Text

Application to Incorporate Licensing Topical Report NEDE-33885P-A, Revision 1, GNF CRDA Application Methodology
	ML21041A490
Person / Time
Site:	LaSalle
Issue date:	02/10/2021
From:	Demetrius Murray; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RS-21-006 NEDE-33885P-A, Rev. 1
	Download: ML21041A490 (62)
	v • d • e

4300 Winfield Road Warrenville , IL 60555 Exelon Generation 630 657 2000 Office RS-21-006 10 CFR 50.90 February 10, 2021 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374

Subject:

Application to Incorporate Licensing Topical Report NEDE-33885P-A, Revision 1, "GNF CRDA Application Methodology"

References:

1. Licensing Topical Report, "GNF CRDA Application Methodology," NEDE-33885P-A, Revision 1, dated March 2020

2. Final Safety Evaluation for Global Nuclear Fuel - Americas LLC (GNF),

Licensing Topical Report NEDE-33885P, Revision 0, "GNF Control Rod Drop Accident Application Methodology," dated January 16, 2020 (EPID L-2018-TOP-0006)

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) is submitting a license amendment request for LaSalle County Station, Units 1 and 2 (LSCS) to incorporate the NRC-approved methodology described in Licensing Topical Report, "GNF CRDA Application Methodology," NEDE-33885P-A, Revision 1 (Reference 1).

The proposed change revises the LSCS Technical Specifications (TS) to incorporate to the methodology in Reference 1 by modifying LSCS TS Sections 3.1.3, "Control Rod Operability,"

3.1.6, "Rod Pattern Control," and 3.3.2.1, "Control Rod Block Instrumentation," to allow for greater flexibility in rod control operations during various stages of reactor power operation at LSCS. The proposed amendment will modify the requirements on control rod withdrawal order and conditions to protect against a postulated control rod drop accident (CRDA) during startup and low power conditions .

The proposed amendment also eliminates the control rod operability separation criterion while operating in those conditions. The changes are being considered to align the plant startup sequences with the calculated control rod reactivity worths during the control rod withdrawal process. This methodology incorporates the characteristics of advanced fuel products and the latest analytical methods into the design basis for the CRDA to meet the requirements for fuel cladding failure thresholds and allow more flexibility during plant startups.

February 10, 2021 U.S. Nuclear Regulatory Commission Page 2 These TS changes are consistent with the NRC-approved methodology in Reference 1.

The proposed changes have been reviewed by the LSCS Plant Operations Review Committee in accordance with the EGC Quality Assurance Program. provides a description and assessment of the proposed changes . Attachment 2 provides the proposed TS markups for this amendment request. Attachment 3 provides the proposed TS Bases markups for information only.

Approval of the proposed amendment is requested by February 10, 2022 to support the upcoming LaSalle County Station, Unit 1 refueling outage. Once approved, the amendments shall be implemented prior to achieving MODE 2 following the Unit 1 refueling outage.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State Officials.

There are no regulatory commitments contained within this letter. Should you have any questions concerning this letter, please contact Mr. Jason Taken at (630) 806-9804.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 101h day of February, 2021.

Dwi urray Sr. Manager - Licensing Exelon Generation Company, LLC Attachments: 1. Description and Assessment

2. LSCS TS Markups for Proposed Changes to Incorporate CRDA LTR

3. LSCS TS Bases Markups for Proposed Changes to Incorporate CRDA LTR (For Information Only) cc: NRC Regional Administrator, Region Ill NRC Senior Resident Inspector - LaSalle County Station NRC Project Managers - LaSalle Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1 LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374 Description and Assessment

ATTACHMENT 1 Description and Assessment

Subject:

Application to Incorporate Licensing Topical Report NEDE-33885P-A, Revision 1, "GNF CRDA Application Methodology" 1.0 Summary Description 2.0 Detailed Description 2.1 Variations 3.0 Techn ical Evaluation 3.1 Applicability of Safety Evaluation 3.2 Technical Justification 4.0 Regulatory Evaluation 4.1 Applicable Regulatory Requirements 4.2 No Significant Hazards Consideration Analysis 4.3 Conclusion 5.0 Environmental Evaluation 6.0 References

ATTACHMENT 1 Description and Assessment 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) is submitting a license amendment request for LaSalle County Station , Units 1 and 2 (LSCS) to incorporate the NRC-approved methodology described in Licensing Topical Report (L TR), "GNF CRDA Application Methodology," NEDE-33885P-A, Revision 1 (Reference 1).

The proposed change revises the LSCS Technical Specifications (TS) to incorporate to the methodology in Reference 1 by modifying LSCS TS Sections 3.1.3, "Control Rod Operability,"

3.1 .6, "Rod Pattern Control," and 3.3.2.1, "Control Rod Block Instrumentation," to allow for greater flexibility in rod control operations during various stages of reactor power operation at LSCS. The proposed amendment will modify the requirements on control rod withdrawal order and conditions to protect against a postulated control rod drop accident (CRDA) during startup and low power conditions .

The proposed amendment also eliminates the control rod operability separation criterion while operating in those conditions. The changes are being considered to align the plant startup sequences with the calculated control rod reactivity worths during the control rod withdrawal process. This methodology incorporates the characteristics of advanced fuel products and the latest analytical methods into the design basis for the CRDA to meet the requirements for fuel cladding failure thresholds and allow more flexibility during plant startups.

2.0 DETAILED DESCRIPTION The proposed amendment incorporates the requirements of Reference 1 into the LSCS licensing basis.

This CRDA methodology supersedes the Banked Position Withdrawal Sequence (BPWS) methodology that is currently in use at LSCS. By incorporating the Reference 1 methodology, changes to the applicability region and the out-of-sequence control rod requirements are made.

A description of the specific Technical Specification (TS) changes is provided below. The proposed TS changes are provided in Attachment 2 of this letter.

The proposed amendment deletes TS 3.1 .3 Condition D for "Two or more inoperable control rods not in compliance with analyzed rod position sequence and not separated by two or more OPERABLE control rods." This was a generic BPWS requirement that is superseded in the Reference 1 methodology. There are no generic separation criteria required with the Reference 1 methodology.

The proposed amendment replaces the minimum Thermal Power requirement above which there are no constraints on control rod withdrawal order required to protect against a CRDA with a new minimum thermal power or a reactor steam dome pressure threshold as follows:

THERMAL POWER less than or equal to 5% Rated Thermal Power (RTP) or REACTOR 1 of 9

ATTACHMENT 1 Description and Assessment STEAM DOME PRESSURE less than or equal to 300 psig. The reactor steam dome pressure condition would be exceeded in MODE 2, so there is no longer a need for MODE 1 applicability.

The proposed amendment modifies TS 3.1 .6, "Rod Pattern Control ," to apply to all control rods.

Additionally, TS 3.1.6 is being modified to align with the new minimum thermal power limit and reactor steam dome pressure limit as described above.

TS 3.3.2.1, "Control Rod Block Instrumentation ," is being modified to incorporate the new minimum thermal power limit and reactor steam dome pressure limit described above.

2.1 Variations EGC has identified variations between the approved LTR TS pages and LSCS TS pages. These variations are editorial in nature and do not affect the justification for incorporation of this LTR for LSCS.

The change to Surveillance Requirement (SR) 3.3.2.1.6 in LSCS TS pages will vary from the proposed changes in the approved LTR TS pages. Specifically, the LTR TS pages delete the "when THERMAL POWER is~ 10% RTP." LSCS TS SR will deviate from this change as the Rod Worth Minimizer is required to not be bypassed when THERMAL POWER is less than or equal to 5%, which is the new thermal power limit described in this letter. EGC concludes this variation from the LTR does not affect the justification for incorporation of this methodology.

Wording changes throughout the LTR where "BPWS" was replaced with "analyzed rod position sequence" will not be incorporated through the proposed amendment because these changes have already been implemented in the LSCS TS as of Amendment No. 147 for Unit 1 (ML011130202) and Amendment No. 133 for Unit 2 (ML011130202). EGC concludes this variation from the LTR does not affect the justification for incorporation of this methodology.

3.0 TECHNICAL EVALUATION

3.1 Applicability of Safety Evaluation EGC has reviewed the safety evaluation for the CRDA LTR provided to GE-Hitachi Nuclear Energy Americas, LLC in a letter dated January 16, 2020 (Reference 4). This review included a review of the NRC staffs evaluation, as well as the information provided in the CRDA LTR. As described herein, EGC concludes that the methodology as provided in the CRDA LTR along with the limitations and conditions provided in the staff's evaluation are applicable to LSCS , and justify this amendment for the incorporation of Reference 1 along with the proposed TS changes described in this letter.

3.2 Technical Justification General Design Criteria (GDC) 28, "Reactivity Limits," of 10 CFR 50, Appendix A requires reactivity control systems to be designed with appropriate limits on the potential amount and rate of reactivity increase to ensure that the effects of postulated reactivity accidents can neither result in damage to the reactor coolant pressure boundary greater than local yielding value , nor sufficiently disturb the core, its support structures, or other reactor pressure vessel internals so 2 of 9

ATTACHMENT 1 Description and Assessment as to impair significantly the capability to cool the core . GDC 28 also requires that these postulated reactivity accidents include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold-water addition .

In Boiling Water Reactors (BWRs), the CRDA is a postulated design basis reactivity insertion accident. Reference 1 provides a method for analyzing the effects of such an event. Reference 1 describes and demonstrates a methodology for assuring compliance with the applicable BWR CRDA licensing acceptance criteria. The CRDA consequences are evaluated based on the fuel enthalpy response during the event. The proposed methodology evaluates these enthalpy responses in relation to the NRG-provided guidance on the fuel cladding failure thresholds and other related failure mechanisms to confirm that no cladding failures occur during the event.

