reactor coolant within the loop not in operation and the operating loop.

The differential temperature requirements of Specification 4.4.1.1.5b. and c.

do not apply when the loop not in operation is isolated from the reactor pressure vessel.

LR -UNIT23/44-2 Am nt 2000 Amendment No.37,38,51,104 LIMERICK - UNIT 2 3/4 4-2

0s LIMERICK - UNIT 2 FEB 1 6 1995 Amendment No. 51 3/4 4-3

-3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

a.

Preserve the integrity of the fuel cladding.

b.

Preserve the integrity of the reactor coolant system.

c.

Minimize the energy which must be adsorbed following a loss-of-coolant accident, and

d.

Prevent inadvertent criticality.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.

The reactor protection system is made up of two independent trip systems.

There are usually four channels to monitor each parameter with two channels in each trip system.

The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system.

The tripping of both trip systems will produce a reactor scram. The APRM system is divided into four APRM channels and four 2-Out-Of-4 Voter channels.

Each APRM channel provides inputs to each of the four voter channels.

The four voter channels are divided into two groups of two each, with each group of two providing inputs to one RPS trip system.

The system is designed to allow one APRM channel, but no voter channels, to be bypassed.

The system meets the intent of JEEE-279 for nuclear power plant protection systems. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System" and NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function." The bases for the trip settings of the RPS discussed in the bases for Specification 2.2.1.

a S Ez ctions a, b and c define th._

s) required when RPS channels are iscovered to be inoperable. For tho ions, separate entry condition is allowed for each inoperable RPS channel.

Sepaiate entry means that the allowable time clock(s) for Actions a, b or c start upon discovery of inoperability for that specific channel. Restoration of an inoperable RPS channel satisfies only the action statements for that particular channel. Action statement(s) for remaining inoperable channel(s) must be met according to their original entry time.

Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> has been shown to be acceptable (NEDC-30851P-A and NEDC-32410P-A) to permit restoration of any inoperable channel to OPERABLE status.

However, this out of service time is only acceptable provided that the associated Function's (identified as a "Functional Unit" in Table 3.3.1-1) inoperable channel is in one trip system and the Function still maintains RPS trip capability.

LIMERICK - UNIT 2 B 3/4 3-1 Abefdnmt b-. 17l5og93. 19

TECH SPEC MARKUP INSERT 14:

The APRM Functions include five Functions accomplished by the four APRM channels (Functions 2.a, 2.b, 2.c, 2.d, and 2.f) and one accomplished by the four 2-Out-Of-4 Voter channels (Function 2.e). Two of the five Functions accomplished by the APRM channels are based on neutron flux only (Functions 2.a and 2.c), one Function is based on neutron flux and recirculation drive flow (Function 2.b) and one is based on equipment status (Function 2.d). The fifth Function accomplished by the APRM channels is the Oscillation Power Range Monitor (OPRM) Upscale trip Function 2.f, which is based on detecting oscillatory characteristics in the neutron flux. The OPRM Upscale Function is also dependent on average neutron flux (Simulated Thermal Power) and recirculation drive flow, which are used to automatically enable the output trip.

The Two-Out-Of-Four Logic Module includes 2-Out-Of-4 Voter hardware and the APRM Interface hardware. The 2-Out-Of-4 Voter Function 2.e votes APRM Functions 2.a, 2.b, 2.c, and 2.d independently of Function 2.f. This voting is accomplished by the 2-Out-Of-4 Voter hardware in the Two-Out-Of-Four Logic Module. The voter includes separate outputs to RPS for the two independently voted sets of Functions, each of which is redundant (four total outputs). The analysis in Reference 2 took credit for this redundancy in the justification of the 12-hour allowed out-of-service time for Action b, so the voter Function 2.e must be declared inoperable if any of its functionality is inoperable. The voter Function 2.e does not need to be declared inoperable due to any failure affecting only the APRM Interface hardware portion of the Two-Out-Of-Four Logic Module.

Three of the four APRM channels and all four of the voter channels are required to be OPERABLE to ensure that no single failure will preclude a scram on a valid signal. To provide adequate coverage of the entire core, consistent with the design bases for the APRM Functions 2.a, 2.b, and 2.c, at least 20 LPRM inputs, with at least three LPRM inputs from each of the four axial levels at which the LPRMs are located, must be OPERABLE for each APRM channel. In addition, no more than 9 LPRMs may be bypassed between APRM calibrations (weekly gain adjustments). For the OPRM Upscale Function 2.f, LPRMs are assigned to "cells" of 3 or 4 detectors. A minimum of 23 cells (Reference 9), each with a minimum of 2 OPERABLE LPRMs, must be OPERABLE for each APRM channel for the OPRM Upscale Function 2.f to be OPERABLE in that channel.

TECH SPEC MARKUP INSERT 14 (continued):

References 4, 5 and 6 describe three algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. All three are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period based detection algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY for Technical Specification purposes is based only on the period based detection algorithm.

An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in any cell, with period confirmations and relative cell amplitude exceeding specified setpoints. One or more cells in a channel exceeding the trip conditions will result in a channel trip. An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithms detect growing oscillatory changes in the neutron flux for one or more cells in that channel.

The OPRM Upscale Function is required to be OPERABLE when the plant is at 2 25% RATED THERMAL POWER. The 25% RATED THERMAL POWER level is selected to provide margin in the unlikely event that a reactor power increase transient occurring while the plant is operating below 30% RATED THERMAL POWER causes a power increase to or beyond the 30% RATED THERMAL POWER OPRM Upscale trip auto-enable point without operator action. This OPERABILITY requirement assures that the OPRM Upscale trip automatic-enable function will be OPERABLE when required.

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

The requirements of Action a are intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same trip system for the same Function result in the Function not maintaining RPS trip capability. A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal.

For the typical Function with one-out-of-two taken twice logic, including the IRM Functions and APRM Function 2.e (trip capability associated with APRM Functions 2.a, 2.b, 2.c,<93)2.d ar discussed below), this would require both trip systems to have one channel OPERABEor i trl (or the associated trip system in trip).

F sn 2,$

For Function 5 W Steam Isolation Valve--Closure), this would require both trip systems to have each channel associated with the MSIVs in three main steam lines (not necessarily the same main steam lines for both trip systems) OPERABLE or in trip (or the associated trip system in trip).

For Function 9 (Turbine Stop Valve-Closure), this would require both trip systems to have three channels,, each OPERABLE or in trip (or the associated trip system in trip).

The completion time to satisfy the requirements of Action a is intended to allow the operator time to evaluate and repair any discovered inoperabilities.

The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.

With trip capability maintained, i.e., Action a satisfied, Actions b and c as applicable must still be satisfied.

If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Action b requires that the channel or the associated trip system must be placed in the tripped condition.

Placing the inoperable channel in trip (or the associated trip system in trip) would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue As noted, placing the trip system in iS not applicable to satisfy Action b for APRM Functions 2.a, 2.b, 2.c,4H 2.c Inoperability of one required APRM channel affects both trip systems.

For that condition, the Action b requirements can only be satisfied by placing the inoperable APRM channel in trip.

Restoring OPERABILITY or placing the inoperable APRM channel in trip are the only actions that will restore capability to accommodate a single APRM channel failure.

Inoperability of more than one required APRM channel of the same trip function results in loss of trip capability and the requirement to satisfy Action a.

The requirements of Action c must be satisfied when, for any one or more Functions, at least one required channel is inoperable in each trip system.

In this condition, provided at least one channel per trip system is OPERABLE, normally the RPS still maintains trip capability for that Function, but cannot accommodate a single failure in either trip system (see additional bases discussion above related to loss of trip capability and the requirements of Action a, and special cases for Functions 2.a, 2.b, 2.c, 2.d 5 and 9).

2.,n LIMERICK - UH4IT 2 B 3/4 3-la in-rt NM. 109 I

-3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

The requirements of Action c limit the time the RPS scram logic, for any Function, would not accommodate single failure in both trip systems (e.g., one-out-of-one and one-out-of-one arrangement for a typical four channel Function).

The reduced reliability of this logic arrangement was not evaluated in NEDC-30851P-A for the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time.

Within the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance, the associated Function must have all required channels OPERABLE or in trip (or any combination) in one trip system.

Completing the actions required by Action c restores RPS to a reliability level equivalent to that evaluated in NEDC-30851P-A, which justified a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowable out of service time as allowed by Action b.

To satisfy the requirements of Action c, the trip system in the more degraded state should be placed in trip or, alternatively, al the inoperable channels in that trip system-should be placed in trip (e.g., a trip system with two inoperable channels could be in a more degraded state than a trip system with four inoperable channels if the two inoperable channels are in the same Function while the four inoperable channels are all in different Functions).

The decision of which trip system is in the more degraded state should be based on prudent judgment and take into account current plant conditions (i.e.,

what OPERATIONAL CONDITION the plant is in).

If this action would result in a scram or RPT, it is permissible to place the other trip system or its inoperable channels in trip.

The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowable out of service time is judged acceptable based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram.

,,_______;7>

Q >orZ2. J As noted, Action c is not applicable for APRM Functions 2.a, 2.b, 2.c,2 JT2.d Inoperability of an APRM channel affects both trip systems and is not associated with a specific trip system as are the APRM 2-Out-Of-4 voter and other non-APRM channels for which Action c applies.

For an inoperable APRM channel, the requirements of Action b can only be satisfied by tripping the inoperable APRM channel.

Restoring OPERABILITY or placing the inoperable APRM channel in trip are the only actions that will restore capability to accommodate a single APRM channel failure.

If it is not desired to place the channel (or trip system) in trip to satisfy the requirements of Action a, Action b or Action c (e.g., as in the case where placing the inoperable channel in trip would result in a full scram), Action d requires that the Action defined by Table 3.3.1-1 for the applicable Function be initiated immediately upon expiration of the allowable out of service time.

7

/ Jfie TwoSOu;-Of-jPtwr Lonic Ifidule ~includesV2-out-of-C voter hardware and APRM LHIERICK - UNIT 2 B 3/4 3-1b U 2B 3brrd t No. log

TECH SPEC MARKUP INSERT 15:

Table 3.3.1-1, Function 2.f, references Action 10, which defines the action required if OPRM Upscale trip capability is not maintained. Action 1 Ob is required to address identified equipment failures. Action 1 Oa is to address common mode vendor/industry identified issues that render all four OPRM channels inoperable at once. For this condition, References 2 and 3 justified use of alternate methods to detect and suppress oscillations for a limited period of time, up to 120 days. The alternate methods are procedurally established consistent with the guidelines identified in Reference 7 requiring manual operator action to scram the plant if certain predefined events occur. The 12-hour allowed completion time to implement the alternate methods is based on engineering judgment to allow orderly transition to the alternate methods while limiting the period of time during which no automatic or alternate detect and suppress trip capability is formally in place. The 120-day period during which use of alternate methods is allowed is intended to be an outside limit to allow for the case where design changes or extensive analysis might be required to understand or correct some unanticipated characteristic of the instability detection algorithms or equipment.

The evaluation of the use of alternate methods concluded, based on engineering judgment, that the likelihood of an instability event that could not be adequately handled by the alternate methods during the 120-day period was negligibly small. Plant startup may continue while operating within.the allowed completion time of Action 1 Oa. The primary purpose of this is to allow an orderly completion, without undue impact on plant operation, of design and verification activities in the event of a required design change to the OPRM Upscale function. This exception is not intended as an alternative to restoring inoperable equipment to OPERABLE status in a timely manner.

Action 1 Oa is not intended and was not evaluated as a routine alternative to returning failed or inoperable equipment to OPERABLE status. Correction of routine equipment failure or inoperability is expected to be accomplished within the completion times allowed for LCO 3.3.1 Action a or Action b, as applicable. Action 1 Ob applies when routine equipment OPERABILTY cannot be restored within the allowed completion times of LCO 3.3.1 Actions a or b, or if a common mode OPRM deficiency cannot be corrected and OPERABILTY of the OPRM Upscale Function restored within the 120-day allowed completion time of Action 1 Oa.

TECH SPEC MARKUP INSERT 15 (continued):

The OPRM Upscale trip output shall be automatically enabled (not-bypassed) when APRM Simulated Thermal Power is > 30% and recirculation drive flow is < 60%

as indicated by APRM measured recirculation drive flow. NOTE: 60% recirculation drive flow is the recirculation drive flow that corresponds to 60% of rated core flow. This is the operating region where actual thermal-hydraulic instability and related neutron flux oscillations may occur. As noted in Table 4.3.1.1-1, Note c, CHANNEL CALIBRATION for the OPRM Upscale trip Function 2.f includes confirming that the auto-enable (not-bypassed) setpoints are correct. Other surveillances ensure that the APRM Simulated Thermal Power properly correlates with THERMAL POWER (Table 4.3.1.1-1, Note d) and that recirculation drive flow properly correlates with core flow (Table 4.3.1.1-1, Note g).

If any OPRM Upscale trip auto-enable setpoint is exceeded and the OPRM Upscale trip is not enabled, i.e., the OPRM Upscale trip is bypassed when APRM Simulated Thermal Power is 2 30% and recirculation drive flow is < 60%, then the affected channel is considered inoperable for the OPRM Upscale Function.

Alternatively, the OPRM Upscale trip auto-enable setpoint(s) may be adjusted to place the channel in the enabled condition (not-bypassed). If the OPRM Upscale trip is placed in the enabled condition, the surveillance requirement is met and the channel is considered OPERABLE.

As noted in Table 4.3.1.1-1, Note g, CHANNEL CALIBRATION for the APRM Simulated Thermal Power - Upscale Function 2.b and the OPRM Upscale Function 2.f, includes the recirculation drive flow input function. The APRM Simulated Thermal Power-Upscale Function and the OPRM Upscale Function both require a valid drive flow signal. The APRM Simulated Thermal Power - Upscale Function uses drive flow to vary the trip setpoint. The OPRM Upscale Function uses drive flow to automatically enable or bypass the OPRM Upscale trip output to RPS. A CHANNEL CALIBRATION of the APRM recirculation drive flow input function requires both calibrating the drive flow transmitters and establishing a valid drive flow / core flow relationship. The drive flow / core flow relationship is established once per refuel cycle, while operating within 10% of rated core flow and within 10% of RATED THERMAL POWER. Plant operational experience has shown that this flow correlation methodology is consistent with the guidance and intent in Reference 8. Changes throughout the cycle in the drive flow / core flow relationship due to the changing thermal hydraulic operating conditions of the core are accounted for in the margins included in the bases or analyses used to establish the setpoints for the APRM Simulated Thermal Power - Upscale Function and the OPRM Upscale Function.

TECH SPEC MARKUP INSERT 15 (continued):

As noted in Table 3.3.1-2, Note "*", the redundant outputs from the 2-Out-Of-4 Voter channel are considered part of the same channel, but the OPRM and APRM outputs are considered to be separate channels, so N = 8 to determine the interval between tests for application of Specification 4.3.1.3 (REACTOR PROTECTION SYSTEM RESPONSE TIME). The note further requires that testing of OPRM and APRM outputs shall be alternated.

Each test of an OPRM or APRM output tests each of the redundant outputs from the 2-Out-Of-4 Voter channel for that function, and each of the corresponding relays in the RPS. Consequently, each of the RPS relays is tested every fourth cycle. This testing frequency is twice the frequency justified by References 2 and 3.

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

Automatic reactor trip upon receipt of a high-high radiation signal from the Main Steam Line Radiation Monitoring System was removed-as the result of an analysis performed by General Electric in NEDO-31400A.

The NRC approved the results of this analysis as documented in the SER (letter to George J. Beck, BWR Owner's Group from A.C. Thadani NRC dated May 15, 1991).

averms~

2.*)

The measurement o response time at the specified frequencies provides assurance that the protective unctions associated with each channel are completed within the time lim' assumed in the safety analyses.

No credit was taken for those channels wit response times indicated as not applicable except for APRM Simulated Thermal Power U scale and Neutron Flux - Upscale trip functions (Table 3.3.1-2, Items 2.b Response time may be demonstrated by any series of sequential, overlapping or ota channel test measurement, provided such tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times.

Response time testing for the sensors as noted in Table 3.3.1-2 is not required based on the analysis in NEDO-32291-A. Response time testing for the remaining channel components is required as noted.

For the digital electronic portions of the APRM functions, performance characteristics that determine response time are checked by a combination of automatic self-test, calibration activities, and response time tests of the 2-Out-Of-4 Voter (Table 3.3.1-2, Item 2.e).

K aft 4J6, A CPPRM V sca leA You 7 L LIMERICK -

UNIT 2 B 3/4 3-1c AraYhmt Nb. 109 1

INSTRUMENTATION BASES 3/4.3.7.10 (Deleted)

I 3/4.3.7.11 (Deleted)

- INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.

3/4.3.7.12 OFFGAS MONITORING INSTRUMENTATION This instrumentation includes provisions for monitoring the concentrations of potentially explosive gas mixtures and noble gases in the off-gas system.

3/4.3.

(Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE UFSAR.

3/4,3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION The feedwater/main turbine trip system actuation instrumentation is provided to initiate action of the feedwater system/main turbine trip system in the event of failure of feedwater controller under maximum demand.

LIMERICK - UNIT 2 B 3/4 3-7 Amendment No. 44, 2, 33, 14, 67, 117

TECH SPEC MARKUP INSERT 16:

REFERENCES:

1.

NEDC-30851 P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System", March 1988.

2.

NEDC-3241OP-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option IlIl Stability Trip Function", October 1995.

3.

NEDC-3241OP-A, Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option IlIl Stability Trip Function", November 1997.

4.

NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

5.

NEDO-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

6.

NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

7.

Letter, L. A. England (BWROG) to M. J. Virgilio, UBWR Owners' Group Guidelines for Stability Interim Corrective Action", June 6, 1994.

8.

GE Service Information Letter No. 516, "Core Flow Measurement - GE BWR/3, 4, 5 and 6 Plants", July 26, 1990.

9.

GE Letter NSA 00-433, Alan Chung (GE) to Sujit Chakraborty (GE),

"Minimum Number of Operable OPRM Cells for Option IlIl Stability at Limerick 1 & 2," May 02, 2001.

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM The impact of single recirculation loop operation upon plant safety is assessed and shows that single-loop operation is permitted if the MCPR fuel cladding safety limit is increased as noted by Specification 2.1.2, APRM scram and control rod block setpoints are adjusted as noted in Tables 2.2.1-1 and 3.3.6-2, respectively.

Additionally, surveillance on the pump speed of the operating recirculation loop is imposed to exclude the possibility of excessive internals vibration. The surveillance on differential temperatures below 30%

RATED THERMAL POWER or 50% rated recirculation loop flow is to mitigate the undue thermal stress on vessel nozzles, recirculation pump and vessel bottom head during the extended operation of the single recirculation loop mode.

An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the-core; thus, the requirement for shutdown of the facility with a jet pump inoperable.

Jet pump failure can be detected by monitoring jet pump performance on a pre-scribed schedule for significant degradation.

Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteria for two recirculation loop operation.

The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA. In the-case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single recirculation loop mode.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50'F of each other prior to startup of an idle loop. The loop temperature must also be within 500F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Sudden equalization of a temperature difference > 145'F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.

heobjective o'f BWR plant and fuel sign i to pr vide sta re opera-ti wi mag in verhe n mal op/rating main.. owever at the igh wer ow ow orn of e oper ing do in, a all pr ability f limi cy e n ro flux scil ations xists pendin on comr nations oper ing ndi ons e.g.

rod ttern 0power sape).

provi e assura e that eutron flu imi cycl osc latio a re de cted a suppr sed, AP and LP N

n ron lux oise evels ould b monito d whil operatin in thi region.

Sta lity ests opera ng BWRs ere rev ewed to.termin a gener regon a the ower/f w map which rveill ce of ne ron fl x noise evel s ould e pe formed A con rvative ecay ra io of 0. was ch en as tye bas s or term ing th generi region r surv ilance t accoun for th plan to pla var abilit of dec ratio w th core nd fuel esigns. This neri rgion s been determiyed to cori espond a core low of ess th or ual 455'of rat core fyow and a HERMAL WER grea er than that s cifi d in LIMERICK - UNIT 2 B 3/4 4-1 Amendnent No. 48 JAN 3 1 1995

REACTOR COOLANT SYSTEM BASES RE CUL TION ZYSTEM (Continued) lan speci c ca la ons ca be erformed to determie an a licable regi n fo monit nfl eutr flux sais evel.

