Text

Technical Specifications Revision Request to Adopt TSTF-476, Improved BPWS Control Rod Insertion Process
	ML072820527
Person / Time
Site:	Hatch
Issue date:	10/03/2007
From:	Stinson L; Southern Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NL-07-1709
	Download: ML072820527 (54)
	v • d • e

L. M. Stinson (Mike) Southern Nuclear Vice President Operating Company, Inc.

Fleet Operations Support 40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.5181 Fax 205.992.0341 SOUTHERN October 3, 2007 COMPANY Energy to Serve Your World' Docket Nos.: 50-321 NL-07-1709 50-366 U, S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant - Units 1 and 2 Technical Specifications Revision Request to Adopt TSTF-476 Improved BPWS Control Rod Insertion Process Ladies and Gentlemen:

In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations, Southern Nuclear Operating Company is submitting a request for an amendment to the Technical Specifications (TS) for Plant Hatch, Units 1 and 2. The proposed changes would revise Section 3.1.6, "Rod Pattern Control," and 3.3.2.1, "Control Rod Block Instrumentation," to allow Plant Hatch to reference a new Banked Position Withdrawal Sequence (BPWS) in the TS Bases. In addition, a footnote is added to Table 3.3.2.1-1, "Control Rod Block Instrumentation."

The changes are consistent with NRC approved Industry Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-476, Revision 1, Improved BPWS Control Rod Insertion Process (NEDO-33091)." The availability of this TS improvement was announced in the Federal Register on May 23, 2007 (72 FR 29004) as part of the consolidated line item improvement process (CLIIP).

Enclosure 1 provides a description and assessment of the proposed changes, as well as confirmation of applicability. Enclosure 2 provides the existing TS pages and TS Bases marked up to show the proposed changes. Enclosure 3 provides the final TS and TS Bases pages.

Southern Nuclear requests approval of the proposed license amendment by January 30, 2008 with the amendment being implemented within 30 days. In accordance with 10 CFR 50.91, a copy of this application with enclosures, is being provided to the designated Georgia state official.'

(Affirmation and signature are provided on the following page.)

Mr. L. M. Stinson states he is a Vice President of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear

U. S, Nuclear Regulatory Commission NL-07-1709 Page. 2 Operating Company and to the best of his knowledge and belief, the facts set forth in this letter are true.

The NRC commitments contained in this letter are provided as a table in . If you have any questions, please advise.

Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY L. M, Stinson Vice President Fleet Operations Support Swo to andjubscribed before me thisjŽ. day of . c, 2007.

Notary Public My commission expires: La.lio LMS/OCV/daj

Enclosures:

1. Basis for Proposed Change

2. Technical Specifications and Bases Markup Pages

3. Technical Specifications and Bases Clean Typed Pages

4. List of Regulatory Commitments cc: Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. D. R. Madison., Vice President - Hatch Mr. D. H. Jones, Vice President - Engineering RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Dr. W. D. Travers, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Hatch Mr. J. A. Hickey, Senior Resident Inspector - Hatch State of Georgia Mr. N. Holcomb, Commissioner - Department of Natural Resources

Edwin I. Hatch Nuclear Plant Enclosure 1 Basis for Proposed Change

Edwin I. Hatch Nuclear Plant Enclosure 1 Basis for Proposed Change 1.0 Description This is a request to amend Operating Licenses DPR-57 and NPF-5 for Plant Hatch Units 1 and 2, respectively. The proposed changes would revise the Bases section of Technical Specification 3.1.6, "Rod Pattern Control," and 3.3.2.1, "Control Rod Block Instrumentation,": along with TS Table 3.3.2.1-1, "Control Rod Block Instrumentation," to allow reference to an improved, optional Banked Position Withdrawal Sequence (BPWS) for use during reactor shutdown.

The new BPWS is described in Topical Report NEDO-33091-A, Revision 2, "Improved BPWS Control Rod Insertion Process," dated July 2004, and approved by the NRC by Safety Evaluation (SE) dated June 16, 2004. Technical Specification Task Force (TSTF) change traveler TSTF-476, revision 1, "Improved BPWS Control Rod Insertion Process (NEDO-33091)" was announced for availability in the Federal Register on May, 23, 2007 as part of the Consolidated Line Item Improvement Process (CLIIP).

2.0 Proposed Changes Consistent with NRC-approved TSTF-476, Revision 1, the proposed TS and Bases changes include:

Revised TS Section 3.1.6 Bases to allow use of an optional BPWS during plant shutdown.

Revised TS Section 3.3.2.1 Bases to allow reprogramming of the rod worth minimizer during optional BPWS shutdown sequence.