The historical basis for GNF-A analysis methodologies for the CRDA event is the Banked Position Withdrawal Sequence (BPWS), as described in Reference 2. The intent of this approach is to establish a generic control rod withdrawal sequence that would ensure that control rod worths from a dropped rod would, in all cases, be sufficiently limited to meet the legacy NRC CRDA acceptance criteria (a peak enthalpy of no greater than 280 calories (cal)/gram (g), and rarely exceeding 170 cal/g for fuel cladding failure). The control rod worths are minimized through banking of control rod banks at specified positions, and generic analyses are used to demonstrate that the fuel rod enthalpies will be adequately limited by the given control rod worths.

Since Reference 2 was approved by the NRC, additional research in Reactivity Initiated Accidents (RIAs) has identified that the previously mentioned legacy acceptance criteria (e.g .,

280 cal/g peak enthalpy) are not adequate. In particular, two separate failure mechanisms were identified , high temperature cladding failure (HTCF) and pellet-clad mechanical interaction (PCMI). The former mechanism is sensitive to the differential pressure across the cladding, while the latter mechanism is sensitive to the hydrogen concentration within the cladding. This information was used to develop new interim CRDA acceptance criteria, as captured in Appendix B, "Interim Acceptance Criteria and Guidance for the Reactivity Initiated Accidents ," to Chapter 4.2, "Fuel System Design," of the Standard Review Plan (SRP). These criteria were refined using more updated knowledge and published as part of a proposed draft guide, DG-1327, "USNRC Draft Regulatory Guide DG-1327, 'Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Drop Accidents."'

Reference 1 describes a new methodology for analysis of the CRDA event, including evaluation against the more recent acceptance criteria. The approval of Reference 1 allows licensees to utilize this methodology in their licensing basis and in development of their own rod withdrawal sequences that can be demonstrated to comply with the revised CRDA acceptance criteria , in lieu of the BPWS. At the time that the Safety Evaluation was written, the NRC staff had not yet completed the process of issuing DG-1327 as a final regulatory guide. However, the form of the acceptance criteria in DG-1327 is very similar to the interim acceptance criteria currently captured in Appendix B of SRP 4.2. As part of the review of Reference 1, the NRC staff utilized both SRP 4.2 Appendix Band DG-1327, to the extent possible. It is noted that the NRC reviewed and approved NEDE-33885P-A Revision 0, but that the issued NEDE-33885P-A Revision 1 adds the NRC Safety Evaluation and the associated Requests for Addition 3 of 9

ATTACHMENT 1 Description and Assessment Information (RAls) on the Revision 0 document. No physical changes to the plant will be required to implement this proposed amendment.

The methodology for evaluation of a postulated BWR CRDA applies methodologies from three main technology areas: (1) core neutronics modeling, (2) hydraulic modeling, and (3) fuel thermal/mechanical modeling. Core modeling performed using the PANACEA BWR core simulator depends on lattice physics inputs generated using the TGBLA lattice physics program.

Hydraulic modeling of the reactor system, core, and fuel channels is accomplished using TRACG, which relies on the nuclear modeling from TGBLA and PANACEA to calculate the transient power distribution in the core. TRACG calculates the thermal response of the fuel rods using specific inputs from the PRIME fuel thermal/mechanical program to represent the initial conditions of the fuel pellets and rods . This method for performing the CRDA calculations is included in GESTAR-11.

Enthalpy calculations with TRACG form the basis for demonstrating that no cladding failures occur during a postulated CRDA. The calculations are performed consistent with the models described in Reference 1.

Historically, the enthalpy limit for the CRDA is 280 cal/gm. The enthalpy limit is now dependent on two parameters, HTCF and PCMI. These parameters are dependent on the differential pressure across the cladding and the hydrogen pickup fraction for the cladding material, respectively. The figures are provided in Reference 1 and changes the design basis limit for a fission product barrier (DBLFPB) for the CRDA.

During reactor startup and power ascension, there are various reasons to deviate from the analyzed startup sequence. For example, a control rod drive (CRD) may become inoperable for various mechanical or electrical reasons . In such an instance, the usual operator action is to fully insert and disarm such an inoperable control rod. A plant may wish to deliberately deviate from an analyzed startup sequence by leaving a control rod fully inserted. One example is when a control rod adjacent is left inserted for power suppression of a specific fuel bundle. Plant operation may continue provided that the TS Limiting Condition for Operation (LCO) are still met.

LSCS TS currently permit up to eight inoperable control rods. This allowance for eight inoperable control rods has a wider basis than CRDA. It is customary, however, for the CRDA analysis to allow for eight sequence deviations. For example, the previous GE generic CRDA evaluation (BPWS) assumed a maximum of eight inoperable control rods based on plant TS. In principle, more than eight sequence deviations could be allowed, but the number eight is judged to be more than adequate for actual plant operation . BPWS further required inoperable control rods be separated by two or more operable control rods.

CRDA evaluations need to account for out-of-sequence control rods. However, the number and location of these out-of-sequence control rods are not known in advance. The standard process described in NEDE-33885P-A evaluates eight out-of-sequence control rods in any location.

Therefore, the TS 3.1.3 separation criterion is being removed.

The Low Power Set Point (LPSP) has historically been used to define the reactor power level above which there are no constraints on control rod withdrawal order required to protect against 4 of 9

ATTACHMENT 1 Description and Assessment a CRDA. The new method provides a new pressure threshold, defined as the Lower Dome Pressure Set Point (LDPSP), and a revised thermal power for the LPSP, above which there are no constraints on control rod withdrawal order required to protect against a CRDA. The plant operator will constrain the withdrawal order for other reasons, but above the LPSP even the worst possible CRDA does not challenge the fuel cladding failure criteria, as shown in Reference 1.

The LDPSP is set to 300 psig, which is much lower than the operating pressure required to enter MODE 1 operation. For example, in Table 1 of TS 3.3.6.1, "Primary Containment Isolation Instrumentation," Function 1.b, "Main Steam Line Pressure - Low," is applicable in MODE 1 and has an allowable value of~ 826.5 psig. Therefore, the applicability for the CRDA is only in MODE 2.

The Safety Evaluation (SE) for NEDE-33885P-A provides several conditions and limitations that are addressed below:

1) Since the NRC review of Reference 1 depends, in part, on the assumption that the technical models for the PANACEA, TRACG, and PRIME codes have been previously reviewed and approved by the NRC for general neutronics, transient analysis, and fuel thermal performance applications, any limitations and conditions associated with these analysis codes remain applicable. This is expected to be controlled as part of the overall GESTAR-11 methodology as maintained by GNF-A.

EGC will continue to remain consistent with the overall GESTAR-11 methodology maintained by GNF-A.

2) For each application of this methodology to perform licensing basis evaluations of the CRDA event, the maximum drop speed for all control rods shall be confirmed to be bounded by the 3.11 ft/s speed assumed in this LTR or the actual maximum drop speed shall be applied.

LSCS Updated Final Safety Analysis Report (UFSAR) Section 4.2.2.7.2, "Velocity Limiter," describes the use of the control rod blade (CRB) velocity limiter. Each CRB in use at LSCS is equipped with a velocity limiter that limits the drop speed to a maximum of 3.11 ft/s.

3) Section 5.0 of Reference 1 provides a Condition and Limitation on the use of Option 2 for a certain set of circumstances that are proprietary to the methodology.

If used, the Option 2 method for prescribing control rod withdrawal order within a group, as described in Reference 1 Section 4.3.5.1, will include applicable consideration of the potential conditions identified in SE, Section 5.0, Item 2 of Reference 1.

4) When utilizing Option 3 in prescribing the control rod withdrawal order within a group, as described in Section 4.3.5. 1 of this LTR, all control rods within a group are withdrawn to the same intermediate position before any control rod is withdrawn past that position .

LSCS utilizes the Rod Worth Minimizer and plant procedures to control the withdraw of control rods at all power levels. The sequence of control rod withdrawal order is 5 of 9

ATTACHMENT 1 Description and Assessment developed by using procedures that describe the conditions and limitations required to meet the CRDA requirements . The Rod Worth Minimizer enforces the sequence, including rod withdraw order and notch position. Procedures will be modified to capture the limitations associated with Reference 1, but the Rod Worth Minimizer will not require modification and will continue to enforce sequences, as it does currently.

5) If updated models, elements, or codes are used with this methodology as described in Section 7.3 of this LTR, the validation results shall be similar to the results for the specific models, elements, and codes referenced in this LTR.

EGC confirms that any updates to models, elements or codes associated with Reference 1 will be validated to be similar to the results for the specific models, elements and codes referenced in Reference 1.

In summary, EGC requests this amendment to implement the new GNF-A CRDA licensing methodology described in Reference 1. The methodology provides the method for analyzing control rod sequences that are less complicated than those generically evaluated in the current methodology. The method provides LSCS with flexibility in the control rod operations for plant startups while implementing a more restrictive failure threshold consistent with Reference 1.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements The proposed amendment has been evaluated to ensure the applicable regulations and requirements, noted below, continue to be met.

Title 10 of the Code of Federal Regulations (10 CFR) 50.36(c)(2)(ii), paragraph (C), Criterion 3, states that a technical specification limiting condition for operation of a nuclear reactor must be established for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Title 10 of the Code of Federal Regulations (10 CFR) 50.36(c)(3), "Surveillance Requirements ,"

requires that requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

General Design Criterion (GDC) 13, "Instrumentation and Control," of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," addresses the availability of instrumentation to monitor variables and systems over their anticipated ranges to assure adequate safety, and of appropriate controls to maintain these variables and systems within prescribed operating ranges. This regulatory requirement primarily applies to ensuring that the limiting system operating parameters and other controls in place (i.e., rod withdrawal limitations) are sufficient to ensure that the CRDA acceptance criteria are not exceeded. This is satisfied by ensuring that the initial conditions and limitations on rod withdrawal represented in the CRDA analyses are sufficiently representative of the most conservative condition allowed by the aforementioned 6 of 9

ATTACHMENT 1 Description and Assessment controls.

GDC 28, "Reactivity Limits," of 10 CFR Part 50, Appendix A, requires that the effects of postulated reactivity accidents result in neither damage to the reactor coolant pressure boundary greater than limited local yielding nor result in sufficient damage to impair significantly core cooling capacity.