In is c e the egree of onse atis can re ced si e ant t lant ariabity w id be e mined.

thi cas, adeq te argin ii b assur by ma itori the egio whic has dec rati gre er th or e al to.8.

Neu on f ux ise li its re alo esta ished o ens e ear dete ion of0 imi cyci neu on fl illat ns. B cor typic 1 op0 ate wih utro flux ois cause by ando ailim and f w nois.

cal n tron lux ise eve of 1- % f rat power (peak-o-peak have en re rted for the Ian0ef 1

to ha rcirc ation 1 op fl durin both invle nd du re io - op a ra on.

eutran lux n ise le s wh'h sig fican y bgnd t ese alues e onsid red in e th mal/me anica desigj of G WR el d a foun to e of egligi e con quenc

'In ditia, stab ity est at perati B s ha demon rated hat w n sta lity lated eutr n flu iit cycl osiillatons occ they resul in pe

-to-pe neu on f x iicit ses 5- 0 ti ts the pical alues Then ore, tions akeno r duc neutr f x no e leve excee ding t ee (3 times he ty ical alue re ffici nt a ens e early'dete ion o limit ycle n tron ux oscillatisf./

//

Z T pic ly, utnon ux n se le ls sh a gra al in rease in ab lute agni de s cargflow i incrsed onstan contro rod tter with wo reac or r circu ation ops i open ion.

erefor, the aselVe ne ron fl noi e I el ob ained a s cific ore flw can b appled ov a r ge of cSe f ows.

0 mai am a eason le va iation twee the 1 w fl and h h ow nd of he fl ran

, the ange er whic a sp ific aselVe is alied sho d not xceed 0% a rated ore f w with wa re ircul tion aops in op ation Dat from ests d oper ting p1 ts i icat that range f 20' o rate core ow wi res t in proxima ely a 0% i creas in neu on f ux aise vel d ring erati with wo rec'cula on 1 ps.

aseline ata shoul be t en ne the Paximum ad lin at w ich t>

najo ity of erat on will occur How er, b eline a ta n at ower ad ii s (i.e. ower power) wi resu in conse ative alue s ce e neu on fl noise evel s proporti nal t the p er lev at a give core flow.

3/4.4.2 SAFETY/RELIEF VALVES The safety valve function of the safety/relief valves operates to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code. A total of 12 OPERABLE safety/

relief valves is required to limit reactor pressure to within ASME III allow-able values for the worst case upset transient.

Demonstration of the safety/relief valve lift settings will occur only during shutdown. The safety/relief valves will be removed and either set pressure tested or replaced with spares which have been previously set pres-sure tested and stored in accordance with manufacturers recommendations in the specified frequency.

LIMERICK -

UNIT 2 B 3/4 4-2 Amendment No. 98 MAY 2 0 1999

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.9.1.9 Core Operating Limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the CORE OPERATING LIMITS REPORT for the following:

a.

The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1,

b.

MAPFAC(P) and MAPFAC(F) factors for Specification 3.2.1,

c.

The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.3,

d.

The MCPR(P) and MCPR(F) adjustment factors for specification 3.2.3,

e.

The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4,

f.

The power biased Rod Block Monitor setpoints and the Rod Block Monitor MCPR OPERABILITY limits of Specification 3.3.6.

9.

The Reactor Coolant System Recirculation Flow upscale trip setpoint and allowable value for Specification 3.3.6, 6.9.1.10 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document

)

a.

NEDE-24011-P-A "General Electric Standard Application for Reactor Fuel" (Latest approved revisional 6.9.1.11 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.

6.9.1.12 The CORE.OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

MAY.o 8 2000 LIMERICK - UNIT 2 6-18a Amendment No.

,*-.48,104

TECH SPEC MARKUP INSERT 17:

h.

The Oscillation Power Range Monitor (OPRM) period based detection algorithm (PBDA) setpoints for Specification 2.2.1.

INSERT 18:

b, NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

ATTACHMENT 3 LICENSE AMENDMENT REQUEST LIMERICK GENERATING STATION, UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 ACTIVATION OF THE TRIP OUTPUTS OF THE OSCILLATION POWER RANGE MONITOR PORTION OF THE POWER RANGE NEUTRON MONITORING SYSTEM TYPED TECHNICAL SPECIFICATIONS AND BASES PAGES FOR PROPOSED CHANGES UNIT 1 xi 2-4 B2-6 B2-7 B2-7a B2-10 3/4 3-1 3/43-la 3/43-2 3/4 3-4 3/4 3-5 3/4 3-6 3/ 4 3-7 3/43-8 3/ 4 3-58 3/4 3-60 3/4 3-60a UNIT 1 3/4 3-61 3/44-1 3/44-1a 3/4 4-2 3/4 4-3 B3/4 3-1 B3/ 4 3-1a B3/ 4 3-1b B3/4 3-1c B3/ 4 3-1d B3/ 4 3-1e B3/ 4 3-7 B3/4 4-1 B3/ 4 4-2 6-18a UNIT 2 xi 2-4 B2-6 B2-7 B2-7a B2-10 3/4 3-1 3/4 3-1a 3/43-2 3/4 3-4 3/43-5 3/43-6 3/43-7 3/4 3-8 3/ 4 3-58 3/4 3-60 3/4 3-60a UNIT 2 3/ 4 3-61 3/44-1 3/44-la 3/44-2 3/4 4-3 B3/ 4 3-1 B3/4 3-1 a B3/43-1b 83/43-1c B3/43-1d 13/43-1e B3/ 4 3-7 B3/4 4-1 B3/4 4-2 6-18a

INDEX IM LTI.1Q CONDITIONSEDQRAERATI-O.NA__SURVEILLANCE REQUIREMENTS 5ECTION PAGE REACTOR COOLANT SYSTEM (Continued)

Figure 3.4.1.1-1 Deleted.........

..................... 3/4 4-3 Jet Pumps.................................................. 3/4 4-4 Recirculation Pumps.................

........................ 3/4 4-5 Idle Recirculation Loop Startup........

..................... 3/4 4-6 3/4.4.2 SAFETY/RELIEF VALVES.

......................................... 3/4 4-7 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE' Leakage Detection Systems...........

........................ 3/4 4-8 Operational Leakage..................

....................... 3/4 4-9 Table 3.4.3.2-1 Reactor Coolant System Pressure Isolation Valves......

............... 3/4 4-11 3/4.4.4 CHEMISTRY.................................................. 3/4 4 -.12 Table 3.4.4-1 Reactor Coolant System Chemistry Limits................................... 3/4 4-14 3/4.4.5 SPECIFIC ACTIVITY..............

3/4 4-15 Table z.4.5-1 Primary Coolant Specific Activity Sample and Analysis Program.3/4 4-17 3/4.4.6 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System.....

3/4 4-18 Figure 3.4.6.1-1 Minimum Reactor Pressure Vessel Metal Temperature Vs. Reactor Vessel Pressure.....

3/4 4-20 Table 4.4.6.1.3-1 Deleted........

3/4 4-21 Reactor Steam Dome............................. 3/4 4-22 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES............................ 3/4 4-23 3/4.4.8 STRUCTURAL INTEGRITY..........

................... 3/4 4-24 I

LIMERICK - UNIT I xi Amendment No. 4-6.7,

TABLE-2-.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS ALLOWABLE VALUES FUNCTIONAL UNIT

1. Intermediate Range Monitor, Neutron Flux-High

2. Average Power Range Monitor:

a. Neutron Flux-Upscale (Setdown)

b. Simulated Thermal Power - Upscale:

- Two Recirculation Loop Operation

- Single Recirculation Loop Operation***

c. Neutron Flux - Upscale TRIP SETPOINT

• 120/125 divisions of full scale

• 15.02 of RATED THERMAL POWER

• 0.66 W + 62.8% and

• 116.62 of RATED THERMAL POWER

• 0.66 (W-7.6%) + 62.8% and

• 116.6% of RATED THERMAL POWER 118.32 of RATED THERMAL POWER

• 122/125 divisions of full scale

• 20.0% of RATED THERMAL POWER

• 0.66 W + 63.32 and

• 117.0% of RATED THERMAL POWER

• 0.66 (W-7.6%) + 63.3% and 5 117.0% of RATED THERMAL POWER 118.7% of RATED THERMAL POWER I

3.

4.

5.

6.

7.

8.

9.

10.

11.

12.

d. Inoperative

e. 2-Out-Of-4 Voter

f. OPRM Upscale Reactor Vessel Steam Dome Pressure - High Reactor Vessel Water Level - Low, Level 3 Main Steam Line Isolation Valve - Closure DELETED Drywell Pressure - High Scram Discharge Volume Water Level - High

a.

Level Transmitter

b.

Float Switch Turbine Stop Valve - Closure Turbine Control Valve Fast Closure, Trip Oil Pressure -

Low Reactor Mode Switch Shutdown Position Manual Scram N.A.

N.A.

N.A.

I

• 1096 psig 2 12.5 inches zero*

• 8% closed DELETED

• 1.68 psig

• 260' 9 5/8"

• 260' 9 5/8"

• 5% closed above instrument elevation**

elevation**

N.A.

N.A.

• 1103 psig 2 11.0 inches above instrument zero

• 12% closed DELETED

• 1.88 psig

• 261' 5 5/8" elevation

• 261' 5 5/8" elevation

• 7% closed 2 465 psig N.A.

N.A.

2 500 psig N.A.

N.A.

• See Bases Figure B 3/4.3-1.

• • Equivalent to 25.45 gallons/scram discharge volume.

• • • The 7.6% flow "offset" for Single Loop Operation (SLO) is applied for W 2 7.6%.

For flows W < 7.6%, the (W-7.6%) term is set equal to zero.

• • • • See COLR for OPRM period based detection algorithm trip setpoints.

OPRM Upscale trip output auto-enable (not bypassed) setpoints shall be APRM Simulated Thermal Power 2 30% and recirculation drive flow < 60%.

LIMERICK - UNIT 1 2-4 Amendment No. ;, 4O, &6, 89, 1-06, 441,

2.2 LIMITING SAFETY SYSTEM SETTTNGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System instrumentation setpoints specified in Table 2.2.1-1 are the values at which the reactor trips are set for each para-meter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist in mitigating the consequences of accidents.

Operation with-a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

1.

Intermediate Range Monitor. Neutron Flux - High The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5 decade 10 range instrument. The trip setpoint of 120 divisions of scale is active in each of the 10 ranges. Thus as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up.

The IRM instruments provide for overlap with both the APRM and SRM systems.

The most significant source of reactivity changes during the power increase is due to control rod withdrawal.

In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed.

The results of these analyses are in Section 15.4 of the FSAR. The most severe case involves an initial condition in which THERMAL POWER is at approximately 1% of RATED THERMAL POWER.

Additional conservatism was taken in this analysis by assuming the IRM channel closest to the control rod being withdrawn is bypassed.

The results of this analysis show that the reactor is shutdown and peak power is limited to 21% of RATED THERMAL POWER with the peak fuel enthalpy well below the fuel failure threshold of 170 cal/gm.

Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

2. Averaae Power Range Monitor The APRM system is divided into four APRM channels and four 2-Out-Of-4 Voter channels.

The four voter channels are divided into two groups of two each, with each group of two providing inputs to one RPS trip system. All four voters will trip (full scram) when any two unbypassed APRM channels exceed their trip setpoints.

APRM trip Functions 2.a, 2.b, 2.c, and 2.d are voted independently from OPRM Upscale Function 2.f. Therefore, any Function 2.a, 2.b, 2.c, or 2.d trip from any two unby passed APRM channels will result in a full trip in each of the four voter channels. Similarly, a Function 2.f trip from any two unbypassed APRM channels will result in a full trip from each of the four voter channels.

For operation at low pressure and low flow during STARTUP, the APRM Neutron Flux-Upscale (Setdown) scram setting of 15% of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits. The margin accommodates the anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system. Tempera-ture coefficients are small and control rod patterns are constrained by the RWM.

Of all the possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power increase.

LIMERICK - UNIT 1 B 2-6 Amendment No. -14, -14,

LIMITING SAFETY SYSTEM SETTINGS LASES_

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Average Power Range Monitor (Continued)

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant amount, the rate of power rise is very slow.

Generally the heat-flux is in near equilibrium with the fission rate.

In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit.

The 15% Neutron Flux - Upscale (Setdown) trip remains active until the mode switch is placed in the Run position.

The APRM trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the system and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Neutron Flux - Upscale setpoint; i.e.,

for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel.

For the Simulated Thermal Power - Upscale setpoint, a time constant of 6 +/- 0.6 seconds is introduced into the flow-biased APRM in order to simulate the fuel thermal transient characteristics. A more conservative maximum value is used for the flow-biased setpoint as shown in Table 2.2.1-1.

A reduced Trip Setpoint and Allowable Value is provided for the Simulated Thermal Power - Upscale Function, applicable when the plant is operating in Single Loop Operation (SLO) per LCO 3.4.1.1.

In SLO, the drive flow values (W) used in the Trip Setpoint and Allowable Value equations is reduced by 7.6%.

The 7.6% value is established to conservatively bound the inaccuracy created in the core flow/drive flow correlation due to back flow in the jet pumps associated with the inactive recirculation loop.

The Trip Setpoint and Allowable Value thus maintain thermal margins essentially unchanged from those for two-loop operation.

The Trip Setpoint and Allowable Value equations for single loop operation are only valid for flows down to W = 7.6%.

The Trip Setpoint and Allowable Value do not go below 62.8% and 63.3% RATED THERMAL POWER, respectively.

This is acceptable because back flow in the inactive recirculation loop is only an issue with drive flows of approximately 40% or greater (Reference 1).

The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unneces-sary shutdown.

The APRM channels also include an Oscillation Power Range Monitor (OPRM) Upscale Function.

The OPRM Upscale Function provides compliance with GDC 10 and GDC 12, thereby providing protection from exceeding the fuel MCPR Safety Limit due to anticipated thermal-hydraulic power oscillations. The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into "cells" for evaluation by the OPRM algorithms.

References 2, 3 and 4 describe three algorithms for detecting thermal-hydraulic instability related neutron flux oscillations:

the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. All three are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period based detection algorithm.

The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations.

OPRM Upscale Function OPERABILITY for Technical Specification purposes is based only on the period based detection algorithm.

LIMERICK - UNIT I B 2-7 Amendment No.

6X, 444,

LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Average Power Range Monitor (Continued)

The OPRM Upscale trip output shall be automatically enabled (not bypassed) when APRM Simulated Thermal Power is 2 30% and recirculation drive flow is < 60%

as indicated by APRM measured recirculation drive flow. (NOTE: 60% recirculation drive flow is the recirculation drive flow that corresponds to 60% of rated core flow.

Refer to TS Bases 3/4.3.1 for further discussion concerning the recirculation drive flow/core flow relationship.)

This is the operating region where actual thermal-hydraulic instability and related neutron flux oscillations may occur. See Reference 5 for additional discussion of OPRM Upscale trip enable region limits. These setpoints, which are sometimes referred to as the "auto-bypass" setpoints, establish the boundaries of the OPRM Upscale trip enabled region. The APRM Simulated Thermal Power auto-enable setpoint has 1% deadband while the drive flow setpoint has a 2% deadband. The deadband for these setpoints is established so that it increases the enabled region.

An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in a cell, with period confirmations and relative cell amplitude exceeding specified setpoints.

One or more cells in a channel exceeding the trip conditions will result in a channel trip.

An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithms detect oscillatory changes in the neutron flux for one or more cells in that channel.

There are four "sets" of OPRM related setpoints or adjustment parameters:

a) OPRM trip auto-enable setpoints for APRM Simulated Thermal Power (30%) and recirculation drive flow (60%); b) period based detection algorithm (PBDA) confirmation count and amplitude setpoints; c) period based detection algorithm tuning parameters; and d) growth rate algorithm (GRA) and amplitude based algorithm (ABA) setpoints.

The first set, the OPRM auto-enable region setpoints, are treated as nominal setpoints with no additional margins added as discussed in Reference 5.

The settings, 30% APRM Simulated Thermal Power and 60% recirculation drive flow, are defined (limit values) in a note to Table 2.2.1-1. The second set, the OPRM PBDA trip setpoints, are established in accordance with methodologies defined in Reference 4, and are documented in the COLR.

There are no allowable values for these setpoints. The third set, the OPRM PBDA "tuning" parameters, are established or adjusted in accordance with and controlled by station procedures.

The fourth set, the GRA and ABA setpoints, in accordance with References 2 and 3, are established as nominal values only, and controlled by station procedures.

3.

Reactor Vessel Steam Dome Pressure-High High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will a so tend to increase the power of the reactor by compressing voids thus adding reactivity.

The trip will quickly reduce the neutron flux, counteracting the pressure increase.

The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient.

This trip setpoint is effective at low power/flow conditions when the turbine stop valve and control fast closure trips are bypassed.

For a turbine trip or load rejection under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.

LIMERICK - UNIT 1 B 2-7a Amendment No.

X6, 441, l

LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

REFERENCES:

1.

NEDC-31300, "Single-Loop Operation Analysis for Limerick Generating Station, Unit 1," August 1986.

2.

NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

3.

NEDO-31960-A, Supplement 1, `BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

4.

NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

5.

BWROG Letter 96113, K. P. Donovan (BWROG) to L. E. Phillips (NRC),

"Guidelines for Stability Option III 'Enable Region' (TAC M92882),"

September 17, 1996.

LIMERICK - UNIT 1 B 2-10 Amendment No.

I

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

APPLICABILITY:

As shown in Table 3.3.1-1.

ACTION:

Note:

Separate condition entry is allowed for each channel.

a.

With the number of OPERABLE channels in either trip system for one or more Functional Units less than the Minimum OPERABLE Channels per Trip System required by Table 3.3.1-1, within one hour for each affected functional unit either verify that at least one* channel in each trip system is OPERABLE or tripped or that the trip system is tripped, or place either the affected trip system or at least one inoperable channel in the affected trip system in the tripped condition.

b.

With the number of OPERABLE channels in either tip system less than the Minimum OPERABLE Channels per Trip System required by Table 3.3.1-1, place either the inoperable channel(s) or the affected trip system** in the tripped conditions within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.***

c.

With the number of OPERABLE channels in both trip systems for one or more Functional Units less than the Minimum OPERABLE Channels per Trip System required by Table 3.3.1-1, place either the inoperable channel(s) in one trip system or one trip system in the tripped condition within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s**.***

d.

If within the allowable time allocated by Actions a, b or c, it is not desired to place the inoperable channel or trip system in trip (e.g., full scram would occur), Then no later than expiration of that allowable time initiate the action identified in Table 3.3.1-1 for the applicable Functional Unit.

• For Functional Units 2.a, 2.b, 2.c, 2.d, and 2.f, at least two channels shall be OPERABLE or tripped.

For Functional Unit 5, both trip systems shall have each channel associated with the MSIVs in three main steam lines (not necessarily the same main steam lines for both trip systems) OPERABLE or tripped.

For Function 9, at least three channels per trip system shall be OPERABLE or tripped.

• • For Functional Units 2.a, 2.b, 2.c, 2.d, and 2.f, inoperable channels shall be placed in the tripped condition to comply with Action b. Action c does not apply for these Functional Units.

• • • A channel or trip system which has been placed in the tripped condition to satisfy Action b. or c. may be returned to the untripped condition under administrative control for up to two hours solely to perform testing required to demonstrate its operability or the operability of other equipment provided Action a. continues to be satisfied.

LIMERICK - UNIT 1 3/4 3-1 Amendment No..&3, 74, 141,

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 24 months, except Table 4.3.1.1-1 Functions 2.a, 2.b, 2.c, 2.d, 2.e and 2.f.

Functions 2.a, 2.b, 2.c, 2.d, and 2.f do not require separate LOGIC SYSTEM FUNCTIONAL TESTS.

For Function 2.e, tests shall be performed at least once per 24 months.

LOGIC SYSTEM FUNCTIONAL TEST for Function 2.e includes simulating APRM and OPRM trip conditions at the APRM channel inputs to the voter channel to check all combinations of two tripped inputs to the 2-Out-Of-4 voter logic in the voter channels.

4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its limit at least once per 24 months.

Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 24 months where N is the total number of redundant channels in a specific reactor trip system.

LIMERICK - UNIT 1 3/4 3-la Amendment No.

1-4F,

TABLE 3.3.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION APPLICABLE OPERATIONAL CONDITIONS FUNCTIONAL UNIT

1.