Revised TS Table 3.3.2.1-1 "Control Rod Block Instrumentation," which adds a footnote that allows operators to bypass the rod worth minimizer if conditions for the optional BPWS shutdown process are satisfied.

3.0 Background

The background for this application is as stated in the model SE in NRC's Notice of Availability published on May 23, 2007, Federal Register, May 23, 2007 (72 FR 29004), the NRC Notice for Comment published on May 3, 2006 (71 FR 26118),

and TSTF-476, Revision 1.

4.0 Technical Analysis Southern Nuclear Plant Hatch has reviewed NEDO-33091-A, Revision 2, and the staff's SE dated June 16, 2004, as well as TSTF-476, Revision 1, and the model SE published on May 23, 2007 (72 FR 29004) as part of the CLIIP Notice for Comment. Southern Nuclear has applied the methodology in NEDO-33091 -A, Revision 2 to develop the proposed TS changes. Southern Nuclear has also concluded that the justifications presented in TSTF-476, Revision 1 and the model SE prepared by the NRC staff are applicable to Plant Hatch Units 1 and 2, El - 1

Edwin I. Hatch Nuclear Plant Enclosure 1 Basis for Proposed Change and justify this amendment for the incorporation of the changes to the Plant Hatch Technical Specifications.

5.0 Regulatory Analysis A description of this proposed change and its relationship to applicable regulatory requirements and guidance was provided in the NRC Notice of Availability published on May 23, 2007 (72 FR 29004), the NRC Notice for Comment published on May 3, 2006 (71 FR 26118), and TSTF-476, Revision 1.

5.1 Regulatory Commitments As discussed in the model SE published in the Federal Register on May 23, 2007 for this Technical Specifications improvement, the following plant-specific verification/commitments were performed. The safety evaluation for NEDO-33091-A explained that the potential for a control rod drop accident (CRDA) will be eliminated by the following changes to the operational procedures, which Plant Hatch will commit to make prior to implementation:

1. Before reducing power to the low power setpoint (LPSP), operators shall confirm control rod coupling integrity for all rods that are fully withdrawn. Control rods that have not been confirmed coupled and which are in intermediate positions must be fully inserted prior to power reduction to the LPSP. No action is required for fully-inserted control rods. If shutdown is required and all rods which are not confirmed coupled cannot be fully inserted prior to the power dropping below the LPSP, then the original standard BPWS must be adhered to. The original/standard BPWS can be found in Licensing Topical Report (LTR) NEDO-21231, "Banked Position Withdrawal Sequence,"

January 1977, and is referred to in NUREG-1433 and NUREG-1434.

1

2. After reactor power drops below the LPSP, rods may be inserted from notch position 48 to notch position 00 without stopping at intermediate positions. However, GE Nuclear Energy recommends that operators insert control rods in the same order as specified for the original/standard BPWS as much as reasonably possible. If a plant is in the process of shutting down following improved BPWS with the power below the LPSP, no control rod shall be withdrawn unless the control rod pattern is in compliance with standard BPWS requirements.

In addition to the procedure changes specified above, the staff previously concluded, based on its review of NEDO-33091-A, that no single failure of the boiling water reactor control rod drive (CRD) mechanical or hydraulic system can cause a control rod to drop completely out of the reactor core during the shutdown process. Therefore, the proper use of the improved BPWS will prevent El -2

Edwin I. Hatch Nuclear Plant Enclosure 1 Basis for Proposed Change a CRDA from occurring while power is below the LPSP. Southern Nuclear has verified, in accordance with NEDO-33091-A, Revision 2, that no single failure of the boiling water reactor CRD mechanical or hydraulic system can cause a control rod drop completely out of the reactor core during the shutdown process.

6.0 No Significant Hazards Consideration Southern Nuclear has reviewed the proposed no significant hazards consideration determination published on May 23, 2007 (72 FR 29004) as part of the CLIIP. Southern Nuclear has concluded that the proposed determination presented in the notice is applicable to Plant Hatch and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91 (a).

7.0 Environmental Evaluation Southern Nuclear has reviewed the environmental consideration included in the model SE published on May 23, 2007 (72 FR 29004) as part of the CLIIP.

Southern Nuclear has concluded that the staff's findings presented therein are applicable to Plant Hatch and the determination is hereby incorporated by reference for this application.