EGC has determined that the proposed amendment remain in conformance with the regulatory requirements above along with the regulatory requirements identified in Reference 1.

4.2 No Significant Hazards Consideration Analysis In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) is submitting a license amendment request for LaSalle County Station, Units 1 and 2 (LSCS) to incorporate the NRC-approved methodology described in Licensing Topical Report (LTR), "GNF CRDA Application Methodology," NEDE-33885P-A, Revision 1. The proposed change revises the LSCS Technical Specifications (TS) to incorporate to this new methodology by modifying LSCS TS Sections 3.1.3, "Control Rod Operability," 3.1.6, "Rod Pattern Control ," and 3.3.2.1, "Control Rod Block Instrumentation," to allow for greater flexibility in rod control operations during various stages of reactor power operation at LSCS.

The proposed amendment will modify the requirements on control rod withdrawal order and conditions to protect against a postulated control rod drop accident (CRDA) during startup and low power conditions . The proposed amendment also eliminates the control rod operability separation criterion while operating in those conditions. The changes are being considered to align the plant startup sequences with the calculated control rod reactivity worths during the control rod withdrawal process. This methodology incorporates the characteristics of advanced fuel products and the latest analytical methods into the design basis for the CRDA to meet the requirements for fuel cladding failure thresholds and allow more flexibility during plant startups.

EGC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50 .92, "Issuance of amendment," as discussed below:

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

Response: No The proposed change revises the LSCS Technical Specifications (TS) to incorporate the methodology in Reference 1 by modifying LSCS TS Sections 3.1.3, "Control Rod Operability," 3.1.6, "Rod Pattern Control," and 3.3.2.1, "Control Rod Block Instrumentation,"

to allow for greater flexibility in rod control operations during various stages of power operation at LSCS.

The methodology incorporates the latest analytical codes from GNF-A to evaluate the Control Rod Drop Accident (CRDA). No physical plant changes are being made, so there is 7 of 9

ATTACHMENT 1 Description and Assessment no change to the probability of the accident. The consequences associated with the CRDA are lowered with this methodology, as it shows that no fuel failures could occur.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No The proposed change revises the LSCS Technical Specifications (TS) to incorporate the methodology in Reference 1 by modifying LSCS TS Sections 3.1.3, "Control Rod Operability," 3.1 .6, "Rod Pattern Control," and 3.3.2 .1, "Control Rod Block Instrumentation,"

to allow for greater flexibility in rod control operations during various stages of power operation at LSCS.

The CRDA is continues to be analyzed as a Design Basis Accident (OBA). The methodology does not require any physical changes to the plant; therefore no new accidents could be introduced.

No credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases are introduced . The proposed changes do not alter assumptions made in the safety analysis.

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

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The proposed change revises the LSCS Technical Specifications (TS) to incorporate to the methodology in Reference 1 by modifying LSCS TS Sections 3.1.3, "Control Rod Operability," 3.1.6, "Rod Pattern Control," and 3.3.2.1, "Control Rod Block Instrumentation,"

to allow for greater flexibility in rod control operations during various stages of reactor power operation at LSCS.

The methodology provides a means to evaluate control rod worths analytically with GNF-A methods that have been previously approved by the NRC. As such, control rod worths will be analyzed explicitly and will be maintained below a fuel failure threshold, even in the accident scenario.

The proposed change does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed change.

8 of 9

ATTACHMENT 1 Description and Assessment Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

4.3 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations ,

and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL EVALUATION The proposed change 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 change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51 .22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

6.0 REFERENCES

1. Licensing Topical Report, "GNF CRDA Application Methodology," NEDE-33885P-A, Revision 1, dated March 2020(ML18059A874)

2. NED0-21231, "Banked Rod Withdrawal Sequence," date January 1977 (ML090771242)

(Non-Public)

3. NUREG-0800, Section 4.2, Revision 3, "Fuel System Design," dated March 2007 (ML070660036)

4. Final Safety Evaluation for Global Nuclear Fuel - Americas LLC (GNF), Licensing Topical Report NEDE-33885P, Revision 0, "GNF Control Rod Drop Accident Application Methodology," dated January 16, 2020 (EPID L-2018-TOP-0006) 9 of 9

ATTACHMENT 2 LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374 Proposed TS Markups

Contro l Rod OPERA BILIT Y 3 .1. 3 AC TIO NS CON DIT ION REQU I RED ACTION COMP LE TION TI ME D. ~IGH 0.1 Rests Fe E8F!l[3~iaAEe 4 R8blFS

~let ar:ir:i~ i EBB~ e liReA 1. i tt:i a Aa~ :~li"eEI F8EI 1

n1ERMAL PG\~ ER r:iesitieA seqbleAEe.

1G?6 RH.

" G-R-

+.1e 1 8F F!l8 Fe iAe[:leFae~e D.~ Rests Fe E8AtF8~ F8EI 4 R8b1FS E8AtF8~ F8EIS A8t i A te GPER,6,ElLE stat bis.

E8F!l[:l~ i a AEe 1iiU1 aAa~:)':!"eEI Fe El r:iesiti eA seqbleAEe a REI A8t ser:iaFateEI ey t 1.1e 8F F!l8 Fe GPER,6,g Li; E8AtF8~

-F-W-5-.--

.f.D . Requ i red Ac ti on and f.D.1 Be i n MO DE 3 . 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months assoc i ated Comp l eti on Ti me of Cond iti on A or, C, 8F D not met.

OR Nine or more co ntrol rods inoperab l e .

LaSal l e 1 and 2 3.1.3 - 3 Amendmen t No. 147/133

Rod Pattern Control 3 .1. 6 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Rod Pattern Control LCO 3.1.6 OP[RABL[ cControl rods sha ll comply with the requirements of the analyzed rod position sequence .

APPLICABILITY: MODE ~ 1 and 2 with THERMAL POWER~ -+/--G-5% RTP and reactor steam dome pressure ~ 300 psig .

ACTIONS CONDITION REQUIRED ACT ION COMPLETION TIME A. One or more QP[RABL[ A. 1 - - - - - - - - NOTE- - - - - - - - -

control rods not in Rod Worth Minimizer comp li ance with the (RWM) may be bypassed analyzed rod position as allowed by sequence. LCO 3.3.2.1, "Control Rod Block In strumentation."

Move associated 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months control rod(s) to correct position.

A.2 Declare 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months associated Fully insert associated control rod(s) inoperable .

(continued)

LaSalle 1 and 2 3.1.6-1 Amendment No. 147/133

Rod Pattern Control 3 .1. 6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Nine or more fully B.l - - - - - - - - NOTE- - - - - - - - -

inserted OPtRAglt RWM may be bypassed control rods not in as allowed by compliance with the LCO 3.3.2.1.

analyzed rod position sequence.

Suspend withdrawal of Immediately control rods.

B.2 Place the reactor 1 hour mode switch in the shutdown position.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify all OPrnAglt control rods comply In accordance with the analyzed rod position sequence. with the Surveillance Frequency Control Program LaSalle 1 and 2 3.1.6-2 Amendment No. 200/187

Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.2.1.2 - - - - - - - - - - - - - - - - - -NOTE- - - - - - - - - - - - - - - - - - -

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after any control rod is withdrawn at

~ -+/--05% RTP and~ 300 psig reactor steam dome pressure in MODE 2.

In accordance Perform CHANNEL FUNCTIONAL TEST. with the Surveillance Frequency Control Program SR 3.3.2.1.3 - - - - - - - - - - - - - - - - - -NOTE- - - - - - - - - - - - - - - - - - -

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after THERMAL POWER is ~ -+/--05% RTP and reactor steam dome pressure is~ 300 psig in MODE -+/--2.

In accordance Perform CHANNEL FUNCTIONAL TEST. with the Surveillance Frequency Control Program SR 3.3.2.1.4 - - - - - - - - - - - - - - - - - -NOTE- - - - - - - - - - - - - - - - - - -

Neutron detectors are excluded.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program (continued)

LaSalle 1 and 2 3.3.2.1-4 Amendment No. 200/187

Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.2.1.5 - - - - - - - - - - - - - - - - - -NOTE- - - - - - - - - - - - - - - - - - -

Neutron detectors are excluded.

Verify the RBM is not bypassed when In accordance THERMAL POWER is~ 30% RTP and a with the peripheral control rod is not selected. Surveillance Frequency Control Program SR 3.3.2.1.6 Verify the RWM is not bypassed when In accordance THERMAL POWER is~ 5-+/--G% RTP. with the Surveillance Frequency Control Program SR 3.3.2.1.7 - - - - - - - - - - - - - - - - - -NOTE- - - - - - - - - - - - - - - - - - -

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after reactor mode switch is in the shutdown position.

Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.2.1.8 Verify control rod sequences input to the Prior to RWM are in conformance with analyzed rod declaring RWM position sequence. OPERABLE following loading of sequence into RWM SR 3.3.2.1.9 Verify the bypassing and position of Prior to and control rods required to be bypassed in during the RWM by a second licensed operator or movement of other qualified member of the technical control rods staff. bypassed in RWM LaSalle 1 and 2 3.3.2.1-5 Amendment No.

200/1@7

Control Rod Block I nstrumentation 3.3.2.1 Table 3 . 3 . 2 . 1-1 (page 1 of ll Control Rod Block Instrumentat i on APPLICABLE MODES OR OTHER SPEC ! FI ED REQUI RED SURV EILLANCE ALLOWABL E FUNCTION CONDITIONS CHANNELS REQUI REMEN TS VALUE

1. Rod Block Monitor a . Upscale (a) SR 3 . 3 . 2 . 1.1 As specified i n SR 3 . 3 . 2 . 1.4 the CO LR SR 3 . 3 . 2 . 1.5

b. In op (a) SR 3 . 3 . 2 . 1.1 NA SR 3 . 3 . 2 . 1.5 c . Downsca l e (a) SR 3 . 3 . 2 . 1.1 2 1. 25% RTP SR 3 . 3 . 2.1.4 SR 3 . 3.2.1.5

2. Rod Worth Minimizer SR 3 . 3 . 2 . 1.2 NA SR 3 . 3.2.1.3 SR 3 . 3 . 2.1.6 SR 3 . 3 . 2 . 1.8 SR 3 . 3 . 2.1.9

3. Reactor Mode Switch-Sh utdown ( c) SR 3 . 3 . 2 . 1.7 NA Posit i on (a) THERMAL POWER 2 30% RTP and no peripheral control rod selected .