Intermediate Range Monitors"b:

a.

Neutron Flux - High 2

3(i),

5(i) 2 3(i),

5(i )

MINIMUM OPERABLE CHANNELS PER TRIP SYSTEM (a) 3 3

3(d) 3 3

3(d)

ACTION 4(i) 4(i) 1 2

3 1

2 3

b.

Inoperative

2.

Average Power Range Monitor(e):

a.

Neutron Flux - Upscale (Setdown)

b.

Simulated Thermal Power - Upscale

c.

Neutron Flux - Upscale

d.

Inoperative

e.

2-Out-Of-4 Voter

f.

OPRM Upscale

3.

Reactor Vessel Steam Dome Pressure - High

4.

Reactor Vessel Water Level - Low, Level 3

5.

Main Steam Line Isolation Valve-Closure 2

1 1

1, 2 1, 2((

1(o)(p) 3(m) 3(m) 3(m) 3(m) 2 3(m) 1 4

4 1

1 10 I

1, 2(f) 2 1

1, 2 2

1 1(g) 1/valve 4

LIMERICK - UNIT 1 3/4 3 -2 Amendment No. 28, 4-,

1414, 449,

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION STATEMENTS ACTION 1 Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 2 Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the SHUTDOWN position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 3 Suspend all operations involving CORE ALTERATIONS and insert all insertable control rods within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 4 Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 5 Be in STARTUP with the main steam line isolation valves closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 6 Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure until the function is automatically bypassed, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

ACTION 7 Verify all insertable control rods to be inserted within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 8 Lock the reactor mode switch in the Shutdown position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 9 Suspend all operations involving CORE ALTERATIONS, and insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 10

a.

If the condition exists due to a common-mode OPRM deficiency*, then initiate alternate method to detect and suppress thermal-hydraulic instability oscillations within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND restore required channels to OPERABLE status within 120 days, OR

b.

Reduce THERMAL POWER to < 25X RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Unanticipated characteristic of the instability detection algorithm or equipment that renders all OPRM channels inoperable at once.

LIMERICK - UNIT 1 3/4 3-4 Amendment No..141, -149,

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

(b)

This function shall be automatically bypassed when the reactor mode switch is in the Run position.

(c)

DELETED (d)

The noncoincident NMS reactor trip function logic is such that all channels go to both trip systems.

Therefore, when the "shorting links" are removed, the Minimum OPERABLE Channels Per Trip System is 6 IRMs.

(e)

An APRM channel is inoperable if there are less than 3 LPRM inputs per level or less than 20 LPRM inputs to an APRM channel, or if more than 9 LPRM inputs to the APRM channel have been bypassed since the last APRM calibration (weekly gain calibration).

(f)

This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

(g)

This function shall be automatically bypassed when the reactor mode switch is not in the Run position.

(h)

This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

Ci)

With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

Ci)

This function shall be automatically bypassed when turbine first stage pressure is equivalent to a THERMAL POWER of less than 30% of RATED THERMAL POWER.

Ck)

Also actuates the EOC-RPT system.

(l)

DELETED (m)

Each APRM channel provides inputs to both trip systems.

(n)

DELETED (o)

With THERMAL POWER 2 25% RATED THERMAL POWER.

The OPRM Upscale trip output shall be automatically enabled (not bypassed) when APRM Simulated Thermal Power is 2 30% and recirculation drive flow is < 60%.

The OPRM trip output may be automatically bypassed when APRM Simulated Thermal Power is < 30% or recirculation drive flow is 2 60%.

(p) A minimum of 23 cells, each with a minimum of 2 OPERABLE LPRMs, must be OPERABLE for an OPRM channel to be OPERABLE.

LIMERICK - UNIT 1 3/4 3-5 Amendment No. 44,

1 4144,

TABLE 3.3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES RESPONSE TIME FUNCTIONAL UNIT (Seconds)

1.

Intermediate Range Monitors:

a.

Neutron Flux - High N.A.

b.

Inoperative N;A.

2. Average Power Range Monitor*:

a.

Neutron Flux - Upscale (Setdown)

N.A.

b.

Simulated Thermal Power - Upscale N.A.

c.

Neutron Flux - Upscale N.A.

d.

Inoperative N.A.

e.

2-Out-Of-4 Voter

• 0.05*

f.

OPRM Upscale N.A.

3.

Reactor Vessel Steam Dome Pressure - High

• 0.55

4.

Reactor Vessel Water Level - Low, Level 3

• 1.05#

5. Main Steam Line Isolation Valve - Closure

• 0.06

6.

DELETED DELETED

7.

Drywell Pressure - High N.A.

8.

Scram Discharge Volume Water Level - High

a.

Level Transmitter N.A.

b.

Float Switch N.A.

9.

Turbine Stop Valve - Closure

• 0.06

10.

Turbine Control Valve Fast Closure, Trip Oil Pressure - Low

<0.08**

11.

Reactor Mode Switch Shutdown Position N.A.

12.

Manual Scram N.A.

• Neutron detectors, APRM channel and 2-Out-Of-4 Voter channel digital electronics are exempt from response time testing. Response time shall be measured from activation of the 2-Out-Of-4 Voter output relay For applications of Specification 4.3.1.3, the redundant outputs from each 2-Out-Of-4 Voter channel are considered part of the same channel, but the OPRM and APRM outputs are considered to be separate channels, so N = 8. Testing of OPRM and APRM outputs shall alternate.

• • Measured from start of turbine control valve fast closure.

1. Sensor is eliminated from response time testing for the RPS circuits.

Response time testing and conformance to the administrative limits for the remaining channel including trip unit and relay logic are required.

LIMERICK - UNIT 1 3/4 3-6 Amendment No. 89, 4,22., 44-1,

TABLE 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL FUNCTIONAL CHANNEL CHECK TEST CALIBRATION(a)

FUNCTIONAL UNIT

1. Intermediate Range Mon

a.

Neutron Flux -

OPERATIONAL CONDITIONS FOR WHICH SURVEILLANCE REQUIRED itors:

High

b.

Inoperative

2.

Average Power Range Monitor(f):

a.

Neutron Flux - Upscale (Setdown)

b.

Simulated Thermal Power -

Upscale

c.

Neutron Flux - Upscale

d.

Inoperative

e.

2-Out-Of-4 Voter

f.

OPRM Upscale

3.

Reactor Vessel Steam Dome Pressure - High

4.

Reactor Vessel Water Level-Low, Level 3

5.

Main Steam Line Isolation Valve - Closure

6.

DELETED

7.

Drywell Pressure - High

8.

Scram Discharge Volume Water Level - High

a. Level Transmitter

b. Float Switch S(b)

S N.A.

D(b)

D D

N.A.

D D

W W(j)

W(j)

SA(l)

SACe)

SA SA SA SA(e)

R R

N.A.

R W(d), R(g)

W(d), R N.A.

N.A.

R(c)(g)

R 2

3(i), 4(i), 5(i) 2, 3(i), 4(i), 5(i) 2 1

1 1, 2 1, 2 1(m)

I S

Q S

Q R

1, 2(h) 1, 2 1

DELETED 1, 2 N.A.

DELETED S

S N.A.

0 DELETED Q

Q Q

R DELETED R

R R

1, 2, 5(i) 1, 2, 5(i)

LIMERICK UNIT I 3/4 3-7 Amendment No. 44, 53, 89, 4-9,

TABLE 4.3.1.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION(a)

SURVEILLANCE REOUIRED

9. Turbine Stop Valve - Closure N.A.

Q R

1

10.

Turbine Control Valve Fast Closure, Trip Oil Pressure - Low N.A.

Q R

1

11.

Reactor Mode Switch Shutdown Position N.A.

R N.A.

1, 2, 3, 4, 5

12.

Manual Scram N.A.

W N.A.

1, 2, 3, 4, 5 (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for at least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days.

(c) Calibration includes verification that the OPRM Upscale trip auto-enable (not-bypass) setpoint for APRM Simulated Thermal Power is 2 30% and for recirculation drive flow is < 60%.

(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER 225% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER.

(e) CHANNEL FUNCTIONAL TEST shall include the flow input function, excluding the flow transmitter.

(f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH).

(g) Calibration includes the flow input function.

(h) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

(i) With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(j) If the RPS shorting links are required to be removed per Specification 3.9.2, they may be reinstalled for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance. During this time, CORE ALTERATIONS shall be suspended, and no control rod shall be moved from its existing position.

Ck)

DELETED (l) Not required to be performed when entering OPERATIONAL CONDITION 2 from OPERATIONAL CONDITION 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering OPERATIONAL CONDITION 2.

(m) With THERMAL POWER 2 25% of RATED THERMAL POWER.

LIMERICK - UNIT 1 3/4 3-8 Amendment No. 29, 4A, 63, 66, 4113, 417, ACT, -11,

TABLE 3.3.6-1 CONTROL ROD BLOCK INSTRUMENTATION MINIMUM APPLICABLE OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION PER TRIP FUNCTION CONDITIONS ACTION

1. ROD BLOCK MONITOR (a)

a. Upscale 2

1*

60

b. Inoperative 2

1*

60

c. Downscale 2

1*

60

2. APRM

a. Simulated Thermal Power - Upscale 3

1 61

b. Inoperative 3

1, 2 61

c. Neutron Flux - Downscale 3

1 61

d. Simulated Thermal Power - Upscale (Setdown) 3 2

61

e.

Recirculation Flow - Upscale 3

1 61

f.

LPRM Low Count 3

1, 2 61

3.

SOURCE RANGE MONITORS ***

a.

Detector not full in(b')

3 2

61 2

5 61

b. Upscale(c) 3 2

61 2

5 61

c.

Inoperativee')

3 2

61 2

5 61

d. Downscale(d) 3 2

61 2

5 61

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in 6

2, 5**

61

b. Upscale 6

2, 5**

61

c. Inoperative 6

2, 5**

61

d. Downscale(e) 6 2, 5**

61

5. SCRAM DISCHARGE VOLUME

a. Water Level-High 2

1, 2, 5**

62

6. DELETED DELETED DELETED DELETED

7. REACTOR MODE SWITCH SHUTDOWN POSITION 2

3, 4 63 LIMERICK - UNIT I 3/4 3 -58 Amendment No. 4, 41, 144,

TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP SETPOINT TRIP FUNCTION ALLOWABLE VALUE

1. ROD BLOCK MONITOR

a. Upscal el,"

1) Low Trip Setpoint (LTSP)

2) Intermediate Trip Setpoint (ITSP)

3) High Trip Setpoint (HTSP)

b. Inoperative

c. Downscale (DTSP)

d. Power Range Setpoint'b)

1) Low Power Setpoint (LPSP)

2) Intermediate Power Setpoint (IPSP)

3) High Power Setpoint (HPSP)

2. APRM

a. Simulated Thermal Power - Upscale Two Recirculation Loop Operation N/A N/A 28.1% RATED THERMAL POWER 63.1% RATED THERMAL POWER 83.1% RATED THERMAL POWER

• 0.66 W + 55.2% and

• 108.0% of RATED THERMAL POWER 28.4% RATED THERMAL POWER 63.4% RATED THERMAL POWER 83.4% RATED THERMAL POWER

• 0.66 W + 55.7%

• 108.4% of RATED THERMAL POWER Single Recirculation Loop Operation****

• 0.66 (W-7.6%) + 55.2% and

• 108.0% of RATED THERMAL POWER 5 0.66 (W-7.6%) +

• 108.4% of RATED THERMAL POWER 55.7% and

b. Inoperative

c. Neutron Flux - Downscale

d. Simulated Thermal Power - Upscale (Setdown)

e.

Recirculation Flow - Upscale

f.

LPRM Low Count N.A.

N.A.

2 3.2% of RATED THERMAL POWER

• 12.0% of RATED THERMAL POWER 2 2.8% of RATED THERMAL POWER

• 13.0% of RATED THERMAL POWER 20 per channel

< 3 per axial level

3. SOURCE RANGE MONITORS

a. Detector not full in

b.

Upscale

c.

Inoperative

d. Downscale N.A.

• 1 x 105 cps N.A.

2 3 cps**

20 per channel

< 3 per axial level N.A.

• 1.6 x 105 cps N.A.

2 1.8 cps**

LIMERICK - UNIT 1 3/4 3-60 Amendment No..', 4-9,

-N, 6X, 406, 441,

TABLE 3.3.6-2 (continued)

CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in

b. Upscale c.

d.

Inoperative Downscale N.A.

• 108/125 divisions of full scale N.A.

2 5/125 divisions of full scale

• 257' 5 9/16" elevation***

N.A.

• 110/125 divisions of full scale N.A.

2 3/125 divisions of full scale

• 257' 7 9/16" elevation

5. SCRAM DISCHARGE VOLUME

a. Water Level-High

a. Float Switch

6. DELETED

7. REACTOR MODE SWITCH SHUTDOWN POSITION

• Refer to the COLR for these setpoints.

DELETED DELETED N.A.

N.A.

May be reduced provided the Source Range above the curve shown in Figure 3.3.6-1.

Monitor has an observed count rate and signal-to-noise ratio on or Equivalent to 13 gallons/scram discharge volume.

• • • • The 7.6% flow "offset" for Single Loop Operation (W-7.6%)

term is set equal to zero.

(SLO) is applied for W 2 7.6%.

For flows W < 7.6%, the I

(a)

There are three upscale trip levels. Each is applicable only over its specified operating core thermal power range.

All RBM trips are automatically bypassed below the low power setpoint (LPSP).

The upscale LTSP is applied between the low power setpoint (LPSP) and the intermediate power setpoint (IPSP). The upscale ITSP is applied between the intermediate power setpoint and the high power setpoint (HPSP).

The HTSP is applied above the high power setpoint.

(b)

Power range setpoints control enforcement of appropriate upscale trips over the proper core thermal power ranges.

The power signal to the RBM is provided by the APRM.

LIMERICK - UNIT 1 3/4 3-60a Amendment No. 3, 30, 34, 3-X, X,

1-4-,

TABLE 4.a36-1 CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP FUNCTION CHANNEL CHECK CHANNEL FUNCTIONAL TEST CHANNEL CALIBRATION(a)

OPERATIONAL CONDITIONS FOR WHICH SURVEILLANCE REOUIRED

1. ROD BLOCK MONITOR a.

b.

c.

Upscale Inoperative Downscale N.A.

N.A.

N.A.

Q(C)

Q(C)

Q(C)

R N.A.

R 1*

1*

1*

2.

APRM

a. Simulated Thermal Power-Upscale

b. Inoperative

c. Neutron Flux - Downscale

d. Simulated Thermal Power -

Upscale (Setdown)

e. Recirculation Flow - Upscale

f. LPRM Low Count N.A.

N.A.

N.A.

SA SA SA R

N.A.

R 1

1, 2 1

2 1

1, 2 N.A.

N.A.

N.A.

SA SA SA R

R R

3. SOURCE RANGE MONITORS a.

b.

c.

d.

Detector not full Upscale Inoperative Downscale in N.A.

N.A.

N.A.

N.A.

M(d)(e) W(f)

M(d)(e)

W(f)

M(d)(e) w(f)

N.A.

R N.A.

R 2,

2,

2,

2,

5 5

5 5

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in

b. Upscale

c. Inoperative

d. Downscale

5. SCRAM DISCHARGE VOLUME N.A.

N.A.

N.A.

N.A.

W W.-

W W

N.A.

R N.A.

R 2,

2,

2,

2,

a. Water Level - High N.A.

Q R

6. DELETED DELETED DELETED DELETED 1, 2, 5**

DELETED 3, 4

7. REACTOR MODE SWITCH SHUTDOWN POSITION N.A.

N.A.

LIMERICK - UNIT 1 3/4 3-61 Amendment No. 41, 6.3, 6X, 99,

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LTMITING CONDITION FOR OPFRATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.

APPLICABILITY:

OPERATIONAL CONDITIONS 1* and 2*.

ACTION:

a.

With one reactor coolant system recirculation loop not in operation:

1.

Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

a.

Place the recirculation flow control system in the Local Manual mode, and

b.

Reduce THERMAL POWER to

• 76.2% of RATED THERMAL POWER,

and,

c.

Limit the speed of the operating recirculation pump to less than or equal to 90% of rated pump speed, and

d.

Verify that the differential temperature requirements of Surveillance Requirement 4.4.1.1.5 are met if THERMAL POWER is < 30% of RATED THERMAL POWER or the recirculation loop flow in the operating loop is < 50% of rated loop flow, or suspend the THERMAL POWER or recirculation loop flow increase.

• See Special Test Exception 3.10.4.

LIMERICK - UNIT 1 3/4 4-1 Amendment No. 30, 6X, 4A6,

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

(Continued)

2.

Within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s:

Reduce the Average Power Range Monitor (APRM) Simulated Thermal Power Upscale Scram and Rod Block Trip Setpoints and Allowable Values, to those applicable for single recirculation loop operation per Specifications 2.2.1 and 3.3.6, or declare the associated channel(s) inoperable and take the actions required by the referenced specifications.

3.

Otherwise be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

With no reactor coolant system recirculation loops in operation, initiate measures to place the unit in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

LIMERICK - UNIT 1 3/4 4-la Amendment No. 3O, 66, 4414, 6-9,

REACTOR COOLANT SYSTEM SURVEILLANCE R-EQUIREME N___

4.4.1.1.1 4.4.1.1.2 4.4.1.1.3 DELETED DELETED DELETED 4.4.1.1.4 With one reactor coolant system recirculation loop not in operation, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verify that:

a.

Reactor THERMAL POWER is

• 76.2% of RATED THERMAL POWER,

b.

The recirculation flow control system is in the Local Manual mode, and

c.

The speed of the operating recirculation pump is

• 90% of rated pump speed.

I 4.4.1.1.5 With one reactor coolant system recirculation loop not in operation, within 15 minutes prior to either THERMAL POWER increase or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is

• 30% of RATED THERMAL POWER or the recirculation loop flow in the operating recirculation loop is

• 50% of rated loop flow.

a.

• 1450F between reactor vessel steam space coolant and bottom head drain line coolant,

b.

• 50'F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and

c.

< 50'F between the reactor coolant within the loop not in operation and the operating loop.

The differential temperature requirements of Specification 4.4.1.1.5b. and c.

do not apply when the loop not in operation is isolated from the reactor pressure vessel.

LIMERICK - UNIT 1 3/4 4-2 Amendment No. 3 4-4,

CONTENTS OF THIS PAGE HAVE BEEN DELETED LIMERICK - UNIT 1 3/4 4-3 Amendment No. 3G, 4-6,

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

a.

Preserve the integrity of the fuel cladding.

b.

Preserve the integrity of the reactor coolant system.

c.

Minimize the energy which must be adsorbed following a loss-of-coolant accident, and

d.

Prevent inadvertent criticality.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance.

When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.

The reactor protection system is made up of two independent trip systems.

There are usually four channels to monitor each parameter with two channels in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system.

The tripping of both trip systems will produce a reactor scram.

The APRM system is divided into four APRM channels and four 2-Out-Of-4 Voter channels.

Each APRM channel provides inputs to each of the four voter channels.

The four voter channels are divided into two groups of two each, with each group of two providing inputs to one RPS trip system. The system is designed to allow one APRM channel, but no voter channels, to be bypassed.

The system meets the intent of IEEE-279 for nuclear power plant protection systems.

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System" and NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function." The bases for the trip settings of the RPS are discussed in the bases for Specification 2.2.1.

The APRM Functions include five Functions accomplished by the four APRM channels (Functions 2.a, 2.b, 2.c, 2.d, and 2.f) and one accomplished by the four 2-Out-Of-4 Voter channels (Function 2.e).

Two of the five Functions accomplished by the APRM channels are based on neutron flux only (Functions 2.a and 2.c), one Function is based on neutron flux and recirculation drive flow (Function 2.b) and one is based on equipment status (Function 2.d).

The fifth Function accomplished by the APRM channels is the Oscillation Power Range Monitor (OPRM) Upscale trip Function 2.f, which is based on detecting oscillatory characteristics in the neutron flux. The OPRM Upscale Function is also dependent on average neutron flux (Simulated Thermal Power) and recirculation drive flow, which are used to automatically enable the output trip.

The Two-Out-Of-Four Logic Module includes 2-Out-Of-4 Voter hardware and the APRM Interface hardware.

The 2-Out-Of-4 Voter Function 2.e votes APRM Functions 2.a, 2.b, 2.c, and 2.d independently of Function 2.f.