(b) With THERMAL POWER$ -+/--05 % RTP and reactor steam dome pressure~ 300 psig .

(c) Reactor mode switch in the shutdown position .

LaSalle 1 and 2 3.3.2.1-7 Amendmen t No. 10/133

ATTACHMENT 3 LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374 Proposed TS Bases Markups (For Information Only)

Control Rod OPERABILITY B 3.1.3 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Control Rod OPERABILITY BASES BACKGROUND Control rods are components of the Control Rod Drive (CRD)

System, which is the primary reactivity contro l system for the reactor. In conjunction with the Reactor Protection System, the CRD System provides the means for the reliable control of reactivity changes to ensure that under conditions of normal operation, including anticipated operational occurrences, specified acceptable fuel design limits are not exceeded. In addition, the control rods provide the capability to hold the reactor core subcritical under all conditions and to limit the potentia l amount and rate of reactivity increase caused by a malfunction in the CRD System. The CRD System is designed to sat i sfy the requirements of GDC 26, GDC 27, GDC 28, and GDC 29, (Ref. 1).

The CRD System consists of 185 locking piston control rod drive mechanisms (CRDMs) and a hydraulic control unit for each drive mechanism. The locking piston type CROM is a double acting hydraulic piston, which uses condensate water as the operating fluid. Accumulators provide additional energy for scram. An index tube and piston, coupled to the control rod, are locked at fixed increments by a coll et mechanism. The col let fingers engage notches i n the index tube to prevent unintentional withdrawal of the control rod, but without restricting insertion.

This Specification, along with LCO 3.1.4, "Control Rod Scram Times," LCO 3.1.5, "Control Rod Scram Accumulators,"

and LCO 3.1.6, "Rod Pattern Control," ensure that the performance of the control rods in the event of a Design Basis Accident (OBA) or transient meets the assumptions used in the safety analyses of References 2, 3, 4, 5, and 6.

APPLICABLE The analytical methods and assumptions used in the SAFETY ANALYSES evaluations involving control rods are presented in References 2, 3, 4, 5, and 6. The control rods provide the primary means for rapid reactivity control (reactor scram),

for maintaining the reactor subcritical, and for limiting the potential effects of reactivity insertion events caused by malfunctions in the CRD System.

LaSalle 1 and 2 B 3.1.3-1 Revision 0

Control Rod OPERABILITY B 3.1.3 (continued)

La Sal le 1 and 2 B 3.1.3-2 Revision 0

Control Rod OPERABILITY B 3.1.3 BASES APPLICABLE The capability of inserting the control rods provides SAFETY ANALYSES assurance that the assumptions for scram react i vity in the (continued) OBA and transient analyses are not violated. Since the SOM ensures the reactor will be subcritical with the highest worth control rod withdrawn (assumed single fa i lure), the additional failure of a second control rod to i nsert could invalidate the demonstrated SOM and potentially limit the ability of the CRO System to hold the reactor subcritical.

If the control rod is stuck at an inserted pos i tion and becomes decoupled from the CRO, a control rod drop accident (CROA) can possibly occur. Therefore, the requirement that all control rods be OPERABLE ensures the CRO System can perform its intended function.

The control rods al so protect the fuel from da mage that results in release of radioactivity. The limits protected are the MCPR Safety Limit (SL) (see Bases for SL 2.1.1, "Reactor Core SLs," and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"), the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLGHR)," and LCO 3.2.3, "LINEAR HEAT GEN ERA TIO N RA TE ( LHG R) " ) , and the fuel des i gn l i mit ( see Bases for LCO 3.1.6, "Rod Pattern Control") during reactivity insertion events.

The negative reactivity insertion (scram) prov i ded by the CRO System provides the analytical basis for determination of plant thermal limits and provides protection against fuel design limits during a CROA. Bases for LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6 discuss in more detai l how the SLs are protected by the CRD System.

Control rod OPERABILITY satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO OPERABILITY of an individual control rod is based on a combination of factors, primarily the scram insertion times, the control rod coupling integrity, and the ability to determine the control rod position. Accumu l ator OPERABILITY is addressed by LCO 3.1.5. The associated scram accumulator status for a control rod only affects the scram insertion times and therefore an inoperable accumulator does not (continued)

LaSalle 1 and 2 B 3.1.3-3 Revision 0

Control Rod OPERABILITY B 3.1.3 BASES LCO immediately require declaring a control rod inoperable.

(continued) Although not all control rods are required to be OPERABLE to satisfy the intended reactivity control requirements, strict control over the number and distribution of inoperable control rods is required to satisfy the assumptions of the DBA and transient analyses.

OPERABILITY requirements for control rods also includes correct assembly of the CRD housing supports.

APPLICABILITY In MODES 1 and 2, the control rods are assumed to function during a DBA or transient and are therefore required to be OPERABLE in these MODES. In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is appl i ed. This provides adequate requirements for control rod OPERABILITY during these conditions. Control rod requirements in MODE 5 are located in LCO 3.9.5, "Control Rod OPERABILITY-Refueling."

ACTIONS The ACTIONS Table is modified by a Note indicating that a separate Condition entry is allowed for each control rod.

This is acceptable, since the Required Actions for each Condition provide appropriate compensatory act i ons for each inoperable control rod. Complying with the Required Actions may allow for continued operation, and subsequent inoperable control rods are governed by subsequent Condition entry and application of associated Required Actions.

A.1. A.2. A.3. and A.4 A control rod is considered stuck if it will not insert by either CRD drive water or scram pressure. The Required Actions are modified by a Note that allows the Rod Worth Minimizer (RWM) to be bypassed if required to allow continued operation. LCO 3.3.2.1, "Control Rod Block Instrumentation," provides additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis. With one withdrawn control rod stuck, the local scram reactivity rate assumptions may not be met if the stuck (continued)

LaSalle 1 and 2 B 3.1.3-4 Revision 0

Control Rod OPERABILITY B 3.1.3 BASES ACTIONS A.l A.2 A.3 and A.4 (continued) control rod separation criteria are not met. Therefore, a verification that the separation criteria are met must be performed immediately. The separation criteria are not met if: a) the stuck control rod occupies a location adjacent to two "slow" control rods, b) the stuck control rod occupies a location adjacent to one "slow" control rod, and the one "slow" control rod is also adjacent to another "slow" control rod, or c) if the stuck control rod occupies a location adjacent to one "slow" control rod when there is another pair of "slow" control rods elsewhere i n the core adjacent to one another. The description of "slow" control rods is provided in LCO 3.1.4, "Control Rod Scram Times."

In addition, the associated control rod drive must be disarmed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The allowed Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is acceptable, considering the reactor can still be shut down, assuming no additional control rods fail to insert, and provides a reasonable amount of ti me to perform the Required Action in an orderly manner. The control rod must be isolated from both scram and normal insert and withdraw pressure. Isolating the control rod from scram and normal insert and withdraw pressure prevents damage to the CROM or reactor internals. The control rod isolation method should also ensure cooling water to the CRD is maintained.

Monitoring of the insertion capability for each withdrawn control rod must also be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from discovery of Condition A concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM.

SR 3.1.3.3 performs periodic tests of the control rod insertion capability of withdrawn control rods. Testing each withdrawn control rod ensures that a generic problem does not exist. This Completion Time also allows for an exception to the normal "time zero" for beginn i ng the allowed outage time "clock." The Required Act i on A.3 Completion Time only begins upon discovery of Condition A concurrent with THERMAL POWER greater than the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of rod pattern control (LCO 3.1.6) and the RWM (LCO 3.3.2.1). The al l owed Completion Time provides a reasonable time to test the control rods, considering the potential for a need to reduce power to perform the tests.

(continued)

LaSalle 1 and 2 B 3.1.3-5 Revision 42

Control Rod OPERABILITY B 3.1.3 BASES ACTIONS A.l A.2 A.3 and A.4 (continued)

To allow continued operation with a withdrawn control rod stuck, an evaluation of adequate SOM is also required within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Should a OBA or transient require a shutdown, to preserve the single failure criterion an additional control rod would have to be assumed to have failed to insert when required. Therefore, the original SOM demonstration may not be valid. The SOM must therefore be evaluated (by measurement or analysis) with the stuck control rod at its stuck position and the highest worth OPERABLE control rod assumed to be fully withdrawn.

The allowed Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to verify SOM is adequate, considering that with a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity. Failure to reach MODE 4 is only likely if an additional control rod adjacent to the stuck control rod also fails to insert during a requ i red scram.

Even with the postulated additional single fai l ure of an adjacent control rod to insert, sufficient reactivity control remains to reach MODE 3 conditions.

With two or more withdrawn control rods stuck, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The occurrence of more than one control rod stuck at a withdrawn position increases the probability that the reactor cannot be shut down if required. Insertion of all insertable control rods eliminates the possibility of an additional fa i lure of a control rod to insert. The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

C.l and C.2 With one or more control rods inoperable for reasons other than being stuck in the withdrawn position, operation may continue, provided the control rods are fully i nserted (continued)

LaSalle 1 and 2 B 3.1.3-6 Revision 0

Control Rod OPERABILITY B 3.1.3 BASES ACTIONS C.1 and C.2 (continued) within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and disarmed (electrically or hydraulically) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected.

The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves. Electrica l ly, the control rods can be disarmed by disconnecting power from all four directional control valve solenoids. Required Action C.l is modified by a Note that allows the RWM to be bypassed if required to allow insertion of the inoperable control rods and continued operation. LCO 3.3.2.1 provides additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis.