This voting is accomplished by the 2-Out-Of-4 Voter hardware in the Two-Out-Of-Four Logic Module.

The voter includes separate outputs to RPS for the two independently voted sets of Functions, each of which is redundant (four total outputs). The analysis in Reference 2 took credit for this redundancy in the justification of the 12-hour allowed out-of-service time for LIMERICK - UNIT 1 B 3/4 3-1 Amendment No. 63, 89, 41321, 441,

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

Action b, so the voter Function 2.e must be declared inoperable if any of its functionality is inoperable.

The voter Function 2.e does not need to be declared inoperable due to any failure affecting only the APRM Interface hardware portion of the Two-Out-Of-Four Logic Module.

Three of the four APRM channels and all four of the voter channels are required to be OPERABLE to ensure that no single failure will preclude a scram on a valid signal.

To provide adequate coverage of the entire core, consistent with the design bases for the APRM Functions 2.a, 2.b, and 2.c, at least 20 LPRM inputs, with at least three LPRM inputs from each of the four axial levels at which the LPRMs are located, must be operable for each APRM channel.

In addition, no more than 9 LPRMs may be bypassed between APRM calibrations (weekly gain adjustments).

For the OPRM Upscale Function 2.f, LPRMs are assigned to "cells" of 3 or 4 detectors. A minimum of 23 cells (Reference 9), each with a minimum of 2 OPERABLE LPRMs, must be OPERABLE for each APRM channel for the OPRM Upscale Function 2.f to be OPERABLE in that channel.

References 4, 5 and 6 describe three algorithms for detecting thermal-hydraulic instability related neutron flux oscillations:

the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm.

All three are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period based detection algorithm.

The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations.

OPRM Upscale Function OPERABILITY for Technical Specification purposes is based only on the period based detection algorithm.

An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in any cell, with period confirmations and relative cell amplitude exceeding specified setpoints.

One or more cells in a channel exceeding the trip conditions will result in a channel trip.

An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithms detect growing oscillatory changes in the neutron flux for one or more cells in that channel.

The OPRM Upscale Function is required to be OPERABLE when the plant is at 2 25% RATED THERMAL POWER.

The 25% RATED THERMAL POWER level is selected to provide margin in the unlikely event that a reactor power increase transient occurring while the plant is operating below 30% RATED THERMAL POWER causes a power increase to or beyond the 30% RATED THERMAL POWER OPRM Upscale trip auto-enable point without operator action. This OPERABILITY requirement assures that the OPRM Upscale trip automatic-enable function will be OPERABLE when required.

Actions a, b and c define the Action(s) required when RPS channels are discovered to be inoperable.

For those Actions, separate entry condition is allowed for each inoperable RPS channel.

Separate entry means that the allowable time clock(s) for Actions a, b or c start upon discovery of inoperability for that specific channel.

Restoration of an inoperable RPS channel satisfies only the action statements for that particular channel.

Action statement(s) for remaining inoperable channel(s) must be met according to their original entry time.

Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> has been shown to be acceptable (NEDC-30851P-A and NEDC-32410P-A) to permit restoration of any inoperable channel to OPERABLE status.

However, this out of service time is only acceptable provided that the associated Function's (identified as a "Functional Unit" in Table 3.3.1-1) inoperable channel is in one trip system and the Function still maintains RPS trip capability.

LIMERICK - UNIT 1 B 3/4 3 -1 a Amendment No. -53, 89, 4-3k, 141,

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

The requirements of Action a are intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same trip system for the same Function result in the Function not maintaining RPS trip capability.

A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal.

For the typical Function with one-out-of-two taken twice logic, including the IRM Functions and APRM Function 2.e (trip capability associated with APRM Functions 2.a, 2.b, 2.c, 2.d, and 2.f are discussed below), this would require both trip systems to have one channel OPERABLE or in trip (or the associated trip system in trip).

For Function 5 (Main Steam Isolation Valve--Closure), this would require both trip systems to have each channel associated with the MSIVs in three main steam lines (not necessarily the same main steam lines for both trip systems) OPERABLE or in trip (or the associated trip system in trip).

For Function 9 (Turbine Stop Valve-Closure), this would require both trip systems to have three channels, each OPERABLE or in trip (or the associated trip system in trip).

The completion time to satisfy the requirements of Action a is intended to allow the operator time to evaluate and repair any discovered inoperabilities.

The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.

With trip capability maintained, i.e., Action a satisfied, Actions b and c as applicable must still be satisfied.

If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Action b requires that the channel or the associated trip system must be placed in the tripped condition.

Placing the inoperable channel in trip (or the associated trip system in trip) would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.

As noted, placing the trip system in trip is not applicable to satisfy Action b for APRM functions 2.a, 2.b, 2.c, 2.d, or 2.f.

Inoperability of one required APRM channel affects both trip systems.

For that condition, the Action b requirements can only be satisfied by placing the inoperable APRM channel in trip.

Restoring OPERABILITY or placing the inoperable APRM channel in trip are the only actions that will restore capability to accommodate a single APRM channel failure.

Inoperability of more than one required APRM channel of the same trip function results in loss of trip capability and the requirement to satisfy Action a.

The requirements of Action c must be satisfied when, for any one or more Functions, at least one required channel is inoperable in each trip system.

In this condition, provided at least one channel per trip system is OPERABLE, normally the RPS still maintains trip capability for that Function, but cannot accommodate a single failure in either trip system (see additional bases discussion above related to loss of trip capability and the requirements of Action a, and special cases for Functions 2.a, 2.b, 2.c, 2.d, 2.f, 5 and 9).

LIMERICK - UNIT 1 B 3/4 3-lb Amendment No. 4-44,

3/4.3 INSTRUMENTATION BASES_

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

The requirements of Action c limit the time the RPS scram logic, for any Function, would not accommodate single failure in both tip systems (e.g., one-out-of-one and one-out-of-one arrangement for a typical four channel Function).

The reduced reliability of this logic arrangement was not evaluated in NEDC-30851P-A for the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time.

Within the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance, the associated Function must have all required channels OPERABLE or in trip (or any combination) in one trip system.

Completing the actions required by Action c restores RPS to a reliability level equivalent to that evaluated in NEDC-30851P-A, which justified a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowable out of service time as allowed by Action b. To satisfy the requirements of Action c, the trip system in the more degraded state should be placed in trip or, alternatively, all the inoperable channels in that trip system should be placed in trip (e.g., a trip system with two inoperable channels could be in a more degraded state than a trip system with four inoperable channels if the two inoperable channels are in the same Function while the four inoperable channels are all in different Functions).

The decision of which trip system is in the more degraded state should be based on prudent judgment and take into account current plant conditions (i.e.,

what OPERATIONAL CONDITION the plant is in).

If this action would result in a scram or RPT, it is permissible to place the other trip system or its inoperable channels in trip.

The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowable out of service time is judged acceptable based on the remaining capability to trip, the diversity of the sensors-available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram.

As noted, Action c is not applicable for APRM Functions 2.a, 2.b, 2.c, 2.d, or 2.f.

Inoperability of an APRM channel affects both trip systems and is not associated with a specific trip system as are the APRM 2-Out-Of-4 voter and other non-APRM channels for which Action c applies.

For an inoperable APRM channel, the requirements of Action b can only be satisfied by tripping the inoperable APRM channel.

Restoring OPERABILITY or placing the inoperable APRM channel in trip are the only actions that will restore capability to accommodate a single APRM channel failure.

If it is not desired to place the channel (or trip system) in trip to satisfy the requirements of Action a, Action b or Action c (e.g., as in the case where placing the inoperable channel in trip would result in a full scram), Action d requires that the Action defined by Table 3.3.1-1 for the applicable Function be initiated immediately upon expiration of the allowable out of service time.

Table 3.3.1-1, Function 2.f, references Action 10, which defines the action required if OPRM Upscale trip capability is not maintained. Action 10b is required to address identified equipment failures. Action 10a is to address common mode vendor/industry identified issues that render all four OPRM channels inoperable at once.

For this condition, References 2 and 3 justified use of alternate methods to detect and suppress oscillations for a limited period of time, up to 120 days. The alternate methods are procedurally established consistent with the guidelines identified in Reference 7 requiring manual operator action to scram the plant if certain predefined events occur.

The 12-hour allowed completion time to implement the alternate methods is based on engineering judgment to allow orderly transition to the alternate methods while limiting the period of time during which no automatic or alternate detect and suppress trip capability is formally in place.

The 120-day period during which use of alternate methods is allowed is intended to be an outside LIMERICK - UNIT 1 B 3/4 3 -1c Amendment No.

1441,

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued) limit to allow for the case where design changes or extensive analysis might be required to understand or correct some unanticipated characteristic of the instability detection algorithms or equipment.

The evaluation of the use of alternate methods concluded, based on engineering judgment, that the likelihood of an instability event that could not be adequately handled by the alternate methods during the 120-day period was negligibly small.

Plant startup may continue while operating within the allowed completion time of Action 10a.

The primary purpose of this is to allow an orderly completion, without undue impact on plant operation, of design and verification activities in the event of a required design change to the OPRM Upscale function.

This exception is not intended as an alternative to restoring inoperable equipment to OPERABLE status in a timely manner.

Action 10a is not intended and was not evaluated as a routine alternative to returning failed or inoperable equipment to OPERABLE status.

Correction of routine equipment failure or inoperability is expected to be accomplished within the completion times allowed for LCO 3.3.1 Action a or Action b, as applicable.

Action 10b applies when routine equipment OPERABILITY cannot be restored within the allowed completion times of LCO 3.3.1 Actions a or b, or if a common mode OPRM deficiency cannot be corrected and OPERABILITY of the OPRM Upscale Function restored within the 120-day allowed completion time of Action 10a.

The OPRM Upscale trip output shall be automatically enabled (not-bypassed) when APRM Simulated Thermal Power is 2 30% and recirculation drive flow is < 60%

as indicated by APRM measured recirculation drive flow.

NOTE:

60% recirculation drive flow is the recirculation drive flow that corresponds to 60% of rated core flow. This is the operating region where actual thermal-hydraulic instability and related neutron flux oscillations may occur. As noted in Table 4.3.1.1-1, Note c, CHANNEL CALIBRATION for the OPRM Upscale trip Function 2.f includes confirming that the auto-enable (not-bypassed) setpoints are correct.

Other surveillances ensure that the APRM Simulated Thermal Power properly correlates with THERMAL POWER (Table 4.3.1.1-1, Note d) and that recirculation drive flow properly correlates with core flow (Table 4.3.1.1-1, Note g).

If any OPRM Upscale trip auto-enable setpoint is exceeded and the OPRM Upscale trip is not enabled, i.e., the OPRM Upscale trip is bypassed when APRM Simulated Thermal Power is 2 30% and recirculation drive flow is < 60%, then the affected channel is considered inoperable for the OPRM Upscale Function.

Alternatively, the OPRM Upscale trip auto-enable setpoint(s) may be adjusted to place the channel in the enabled condition (not-bypassed).

If the OPRM Upscale trip is placed in the enabled condition, the surveillance requirement is met and the channel is considered OPERABLE.

As noted in Table 4.3.1.1-1, Note g, CHANNEL CALIBRATION for the APRM Simulated Thermal Power - Upscale Function 2.b and the OPRM Upscale Function 2.f, includes the recirculation drive flow input function. The APRM Simulated Thermal Power - Upscale Function and the OPRM Upscale Function both require a valid drive flow signal.

The APRM Simulated Thermal Power - Upscale Function uses drive flow to vary the trip setpoint.

The OPRM Upscale Function uses drive flow to automatically enable or bypass the OPRM Upscale trip output to RPS.

A CHANNEL CALIBRATION of the APRM recirculation drive flow input function requires both calibrating the drive flow transmitters and establishing a valid drive flow /

core flow relationship.

The drive flow / core flow relationship is established once per refuel cycle, while operating within 10% of rated core flow and within LIMERICK - UNIT 1 B 3/4 3 -1d Amendment No.

I

3/4.3 INSTRUMENTATION

_ASE_

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued) 10% of RATED THERMAL POWER.

Plant operational experience has shown that this flow correlation methodology is consistent with the guidance and intent in Reference 8. Changes throughout the cycle in the drive flow / core flow relationship due to the changing thermal hydraulic operating conditions of the core are accounted for in the margins included in the bases or analyses used to establish the setpoints for the APRM Simulated Thermal Power - Upscale Function and the OPRM Upscale Function.

As noted in Table 3.3.1-2, Note "*", the redundant outputs from the 2-Out-Of-4 Voter channel are considered part of the same channel, but the OPRM and APRM outputs are considered to be separate channels, so N = 8 to determine the interval between tests for application of Specification 4.3.1.3 (REACTOR PROTECTION SYSTEM RESPONSE TIME). The note further requires that testing of OPRM and APRM outputs shall be alternated.

Each test of an OPRM or APRM output tests each of the redundant outputs from the 2-Out-Of-4 Voter channel for that function, and each of the corresponding relays in the RPS.

Consequently, each of the RPS relays is tested every fourth cycle. This testing frequency is twice the frequency justified by References 2 and 3.

Automatic reactor trip upon receipt of a high-high radiation signal from the Main Steam Line Radiation Monitoring System was removed as the result of an analysis performed by General Electric in NEDO-31400A. The NRC approved the results of this analysis as documented in the SER (letter to George J. Beck, BWR Owner's Group from A.C. Thadani, NRC, dated May 15, 1991).

The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are completed within the time limit assumed in the safety analyses.

No credit was taken for those channels with response times indicated as not applicable except for the APRM Simulated Thermal Power - Upscale and Neutron Flux - Upscale trip functions and the OPRM Upscale trip function (Table 3.3.1-2, Items 2.b, 2.c, and 2.f).

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurement, provided such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times.

Response time testing for the sensors as noted in Table 3.3.1-2 is not required based on the analysis in NEDO-32291-A.

Response time testing for the remaining channel components is required as noted.

For the digital electronic portions of the APRM functions, performance characteristics that determine response time are checked by a combination of automatic self-test, calibration activities, and response time tests of the 2-Out-Of-4 Voter (Table 3.3.1-2, Item 2.e).

LIMERICK - UNIT 1 B 3/4 3-le Amendment No. 1441,

INSTRUMENTATION BASES 3/4.3.7.10 (Deleted) 3/4.3.7.11 (Deleted) -

INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.

3/4.3.7.12 OFFGAS MONITORING INSTRUMENTATION This instrumentation includes provisions for monitoring the concentrations of potentially explosive gas mixtures and noble gases in the off-gas system.

3/4.3.8 (Deleted) -

INFORMATION FROM THIS SECTION RELOCATED TO THE UFSAR.

3/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION The feedwater/main turbine trip system actuation instrumentation is provided to initiate action of the feedwater system/main turbine trip system in the event of failure of feedwater controller under maximum demand.

REFERENCES:

1.

NEDC-30851P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System," March 1988.

2.

NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," October 1995.

3.

NEDC-32410P-A, Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," November 1997.

4.

NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

5.

.NEDO-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

6.

NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

7.

Letter, L. A. England (BWROG) to M. J. Virgilio, "BWR Owners' Group Guidelines for Stability Interim Corrective Action," June 6, 1994.

8.

GE Service Information Letter No. 516, "Core Flow Measurement - GE BWR/3, 4, 5 and 6 Plants," July 26, 1990.

9.

GE Letter NSA 00-433, Alan Chung (GE) to Sujit Chakraborty (GE),

"Minimum Number of Operable OPRM Cells for Option III Stability at Limerick 1 & 2," May 02, 2001.

LIMERICK - UNIT 1 B 3/4 3-7 Amendment No..33, 48,.7-0, 4-0, 4-04, 453

3/4.4.REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM The impact of single recirculation loop operation upon plant safety is assessed and shows that single-loop operation is permitted if the MCPR fuel cladding safety limit is increased as noted by Specification 2.1.2, APRM scram and control rod block setpoints are adjusted as noted in Tables 2.2.1-1 and 3.3.6-2, respectively.

Additionally, surveillance on the pump speed of the operating recirculation loop is imposed.to exclude the possibility of excessive internals vibration.

The surveillance on differential temperatures below 30% RATED THERMAL POWER or 50% rated recirculation loop flow is to mitigate the undue thermal stress on vessel nozzles, recirculation pump and vessel bottom head during the extended operation of the single recirculation loop mode.

An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable. Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation.

Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteria for two recirculation loop operation.

The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.

In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single recir-culation loop mode.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50'F of each other prior to startup of an idle loop.

The loop temperature must also be within 50F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Sudden equalization of a temperature difference > 1450F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.

LIMERICK - UNIT 1 B 3/4 4-1 Amendment No.

30, 66,

REACTOR COOLANT SYSTEM BASES age__

3/4.4.2 SAFETY/RELIEF VALVES The safety valve function of the safety/relief valves operates to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code. A total of 12 OPERABLE safety/

relief valves is required to limit reactor pressure to within ASME III allow-able values for the worst case upset transient.

Demonstration of the safety/relief valve lift settings will occur only during shutdown. The safety/relief valves will be removed and either set pressure tested or replaced with spares which have been previously set pres-sure tested and stored in accordance with manufacturers recommendations in the specified frequency.

Corrected by Ltr. Dated 3/10/00 LIMERICK - UNIT 1 B 3/4 4-2 Amendment No.

0-- 4-3;7,

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.9.1.9 Core Operating Limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the CORE OPERATING LIMITS REPORT for the following:

a.

The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1,

b.

MAPFAC(P) and MAPFAC(F) factors for Specification 3.2.1,

c.

The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.3,

d.

The MCPR(P) and MCPR(F) adjustment factors for specification 3.2.3,

e.

The LINEAR HEAT GENERATION RATE.(LHGR) for Specification 3.2.4,

f.

The power biased Rod Block Monitor setpoints and the Rod Block Monitor MCPR OPERABILITY limits of Specification 3.3.6,

9.

The Reactor Coolant System Recirculation Flow upscale trip setpoint and allowable value for Specification 3.3.6,

h.

The Oscillation Power Range Monitor (OPRM) period based detection algorithm (PBDA) setpoints for Specification 2.2.1.

6.9.1.10 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

a.

NEDE-24011-P-A "General Electric Standard Application for Reactor Fuel" (Latest approved revision),*

b.

NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

6.9.1.11 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.

6.9.1.12 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

For Cycle 8, specific documents were approved in the Safety Evaluation dated (5/4/98) to support License Amendment No.

(127).

LIMERICK - UNIT 1 6-18a Amendment No. 127, 442,

INDEX LL-IDOT IT LT-L ED RO L

ATAt ND SURVE ILLANC EREOUIRELNIS SECTION PAGE REACTOR COOLANT SYSTEM (Continued)

Figure 3.4.1.1-1 Deleted........

...................... 3/4 4-3 Jet Pumps.................................................. 3/4 4-4 Recirculation Pumps.

......................................... 3/4 4-5 Idle Recirculation Loop Startup........

..................... 3/4 4-6 3/4.4.2 SAFETY/RELIEF VALVES................

........................ 3/4 4-7 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems...........

........................ 3/4 4-8 Operational Leakage.

......................................... 3/4 4-9 Table 3.4.3.2-1 Reactor Coolant System Pressure Isolation Valves......

................ 3/4 4-11 3/4.4.4 CHEMISTRY.................................................. 3/4 4-12 Table 3.4.4-1 Reactor Coolant System Chemistry Limits..........

........................ 3/4 4-14 3/4.4.5 SPECIFIC ACTIVITY.

........................................... 3/4 4-15 Table 4.4.5-1 Primary Coolant Specific Activity Sample and Analysis Program............. 3/4 4-17 3/4.4.6 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System.............

......................... 3/4 4-18 Figure 3.4.6.1-1 Minimum Reactor Pressure Vessel Metal Temperature Vs. Reactor Vessel Pressure......

................ 3/4 4-20 Table 4.4.6.1.3-1 Deleted........

..................... 3/4 4-21 Reactor Steam Dome..............

3/4 4-22 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES............................ 3/4 4-23 3/4.4.8 STRUCTURAL INTEGRITY..........

................... 3/4 4-24 I

LIMERICK - UNIT 2 xi Amendment No. 430,

TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOTNTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

1. Intermediate Range Monitor, Neutron Flux-High

2. Average Power Range Monitor:

a. Neutron Flux-Upscale (Setdown)

b. Simulated Thermal Power -

Upscale:

- Two Recirculation Loop Operation

- Single Recirculation Loop Operation***

c.