The allowed Completion Times are reasonable, considering the small number of allowed inoperable control rods, and provide time to insert and disarm the control rods in an orderly manner and without challenging plant systems.

D.l and D.2 Obit of seqblence control rods may increase the potential reactivity 1rnrth of a dropped control rod dblrin§ a CRDA.

At < 10% RTP, the analyzed rod position seqblence analysis (Refs. 7 and g) reqblires inserted control rods not in co FR pl i an c e I<' i U1 the an a l y zed rod po s i ti on seq bl enc e to be separated tiy at least t1rn OPER/\i;iLE control rods in all directions, inclbldin§ the dia§Jonal (i.e., all other control rods in a five tiy five array centered on the inoperatile control rod are OPERMlLE). Hlerefore, if tl.*o or more inoperable control rods are not in compliance *..*ith the analyzed rod position seqblence and not separated tiy at least t*u10 OPERA!;iLE control rods in all directions, action FR bl st tie ta ken to restore co FR pl i an c e *,1 i th the anal y zed rod position seqblence or restore the control rods to OPERABLE statbls. A Note has tieen added to the Condition to clarify that the Condi ti on is not appl i catil e *,1hen > 10% RTP si nee the analyzed rod position seqblence is not reqbl i red to tie follm1ed binder these conditions, as described i n the Bases for LCO 2.1.§. The allo*,1ed Completion Time of 4 hoblrs is acceptatile, considerin§ the 1011 protiatiility of a CRDA ocCblrrin§.

LaSalle 1 and 2 B 3.1.3-7 Revision 0

Control Rod OPERABILITY B 3.1.3 (continued)

LaSalle 1 and 2 B 3.1.3-8 Revision 0

Control Rod OPERABILITY B 3.1.3 BASES ACTIONS (continued)

If any Required Action and associated Completion Time of Condition A, C, or D or C are not met or nine or more inoperable control rods exist, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This ensures all insertable control rods are inserted and places the reactor in a condition that does not require the active function (i.e., scram) of the control rods. The number of control rods perm i tted to be inoperable when operating above +.G-5% RTP or 300 psig reactor steam dome pressure (i.e., no CRDA considerations as described in the Bases for LCO 3.1.6 ) could be more than the value specified, but the occurrence of a large number of inoperable control rods could be indicative of a generic problem, and investigation and resolution of the potential problem should be undertaken. The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.3.1 REQUIREMENTS The position of each control rod must be deter mined, to ensure adequate information on control rod pos i tion is available to the operator for determining control rod OPERABILITY and controlling rod patterns. Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods by single notch movement to a posit i on wi th an 0 PERA BL E i ndi cat or ( f ul l - i n , f ul l -

out, or numeric indicator) and then returning the control rods by single notch movement to their origina l position, or by the use of other appropriate methods. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(continued)

LaSalle 1 and 2 B 3.1.3-9 Revision 51

Control Rod OPERABILITY B 3.1.3 BASES SURVEILLANCE SR 3.1.3.2 REQUIREMENTS (continued) DELETED SR 3.1.3.3 Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves.

The control rod may then be returned to its or i ginal position. This ensures the control rod is not stuck and is free to insert on a scram signal. This Surveillances is not required when THERMAL POWER is less than or equal to the actual LPSP of the RWM since the notch insertions may not be compatible with the requirements of the analyzed rod position sequence (LCO 3.1.6) and the RWM (LCO 3.3.2.1).

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

This SR is modified by a Note that allows 31 days, after withdrawal of the control rod and increasing power to above the LPSP, to perform the Surveillance. This acknowledges that the control rod must be first withdrawn and THERMAL POWER must be increased to above the LPSP before performance of the Surveillance, and therefore, the Note avoids potential conflicts with SR 3.0.3 and SR 3.0.4.

(continued)

LaSalle 1 and 2 B 3.1.3-10 Revision 51

Control Rod OPERABILITY B 3.1.3 BASES SURVEILLANCE SR 3.1.3.4 REQUIREMENTS (continued) Verifying the scram time for each control rod to notch position 05 is~ 7 seconds provides reasonable assurance that the control rod will insert when required during a DBA or transient, thereby completing its shutdown function.

This SR is performed in conjunction with the control rod scram time testing of SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3. 3 .1.1, "Reactor Protection System (RPS)

Instrumentation," and the functional testing of SDV vent and drain valves in LCO 3.1.8, "Scram Discharge Volume (SDV) Vent and Drain Valves," overlap this Surveillance to provide complete testing of the assumed safety function.

The associated Frequencies are acceptable, considering the more frequent testing performed to demonstrate other aspects of control rod OPERABILITY and operating experience, which shows scram times do not significantly change over an operating cycle.

SR 3.1.3.5 Coupling verification is performed to ensure the control rod is connected to the CRDM and will perform i ts intended function when necessary. The Surveillance requires verifying that a control rod does not go to the withdrawn overtravel position when it is fully withdrawn. The overtravel position feature provides a positive check on the coupling integrity, since only an uncoupled CRD can reach the overtravel position. The verification is required to be performed anytime a control rod is withdrawn to the "full out" position (notch position 48) or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This includes control rods inserted one notch and then returned to the "full out" position during the performance of SR 3.1.3.2. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved and operating exper i ence related to uncoupling events.

The performance of SR 3.1.3.5 is an assumption of Reference 8.

(continued)

LaSalle 1 and 2 B 3.1.3-11 Revision 0

Control Rod OPERABILITY B 3.1.3 BASES (continued)

REFERENCES 1. 10 CFR 50, Appendix A, GDC 26, GDC 27, GDC 28, and GDC 29.

2. UFSAR, Section 4.3.2.5.

3. UFSAR, Section 4.6.1.1.2.

4. UFSAR, Section 5.2.2.2.

5. UFSAR, Section 15.4.

6. UFSAR, Section 15.4.9.

7. ~IEDO 21231, "Banked Position '.~ithdra\1al ?;equence,"

?;ection 7.2, January 1977.

7g . NFSR-0091, Commonwealth Edison Topical Report, Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods, (as specified in Technical Specification

5. 6. 5).

8. NEDE 33885P-A, "GNF CRDA Application Methodology,"

Revision 1, March 2020 LaSalle 1 and 2 B 3.1.3-12 Revision 0

Rod Pattern Control B 3.1.6 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Rod Pattern Control BASES BACKGROUND Control rod patterns during startup conditions are controlled by the operator and the Rod Worth Mi nimizer (RWM) (LCO 3.3.2.1, "Control Rod Block Instrumentation"),

so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods inserted to 5-+/--G-% RTP or 300 psig reactor steam dome pressure . The sequences effectively limit the potential amount of reactivity addition that could occur in the event of a control rod drop accident (CRDA).

This Specification assures that the control rod patterns are consistent with the assumptions of the CRDA analyses of References 1, 2, and 3.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the CRDA are summarized in References 1, 2, 3, 4, and 5.

CRDA analyses assume that the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analysis.

The RWM (LCO 3.3.2.1) provides backup to operator control of the withdrawal sequences to ensure that the initial conditions of the CRDA analysis are not violated.

Prevention or mitigation of positive reactivity insertion events is necessary to limit the energy deposition in the fuel, thereby preventing significant fuel damage, which could result in undue release of radioactivity. ~ince the fa i l bl re cons e q bl enc es for U0~ have been sh Olm to be i Asi §)Ai fi cant bel O'.J fblel eAer§ly deposi ti OAS of JOO cal /§JFR (Ref. 6), the fblel desi§ln liFRit of 280 cal/§JFR provides a FRar§liA of safety froFR si§JAificaAt core daFRa§le, 11hich h'Oblld resbllt iA release of radioactivity (Ref. 7). GeAeric evalblations (Refs. 8 and 9) of a desi§JA basis CRD/\ (i.e., a CRDA resblltiA§l iA a peak fblel eAer§ly depositioA of 280 cal/§JFR) have sh01m that if the peak fblel eAthalpy reFRains beloi,1 280 cal/§JFR, theA the FRa>dFRblFR reactor pressblre 11i 11 be less thaA the reqbli red Control rod patterns analyzed in the cycle-specific analyses are developed in accordance with Reference 12. The Technical Specifications refer to these patterns as the "analyzed rod position sequence(s)."

Per Ref. 12, use of the analyzed rod position sequence LaSalle 1 and 2 B 3.1.6-1 Revision 0

Rod Pattern Control B 3.1.6 ensures ASME Code limits (Ref. 10) and the calculated offsite doses will be well within the (continued)

LaSalle 1 and 2 B 3.1.6-2 Revision 0

Rod Pattern Control B 3.1.6 BASES APPLICABLE required limits (Ref. 11). Cycle specific CRDA analyses F-C-SAF ET Y ANALYSES perforrned that assurne eight inoperable control rods .1ith at 1 (continued) least t\10 cell separation and confirrn the fuel energy deposition is less that 280 cal/grn.

Control rod patterns analyzed in the cycle specific analyses follo .1 predeterrnined seEJblencing rules (analyzed 1

rod position sec:iuence). The analyzed rod position seEJblence is applicable frorn the condition of all control rods fbll ly inserted to 10~6 RTP (Ref. 5). The control rods are reEJbli red to be rnoved in groups, li'i th all control rods assigned to a specific groblp rec:iuired to be .1ithin 1

specified banked positions (e.g., bet .1een notches 08 1

and 12). The banked positions are defined to rninirnize the rnaxiFRblFR incrernental control rod 1rnrths \1itho1:Jt being overly restrictive during norrnal plant operation. Cycle specific analyses ensblre that the 280 cal/grn fblel design lirnit .1ill 1 not be violated dblring a CRD/', binder *orst case scenarios.

1 The cycle specific analyses (Refs. 1, 2, 3, 4, and 5) also eval1:Jate the effect of fbllly inserted, inoperable control rods not in cornpl i ance li'i th the sec:iuence, to al l O'd a lirnited nblFRber (i.e., eight) and distribution of fbllly inserted, inoperable control rods. ~pecific analyses rnay also be perforrned for atypical operating conditions (e.g.,

fuel leaker suppression).

When performing a shutdown of the plant, an optional rod position sequence (Ref. ~ 12 ) may be used prov i ded that all withdrawn control rods have been confirmed to be coupled.

The rods may be inserted without the need to stop at intermediate positions since the possibility of a CRDA is eliminated by the confirmation that withdrawn control rods are coupled. When using the optional (Ref. 1 2 ~ ) control rod sequence for shutdown, the rod worth minimizer may be reprogrammed to enforce the requirements of the improved control rod insertion process.

In order to use the Reference -l-J-12 shutdown process, an extra check is required in order to consider a control rod to be "confirmed" to be coupled. This extra check ensures that no single operator error can result in an incorrect coupling check. For purposes of this shutdown process, the method for confirming that control rods are coupled varies depending on the position of the control rod in the core.

Details on this coupling confirmation requirement are provided in Reference ~ 12 @ .

LaSalle 1 and 2 B 3.1.6-3 Revision 17

Rod Pattern Control B 3.1.6 (continued)

LaSalle 1 and 2 B 3.1.6-4 Revision 17

Rod Pattern Control B 3.1.6 BASES APPLICABLE The plant is in compliance with the rod position sequence SAFETY ANALYSES required by this LCD when the requirements of Reference 13 (continued) a re met.

Rod pattern control satisfies the requirements of Criterion 3 of 10 CFR 50.36(c)(2)(ii ).

LCO Compliance with the prescribed control rod sequences minimizes the potential consequences of a CRDA by limiting the initial conditions to those consistent with the analyzed rod position sequence. This LCD B-A-f-:y-applies to OPERABLE all control rods. P"or inoperable control rods required to be inserted, separate requirements are specified in LCO 2.1.2, "Control Rod OPER/\i;;ILITY,"

consistent 11i th the all 0 .:ances for inoperable control rods 1

in the analyzed rod position sequence.

APPLICABILITY In MODE S 1 and 2, when THERMAL POWER is ~ 5-+/--G-% RTP and reactor steam dome pressure is s 300 psig , the CRDA is a Design Basis Accident (OBA) and, therefore, compliance with the assumptions of the safety analysis is requ i red. When THERMAL POWER is > 5-+/--G-% RTP or reactor steam dome pressure is> 300 psig , there is no credible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel design limit fuel cladding failure criteria during a CRDA (Ref. 4 and 5). In MODES 3 and 4, the reactor is shutdown and the control rods are not able to be withdrawn since the reactor mode sw i tch is in shutdown and a control rod block is applied, therefore a CRDA is not postulated to occur. In MODE 5, since the reactor is shut down and only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SOM ensures that the consequences of a CRDA are acceptable, since the reactor will remain subcritical with a single control rod withdrawn. Before entering MODE 1, the reactor has completed heat up and pressurization.

Reactor steam dome pressure is therefore above 300 psig, and so constraints on the control rod pattern due to CRDA are not required in MODE 1.

ACTIONS A.l and A.2 With one or more OPERABLE control rods not in compliance with the prescribed control rod sequence, action may be LaSalle 1 and 2 B 3.1.6-5 Revision 17

Rod Pattern Control B 3.1.6 taken to either correct the control rod pattern or declare t-A-e-fully insert the associated control rods inoperable within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Noncompliance with the prescribed sequence ma y be the result of "double notching," drifting from a control rod drive (continued)

BASES ACTIONS A.1 and A.2 (continued) cooling water transient, leaking scram valves, or a power reduction to ~ -+/--G-5% RTP ands 300 psig reactor steam dome pressure before establishing the correct control rod pattern. The number of OPE:RABLt control rods not in compliance with the prescribed sequence is limited to eight to prevent the operator from attempting to correct a control rod pattern that significantly deviates from the prescribed sequence.

Required Action A.1 is modified by a Note, which allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position. LCO 3.3.2.1 requires verification of control rod movement by a second licensed operator (Reactor Operator or Senior Reactor Operator) or by a task qualified member of the technical staff (e.g., a shift technical advisor or reactor engineer). This helps to ensure that the control rods wi 11 be moved to the correct position. A control rod not in compliance with the prescribed sequence is not necessarily considered inoperable except as req~ired by Req~ired Action A.2 . The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonab l e, considering the restrictions on the number of allowed out of sequence control rods and the low probability of a CRDA occurring during the time the control rods are out of sequence.

8.1 and 8.2 If nine or more OPERABLt control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence. Control rod withdrawal should be suspended immediately to prevent the potential for further deviation from the prescribed sequence. Control rod insertion to correct control rods withdrawn beyond their allowed position is allowed since, in general, insertion of control rods has less impact on control rod worth than withdrawals have. Required Action B.1 is modified by a LaSalle 1 and 2 B 3.1.6-6 Revision 17

Rod Pattern Control B 3.1.6 Note that allows the RWM to be bypassed to al lo w the affected contro l rods to be returned to their correct position. LCO 3.3.2.1 requires verification of contro l rod movement by a second li censed operator (Reactor Operator or Senior Reactor Operator) or by a task qualified member of the technical staff (e.g., a sh ift technical advisor or reactor engineer).

(continued)

LaSalle 1 and 2 B 3.1.6-7 Revision 17

Rod Pattern Control B 3.1.6 BASES ACTIONS B.l and B.2 (continued)

With nine or more OPERAgLE control rods not in compliance with analyzed rod position sequence, the reactor mode switch must be placed in the shutdown position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

With the reactor mode switch in shutdown, the reactor is shut down, and therefore does not meet the app l icability requirements of this LCO. The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a CRDA occurring with the control rods out of sequence.

SURVEILLANCE SR 3.1.6.1 REQUIREMENTS The control rod pattern is periodically verified to be in compliance with the analyzed rod position sequence, ensuring the assumptions of the CRDA analyses are met. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The RWM provides control rod blocks to enforce the required control rod sequence and is required to be OPERABLE when operating at~ +G-5% RTP ands 300 psig reactor steam dome pressure .

REFERENCES 1. UFSAR, Section 15.4.10.

2. XN-NF-80-19(P)(A), Volume 1, Supplement 2, Section 7.1, Exxon Nuclear Methodology for Boiling Water Reactor-Neutronics Methods for Design and Analysis, (as specified i n Technical Specification 5.6.5).

3. NEDE-24011-P-A, "GE Standard Application for Reactor Fuel," (as specified in Technical Specification

5. 6. 5).

4. Letter from T.A. Pickens (BWROG) to G.C. Lainas (NRC),

"Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A," BWROG-8644, August 15, 1986.

5. NFSR-0091, Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods, Commonwealth Edison Topical Report, (as specified in Technical Specification 5.6.5).

(continued)

LaSalle 1 and 2 B 3.1.6-8 Revision 51

Rod Pattern Control B 3.1.6 BASES REFERENCES (continued)

6. NUREG-0979, "NRC Safety Evaluation Report for GESSAR II BWR/6 Nuclear Island Design, Docket No. 50-447," Section 4.2.1.3.2, April 1983.

7. NUREG-0800, "Standard Review Plan," Section 15.4.9, "Radiological Consequences of Control Rod Drop Accident CBWR)," Revision 2, July 1981.

8. NED0-21778-A, "Transient Pressure Rises Affected Fracture Toughness Requirements for Boiling Water Reactors," December 1978.

9. NED0-10527, "Rod Drop Accident Analysis for Large BWRs," (including Supplements 1 and 2), March 1972.

10. ASME, Boiler and Pressure Vessel Code.

11. 10 CFR 100.11, "Determination of Exclusion Area Low Population Zone and Popu l ation Center Distance."

12. MEDO 21231, "Banked Position '.~ithdra*.1al Sequence,"

January 1977. NEDE-33885P-A, "GNF CRDA Application Methodology," Revision 1, March 2020

13. MEDO 33091 A, Revision 2, "Improved BP 1.~S Control Rod Insertion Process," July 200~.

LaSalle 1 and 2 B 3.1.6-9 Revision 17

Control Rod Block Instrumentation B 3.3.2.1 B 3.3 INSTRUMENTATION B 3.3.2.1 Control Rod Block Instrumentation BASES BACKGROUND Control rods provide the primary means for control of reactivity changes. Control rod block instrumentation includes channel sensors, logic circuitry, switches, and relays that are designed to ensure that specif i ed fuel design limits are not exceeded for postulated transients and accidents. During high power operation, the rod block monitor (RBM) provides protection for control rod withdrawal error events. During low power operations, control rod blocks from the rod worth minimizer (RWM) enforce specific control rod sequences designed to mitigate the consequences of the control rod drop accident (CRDA).

During shutdown conditions, control rod blocks from the Reactor Mode Switch-Shutdown Position Function ensure that all control rods remain inserted to prevent inadvertent criticalities.

The purpose of the RBM is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint during control rod manipulations (Ref. 1). It is assumed to function to block further control rod withdrawal to preclude a MCPR Safety Limit (SL) violation. The RBM supplies a trip signal to the Rod Control Management System (RCMS) to appropriately inhibit control rod withdrawal during power operation above the 30% RATED THERMAL POWER setpoint when a non peripheral control rod is selected.

The RBM has two channels, either of which can i nitiate a control rod block when the channel output exceeds the control rod block setpoint. Each RBM inputs into both RCMS controllers. The RBM channel signal is generated by averaging a set of local power range monitor (LPRM) signals. One RBM channel averages the signals from LPRM detectors at the A and C positions in the assigned LPRM assemblies. The second RBM channel averages the signals from the LPRM detectors at the B and D positions.

Assignment of LPRM assemblies to be used in RBM averaging is controlled by the selection of control rods. With no control rod selected, the RBM output is set to zero.

However, when a control rod is selected, the gain of each (continued)

LaSalle 1 and 2 B 3.3.2.1-1 Revision 44

Control Rod Block Instrumentation B 3.3.2.1 BASES BACKGROUND RBM channel output is normalized to an assigned average (continued) power range monitor (APRM) channel. The assigned APRM channel is on the same RPS trip system as the RBM channel.