Neutron Flux - Upscale

• 120/125 divisions of full scale

< 15.0% of RATED THERMAL POWER

< 0.66 W + 62.8% and

< 116.6% of RATED THERMAL POWER

< 0.66 (W-7.6%) + 62.8% and

< 116.6% of RATED THERMAL POWER 118.3% of RATED THERMAL POWER

< 122/125 divisions of full scale

< 20.0% of RATED THERMAL POWER

< 0.66 W + 63.3% and

< 117.0% of RATED THERMAL POWER

< 0.66 (W-7.6%) + 63.3% and

< 117.0% of RATED THERMAL POWER 118.7% of RATED THERMAL POWER I

d. Inoperative

e. 2-Out-Of-4 Voter

f.

OPRM Upscale N.A.

N.A.

N.A.

N.A.

N.A.

3.

4.

5.

6.

7.

8.

Reactor Vessel Steam Dome Pressure - High Reactor Vessel Water Level - Low, Level 3 Main Steam Line Isolation Valve - Closure DELETED Drywell Pressure - High Scram Discharge Volume Water Level - High

a.

Level Transmitter

b.

Float Switch

< 1096 psig 2 12.5 inches above instrument zero*

• 8% closed DELETED

< 1.68 psig

< 1103 psig 2 11.0 inches above instrument zero

< 12% closed DELETED

< 1.88 psig

< 261' 9 1/4" elevation

< 261' 9 1/4" elevation

< 261' 1 1/4" elevation**

< 261' 1 1/4" elevation**

• See Bases Figure B 3/4.3-1.

Equivalent to 25.58 gallons/scram discharge volume.

• • • The 7.6% flow "offset" for Single Loop Operation (SLO) is applied for W 2 7.6%.

For flows W < 7.6%, the (W-7.6%) term is set equal to zero.

• • • • See COLR for OPRM period based detection algorithm trip setpoints.

OPRM Upscale trip output auto-enable (not bypassed) setpoints shall be APRM Simulated Thermal Power 2 30% and recirculation drive flow < 60%.

LIMERICK - UNIT 2 2-4 Amendment No. 48, i;, 1521,

419,

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System instrumentation setpoints specified in Table 2.2.1-1 are the values at which the reactor trips are set for each para-meter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coo ant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist in mitigating the consequences of accidents.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

1.

Intermediate Range Monitor. Neutron Flux - HiQh The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5 decade 10 range instrument. The trip setpoint of 120 divisions of scale is active in each of the 10 ranges.

Thus as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up.

The IRM instruments provide for overlap with both the APRM and SRM systems.

The most significant source of reactivity changes during the power increase is due to control rod withdrawal.

In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed. The results of these analyses are in Section 15.4 of the FSAR. The most severe case involves an initial condition in which THERMAL POWER is at approximately 1% of RATED THERMAL POWER.

Additional conservatism was taken in this analysis by assuming the IRM channel closest to the control rod being withdrawn is bypassed. The results of this analysis show that the reactor is shutdown and peak power is limited to 21% of RATED THERMAL POWER with the peak fuel enthalpy well below the fuel failure threshold of 170 cal/gm.

Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

2.

Average Power Range Monitor The APRM system is divided into four APRM channels and four 2-Out-Of-4 Voter channels. The four voter channels are divided into two groups of two each, with each group of two providing inputs to one RPS trip system. All four voters will trip (full scram) when any two unbypassed APRM channels exceed their trip setpoints.

APRM trip Functions 2.a, 2.b, 2.c, and 2.d are voted independently from OPRM Upscale Function 2.f.

Therefore, any Function 2.a, 2.b, 2.c, or 2.d trip from any two unbypassed APRM channels will result in a full trip in each of the four voter channels.

Similarly, a Function 2.f trip from any two unbypassed APRM channels will result in a full trip from each of the four voter channels.

For operation at low pressure and low flow during STARTUP, the APRM Neutron Flux-Upscale (Setdown) scram setting of 15% of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits. The margin accommodates the anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system. Tempera-ture coefficients are small and control rod patterns are constrained by the RWM. Of all the possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power increase.

LIMERICK - UNIT 2 B 2-6 Amendment No. 409,

LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Average Power Range Monitor (Continued)

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant amount, the rate of power rise is very slow. Generally the heat flux is in near equilibrium with the fission rate.

In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit.

The 15% Neutron Flux - Upscale (Setdown) trip remains active until the mode switch is placed in the Run position.

The APRM trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the system and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Neutron Flux - Upscale setpoint; i.e.,

for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel.

For the Simulated Thermal Power - Upscale setpoint, a time constant of 6 +/- 0.6 seconds is introduced into the flow-biased APRM in order to simulate the fuel thermal transient characteristics.

A more conservative maximum value is used for the flow-biased setpoint as shown in Table 2.2.1-1.

A reduced Trip Setpoint and Allowable Value is provided for the Simulated Thermal Power - Upscale Function, applicable when the plant is operating in Single Loop Operation (SLO) per LCO 3.4.1.1.

In SLO, the drive flow values (W) used in the Trip Setpoint and Allowable Value equations is reduced by 7.6%. The 7.6% value is established to conservatively bound the inaccuracy created in the core flow/drive flow correlation due to back flow in the jet pumps associated with the inactive recirculation loop. The Trip Setpoint and Allowable Value thus maintain thermal margins essentially unchanged from those for two-loop operation. The Trip Setpoint and Allowable Value equations for single loop operation are only valid for flows down to W = 7.6%. The Trip Setpoint and Allowable Value do not go below 62.8% and 63.3% RATED THERMAL POWER, respectively. This is acceptable because back flow in the inactive recirculation loop is only an issue with drive flows of approximately 40% or greater (Reference 1).

The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unneces-sary shutdown.

The APRM channels also include an Oscillation Power Range Monitor (OPRM)

Upscale Function.

The OPRM Upscale Function provides compliance with GDC 10 and GDC 12, thereby providing protection from exceeding the fuel MCPR Safety Limit due to anticipated thermal-hydraulic power oscillations. The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into "cells" for evaluation by the OPRM algorithms.

References 2, 3 and 4 describe three.algorithms for detecting thermal-hydraulic instability related neutron flux oscillations:

the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm.

All three are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period based detection algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations.

OPRM Upscale Function OPERABILITY for Technical Specification purposes is based only on the period based detection algorithm.

LIMERICK - UNIT 2 B 2-7 Amendment 48,

N09,

LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Average Power Range Monitor (Continued)

The OPRM Upscale trip output shall be automatically enabled (not bypassed) when APRM Simulated Thermal Power is 2 30% and recirculation drive flow is < 60%

as indicated by APRM measured recirculation drive flow.

(NOTE:

60%

recirculation drive flow is the recirculation drive flow that corresponds to 60%

of rated core flow.

Refer to TS Bases 3/4.3.1 for further discussion concerning the recirculation drive flow/core flow relationship.)

This is the operating region where actual thermal-hydraulic instability and related neutron flux oscillations may occur. See Reference 5 for additional discussion of OPRM Upscale trip enable region limits. These setpoints, which are sometimes referred to as the "auto-bypass" setpoints, establish the boundaries of the OPRM Upscale trip enabled region. The APRM Simulated Thermal Power auto-enable setpoint has 1% deadband while the drive flow setpoint has a 2% deadband. The deadband for these setpoints is established so that it increases the enabled region.

An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in a cell, with period confirmations and relative cell amplitude exceeding specified setpoints.

One or more cells in a channel exceeding the trip conditions will result in a channel trip.

An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithms detect oscillatory changes in the neutron flux for one or more cells in that channel.

There are four "sets" of OPRM related setpoints or adjustment parameters:

a) OPRM trip auto-enable setpoints for APRM Simulated Thermal Power (30%)

and recirculation drive flow (60%); b) period based detection algorithm (PBDA) confirmation count and amplitude setpoints; c) period based detection algorithm tuning parameters; and d) growth rate algorithm (GRA) and amplitude based algorithm (ABA) setpoints.

The first set, the OPRM auto-enable region setpoints, are treated as nominal setpoints with no additional margins added as discussed in Reference 5.

The settings, 30% APRM Simulated Thermal Power and 60% recirculation drive flow, are defined (limit values) in a note to Table 2.2.1-1.

The second set, the OPRM PBDA trip setpoints, are established in accordance with methodologies defined in Reference 4, and are documented in the COLR. There are no allowable values for these setpoints.

The third set, the OPRM PBDA "tuning" parameters, are established or adjusted in accordance with and controlled by station procedures.

The fourth set, the GRA and ABA setpoints, in accordance with References 2 and 3, are established as nominal values only, and controlled by station procedures.

3.

Reactor Vessel Steam Dome Pressure-High High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products.

A pressure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips.

The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient.

This trip setpoint is effective at low power/flow conditions when the turbine stop valve and control fast closure trips are bypassed.

For a turbine trip or load rejection under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.

LIMERICK - UNIT 2 B 2-7a Amendment 48, 449,

LIMITING SAFETY SYSTEM SETTING BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

REFERENCES:

1.

NEDC-31300, "Single-Loop Operation Analysis for Limerick Generating Station, Unit 1," August 1986.

2.

NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

3.

NEDO-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

4.

NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

5.

BWROG Letter 96113, K. P. Donovan (BWROG) to L. E. Phillips (NRC),

"Guidelines for Stability Option III 'Enable Region' (TAC M92882),"

September 17, 1996.

LIMERICK - UNIT 2 B 2-10 Amendment No.

I

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

APPLICABILITY: As shown in Table 3.3.1-1.

ACTION:

Note:

Separate condition entry is allowed for each channel.

a.

With the number of OPERABLE channels in either trip system for one or more Functional Units less than the Minimum OPERABLE Channels per Trip System required by Table 3.3.1-1, within one hour for each affected functional unit either verify that at least one* channel in each trip system is OPERABLE or tripped or that the trip system is tripped, or place either the affected trip system or at least one inoperable channel in the affected trip system in the tripped condition.

b.

With the number of OPERABLE channels in either trip system less than the Minimum OPERABLE Channels per Trip System required by Table 3.3.1-1, place either the inoperable channel(s) or the affected trip system** in the 4ripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.***

c.

With the number of OPERABLE channels in both trip systems for one or more Functional Units less than the Minimum OPERABLE Channels per Trip System required by Table 3.3.1-1, place either the inoperable channel(s) in one trip system or one trip system in the tripped condition within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s**.***

d.

If within the allowable time allocated by Actions a, b or c, it is not desired to place the inoperable channel or trip system in trip (e.g., full scram would occur), Then no later than expiration of that allowable time initiate the action identified in Table 3.3.1-1 for the applicable Functional Unit.

• For Functional Units 2.a, 2.b, 2.c, 2.d, and 2.f, at least two channels shall be OPERABLE or tripped.

For Functional Unit 5, both trip systems shall have each channel associated with the MSIVs in three main steam lines (not necessarily the same main steam lines for both trip systems) OPERABLE or tripped.

For Function 9, at least three channels per trip system shall be OPERABLE or tripped.

• • For Functional Units 2.a, 2.b, 2.c, 2.d, and 2.f, inoperable channels shall be placed in the tripped condition to comply with Action b. Action c does not apply for these Functional Units.

A channel or trip system which has been placed in the tripped condition to satisfy Action b. or c. may be returned to the untripped condition under administrative control for up to two hours solely to perform testing required to demonstrate its operability or the operability of other equipment provided Action a. continues to be satisfied.

LIMERICK - UNIT 2 3/4 3-1 Amendment No. A4t, 34, 4-G9,

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 24 months, except Table 4.3.1.1-1 Functions 2.a, 2.b, 2.c, 2.d, 2.e, and 2.f.

Functions 2.a, 2.b, 2.c, 2.d, and 2.f do not require separate LOGIC SYSTEM FUNCTIONAL TESTS.

For Function 2.e, tests shall be performed at least once per 24 months.

LOGIC SYSTEM FUNCTIONAL TEST for Function 2.e includes simulating APRM and OPRM trip conditions at the APRM channel inputs to the voter channel to check all combinations of two tripped inputs to the 2-Out-Of-4 voter logic in the voter channels.

4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its limit at least once per 24 months.

Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 24 months where N is the total number of redundant channels in a specific reactor trip system.

LIMERICK - UNIT 2 3/4 3-la Amendment No. Add,

TABLE 3.3.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION APPLICABLE MINI OPERATIONAL OPERABLE CONDITIONS PER TRIP S MUM CHANNELS YSTEM (a)

FUNCTIONAL UNIT

1.

Intermediate Range Monitors(b):

a.

Neutron Flux - High

b.

Inoperative ACTION 2

3(i), 4(i) 5(i) 2 3(i), 4(i) 5(i) 3 3

3(d) 3 3

3(d) 1 2

3 1

2 3

2.

Average Power Range Monitor(e):

a.

Neutron Flux - Upscale (Setdown)

b.

Simulated Thermal Power - Upscale

c.

Neutron Flux - Upscale

d.

Inoperative

e.

2-Out-Of-4 Voter

f.

OPRM Upscale

3.

Reactor Vessel Steam Dome Pressure - High

4.

Reactor Vessel Water Level - Low, Level 3

5.

Main Steam Line Isolation Valve-Closure 2

1 1

1, 2 1, 2 1

(0)(p) 3(m) 3(m) 3(m) 3(m) 2 3 (m) 1 4

4 1

1 10 I

1, 2(f) 1, 2 2

1 2

1 1(g) 1/valve 4

LIMERICK - UNIT 2 3/4 3-2 Amendment No. ;, 409, 4719,

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION STATEMENTS ACTION 1 Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 2 Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the SHUTDOWN position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 3 Suspend all operations involving CORE ALTERATIONS and insert all insertable control rods within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 4 Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 5 Be in STARTUP with the main steam line isolation valves closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 6 Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure until the function is automatically bypassed, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

ACTION 7 Verify all insertable control rods to be inserted within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 8 Lock the reactor mode switch in the Shutdown position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 9 Suspend all operations involving CORE ALTERATIONS, and insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 10 -

a. If the condition exists due to a common-mode OPRM deficiency*,

then initiate alternate method to detect and suppress thermal-hydraulic instability oscillations within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND restore required channels to OPERABLE status within 120 days,

b.

Reduce THERMAL POWER to < 25% RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Unanticipated characteristic of the instability detection algorithm or equipment that renders all OPRM channels inoperable at once.

LIMERICK - UNIT 2 3/4 3-4 Amendment No. 109, 4-1d,

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

(b)

This function shall automatically be bypassed when the reactor mode switch is in the Run position.

(c)

DELETED (d)

The noncoincident NMS reactor trip function logic is such that all channels go to both trip systems. Therefore, when the "shorting links" are removed, the Minimum OPERABLE Channels Per Trip System is 6 IRMs.

(e)

An APRM channel is inoperable if there are less than 3 LPRM inputs per level or less than 20 LPRM inputs to an APRM channel, or if more than 9 LPRM inputs to the APRM channel have been bypassed since the last APRM calibration (weekly gain calibration).

(f)

This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

(g)

This function shall be automatically bypassed when the reactor mode switch is not in the Run position.

(h)

This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(i)

With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(j)

This function shall be automatically bypassed when turbine first stage pressure is equivalent to a THERMAL POWER of less than 30% of RATED THERMAL POWER.

Ck)

Also actuates the EOC-RPT system.

(1)

DELETED (m)

Each APRM channel provides inputs to both trip systems.

(n)

DELETED (o)

With THERMAL POWER 2 25% RATED THERMAL POWER. The OPRM Upscale trip output shall be automatically enabled (not bypassed) when APRM Simulated Thermal Power is 2 30% and recirculation drive flow is < 60%.

The OPRM trip output may be automatically bypassed when APRM Simulated Thermal Power is < 30% or recirculation drive flow is 2 60%.

(p) A minimum of 23 cells, each with a minimum of 2 OPERABLE LPRMs, must be OPERABLE for an OPRM channel to be OPERABLE.

LIMERICK - UNIT 2 3/4 3-5 Amendment No. i, 14-,

409,

TABLE 3.3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES RESPONSE TIME FUNCTIONAL UNIT (Seconds)

1.

Intermediate Range Monitors:

a.

Neutron Flux - High N.A.

b.

Inoperative N.A.

2. Average Power Range Monitor*:

a.

Neutron Flux - Upscale (Setdown)

N.A.

b.

Simulated Thermal Power - Upscale N.A.

c.

Neutron Flux - Upscale N.A.

d.

Inoperative N.A.

e.

2-Out-Of-4 Voter

• 0.05*

f.

OPRM Upscale N.A.

3.

Reactor Vessel Steam Dome Pressure - High

• 0.55

4.

Reactor Vessel Water Level - Low, Level 3 s1.05#

5.

Main Steam Line Isolation Valve - Closure

• 0.06

6.

DELETED DELETED

7.

Drywell Pressure - High N.A.

8.

Scram Discharge Volume Water Level - High

a.

Level Transmitter N.A.

b.

Float Switch N.A.

9.

Turbine Stop Valve - Closure

• 0.06

10.

Turbine Control Valve Fast Closure, Trip Oil Pressure - Low

• 0.08**

11.

Reactor Mode Switch Shutdown Position N.A.

12.

Manual Scram N.A.

Neutron detectors, APRM channel and 2-Out-Of-4 Voter channel digital electronics are exempt from response time testing.

Response time shall be measured from activation of the 2-Out-Of-4 Voter output relay.

For application of Specification 4.3.1.3, the redundant outputs from each 2-Out-Of-4 Voter channel are considered part of the same channel, but the OPRM and AP RM outputs are considered to be separate channels, so N = 8. Testing of OPRM and APRM outputs shall alternate.

Measured from start of turbine control valve fast closure.

1. Sensor is eliminated from response time testing for the RPS circuits. Response time testing and conformance to the administrative limits for the remaining channel including trip unit and relay logic are required.

LIMERICK - UNIT 2 3/4 3-6 Amendment No. Em2, -93, 4-9,

TABLE 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL FUNCTIONAL CHANNEL CHECK TEST CALIBRATION(a FUNCTIONAL UNIT

1. Intermediate Range Mor

a.

Neutron Flux -

CONDITIONS FOR WHICH I SURVEILLANCE REQUIRED iitors:

High

b.

Inoperative

2.

Average Power Range Monitor(f):

a.

Neutron Flux - Upscale (Setdown)

b.

Simulated Thermal Power - Upscale

c.

Neutron Flux - Upscale

d.

Inoperative

e.

2-Out-Of-4 Voter

f.

OPRM Upscale

3.

Reactor Vessel Steam Dome Pressure - High

4.

Reactor Vessel Water Level-Low, Level 3

5.

Main Steam Line Isolation Valve - Closure

6.

DELETED

7.

Drywell Pressure - High

8.

Scram Discharge Volume Water Level - High

a. Level Transmitter

b. Float Switch S(b)

S N.A.

D(b)

D D

N.A.

D D

W W(j)

W(j)

SA(l)

SA(e)

SA SA SA SA(e)

R R

N.A.

R W(d), R(g)

WNd), R N.A.

N.A.

R(c) (g) 2 3(i), 4(i), 5(i) 2, 3(i), 4(i), 5(i) 2 1

1 1, 2 1, 2 1(m)

S 0

R 1, 2(h) 1, 2 S

Q R

N.A.

DELETED S

S N.A.

Q DELETED 0

Q 0

R DELETED R

1 DELETED 1, 2 R

R 1, 2, 5(i) 1, 2, 5(i)

LIMERICK - UNIT 2 3/4 3-7 Amendment No. 5, i, F4.9, 4412,

TABLE 4.3.1.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION(a)

SURVEILLANCE REOUIRED

9. Turbine Stop Valve - Closure N.A.

0 R

1

10.

Turbine Control Valve Fast Closure, Trip Oil Pressure -

Low N.A.

Q R

1

11.

Reactor Mode Switch Shutdown Position N.A.

R N.A.

1, 2, 3, 4, 5

12.

Manual Scram N.A.

W N.A.

1, 2, 3, 4, 5 (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for a least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days.

(c) Calibration includes verification that the OPRM Upscale trip auto-enable (not-bypass) setpoint for APRM Simulated Thermal Power is 2 30% and for recirculation drive flow is < 60%.

(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER 2 25% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER.

(e) CHANNEL FUNCTIONAL TEST shall include the flow input function, excluding the flow transmitter.

(f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH).

(g) Calibration includes the flow input function.