The gain setting is held constant during the movement of that particular control rod to provide an indication of the change in the relative local power lev el. If the APRM used to normalize the RBM reading is indicating< 30% or a peripheral control rod is selected, the RBM is zeroed and the RBM is bypassed (Refs. 1 and 2).

If any LPRM detector assigned to an RBM is bypassed, the computed average signal is adjusted automatica l ly to compensate for the number of LPRM signals. The minimum number of LPRM inputs required for each RBM channel to prevent an instrument inoperative alarm is four when using four LPRM assemblies, three when using three LPRM assemblies, and two when using two LPRM assemb l ies. If the normalizing APRM channel is bypassed, a second APRM channel automatically provides the normalizing signal (Refs. 1 and

2) .

In addition, to preclude rod movement with an i noperable RBM, a downscale trip and an inoperable trip are provided.

The purpose of the RWM is to control rod patterns during startup and shutdown, such that only specified control rod sequences and relative positions are allowed over the operating range from all control rods inserted to 5-+/--G-% RTP or 300 psig reactor steam dome pressure . The sequences effectively limit the potential amount and rate of reactivity increase during a CRDA. Prescribed control rod sequences are stored in the RWM, which will in i tiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances from the stored sequence. The RWM determines the actual sequence based on position indication for each control rod. The RWM also uses steam flow signals to determine when the reactor power is above the preset power level at which the RWM is automatically bypassed. The RWM programming is part of the sequence enforcement logic within each of the two RCMS controllers, and so normally operates with two channels. A rod block generated within either channel will prevent control rod withdrawal.

(continued)

LaSalle 1 and 2 B 3.3.2.1-2 Revision 44

Control Rod Block Instrumentation B 3.3.2.1 BASES BACKGROUND With the reactor mode switch in the shutdown position, a (continued) control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained. This Function prevents inadvertent criticality as the result of a control rod withdrawal during MODE 3 or 4, or during MODE 5 when the reactor mode switch is required to be in the shutdown position. The reactor mode switch has two channels, each inputting into a separate rod block circuit.

Each reactor mode switch channel has contacts permitting control rod withdrawal in the reactor mode switch positions of run, startup, and refuel interlocked with other plant conditions. With the reactor mode switch in shutdown, the RCMS circuits do not receive a permissive for control rod withdrawal. A rod block in either RCMS circuit will provide a control rod block to all control rods.

APPLICABLE 1. Rod Block Monitor SAFETY ANALYSES, LCO, and The RBM is designed to prevent violation of the MCPR APPLICABILITY SL and the cladding 1% plastic strain fuel design limit that may result from a single control rod withdrawal error (RWE) event. The analytical methods and assumptions used in evaluating the RWE event are summarized in Reference 4.

The cycle-specific analysis considers the continuous withdrawal of the maximum worth control rod at its maximum drive speed from the reactor, which is operating at rated power with a control rod pattern that results i n the core being placed on thermal design limits. The condition is analyzed to ensure that the results obtained are conservative; the approach also serves to demonstrate the function of the RBM.

The RBM Function satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Two channels of the RBM are required to be OPERABLE, with their setpoints within the appropriate Allowab l e Values in the CORE OPERATING LIMITS REPORT to ensure that no single instrument failure can preclude a rod block from this Function. The actual setpoints are calibrated consistent with applicable setpoint methodology.

Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure (continued)

LaSalle 1 and 2 B 3.3.2.1-3 Revision 44

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE that the setpoints do not exceed the Allowable Values SAFETY ANALYSES, between successive CHANNEL CALIBRATIONS. Operation with a LCO, and trip setpoint less conservative than the nominal trip APPLICABILITY setpoint, but within its Allowable Value, is acceptable.

(continued) Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor power), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for the calibration based errors.

These calibration based errors are limited to reference accuracy, instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Al l owable Values determined in this manner provide adequate protection because instrument uncertainties, process effects, calibration tolerances, instrument dr i ft, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation.

The RBM is assumed to mitigate the consequences of an RWE event when operating ~ 30% RTP and a non-peripheral control rod is selected. Below this power level or if a peripheral control rod is selected, the consequences of an RWE event will not exceed the MCPR SL and, therefore, the RBM is not required to be OPERABLE (Ref. 4).

2. Rod Worth Minimizer The RWM enforces the analyzed rod position sequence to ensure that the initial conditions of the CRDA analysis are not violated. The analytical methods and assumptions used in evaluating the CRDA are summarized in References 5, 6, and 7. The analyzed rod position sequence requires that control rods be FRoved in §lroups, ',1ith all control rods assi§lned to a specific §lroup required to be ',1ithin specified banked positions. Requirements that the control rod sequence is in compliance with the analyzed rod LaSalle 1 and 2 B 3.3.2.1-4 Revision 39

Control Rod Block Instrumentation B 3.3.2.1 position sequence are specified in LCO 3.1.6, "Rod Pattern Control."

(continued)

LaSalle 1 and 2 B 3.3.2.1-5 Revision 39

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 2. Rod Worth Minimizer (continued)

SAFETY ANALYSES, LCO, and When performing a shutdown of the plant, an optional control APPLICABILITY rod sequence (Ref. 9) may be used if the coupling of each withdrawn control rod has been confirmed. The rods may be inserted without the need to stop at intermediate positions. When using the Reference 9 control rod insertion sequence for shutdown, the rod worth minimizer may be reprogrammed to enforce the requirements of the improved control rod insertion process.

The RWM Function satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Since the RWM is a system designed to act as a backup to operator control of the rod sequences, only one channel of the RWM is required to be OPERABLE (Ref. 7). The RWM function is included with the sequence enforce ment logic in each of the two RCMS controllers, and so norma l ly operates with two channels. Special circumstances prov i ded for in the Required Action of LCO 3.1.3, "Control Rod OPERABILITY," and LCO 3.1.6 may necessitate bypassing the RWM to allow continued operation with inoperab l e fully inserted, out-of-sequence control rods, or to allow correction of a control rod pattern not in compliance with the analyzed rod position sequence. The RWM may be bypassed as required by these conditions, but then it must be considered inoperable and the Required Actions of this LCO followed.

Compliance with the analyzed rod position sequence, and therefore OPERABILITY of the RWM, is required i n MODE ~

-a-R-El---2 when THERMAL POWER is~ -+/--G-5% RTP and reactor steam dome pressure~ 300 psig . When THERMAL POWER is> -+/--G-5% RTP or reactor steam dome pressure> 300 psig , there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal/§m f~el desi§n limit fuel cladding failure criteria during a CRDA (Refs. 6 and 7). In MODES 3 and 4, all control rods are required to be inserted into the core; therefore, a CRDA cannot occur. In MODE 5, since only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SOM ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.

Before entering MODE 1, the reactor has completed heat up and pressurization. Reactor steam dome pressure is LaSalle 1 and 2 B 3.3.2.1-6 Revision 51

Control Rod Block Instrumentation B 3.3.2.1 therefore above 300 psig, and so constraints on the control rod pattern due to CRDA are not required in MODE 1.

(continued)

LaSalle 1 and 2 B 3.3.2.1-7 Revision 51

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 3. Reactor Mode Switch-Shutdown Position SAFETY ANALYSES, LCO, and During MODES 3 and 4, and during MODE 5 when the reactor APPLICABILITY mode switch is in the shutdown position, the core is assumed (continued) to be subcritical; therefore, no positive reactivity insertion events are analyzed. The Reactor Mode Switch-Shutdown Position control rod withdrawa l bloc k ensures that the reactor remains subcritical by bloc king control rod withdrawal, thereby preserving the assumptions of the safety analysis.

The Reactor Mode Switch-Shutdown Position Function satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii ).

Two channels are required to be OPERABLE to ensure that no single channel failure will preclude a rod block when required. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on reactor mode switch position.

During shutdown conditions (MODES 3 and 4, and MODE 5 when the reactor mode switch is in the shutdown pos i tion), no positive reactivity insertion events are analyzed because assumptions are that control rod withdrawal blocks are provided to prevent criticality. Therefore, when the reactor mode switch is in the shutdown position, the control rod wi th draw al bl ock i s re qui red to be 0 PERA BL E.

During MODE 5 with the reactor mode switch in the refueling position, the refuel position one-rod-out interlock (LCO 3.9.2, "Refuel Position One-Rod-Out Inter l ock")

provides the required control rod withdrawal bl ocks.

ACTIONS With one RBM channel inoperable, the rema1n1ng OPERABLE channel is adequate to perform the control rod block function; however, overall reliability is reduced because a single failure in the remaining OPERABLE channel can result in no control rod block capability for the RBM. For this reason, Required Action A.1 requires restoration of the inoperable channel to OPERABLE status. The Co mpletion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on the low probability of an event occurring coincident with a failure in the remaining OPERABLE channel.

LaSalle 1 and 2 B 3.3.2.1-8 Revision 39

Control Rod Block Instrumentation B 3.3.2.1 (continued)

LaSalle 1 and 2 B 3.3.2.1-9 Revision 39

Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS (continued)

If Required Action A.1 is not met and the associated Completion Time has expired, the inoperable channel must be placed in trip within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . If both RBM channels are inoperable, the RBM is not capable of performing its intended function; thus, one channel must also be placed in trip. This initiates a control rod withdrawal block, thereby ensuring that the RBM function is met.

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

C.l. C.2.1.1. C.2.1.2 and C.2.2 With the RWM inoperable during a reactor startup, the operator is still capable of enforcing the prescribed control rod sequence. However, the overall re l iability is reduced because a single operator error can result in violating the control rod sequence. Therefore, control rod movement must be immediately suspended except by scram.

Alternatively, startup may continue if at least 12 control rods have already been withdrawn, or a reactor startup with an inoperable RWM during withdrawal of one or more of the first 12 control rods was not performed in the last 12 months. These requirements minimize the number of reactor startups initiated with the RWM inoperable.