(h) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

Ci)

With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(j) If the RPS shorting links are required to be removed per Specification 3.9.2, they may be reinstalled for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance. During this time, CORE ALTERATIONS shall be suspended, and no control rod shall be moved from its existing position.

(k) DELETED (1) Not required to be performed when entering OPERATIONAL CONDITION 2 from OPERATIONAL CONDITION 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering OPERATIONAL CONDITION 2.

(m) With THERMAL POWER 2 25% of RATED THERMAL POWER.

LIMERICK - UNIT 2 3/4 3-8 Amendment No. ;,4,48,;-,79,9,4q9,

TABLE 3.3.6-1 CONTROL ROD BLOCK INSTRUMENTATION MINIMUM APPLICABLE OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION PER TRIP FUNCTION CONDITIONS ACTION

1.

ROD BLOCK MONITORa)

a.

Upscale 2

1*

60

b.

Inoperative 2

1*

60

c.

Downscale 2

1*

60

2.

APRM

a.

Simulated Thermal Power - Upscale 3

1 61

b.

Inoperative 3

1, 2 61

c.

Neutron Flux - Downscale 3

1 61

d.

Simulated Thermal Power - Upscale (Setdown) 3 2

61

e.

Recirculation Flow - Upscale 3

1 61

f.

LPRM Low Count 3

1, 2 61

3.

SOURCE RANGE MONITORS ***

a.

Detector not full in (b) 3 2

61 2

5 61

b.

Upscale(')

3 2

61 2

5 61

c.

Inoperative(c) 3 2

61 2

5 61

d.

Downscale(d) 3 2

61 2

5 61

4.

INTERMEDIATE RANGE MONITORS

a.

Detector not full in 6

2, 5**

61

b.

Upscale 6

2, 5**

61

c.

Inoperative 6

2, 5**

61

d.

Downscale(e) 6 2, 5**

61

5.

SCRAM DISCHARGE VOLUME

a.

Water Level-High 2

1, 2, 5**

62

6.

DELETED DELETED DELETED DELETED

7.

REACTOR MODE SWITCH SHUTDOWN POSITION 2

3, 4 63 LIMERICK - UNIT 2 3/4 3-58 Amendment No. -X, 4-09,

TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP SETPOINT TRIP FUNCTION

1. ROD BLOCK MONITOR

a. Upscale"'

1) Low Trip Setpoint (LTSP)

2) Intermediate Trip Setpoint (ITSP)

3) High Trip Setpoint (HTSP)

ALLOWABLE VALUE

b. Inoperative

c. Downscale (DTSP)

d. Power Range Setpoint'bl

1) Low Power Setpoint (LPSP)

2) Intermediate Power Setpoint (IPSP)

3) High Power Setpoint (HPSP)

2. APRM

a. Simulated Thermal Power - Upscale:

Two Recirculation Loop Operation Single Recirculation Loop Operation****

N/A N/A 28.1% RATED THERMAL POWER 63.1% RATED THERMAL POWER 83.1% RATED THERMAL POWER

• 0.66 W + 55.2% and

• 108.0% of RATED THERMAL POWER

• 0.66 (W-7.6%) + 55.2% and

• 108.0% of RATED THERMAL POWER 28.4% RATED THERMAL 63.4% RATED THERMAL 83.4% RATED THERMAL POWER POWER POWER

• 0.66 W + 55.7% and

< 108.4% of RATED THERMAL POWER

• 0.66 (W-7.6%) + 55.7% and

• 108.4% of RATED THERMAL POWER I

b.

Inoperative N.A.

N.A.

c. Neutron Flux - Downscale

d. Simulated Thermal Power - Upscale (Setdown) 2 3.2% of RATED THERMAL POWER

• 12.0% of RATED THERMAL POWER 2 2.8% of RATED THERMAL POWER

• 13.0% of RATED THERMAL POWER

e.

Recirculation Flow - Upscale

f. LPRM Low Count 20 per channel 3 per axial level

3. SOURCE RANGE MONITORS

a. Detector not full in

b. Upscale

c. Inoperative

d. Downscale 20 per channel 3 per axial level N.A.

• 1.6 x 105 cps N.A.

2 1.8 cps**

N.A.

• 1 x 105 cps N.A.

2 3 cps**

LIMERICK - UNIT 2 3/4 3-60 Amendment No. 48, 5-1, 4-1g,

TABLE 3.3.6-2 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in

b. Upscale c.

d.

Inoperative Downscale N.A.

• 108/125 divisions of full scale N.A.

2 5/125 divisions of full scale

• 257' 7 3/8" elevation***

ALLOWABLE VALUE N.A.

• 110/125 divisions of full scale N.A.

2 3/125 divisions of full scale

• 257' 9 3/8" elevation

5. SCRAM DISCHARGE VOLUME

a. Water Level-High

a. Float Switch

6. DELETED

7. REACTOR MODE SWITCH SHUTDOWN POSITION

• Refer to the COLR for these setpoints.

DELETED DELETED I

N.A.

N.A.

May be reduced, provided the Source Range Monitor has an observed or above the curve shown in Figure 3.3.6-1.

• • • Equivalent to 13.56 gallons/scram discharge volume.

count rate and signal-to-noise ratio on

• • • • The 7.6% flow "offset" for Single Loop Operation (SLO) is applied (W-7.6%) term is set equal to zero.

for W Ž 7.6%.

For flows W < 7.6%, the I

(a)

There are three upscale trip levels. Each is applicable only over its specified operating core thermal power range.

All RBM trips are automatically bypassed below the low power setpoint (LPSP). The upscale LTSP is applied between the low power setpoint (LPSP) and the intermediate power setpoint (IPSP). The upscale ITSP is applied between the intermediate power setpoint and the high power setpoint (HPSP).

The HTSP is applied above the high power setpoint.

(b)

Power range setpoints control enforcement of appropriate upscale trips over the proper core thermal power ranges.

The power signal to the RBM is provided by the APRM.

LIMERICK - UNIT 2 3/4 3-60a Amendment No. 3,4,38,48, 409,

TABLE 4.3.6-1 CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION")

SURVEILLANCE REQUIRED

1. ROD BLOCK MONITOR

a. Upscale N.A.

Q(C)

R 1*

b. Inoperative N.A.

Q(u N.A.

1*

c. Downscale N.A.

QC)

R 1*

2. APRM

a. Simulated Thermal Power -

Upscale N.A.

SA R

1

b. Inoperative N.A.

SA N.A.

1, 2

c. Neutron Flux - Downscale N.A.

SA R

1

d. Simulated Thermal Power -

Upscale (Setdown)

N.A.

SA R

2

e. Recirculation Flow - Upscale N.A.

SA R

1

f. LPRM Low Count N.A.

SA R

1, 2

3. SOURCE RANGE MONITORS

a. Detector not full in N.A.

M(d)(e) W(f)

N.A.

2, 5

b. Upscale N.A.

M(d)(e),W(f)

R 2, 5

c. Inoperative N.A.

M(d)(e),W(f)

N.A.

2, 5

d. Downscale N.A.

M(d)(e), W R

2, 5

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in N.A.

W N.A.

2, 5**

b. Upscale N.A.

W R

2, 5**

c. Inoperative N.A.

W N.A.

2, 5**

d. Downscale N.A.

W R

2, 5**

5. SCRAM DISCHARGE VOLUME

a. Water Level - High N.A.

Q R

1, 2, 5**

6. DELETED DELETED DELETED DELETED DELETED

7. REACTOR MODE SWITCH SHUTDOWN POSITION N.A.

R(9)

N.A.

3, 4 LIMERICK - UNIT 2 3/4 3-61 Amendment No..7,

-4;,

48, 63, 109 Corrected by letter dated May 28, 2002 I

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITIOLN FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.

APPLICABILITY:

OPERATIONAL CONDITIONS 1* and 2*.

ACTION:

a.

With one reactor coolant system recirculation loop not in operation:

1.

Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

a.

Place the recirculation flow control system in the Local Manual mode, and

b.

Reduce THERMAL POWER to < 76.2% of RATED THERMAL POWER, and,

c.

Limit the speed of the operating recirculation pump to less than or equal to 90% of rated pump speed, and

d.

Verify that the differential temperature requirements of Surveillance Requirement 4.4.1.1.5 are met if THERMAL POWER is

• 30% of RATED THERMAL POWER or the recirculation loop flow in the operating loop is < 50% of rated loop flow, or suspend the THERMAL POWER or recirculation loop flow increase.

• See Special Test Exception 3.10.4.

LIMERICK - UNIT 2 3/4 4-1 Amendment No. 48, At4,

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

(Continued)

2.

Within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s:

Reduce the Average Power Range Monitor (APRM) Simulated Thermal Power

- Upscale Scram and Rod Block Trip Setpoints and Allowable Values, to those applicable for single recirculation loop operation per Specifications 2.2.1 and 3.3.6, or declare the associated channel(s) inoperable and take the actions required by the referenced specifications.

3.

Otherwise be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

With no reactor coolant system recirculation loops in operation, initiate measures to place the unit in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

LIMERICK - UNIT 2 3/4 4-la Amendment No. 48, 449, 4-391,

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.1.1 DELETED 4.4.1.1.2 DELETED 4.4.1.1.3 DELETED 4.4.1.1.4 With one reactor coolant system recirculation loop not in operation, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verify that:

a.

Reactor THERMAL POWER is < 76.2% of RATED THERMAL POWER,

b.

The recirculation flow control system is in the Local Manual mode, and

c.

The speed of the operating recirculation pump is < 90% of rated pump speed.

4.4.1.1.5 With one reactor coolant system recirculation loop not in operation, within 15 minutes prior to either THERMAL POWER increase or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is < 30% of RATED THERMAL POWER or the recirculation loop flow in the operating recirculation loop is

• 50% of rated loop flow.

a.

• 1450F between reactor vessel steam space coolant and bottom head drain line coolant,

b.

• 50'F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and

c.

• 50'F between the reactor coolant within the loop not in operation and the operating loop.

The differential temperature requirements of Specification 4.4.1.1.5b. and c.

do not apply when the loop not in operation is isolated from the reactor pressure vessel.

LIMERICK - UNIT 2 3/4 4-2 Amendment No. 3X, 38, -54, 4-04,

CONTENTS OF THIS PAGE HAVE BEEN DELETED LIMERICK - UNIT 2 3/4 4-3 Amendment No. -51,

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

a.

Preserve the integrity of the fuel cladding.

b.

Preserve the integrity of the reactor coolant system.

c.

Minimize the energy which must be adsorbed following a loss-of-coolant accident, and

d.

Prevent inadvertent criticality.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance. When necessary, one channel may.be made inoperable for brief intervals to conduct required surveillance.

The reactor protection system is made up of two independent trip systems.

There are usually four channels to monitor each parameter with two channels in each trip system.

The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The APRM system is divided into four APRM channels and.

four 2-Out-Of-4 Voter channels.

Each APRM channel provides inputs to each of the four voter channels. The four voter channels are divided into two groups of two each, with each group of two providing inputs to one RPS trip system.

The system is designed to allow one APRM channel, but no voter channels, to be bypassed.

The system meets the intent of IEEE-279 for nuclear power plant protection systems. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System" and NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function." The bases for the trip settings of the RPS are discussed in the bases for Specification 2.2.1.

The APRM Functions include five Functions accomplished by the four APRM channels (Functions 2.a, 2.b, 2.c, 2.d, and 2.f) and one accomplished by the four 2-Out-Of-4 Voter channels (Function 2.e). Two of the five Functions accomplished by the APRM channels are based on neutron flux only (Functions 2.a and 2.c), one Function is based on neutron flux and recirculation drive flow (Function 2.b) and one is based on equipment status (Function 2.d). The fifth Function accomplished by the APRM channels is the Oscillation Power Range Monitor (OPRM) Upscale trip Function 2.f, which is based on detecting oscillatory characteristics in the neutron flux.

The OPRM Upscale Function is also dependent on average neutron flux (Simulated Thermal Power) and recirculation drive flow, which are used to automatically enable the output trip.

The Two-Out-Of-Four Logic Module includes 2-Out-Of-4 Voter hardware and the APRM Interface hardware. The 2-Out-Of-4 Voter Function 2.e votes APRM Functions 2.a, 2.b, 2.c, and 2.d independently of Function 2.f. This voting is accomplished by the 2-Out-Of-4 Voter hardware in the Two-Out-Of-Four Logic Module. The voter includes separate outputs to RPS for the two independently voted sets of Functions, each of which is redundant (four total outputs). The analysis in Reference 2 took credit for this redundancy in the justification of the 12-hour allowed out-of-service time for LIMERICK - UNIT 2 B 3/4 3-1 Amendment No. 14, 52, 93, 419,

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

Action b,so the voter Function 2.e must be declared inoperable if any of its functionality is inoperable.

The voter Function 2.e does not need to be declared inoperable due to any failure affecting only the APRM Interface hardware portion of the Two-Out-Of-Four Logic Module.

Three of the four APRM channels and all four of the voter channels are required to be OPERABLE to ensure that no single failure will preclude a scram on a valid signal.

To provide adequate coverage of the entire core, consistent with the design bases for the APRM Functions 2.a, 2.b, and 2.c, at least 20 LPRM inputs, with at least three LPRM inputs from each of the four axial levels at which the LPRMs are located, must be operable for each APRM channel.

In addition, no more than 9 LPRMs may be bypassed between APRM calibrations (weekly gain adjustments).

For the OPRM Upscale Function 2.f, LPRMs are assigned to "cells" of 3 or 4 detectors. A minimum of 23 cells (Reference 9), each with a minimum of 2 OPERABLE LPRMs, must be OPERABLE for each APRM channel for the OPRM Upscale Function 2.f to be OPERABLE in that channel.

References 4, 5 and 6 describe three algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. All three are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period based detection algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations.

OPRM Upscale Function OPERABILITY for Technical Specification purposes is based only on the period based detection algorithm.

An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in any cell, with period confirmations and relative cell amplitude exceeding specified setpoints.

One or more cells in a channel exceeding the trip conditions will result in a channel trip. An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithms detect growing oscillatory changes in the neutron flux for one or more cells in that channel.

The OPRM Upscale Function is required to be OPERABLE when the plant is at 2 25% RATED THERMAL POWER. The 25% RATED THERMAL POWER level is selected to provide margin in the unlikely event that a reactor power increase transient occurring while the plant is operating below 30% RATED THERMAL POWER causes a power increase to or beyond the 30% RATED THERMAL POWER OPRM Upscale trip auto-enable point without operator action.

This OPERABILITY requirement assures that the OPRM Upscale trip automatic-enable function will be OPERABLE when required.

Actions a, b and c define the Action(s) required when RPS channels are discovered to be inoperable.

For those Actions, separate entry condition is allowed for each inoperable RPS channel.

Separate entry means that the allowable time clock(s) for Actions a, b or c start upon discovery of inoperability for that specific channel. Restoration of an inoperable RPS channel satisfies only the action statements for that particular channel.

Action statement(s) for remaining inoperable channel(s) must be met according to their original entry time.

Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> has been shown to be acceptable (NEDC-30851P-A and NEDC-32410P-A) to permit restoration of any inoperable channel to OPERABLE status.

However, this out of service time is only acceptable provided that the associated Function's (identified as a "Functional Unit" in Table 3.3.1-1) inoperable channel is in one trip system and the Function still maintains RPS trip capability.

LIMERICK -

UNIT 2 B 3/4 3-la Amendment No. 4a-7-,

a, 93, 409,

3/4.3 INSTRUMENTATION BASES c__

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

The requirements of Action a are intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same trip system for the same Function result in the Function not maintaining :RPS trip capability. A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal.

For the typical Function with one-out-of-two taken twice logic, including the IRM Functions and APRM Function 2.e (trip capability associated with APRM Functions 2.a, 2.b, 2.c, 2.d, and 2.f are discussed below), this would require both trip systems to have one channel OPERABLE or in trip (or the associated trip system in trip).

For Function 5 (Main Steam Isolation Valve--Closure), this would require both trip systems to have each channel associated with the MSIVs in three main steam lines (not necessarily the same main steam lines for both trip systems) OPERABLE or in trip (or the associated trip system in trip).

For Function 9 (Turbine Stop Valve-Closure), this would require both trip systems to have three channels, each OPERABLE or in trip (or the associated trip system in trip).

The completion time to satisfy the requirements of Action a is intended to allow the operator time to evaluate and repair any discovered inoperabilities.

The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.

With trip capability maintained, i.e., Action a satisfied, Actions b and c as applicable must still be satisfied.

If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Action b requires that the channel or the associated trip system must be placed in the tripped condition.

Placing the inoperable channel in trip (or the associated trip system in trip) would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.

As noted, placing the trip system in trip is not applicable to satisfy Action b for APRM Functions 2.a, 2.b, 2.c, 2.d, or 2.f.

Inoperability of one required APRM channel affects both trip systems.

For that condition, the Action b requirements can only be satisfied by placing the inoperable APRM channel in trip.

Restoring OPERABILITY or placing the inoperable APRM channel in trip are the only actions that will restore capability to accommodate a single APRM channel failure.

Inoperability of more than one required APRM channel of the same trip function results in loss of trip capability and the requirement to satisfy Action a.

The requirements of Action c must be satisfied when, for any one or more Functions, at least one required channel is inoperable in each trip system.

In this condition, provided at least one channel per trip system is OPERABLE, normally the RPS still maintains trip capability for that Function, but cannot accommodate a single failure in either trip system (see additional bases discussion above related to loss of trip capability and the requirements of Action a, and special cases for Functions 2.a, 2.b, 2.c, 2.d, 2.f, 5 and 9).

LIMERICK - UNIT 2 B 3/4 3-lb Amendment No. 4-09,

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

The requirements of Action c limit the time the RPS scram logic, for any Function, would not accommodate single failure in both trip systems (e.g., one-out-of-one and one-out-of-one arrangement for a typical four channel Function).

The reduced reliability of this logic arrangement was not evaluated in NEDC-30851P-A for the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time.

Within the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance, the associated Function must have all required channels OPERABLE or in trip (or any combination) in one trip system.

Completing the actions required by Action c restores RPS to a reliability level equivalent to that evaluated in NEDC-30851P-A, which justified a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowable out of service time as allowed by Action b. To satisfy the requirements of Action c, the trip system in the more degraded state should be placed in trip or, alternatively, all the inoperable channels in that trip system should be placed in trip (e.g., a trip system with two inoperable channels could be in a more degraded state than a trip system with four inoperable channels if the two inoperable channels are in the same Function while the four inoperable channels are all in different Functions). The decision of which trip system is in the more degraded state should be based on prudent judgment and take into account current plant conditions (i.e.,

what OPERATIONAL CONDITION the plant is in).

If this action would result in a scram or RPT, it is permissible to place the other trip system or its inoperable channels in trip.

The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowable out of service time is judged acceptable based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram.

As noted, Action c is not applicable for APRM Functions 2.a, 2.b, 2.c, 2.d, or l 2.f.

Inoperability of an APRM channel affects both trip systems and is not associatedl with a specific trip system as are the APRM 2-Out-Of-4 voter and other non-APRM channels for which Action c applies.

For an inoperable APRM channel, the requirements of Action b can only be satisfied by tripping the inoperable APRM channel.

Restoring OPERABILITY or placing the inoperable APRM channel in trip are the only actions that will restore capability to accommodate a single APRM channel failure.

If it is not desired to place the channel (or trip system) in trip to satisfy the requirements of Action a, Action b or Action c (e.g., as in the case where placing the inoperable channel in trip would result in a full scram), Action d requires that the Action defined by Table 3.3.1-1 for the applicable Function be initiated immediately upon expiration of the allowable out of service time.

Table 3.3.1-1, Function 2.f, references Action 10, which defines the action required if OPRM Upscale trip capability is not maintained.

Action 10b is required to address identified equipment failures.

Action 10a is to address common mode vendor/industry identified issues that render all four OPRM channels inoperable at once.

For this condition, References 2 and 3 justified use of alternate methods to detect and suppress oscillations for a limited period of time, up to 120 days. The alternate methods are procedurally established consistent with the guidelines identified in Reference 7 requiring manual operator action to scram the plant if certain predefined events occur. The 12-hour allowed completion time to implement the alternate methods is based on engineering judgment to allow orderly transition to the alternate methods while limiting the period of time during which no automatic or alternate detect and LIMERICK - UNIT 2 B 3/4 3 -1c Amendment No. 409,

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued) suppress trip capability is formally in place.

The 120-day period during which use of alternate methods is allowed is intended to be an outside limit to allow for the case where design changes or extensive analysis might be required to understand or correct some unanticipated characteristic of the instability detection algorithms or equipment.

The evaluation of the use of alternate methods concluded, based on engineering judgment, that the likelihood of an instability event that could not be adequately handled by the alternate methods during the 120-day period was negligibly small.

Plant startup may continue while operating within the allowed completion time of Action 10a.

The primary purpose of this is to allow an orderly completion, without undue impact on plant operation, of design and verification activities in the event of a required design change to the OPRM Upscale function. This exception is not intended as an alternative to restoring inoperable equipment to OPERABLE status in a timely manner.

Action 10a is not intended and was not evaluated as a routine alternative to returning failed or inoperable equipment to OPERABLE status.

Correction of routine equipment failure or inoperability is expected to be accomplished within the completion times allowed for LCO 3.3.1 Action a or Action b, as applicable.

Action 10b applies when routine equipment OPERABILITY cannot be restored within the allowed completion times of LCO 3.3.1 Actions a or b, or if a common mode OPRM deficiency cannot be corrected and OPERABILITY of the OPRM Upscale Function restored within the 120-day allowed completion time of Action 10a.

The OPRM Upscale trip output shall be automatically enabled (not-bypassed) when APRM Simulated Thermal Power is 2 30% and recirculation drive flow is < 60%

as indicated by APRM measured recirculation drive flow.

NOTE:

60% recirculation drive flow is the recirculation drive flow that corresponds to 60% of rated core flow. This is the operating region where actual thermal-hydraulic instability and related neutron flux oscillations may occur.

As noted in Table 4.3.1.1-1, Note c, CHANNEL CALIBRATION for the OPRM Upscale trip Function 2.f includes confirming that the auto-enable (not-bypassed) setpoints are correct.

Other surveillances ensure that the APRM Simulated Thermal Power properly correlates with THERMAL POWER (Table 4.3.1.1-1, Note d) and that recirculation drive flow properly correlates with core flow (Table 4.3.1.1-1, Note g).

If any OPRM Upscale trip auto-enable setpoint is exceeded and the OPRM Upscale trip is not enabled, i.e., the OPRM Upscale trip is bypassed when APRM Simulated Thermal Power is 2 30% and recirculation drive flow is < 60%, then the affected channel is considered inoperable for the OPRM Upscale Function.

Alternatively, the OPRM Upscale trip auto-enable setpoint(s) may be adjusted to place the channel in the enabled condition (not-bypassed).

If the OPRM Upscale trip is placed in the enabled condition, the surveillance requirement is met and the channel is considered OPERABLE.

As noted in Table 4.3.1.1-1, Note g, CHANNEL CALIBRATION for the APRM Simulated Thermal Power - Upscale Function 2.b and the OPRM Upscale Function 2.f, includes the recirculation drive flow input function.

The APRM Simulated Thermal Power - Upscale Function and the OPRM Upscale Function both require a valid drive flow signal.

The APRM Simulated Thermal Power - Upscale Function uses drive flow to vary the trip setpoint.

The OPRM Upscale Function uses drive flow to automatically enable or bypass the OPRM Upscale trip output to RPS.

A CHANNEL CALIBRATION of the APRM recirculation drive flow input function requires both calibrating the drive flow transmitters and establishing a valid drive flow /

LIMERICK - UNIT 2 B 3/4 3 -1d Amendment No.

I

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued) core flow relationship.

The drive flow / core flow relationship is established once per refuel cycle, while operating within 10% of rated core flow and within 10% of RATED THERMAL POWER.

Plant operational experience has shown that this flow correlation methodology is consistent with the guidance and intent in Reference 8.

Changes throughout the cycle in the drive flow / core flow relationship due to the changing thermal hydraulic operating conditions of the core are accounted for in the margins included in the bases or analyses used to establish the setpoints for the APRM Simulated Thermal Power -

Upscale Function and the OPRM Upscale Function.

As noted in Table 3.3.1-2, Note "*", the redundant outputs from the 2-Out-Of-4 Voter channel are considered part of the same channel, but the OPRM and APRM outputs are considered to be separate channels, so N = 8 to determine the interval between tests for application of Specification 4.3.1.3 (REACTOR PROTECTION SYSTEM RESPONSE TIME). The note further requires that testing of OPRM and APRM outputs shall be alternated.

Each test of an OPRM or APRM output tests each of the redundant outputs from the 2-Out-Of-4 Voter channel for that function, and each of the corresponding relays in the RPS.

Consequently, each of the RPS relays is tested every fourth cycle.

This testing frequency is twice the frequency justified by References 2 and 3.

Automatic reactor trip upon receipt of a high-high radiation signal from the Main Steam Line Radiation Monitoring System was removed as the result of an analysis performed by General Electric in NEDO-31400A.

The NRC approved the results of this analysis as documented in the SER (letter to George J. Beck, BWR Owner's Group from A.C. Thadani, NRC, dated May 15, 1991).

The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are completed within the time limit assumed in the safety analyses.

No credit was taken for those channels with response times indicated as not applicable except for the APRM Simulated Thermal Power - Upscale and Neutron Flux - Upscale trip functions and the OPRM Upscale trip function (Table 3.3.1-2, Items 2.b, 2.c, and 2.f).

Response time may-be demonstrated by any series of sequential, overlapping or total channel test measurement, provided such tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times.

Response time testing for the sensors as noted in Table 3.3.1-2 is not required based on the analysis in NEDO-32291-A.

Response time testing for the remaining channel components is required as noted.

For the digital electronic portions of the APRM functions, performance characteristics that determine response time are checked by a combination of automatic self-test, calibration activities, and response time tests of the 2-Out-Of-4 Voter (Table 3.3.1-2, Item 2.e).

LIMERICK - UNIT 2 B 3/4 3-le Amendment No. 4-09,

INSTRUMENTATION BASES 3/4.3.7.10 (Deleted) 3/4.3.7.11 (Deleted) -

INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.

3/4.3.7.12 OFFGAS MONITORING INSTRUMENTATION This instrumentation includes provisions for monitoring the concentrations of potentially explosive gas mixtures and noble gases in the off-gas system.

3/4.3.8 (Deleted) -

INFORMATION FROM THIS SECTION RELOCATED TO THE UFSAR.

3/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION The feedwater/main turbine trip system actuation instrumentation is provided to initiate action of the feedwater system/main turbine trip system in the event of failure of feedwater controller under maximum demand.

REFERENCES:

1.

NEDC-30851P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System," March 1988.

2.

NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," October 1995.

3.

NEDC-32410P-A, Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," November 1997.

4.

NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

5.

NEDO-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

6.

NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

7.

Letter, L. A. England (BWROG) to M. J. Virgilio, "BWR Owners' Group Guidelines for Stability Interim Corrective Action," June 6, 1994.

8.

GE Service Information Letter No. 516, "Core Flow Measurement - GE BWR/3, 4, 5 and 6 Plants," July 26, 1990.

9.

GE Letter NSA 00-433, Alan Chung (GE) to Sujit Chakraborty (GE),

"Minimum Number of Operable OPRM Cells for Option III Stability at Limerick 1 & 2," May 02, 2001.

LIMERICK - UNIT 2 B 3/4 3-7 Amendment No. 44, 25, 33, 64, 468,

144,

3/4.4 RFACTOR COOLANT SYSTFM BASES 3/4.4.1 RECIRCULATION SYSTEM The impact of single recirculation loop operation upon plant safety is assessed and shows that single-loop operation is permitted if the MCPR fuel cladding safety limit is increased as noted by Specification 2.1.2, APRM scram and control rod block setpoints are adjusted as noted in Tables 2.2.1-1 and 3.3.6-2, respectively.

Additionally, surveillance on the pump speed of the operating recirculation loop is imposed to exclude the possibility of excessive internals vibration. The surveillance on differential temperatures below 30%

RATED THERMAL POWER or 50% rated recirculation loop flow is to mitigate the undue thermal stress on vessel nozzles, recirculation pump and vessel bottom head during the extended operation of the single recirculation loop mode.

An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable.

Jet pump failure can be detected by monitoring jet pump performance on a pre-scribed schedule for significant degradation.

Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteria for two recirculation loop operation.

The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.

In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single recirculation loop mode.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50'F of each other prior to startup of an idle loop.

The loop temperature must also be within 50'F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles.

Sudden equalization of a temperature difference > 1450F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.

LIMERICK - UNIT 2 B 3/4 4-1 Amendment No. 489,

REACTOR COOLANT SYSTEM BASES 3/4.4.2 SAFETY/RELIEF VALVES The safety valve function of the safety/relief valves operates to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code. A total of 12 OPERABLE safety/

relief valves is required to limit reactor pressure to within ASME III allow-able values for the worst case upset transient.

Demonstration of the safety/relief valve lift settings will occur only during shutdown.

The safety/relief valves will be removed and either set pressure tested or replaced with spares which have been previously set pres-sure tested and stored in accordance with manufacturers recommendations in the specified frequency.

LIMERICK - UNIT 2 B 3/4 4-2 Amendment No. 98,

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.9.1.9 Core Operating Limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the CORE OPERATING LIMITS REPORT for the following:

a.

The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1,

b.

MAPFAC(P) and MAPFAC(F) factors for Specification 3.2.1,

c.

The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.3,

d.

The MCPR(P) and MCPR(F) adjustment factor for specification 3.2.3,

e.

The LINEAR HEAT-GENERATION RATE (LHGR) for Specification 3.2.4,

f.

The power biased Rod Block Monitor setpoints and the Rod Block Monitor MCPR OPERABILITY limits of Specification 3.3.6.

g.

The Reactor Coolant System Recirculation Flow upscale trip setpoint and allowable value for Specification 3.3.6,

h.

The Oscillation Power Range Monitor (OPRM) period based detection algorithm (PBDA) setpoints for Specification 2.2.1.

6.9.1.10 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

a.

NEDE-24011-P-A "General Electric Standard Application for Reactor Fuel" (Latest approved revision),

b.

NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

6.9.1.11 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.

6.9.1.12 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

LIMERICK - UNIT 2 6-18a Amendment No.44, 38, 48, 4-04,

ATTACHMENT 4 LICENSE AMENDMENT REQUEST LIMERICK GENERATING STATION, UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 ACTIVATION OF THE TRIP OUTPUTS OF THE OSCILLATION POWER RANGE MONITOR PORTION OF THE POWER RANGE NEUTRON MONITORING SYSTEM PLANT-SPECIFIC RESPONSES REQUIRED BY NUMAC PRNM RETROFIT PLUS OPTION III STABILITY TRIP FUNCTION TOPICAL REPORT (NEDC-3241OP-A) PHASE 2 OPRM TRIP ACTIVATION/DELETION OF ICA'S

Plant-Specific Responses Required by NUMAC PRNM Retrofit Topical Report LGS OPRM Trip Activation LAR Page 1 of 12 This Attachment 4 provides plant-specific responses required by the generic NRC approved General Electric (GE) Nuclear Measurement Analysis and Control (NUMAC) Power Range Neutron Monitor (PRNM) Licensing Topical Report (LTR) NEDC-3241 OP-A (including Supplement 1), "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option IlIl Stability Trip Function" (see Attachment 1, References 2 and 3). The responses below also provide descriptions and justifications for each deviation from the NUMAC PRNM LTRs. The section numbers listed below are the Utility Actions Required from the NUMAC PRNM LTRs, and are only the ones that are unique to the Oscillation Power Range Monitor (OPRM) portion of the PRNM, or items with additional information from that in the original submittal. All other items were addressed in the previous Limerick Generating Station (LGS), Units 1 and 2, submittal covering the overall PRNM installation (see Attachment 1, Reference 20), as approved by the NRC in Amendment approval letters dated April 12, 2000, and January 16, 2001 (see Attachment 1, References 22 and 23, respectively). At that time, the OPRM Upscale Function was not being activated, so OPRM specific responses were deferred.

Section Utility Action Required

Response

7.6 Impact on UFSAR Applicable sections of the UFSAR related to PRNM were reviewed and appropriate revisions of The plant-specific action required for those sections prepared and approved as part of FSAR updates will vary between plants.

the normal design process in support of the PRNM In all cases, however, existing FSAR modification (previously reviewed and approved documents should be reviewed to by the NRC). In support of activation of the identify areas that have descriptions OPRM Upscale Function as part of the normal specific to the current PRNM using the LGS modification process, the UFSAR will be general guidance of Sections 7.2 through reviewed with appropriate additions and any 7.5 to identify potential areas impacted.

needed revisions identified. Following The utility should include in the plant-implementation of the design modification, and specific licensing submittal a statement closure of the design package, the UFSAR of the plans for updating the plant FSAR revisions will be submitted to the NRC as part of for the PRNM proiect.

the routine UFSAR update submittal.

8.3.6.1 APRM-Related RPS Trip Functions -

The Technical Specifications (TS) Table 2.2.1-1 Setpoints and the TS Bases description for the APRM Add to or delete from the appropriate Simulated Thermal Power -- Upscale Function document any changed RPS selpoint were updated as part of the original TS changes information. If ARTS is being for the LGS PRNM implementation. The trip implemented concurrently with the setpoint and allowable value for Function 2.b were PRNM modification, either include the shown in the form "0.66W+offset", with a 5%

related Tech Spec submittal information difference between the offset values for two-loop with the PRNM information in the plant-operation (TLO) and single-loop operation (SLO).

specific submittal, or reference the ARTS The 5% difference is a plant-specific value that submittal in the PRNM submittal. In the was established prior to the PRNM modification.

plant-specific licensing submittal, identify The value and its purpose, to compensate for whant-changific aicenyia subein, ibackflow" in the jet pumps associated with a non-what changes, if any, are being operable recirculation drive flow loop, were not method used for the calculation of affected by the PRNM modification.

setpoints and where the setpoint Consequently, the 5% value and the

.inforano chaes be reore..

representation of the setpoint equation were carried over from the original Power Range Monitor system with only those adjustments in offset value necessary to apply to Simulated Thermal Power.

No change to the allowable value or trip setpoint equations is required for the OPRM Upscale

Plant-Specific Responses Required by NUMAC PRNM Retrofit Topical Report LGS OPRM Trip Activation LAR Page 2 of 12 Section Utility Action Required

Response

Function implementation. However, to more accurately present both the intent of the adjustment (offset the flow) and the way the adjustment is actually accomplished in the PRNM equipment, the SLO equations in Table 2.2.1-1 have been rearranged to the form "0.66 (W-AW) +

offset." AW = (5%/0.66) = 7.6%. This alternate form is mathematically equivalent to the current presentation of the equation.

The AW offset for SLO is only required for drive flows above about 40% for Limerick, but for simplicity of implementation, is applied for flows down to W=AW. For drive flows below AW, the equation is 'clamped' at the offset value. To document this limit of application, a footnote has been added to Table 2.2.1-1.

In addition to the above Table 2.2.1-1 changes, the associated TS Bases have also been expanded to include a short description of the SLO adjustment and the limits of application.

A corresponding change is also being implemented in Table 3.3.6-2 for the APRM Simulated Thermal Power - Upscale rod block function.

See the LGS TS and TS Bases markup for the specific changes. See response for Section 8.4.6.1 for a discussion of OPRM setpoints.

8.4.1.4 OPRM-Related RPS Trip Functions -

An OPRM Upscale Function has been added to Functions Covered by Tech SDecs the LGS TS as an UAPRM function" (Function 2.f) consistent with Appendix H to the NUMAC PRNM Add the OPRM Upscale function as an LTR Supplement 1. Additions to the TS Bases

'APRM function" in the RPS have also been incorporated consistent with the Instrumentation 'function" table. Also NUMAC PRNM LTR but expanded to clarify the add the related surveillance intent. Some of the discussion shown in TS requirements and, if applicable, the Bases 2.2.1 in the NUMAC PRNM LTR has been related setpoint, and the related included in TS Bases 3/4.3.1. TS Bases 3/4.3.1 descriptions in the bases sections.

has been expanded significantly from that shown Perform analysis necessary to establish in the NUMAC PRNM LTR to provide a more setpoints for the OPRM Upscale trip.

complete discussion. The TS Bases additions for Add discussions related to the OPRM the 2-Out-Of-4 Voter Function 2.e are modified function in the bases for the APRM Inop somewhat from those shown in the LTR, and 2-Out-Of-4 Voter functions.

Supplement 1. These modifications are conservative in that they delete any discussion of NOTE: The markups in Appendix H of a partially OPERABLE" Voter Function. In Supplement 1 show the OPRM Upscale addition, the modified TS Bases text includes as an APRM sub-function. However, discussion (not included in the LTR) of the individual plants may determine that for hardware that implements the voter function. The their particular situation, addition of the added wording clarifies that operability of parts of OPRM to the RPS Instrumentation table the hardware that are not related to the voter separate from the APRM, or as a function do not need to be considered in separate Tech Spec, better meets their determining operability of the voter function.

needs. In those cases, the basis See the LGS TS and TS Bases markup for the elements of the Tech Spec as shown in specific changes. See response for Section

Plant-Specific Responses Required by NUMAC PRNM Retrofit Topical Report LGS OPRM Trip Activation LAR Page 3 of 12 Section Utility Action Required

Response

this Supplement would remain, but the 8.4.6.1 for a discussion of OPRM setpoints.

specific implementation would be different.

8.4.2.4 OPRM-Related RPS Trip Functions -

Minimum Number of Onerable OPRM Channels For the OPRM functions added (Section 8.4.1), include in the OPRM Tech Spec a "minimum operable channels" requirement for three OPRM channels, shared by both trip systems.

Add the same action statements as for the APRM Neutron Flux - High function for OPRM Upscale function. In addition, add two new action statements for OPRM Upscale function unavailable per Paragraph 8.4.2.2.

Revise the Bases section as needed to add descriptions of the 4-OPRM system with 2-out-of-4 output voter channels (2 per RPS Trip System), and allowed one OPRM bypass total.

A minimum operable channels requirement of three, shared by both trip systems, has been included in the TS for the OPRM Upscale Function (Function 2.f), consistent with the NUMAC PRNM LTRs. The LCO Actions are modified only to include the OPRM Upscale Function 2.f, consistent with the NUMAC PRNM LTR Supplement 1. [Note that the LGS Actions a., b.,

c., and d. are structured somewhat differently than those shown in the NUMAC PRNM LTRs. That difference was justified along with the original PRNM TS changes. The changes for OPRM, however, are equivalent to those shown in the NUMAC PRNM LTR Supplement 1.]

The added Table 3.3.1-1 Action Statements and TS Bases descriptions are consistent with the NUMAC PRNM LTR and LTR Supplement 1 except that the Action has been structured in two parts, Action Statement 10a and Action Statement 1 Ob, for better clarity.

The wording of Action Statement 1 Oa is also modified from that shown for the corresponding Action Statement in the NUMAC PRNM LTR Supplement 1 to clarify that the "alternate methods" action only applies when a common mode OPRM deficiency (defined as an unanticipated characteristic of the instability detection algorithm or equipment that would render all OPRM channels inoperable at once) is the cause of OPRM Upscale Function unavailability. That was the intent of the LTR Action Statement, but the wording of the Action Statement in the LTR was not specific.

The TS Bases 3/4.3.1 discussion has been modified from the LTR proposed text to reflect that plant startup may continue while operating within the allowed completion time of Action 1 Oa. The primary purpose of this is to allow an orderly completion, without undue impact on plant operation, of design and verification activities in the event of a required design change to the OPRM Upscale Function. This exception is not intended as an alternative to restoring inoperable equipment to OPERABLE status in a timely manner.

See the response for Section 8.4.1.4 for a discussion of the 2-Out-Of-4 Voter function.

See the LGS TS and TS Bases markup for the specific changes.

Plant-Specific Responses Required by NUMAC PRNM Retrofit Topical Report LGS OPRM Trip Activation LAR Page 4 of 12 Section ]

Utility Action Required I

Response

8.4.3.4 OPRM-Related RPS Trip Functions -

Applicable Modes of Operation Add the requirement for operation of the OPRM Upscale function in Mode 1 (run) when Thermal Power is > 25% RTP, and add Bases descriptions as required.

A requirement of Operational Condition 1 > 25%

RTP, consistent with the NUMAC PRNM LTR Supplement 1, has been included in the TS Table 3.3.1-1 along with associated TS Bases descriptions. Similarly, the same requirement has been included as the "Operational Conditions For Which Surveillance Required" in Table 4.3.1.1-1.