Required Actions C.2.1.1 and C.2.1.2 require verification of these conditions by review of plant logs and control room indications. Once Required Action C.2.1.1 or C.2.1.2 is satisfactorily completed, control rod withdrawal may proceed in accordance with the restrictions imposed by Required Action C.2.2. Required Action C.2.2 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other task qualified member of the technical staff (e.g., shift technical advisor or reactor engineer).

The RWM may be bypassed under these conditions to allow continued operations. In addition, Required Actions of LaSalle 1 and 2 B 3.3.2.1-10 Revision 39

Control Rod Block I nstrumentat io n B 3.3.2.1 LCO 3. 1. 3 and LCO 3 .1. 6 may require bypassing the RWM, (continued)

LaSalle 1 and 2 B 3.3.2.1- 11 Revision 39

Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS C.1. C.2.1.1. C.2.1.2. and C.2.2 (continued) during which time the RWM must be considered inoperable with Condition C entered and its Required Actions taken.

With the RWM inoperable during a reactor shutdown, the operator is still capable of enforcing the prescribed control rod sequence. Required Action D.1 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other task qualified member of the technical staff (e.g., shift technical advisor or reactor engineer). The RWM may be bypassed under these conditions to allow the reactor shutdown to continue.

E.l and E.2 With one Reactor Mode Switch-Shutdown Position control rod withdrawal block channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod withdrawal block function. However, since the Required Actions are consistent with the normal action of an OPERABLE Reactor Mode Switch-Shutdown Position Function (i.e., maintaining all control rods inserted), there is no distinction between having one or two channels inoperable.

In both cases (one or both channels inoperable), suspending all control rod withdrawal and initiating action to fully insert all insertable control rods in core cel l s containing one or more fuel assemblies will ensure that the core is subcritical with adequate SDM ensured by LCO 3.1.1.

Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are therefore not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.

SURVEILLANCE As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Control Rod Block instrumentation Function are found in the SRs column of Table 3.3.2.1-1.

(continued)

LaSalle 1 and 2 B 3.3.2.1-12 Revision 39

Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE The Surveillances are modified by a second Note to indicate REQUIREMENTS that when an RBM channel is placed in an inoperable status (continued) solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains control rod block capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.

This Note is based on the reliability analysis (Ref. 8) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that a control rod block will be initiated when necessary.

SR 3.3.2.1.1 A CHANNEL FUNCTIONAL TEST is performed for each RBM channel to ensure that the entire channel will perform the intended function. It includes the Reactor Manual Control Multiplexing System input. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a s i ngle contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.

Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.2.1.2 and SR 3.3.2.1.3 A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the entire system will perform the intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This LaSalle 1 and 2 B 3.3.2.1-13 Revision 51

Control Rod Block Instrumentation B 3.3.2.1 clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This (continued)

LaSalle 1 and 2 B 3.3.2.1-14 Revision 51

Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.2 and SR 3.3.2.1.3 (continued)

REQUIREMENTS is acceptable because all of the other required contacts of the relay are verified by other Technical Spec i fications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The CHANNEL FUNCTIONAL TEST for the RWM is performed by attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a control rod bloc k occurs and by verifying proper annunciation of the selection error of at least one out-of-sequence control rod. As noted in the SRs, SR 3.3.2.1.2 is not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after any control rod is withdrawn at ~ -+/--G-5% RTP and s 300 psig reactor steam dome pressure in MODE 2 and SR 3.3.2.1.3 is not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after THERMAL POWER is ~ -+/--G-5% RTP and s 300 psig reactor steam dome pressure in MODE ~ 2 . The Note to SR 3.3.2.1.2 allows entry into MODE 2 -e-A----a-during a startup and entry into MODE 2 concurrent with a power reduction to ~ +G-5% RTP and reactor steam dome pressure is s 300 psig during a shutdown to perform the required Surveillance if the Frequency is not met per SR 3.0.2. The Note to SR 3.3.2.1.3 allows a THERMAL POWER reduction to

~ -+/--G-5% RTP ands 300 psig reactor steam dome pressure in MODE -+/----2 to perform the required Surveillance if the Frequency is not met per SR 3.0.2. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months allowances are based on operating experience and in consideration of providing a reasonable time in which to complete the SRs.

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

SR 3.3.2.1.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.

As noted, neutron detectors are excluded from the CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.8.

LaSalle 1 and 2 B 3.3.2.1-15 Revision 51

Control Rod Block I nstrumentat ion B 3.3.2.1 (conti nued )

LaSalle 1 and 2 B 3.3.2.1- 16 Revision 51

Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.4 (continued)

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

SR 3.3.2.1.5 The RBM is automatically bypassed when power is below a specified value or if a peripheral control rod is selected.

The power level is determined from the APRM signals input to each RBM channel. The automatic bypass setpoint must be verified periodically to be< 30% RTP. In add i tion, it must also be verified that the RBM is not bypassed when a control rod that is not a peripheral control rod is selected (only one non-peripheral control rod is required to be verified). If any bypass setpoint is nonconservative, then the affected RBM channel is considered inoperable.

Alternatively, the APRM channel can be placed i n the conservative condition to enable the RBM. If placed in this condition, the SR is met and the RBM channel is not considered inoperable. As noted, neutron detectors are excluded from the Surveillance because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.8. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.2.1.6 The RWM is automatically bypassed when power is above a specified value. The power level is determined from steam flow signal. The automatic bypass setpoint must be verified periodically to be> 5-+/--0% RTP .~.~ If the RWM low power setpoint is nonconservative, then the RWM is considered inoperable. Alternately, the low power setpoint channel can be placed in the conservative condition (nonbypass). If placed in the nonbypassed condition, the SR is met and the RWM is not considered inoperable. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(continued)

LaSalle 1 and 2 B 3.3.2.1-17 Revision 51

Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.7 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed for the Reactor Mode Switch-Shutdown Position Function to ensure that the entire channel will perform the intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Spec i fications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The CHANNEL FUNCTIONAL TEST for the Reactor Mode Switch-Shutdown Position Function is performed by attempting to withdraw any control rod with the reactor mode switch in the shutdown position and verifying a control rod block occurs.

As noted in the SR, the Surveillance is not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the reactor mode switch is in the shutdown position, since testing of this interlock with the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads, or movable links. This allows entry into MODES 3 and 4 if the Frequency is not met per SR 3.0.2. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs.

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

SR 3.3.2.1.8 The RWM will only enforce the proper control rod sequence if the rod sequence is properly input into the RWM computer. This SR ensures that the proper sequence is loaded into the RWM so that it can perform its intended function. The Surveillance is performed once prior to declaring RWM OPERABLE following loading of sequence into RWM, since this is when rod sequence input errors are possible.

(continued)

LaSalle 1 and 2 B 3.3.2.1-18 Revision 51

Control Rod Block I nstrumentation B 3.3.2.1 BASES SURVEIL LANCE SR 3 . 3.2.1.9 REQUIREMENTS (continued) LCO 3 .1. 3 and LCO 3.1.6 may require individual control rods to be bypassed (taken out of serv i ce) in the RWM to a l low insertion of an in operab le control rod or correction of a control rod pattern not in compliance with the analyzed rod position sequence . With the control rods bypassed (taken out of service) in the RWM, the RWM wil l prov i de insert and withdr aw blocks for bypassed control rods that are fully inserted and a wit hdraw block for bypassed contro l rods t hat are not fully inserted. To ensure th e proper bypassing and movement of those affected control rods, a second lic ensed operator (Reac t or Operator or Senior Reactor Operator) or other task qualified member of the t echn i cal staf f (e.g., sh ift technical adv i sor or reactor engineer) must verify the bypass i ng and position of these control rods. Comp li ance with this SR all ows the RWM to be OPERABLE wi th these con t ro l rods bypassed.

REFERENCES 1. UF SAR, Section 7. 7. 6. 3 .

2. UF SAR, Section 7.7.2.2 . 3 .

3. UFSAR, Section 7.7.7.2 . 3 .

4. UFSAR, Sect i on 15.4.2.3.

5. UF SAR, Section 15.4.9.

6. "Modifications to the Requirements for Contro l Rod Drop Acc i dent Mitigating Systems, " BWR Owners' Group, July 1986.

7. NRC SER, "Acceptance of Referencing of Licensing Topical Report NED E- 24011- P- A," "Genera l El ectr i c Standard Applicat i on for Reactor Fuel, Revision 8, Amendment 17 , " December 27, 1987.

8. GENE-770-06-1-A, "Addendum to Bases for Changes to Survei ll ance Test I nterva l s and Al l owed Out-of-Service Time s for Selected In strumentat i on Technical Spec ifi cations, " December 1992.

9. MEDO 33091 A, Revision 2, "Improved tlP .~:;; Control Rod 1

Insertion Process," J1:1ly 200~. Deleted.

LaSal le 1 and 2 B 3.3.2.1-19 Revision 71

Control Rod Block I nstrumentation B 3.3.2.1

-+/--G--.9.NEDE-33885P-A, "GNF CRDA Application Methodology,"

Revision 1, March 2020 LaSalle 1 and 2 B 3.3.2.1-20 Revision 71

v t e Letter Sequence Request
EPID:L-2021-LLA-0016, NRR E-mail Capture - (External_Sender) IEMA Response: State Consultation - LaSalle 1 and 2: License Amendment Request Re; Revision of Technical Specifications to Incorporate GNF CRDA Methodology (Approved, Closed)
Initiation Request, Request, Request Acceptance Administration Questions 26 July 2021 Answers 17 August 2021 Results Approval Other:
MONTHYEAR2021-07-26 [Table View]

RS-21-006, Application to Incorporate Licensing Topical Report NEDE-33885P-A, Revision 1, GNF CRDA Application Methodology

Text

Subject:

References:

Subject:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

6.0 REFERENCES

Navigation menu

RS-21-006, Application to Incorporate Licensing Topical Report NEDE-33885P-A, Revision 1, GNF CRDA Application Methodology

Text

Subject:

References:

Subject:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

6.0 REFERENCES

Navigation menu

Search