The wording of the new supporting Note "o" in Table 3.3.1-1 is modified from that in the NUMAC PRNM LTR to make "auto-enable" the operational requirement, and "auto-bypass" an allowed action rather than required action. The revised wording is more consistent with the intent, which is to assure that the trip is enabled inside the limits.

Also, LGS has chosen to include a note (Note "p")

in Table 3.3.1-1 defining the minimum of operable LPRMs per cell and minimum operable cells per OPRM channel requirements. The NUMAC PRNM LTR Supplement 1 included this requirement in the TS Bases only. The LGS notation is consistent with the equivalent minimum LPRM requirement for APRM operability, currently included as a Note "e" in Table 3.3.1-1 in the LGS TS. The actual operability requirement in the LGS TS is the same as in the NUMAC PRNM LTR Supplement 1.

The NUMAC PRNM LTR Supplement 1 repeats the "auto-bypass" requirement as part of the

'Operational Conditions for Which Surveillance Required" note to Table 4.3.1.1-1 along with the

"Ž25%" requirement. LGS has determined that the note to Table 4.3.1.1-1 should include only the

"Ž25%" part of the requirement since the "auto-bypass" requirement does not apply to that table.

The TS Bases discussion for Operational Conditions has been included in TS Bases 3/4.3.1 for LGS vs. TS Bases 2.2.1 in the NUMAC PRNM LTR Supplement 1. The wording has been modified somewhat from the LTR proposed text for improved clarity of the intent.

See the LGS TS and TS Bases markup for the specific changes.

8.4.4.1.4 OPRM-Related RPS Trip Functions -

A Channel Check requirement of once per day Channel Check has been included for the OPRM Upscale Add once per 12-hour or once per day Function in Table 4.3.1.1-1, consistent with the Channel Check or Instrument Check NUMAC PRNM LTR Supplement 1.

requirements for the OPRM Upscale See the LGS TS markup for the specific changes.

I function.

SeeIthe________markupforthespecificchanges

Plant-Specific Responses Required by NUMAC PRNM Retrofit Topical Report LGS OPRM Trip Activation LAR Page 5 of 12 Section ]

Utility Action Required I

Response

8.4.4.2.4 OPRM-Related RPS Trin Functions -

Channel Functional Test Add Channel Functional Test requirements with a requirement for a test frequency of every 184 days (6 months), including the 2-Out-Of-4 Voter function.

Add a "confirm auto-enable region' surveillance on a once per outage basis up to 24 month intervals.

A Channel Functional Test requirement has been included in Table 4.3.1.1-1 consistent with the NUMAC PRNM LTR Supplement 1 except that a reference to Note We" has been added to include the flow input function excluding the flow transmitters. The NUMAC PRNM LTR Supplement 1 includes this requirement only for the APRM Simulated Thermal Power function, but it has been included for the OPRM Upscale Function for LGS since that function uses flow for the auto-enable function.

A 'confirm auto-enable region" surveillance requirement has been included as Note "c" to the Channel Calibration requirement in Table 4.3.1.1 -

1. The note states that Channel Calibration includes confirmation that the OPRM Upscale trip output auto-enable (not-bypassed) setpoints remain correct. The surveillance requirement wording is similar to that in the LTR, except that the 'of rated recirculation drive flow" phrase has been deleted. That terminology is not commonly used at LGS. Since there is only one "flow" in the APRM, this deletion does not change the requirement and avoids potential confusion.

The TS Bases 3/4.3.1 wording has been expanded to include discussion of the flow input function calibration, and to include references to more completely document the bases for the specifications. Use of the term 'of rated recirculation drive flow" has been omitted to avoid potential confusion at LGS where the terminology is not commonly used. These changes have no effect on the actual TS requirements as originally defined in the NUMAC PRNM LTRs, but provide a more complete and clear discussion to reduce the risk of confusion in application of those requirements.

See the LGS TS and TS Bases markup for the specific changes.

-t 8.4.4.3.4 1 OPRM-Related RPS Tric, Functions -

Channel Calibrations Add calibration interval requirement of every 24 months for the OPRM Upscale function.

Revise Bases text as required.

A Channel Calibration requirement for the OPRM Upscale Function has been added to Table 4.3.1.1-1, consistent with the NUMAC PRNM LTR Supplement 1, but also with some additional changes not included in the LTR as discussed further below.

The original PRNM modification includes, via Note Ug" to Table 4.3.1.1-1, a requirement associated with the APRM Simulated Thermal Power -

Upscale to calibrate the associated recirculation loop flow input function. This requirement assures that the recirculation drive flow used by the APRM Simulated Thermal Power flow biased trip was properly calibrated. The NUMAC PRNM LTR.

Plant-Specific Responses Required by NUMAC PRNM Retrofit Topical Report LGS OPRM Trip Activation LAR Page 6 of 12 Section Utility Action Required

Response

Supplement 1 does not identify any equivalent requirement for the OPRM Upscale Channel Calibration. However, LGS has determined that such a requirement should also be included for the OPRM Upscale Function because recirculation drive flow is used for the OPRM Upscale trip output auto-enable function. Therefore, the proposed OPRM Upscale Function LGS TS Channel Calibration addition also includes in Table 4.3.1.1-1 a reference to Note "g."

The wording of the new Note "c" to Table 4.3.1.1-1 (confirmation of setpoints for auto-enable region) has been modified slightly from that in the NUMAC PRNM LTR Supplement 1 to be more consistent with the real intent of the requirement.

It has also been determined that some related TS Bases changes are necessary. Specifically, two aspects should be clarified: 1) the wording should recognize that recirculation drive flow is also used as an input to the OPRM Upscale trip auto-enable function, and 2) the "calibrating the recirculation loop flow channel" must include the recirculation drive flow/ core flow correlation. The TS Bases 3/4.3.1 discussion has been expanded from that shown in the NUMAC PRNM LTRs to include discussion of these two aspects.

See the LGS TS and TS Bases markup for the specific changes.

8.4.4.4.4 l OPRM-Related RPS Trin Functions -

Response Time Testing Modify as necessary the response time testing procedure for the 2-Out-Of-4 Voter function to include the voter OPRM output solid-state relays as part of the response time tests, alternating testing of the voter OPRM output with the voter APRM output.

Function 2.f has been added to Table 3.3.1-2 with response time requirement of 'NA" consistent with the NUMAC PRNM LTR Supplement 1.

The original PRNM modification expanded footnote "." to Table 3.3.1-2 to clarify that for application of RPS Response Time Testing per TS 4.3.1.3, the redundant APRM outputs were considered one channel, so N=4. The OPRM adds another set of redundant outputs for each 2-Out-Of-4 Voter channel. One of the OPRM outputs and one of the APRM outputs are connected in series to the coil of one RPS interface relay. The second OPRM output and the second APRM output from the 2-Out-Of-4 Voter channel are connected in series with the coil to a second RPS interface relay. There are 8 total RPS interface relays.

The NUMAC PRNM LTR Supplement 1 justified response time testing at a rate that tested one RPS Interface relay every plant operating cycle, with tests using the APRM output for one cycle and the OPRM output for the next cycle. This yields a testing rate once per 8 operating cycles for each RPS interface relay and once per every 16 operating cycles for the APRM or OPRM

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output.

When the PRNM modification was installed at LGS, the decision was made to test the RPS interface relays at the rate of 2 per operating cycle, twice the rate justified by the NUMAC PRNM LTR Supplement 1. The two relays tested are those driven from one Voter. Since OPRM was not operable at that time, and since testing an RPS interface relay requires operating a Voter output, one redundant set of APRM outputs from the Voter is also tested each operating cycle. This testing rate is four times the rate justified in the NUMAC PRNM LTR Supplement 1.

The response time testing proposed will retain the frequency of testing of the RPS interface relays, but alternate the "actuating" output from the Voter between the OPRM and APRM outputs. This testing is consistent with the sequencing described in NUMAC PRNM LTR Supplement 1, but at twice the rate for all components. To define this testing rate, Note "*" to Table 3.3.1-2 is being modified to define the APRM and OPRM outputs from the Voter as separate channels so that N=8 for application of TS 4.3.1.3, and that the test of the OPRM and APRM outputs shall alternate. In addition, because this sequencing may be confusing, a description of the RPS Response Time Testing requirement for the Voter Function 2.e has been added to the TS Bases 3/4.3.1. The specific tests will be defined in LGS procedures.

See the LGS TS and TS Bases markup for the specific changes.

8.4.5.4 OPRM-Related RPS Trip Functions -

Logic System Functional Testing Add requirement for Logic System Functional Test every refueling cycle, 18 or 24 months at the utility's option based on which best fits plant scheduling.

An OPRM Upscale Function Logic System Functional Testing (LSFT) has been added consistent with the NUMAC PRNM LTR, Supplement 1. As described in the LTR, the actual OPRM related LSFT is the same as for the APRM, a test of the 2-Out-Of-4 Voter only.

Consistent with the NUMAC PRNM LTR Supplement 1, the only change required to implement the OPRM LSFT is the addition of "OPRM" and "2.f" in Surveillance Requirement 4.3.1.2.

See the LGS TS markup for the specific changes.

Position on Compliance With TS SR 4.3.1.2 for the OPRM Upscale Function:

It is LGS' position that performance of Surveillance Requirement 4.3.1.2 relative to the OPRM Upscale voting function within the 2-Out-Of-4 Voter channel can be considered met via acceptance testing performed at the factory, in-

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plant functional testing of the hardware, and internal self testing performed by the hardware.

The next subsequent performance of this LSFT for the OPRM Upscale function will be during the first refueling outage following activation of the OPRM Upscale trip output. Justification for this position is included following this table.

8.4.6.1 OPRM-Related RPS Trip Functions -

Setpoints Add setpoint information to the appropriate document and identify in the plant-specific submittal the basis or method used for the calculation and where the setpoint information will be recorded.

An OPRM Upscale Function entry has been added to Table 2.2.1-1 consistent with the NUMAC PRNM LTR Supplement 1 except that the "see COLR" reference in the LTR has been expanded as a footnote. The note added to Table 2.2.1-1 addresses the auto-enable setpoints and documents that the Period Based Detection Algorithm (PBDA) setpoints are in the Core Operating Limits Report (COLR). The NUMAC PRNM LTR Supplement 1 markup identified only the COLR with no clarification of what setpoints were defined therein.

The OPRM related setpoints and adjustable parameters, and the basis or method for establishing them, are discussed further below.

Also, to better document this information, the wording in TS Bases 2.2.1 has been expanded and modified from that shown in the NUMAC PRNM LTR Supplement 1.

There are four "sets" of OPRM related setpoints and adjustable parameters: a) OPRM trip auto-enable (not-bypassed) setpoints for Simulated Thermal Power and recirculation drive flow; b)

PBDA confirmation count and amplitude setpoints; c) period based detection algorithm tuning parameters; and d) growth rate algorithm (GRA) and amplitude based algorithm (ABA) setpoints.

The first set, the setpoints for the "auto-enable" region for OPRM, as discussed in the expanded TS Bases 2.2.1, will be treated as nominal setpoints with no additional margins added. The deadband for these setpoints is established so that it increases the enabled region. The settings, 30% APRM Simulated Thermal Power and 60%

recirculation drive flow, are defined (limit values) in the footnote added to Table 2.2.1-1.

The second set, the PBDA trip setpoints, will be established in accordance with the BWROG LTR 32465-A methodology (see Attachment 1, Reference 4), previously reviewed and approved by the NRC. As stated in the footnote added to Table 2.2.1-1, these setpoints are documented in the COLR.

The third set, the PBDA "tuning" parameter values, are established in accordance with and

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controlled by LGS procedures only, within the limits established in the BWROG LTR (see, Reference 4). This is discussed in the TS Bases, but are not addressed in the TS.

The fourth set, the GRA and ABA setpoints, consistent with the BWROG LTRs (see, References 4, 5 and 6), are established as nominal values only. The TS Bases 2.2.1 documents that these setpoints are controlled by LGS procedures.

See the LGS TS and TS Bases markups for the

.specific changes.

None Recirculation Loons Operating LCO 3.4.1.1 currently requires operation in the "Unrestricted" zone of the TS Figure 3.4.1.1-1. This restriction and associated required Actions and Surveillance Requirements were implemented as part of the Interim Corrective Actions in response to NRC Bulletin 88-07, including Supplement 1.

No action identified in the NUMAC PRNM LTR.

Concurrent with activation of the OPRM Upscale Function, LCO 3.4.1.1, its associated Actions and Surveillance Requirements, and the related TS Bases are being revised to delete requirements related to the restricted zone of operation. The implementation of the automatic OPRM Upscale Function eliminates the need for the Interim Corrective Actions (ICAs) and the related administrative requirements implemented in LCO 3.4.1.1. The other LCO conditions limiting operation with one recirculation loop out of service (single loop operation) are retained as are the Action requirements associated with no recirculation loop operation that were not related to ICAs. Action statements and Surveillance Requirements have been modified to delete portions required to support the ICAs.

In addition to deletion of requirements to exit the restricted zone within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, Action b. has been simplified to require only that the plant be in at least Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This change does not increase the total time allowed to reach Hot Shutdown, but removes the requirement to be in Startup within the first 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. With the added protection of the automatic OPRM Upscale Function to detect potential instabilities, this change is judged reasonable in that it allows more flexibility in the operator actions to achieve an orderly plant shutdown and simplifies the Action time tracking. This change makes the LGS TS 3.4.1.1 allowed completion time for this Action consistent with a similar allowed Action completion time for TS 3.4.1.2 Get pumps inoperable) and the Completion Time for the equivalent Required Action in the Improved Standard Technical Specifications (NUREG-1433) (see Attachment 1, Reference 7). The TS Bases have been modified consistent with these TS changes.

See the LGS TS and TS Bases markup for the I specific changes.

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None PRNM Related Control Rod Block Technical Specification 3.3.6 Clarifications No action identified in the NUMAC PRNM LTR.

In the process of preparing the PRNM modification and RPS TS changes previously implemented, a need was identified to clarify or modify some of the associated RPS TS requirements including the Intermediate Range Monitor (IRM) (Table 3.3.1-1, RPS Functions 1.a and 1.b) operability requirements in Operational Condition 5. In conjunction with the PRNM modification, those requirements were modified to limit the operability requirement in Operational Condition 5 by adding a reference to an existing 'Note (i)." That note requires operability only if a control rod is withdrawn with an exception to this requirement for control rods withdrawn per TS 3.9.10.1 and 3.9.10.2. Justification for this change was included with the original PRNM licensing submittal, which has been approved by the NRC.

At the time IRM RPS changes were implemented, it was not recognized that there were similar Operational Condition 5 requirements for operability of the IRM Rod Block functions in TS 3.3.6. To make the TS 3.3.6 Rod Block Operational Condition 5 operability requirements consistent with those for RPS, a change is proposed to add footnote "*" to the Operational Condition 5 requirement for Functions 4.a, 4.b, 4.c and 4.d in Table 3.3.6-1 and Table 4.3.6-1. The existing Note "**" for those tables includes the same restrictions as the Note (i) for Table 3.3.1-1 (RPS). This change will make the Rod Block and RPS Operational Condition 5 operability requirements the same.

Justification:

The primary purpose of the IRM Rod Block is to provide an alarm to the operator prior to reaching an RPS trip point. Therefore, the Operational Condition 5 operability requirement for the Rod Block function should be the same as the Operational Condition 5 operability requirement for the corresponding RPS function.

See the LGS TS markup for the specific changes.

None Core Operating Limits Report Requirements have been added to TS Section Administrative Controls requirements in 6.9.1.9 to include the OPRM PBDA setpoints in TS Sections 6.9.1.9 and 6.9.1.1 0 do not the COLR, and in TS Section 6.9.1.1 0 to identify currently address the OPRM.

the BWROG LTR (see Attachment 1, Reference

4) as the basis.

See the LGS TS markup for the specific changes.

None PRNM Related RPS Technical As part of the original PRNM TS modification, a Specification 3.3.1 Clarifications Note "(n)" was added to Table 3.3.1-1.

Justification for addition of that note was included No action identified in the NUMAC in the original PRNM TS submittal. Operational PRNM LTR experience has shown that the location of that

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note in the TS structure is awkward. To clarify the TS, the content of the existing Note (n) is being moved to a new footnote "***'to the Action statements for TS 3.3.1. A U***a reference to the footnote is being added to Action b. and Action c.

The movement of the notation does not make any technical change to the TS requirements, but more clearly relates the note to the applicable Action statements.

See the LGS TS markup for the specific changes.

9.1.3 Utility Quality Assurance Program As part of the plant-specific licensing The activation of the OPRM trip is accomplished submittal, the utility should document the by removing hardware jumpers in the panel.

established program that is applicable to There are no required firmware changes.

the project modification. The submittal should also document for the project what scope is being performed by the utility and what scope is being supplied by others. For scope supplied by others, document the utility actions taken or planned to define or establish requirements for the project, to assure those requirements are compatible with the plant-specific configuration. Actions taken or planned by the utility to assure compatibility of the GE quality program with the utility program should also be documented.

Utility planned level of participation in the overall V&V process for the project should be documented, along with utility plans for software configuration management and provision to support any required changes after delivery should be documented.

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Logic System Functional Test - as being satisfactorily met for the OPRM Upscale voting function of the 2-Out-Of-4 Voter channel Surveilljance Requirement (SR) 4.3.1.2 is normally performed during an outage because the method for performing the SR creates a full RPS trip (scram). LGS plans to activate the OPRM Upscale Function on-line during full power operation, so the normal method of performing SR 4.3.1.2 can not be used. LGS has evaluated alternative methods for performing the SR on-line.

One alternative requires careful coordination of actions in multiple channels, with a very limited time available. For this alternative, any minor error or delay in the sequence of actions, which would normally have no adverse consequences, will lead to an unintentional scram. A second alternative has been identified that carries a smaller risk of inadvertent scram, but requires disconnecting multiple fiberoptic cables within the cabinet, an action that would normally not be required, and creates a significantly increased risk of causing equipment damage or equipment inoperability.

Since the only identified alternatives for performing the SR while at power carry significant risk of causing problems, LGS has evaluated the overall testing that has been and will be performed for the equipment performing the OPRM Upscale Function. Based on that evaluation, LGS has determined that the intent of the Logic Systerri Functional Test (LSFT) for the OPRM Upscale Function testing will be met and that this SR will not need to be performed until the next refuel outage following trip activation. The basis for that conclusion is described below.

The primary purpose of SR 4.3.1.2 is to reconfirm that the 2-out-of-4 voting logic is still functioning correctly. As stated in the NUMAC PRNM LTR, the test of the voting logic in the LSFT SR is redundant to an automatic self-test function that repetitively injects test signals for all combinations of inputs to confirm that the voting logic continues to function correctly.

Failures detected by the self-test function are reported via the associated APRM channel to the operator. The NUMAC PRNM LTR states that the LSFT SR provides "overlap" between the automatic self-test of the voting logic and the voter output test provided by the Channel Functional Test SR, which will be performed at the time of OPRM Upscale Function activation.

At the request of LGS, GE has re-evaluated the final hardware design and confirmed that the Channel Functional Test SR and the automatic self-test of the voting logic provide full overlap, so the LSFT is not required for coverage. GE further clarified that the primary reason for the NUMAC PRNM LTR recommended LSFT coverage of the voting logic was to provide "defense-in-depth" due to the lack of operational experience with the new equipment. Since the time the NUMAC PRNM LTR was approved, the same voter hardware used at LGS has been installed at 10 BWRs, with over 20 plant-years of operation without any identified failures of the voting logic.

LGS performed the equivalent of the LSFT for the OPRM Upscale Function during the factory acceptance test (FAT) prior to installation for both units. The normal LSFT SR was performed for the APRM Upscale/lnop voting logic after installation of the PRNM equipment prior to start-up for the current cycle. No voting logic problems were found during these tests.

Based on (1) the determination that the LSFT provides no additional hardware test coverage beyond that provided by the automatic self-test and the Channel Functional Test SR to be performed at the time of trip activation, (2) the completion of an equivalent OPRM LSFT during the FAT and the normal LSFT on the APRM Upscale/lnop voting logic without detected problems, and (3) the extensive operating experience at other BWR plants without voting logic failures, LGS has concluded that TS SR 4.3.1.2, as it applies to the OPRM Upscale Function of the 2-Out-Of-4 Voter channel, has already been satisfied and need not be performed until the next refueling outage following activation of the OPRM Upscale Function.