8.0 References Federal Register Notices:

Notice for Comment published on May 3, 2006 (71 FR 26118)

Notice of Availability published on May 23, 2007 (72 FR 29004)

El -3

Edwin I. Hatch Nuclear Plant Enclosure 2 Technical Specifications and Bases Markup Pages

Control Rod Block Instrumentation 3.3.2.1 Table 3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE Rod Block Monitor

a. Low Power Range - Upscale (a) 2 SR 3.3.2.1.1 < 115.5/125 SR 3.3.2.1.4 divisions of full SR 3.3.2.1.7 scale

b. Intermediate Power Range - Upscale (b) 2 SR 3.3.2.1.1 < 109.7/125 SR 3.3.2.1.4 divisions of full SR 3.3.2.1.7 scale

c. High Power Range - Upscale (c) 2 SR 3.3.2.1.1 5 105.9/125 SR 3.3.2.1.4 divisions of full SR 3.3.2.1.7 scale

d. Inop (d) 2 SR 3.3.2.1.1 NA I I

e. Downscale (d) 2 SR 3.3.2.1.1 >93/125 SR 3.3.2.1.7 divisions of full scale

2. Rod Worth Minimizer 1(e), 2(e) 1 SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.5 SR 3.3.2.1.8

3. Reactor Mode Switch - Shutdown Position (f) 2 SR 3.3.2.1.6 NA (a) THERMAL POWER > 29% and < 64% RTP.

(b) THERMAL POWER > 64% and < 84% RTP.

(c) THERMAL POWER > 84%.

(d) THERMAL POWER > 29%.

(e) With THERMAL POWER < 10% RTP (f) Reactor mode switch in the shutdow' positic n.

except during the reactor shutdown process if the coupling of each withdrawn control rod has been confirmed.

HATCH UMIT 1 3.3-19 Amendment No. =

Rod Pattern Control B 3.1.6 mode BASES APPLICABLE banked positions are established to minimi e the maximum SAFETY ANALYSES incremental control rod worth without bein( overly restrictive during (continued) normal plant operation. Generic analysis o the BPWS (Ref. 1) has demonstrated that the 280 cal/gm fuel damge limit will not be violated during a CRDA while following the BPWS R of operation. The evaluation provided by the generic BPWS analysis (Ref. 8) allows a limited number (i.e., eight) and corresponding distribution of fully inserted, inoperable control rods that are not in compliance with the Add insert 1B here. sequence. This analysis may be modified by plant specific

,\evaluations.

Rod pattern control satisfies Criterion 3 of the NRC Policy Statement (Ref. 9).

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 BPWS. This LCO only applies to OPERABLE control rods. For inoperable control rods required to be inserted, separate requirements are specified in LCO 3.1.3, "Control Rod OPERABILITY," consistent with the allowances for inoperable control rods in the BPWS.

APPLICABILITY In MODES 1 and 2, when THERMAL POWER is < 10% RTP, the CRDA is a Design Basis Accident and, therefore, compliance with the assumptions of the safety analysis is required. When THERMAL POWER is > 10% RTP, there is no credible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Ref. 2). In MODES 3, 4, and 5, since the reactor is shutdown and only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will remain subcritical with a single control rod withdrawn.

ACTIONS A.1 and A.2 With one or more OPERABLE control rods not in compliance with the prescribed control rod sequence, actions may be taken to either correct the control rod pattern or declare 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 may be the result of "double notching," drifting from a (continued)

HATCH UNIT 1 B 3.1-31

Rod Pattern Control B 3.1.6 BASES ACTIONS B.1 and B.2 (continued) 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 verified to be in compliance with the BPWS at a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency to ensure the assumptions of the CRDA analyses are met. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency was developed considering that the primary check on compliance with the BPWS is performed by the RWM (LCO 3.3.2.1), which provides control rod blocks to enforce the required sequence and is required to be OPERABLE when operating at < 10% RTP.

REFERENCES 1. NEDE-24011-P-A-US, "General Electric Standard Application for Reactor Fuel, Supplement for United States," (revision specified in the COLR).

2. 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, 1988.

3. NUREG-0979, Section 4.2.1.3.2, April 1983.

4. NUREG-0800, Section 15.4.9, Revision 2, July 1981.

5. 10CFR 100.11.

6. NEDO-21778-A, "Transient Pressure Rises Affected Fracture Toughness Requirements for Boiling Water Reactors,"

10. NEDO 33091-A, December 1978.

Revision 2, "Improved BPWS Control Rod 7. ASME, Boiler and Pressure Vessel Code.

Insertion Process,"

July 2004. 8. NEDO-21231, "Banked Position Withdrawal Sequence,"

January 1977.

9. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

HATCH UNIT 1 B 3.1-33

INSERT lB When performing a shutdown of the plant, an optional BPWS control rod sequence (Ref. 10) may be used provided 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 Reference 10 control rod sequence for shutdown, the rod worth minimizer may be reprogrammed to enforce the requirements of the improved BPWS control rod insertion process, or bypassed in accordance with the allowance provided in the Applicability Note for the Rod Worth Minimizer in Table 3.3.2.1-1.

In order to use the Reference 10 BPWS 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 10. If the requirements for use of the BPWS control rod insertion process contained in Reference 10 are followed, the plant is considered to be in compliance with BPWS requirements, as required by LCO 3.1.6.

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 1. Rod Block Monitor (continued)

SAFETY ANALYSES, LCO, and Two channels of the RBM are required to be OPERABLE, with their APPLICABILITY setpoints within the appropriate Allowable Values, to ensure that no single instrument failure can preclude a rod block from this Function.

The setpoints are calibrated consistent with applicable setpoint methodology (nominal trip setpoint).

Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable.

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 Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.

The RBM is assumed to mitigate the consequences of an RWE event when operating > 29% RTP. Below this power level, the consequences of an RWE event will not exceed the MCPR SL or the 1% plastic strain design limit; therefore, the RBM is not required to be OPERABLE (Ref. 3).

1standard

2. Rod Worth Minimizer , and 14 The RWM enforces the banked p sition withdrawal seq ence (BPWS) to ensure that the initial condition of the CRDA analysi are not violated. The analytical methods nd assumptions use, in evaluating the CRDA are summarized in Ref rences 4, 5, 6, P . In addition, the Reference 6 analysis (Generi BPWS analysis) may be modified by plant specific evaluations. The BPWS requires that control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. Requirements that (continued)

HATCH UNIT 1 B 3.3-44

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

SAFETY ANALYSES, LCO, and the control rod sequence is in compliance with the BPWS are APPLICABILITY specified in LCO 3.1.6, "Rod Pattern Control."

The RWM Function satisfies Criterion 3 of the NRC Policy Statement Add insert 2B here. *" (Ref. 10).

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 available and required to be OPERABLE (Ref. 7). Special circumstances provided 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 inoperable control rods, or to allow correction of a control rod pattern not in compliance with the BPWS. 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 BPWS, and, therefore, OPERABILITY of the RWM, is required in MODES 1 and 2 when THERMAL POWER is

< 10% RTP. When THERMAL POWER is > 10% RTP, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Refs. 5 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 SDM ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.

3. Reactor Mode Switch - Shutdown Position During MODES 3 and 4, and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical; therefore, no positive reactivity insertion events are analyzed. The Reactor Mode Switch - Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.

The Reactor Mode Switch - Shutdown Position Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 10).

(continued)

HATCH UNIT 1 B 3.3-45

INSERT 2B When performing a shutdown of the plant, an optional BPWS control rod sequence (Ref. 14) 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 14 control rod insertion sequence for shutdown, the rod worth minimizer may be reprogrammed to enforce the requirements of the improved BPWS control rod insertion process, or it can be bypassed if it is not programmed to reflect the optional BPWS shutdown sequence, as permitted by the Applicability Note for the RWM in Table 3.3.2.1-1.

Control Rod Block Instrumentation B 3.3.2.1 BASES REFERENCES 7. NRC SER, "Acceptance of Referencing of Licensing Topical (continued) Report NEDE-24011-P-A," "General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17,"

December 27, 1987.

8. NEDC-30851-P-A, "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation,"

October 1988.

9. GENE-770-06-1, "Bases For Changes To Surveillance Test Intervals and Allowed Out-Of-Service Times For Selected Instrumentation Technical Specifications," February 1991.

10. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

11. NEDC-3241 OP-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)

Retrofit Plus Option III Stability Trip Function," October 1995.

12. NRC Safety Evaluation Report for Amendment 232.

13. NRC Safety Evaluation Report for Amendment 234, Quarterly Surveillance Extension.

14. NEDO 33091-A, Revision 2, "Improved BPWS Control Rod Insertion Process," July 2004.

HATCH UNIT 1 B 3.3-52

Control Rod Block Instrumentation 3.3.2.1 Table 3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE 1 Rod Block Monitor

a. Low Power Range - Upscale (a) 2 SR 3.3.2.1.1 5 115.5/125 SR 3.3.2.1.4 divisions of full SR 3.3.2.1.7 scale

b. Intermediate Power Range - Upscale (b) 2 SR 3.3.2.1.1 < 109.7/125 SR 3.3.2.1.4 divisions of full SR 3.3.2.1.7 scale

c. High Power Range - Upscale (c) 2 SR 3.3.2.1.1 5 105.9/125 SR 3.3.2.1.4 divisions of full SR 3.3.2.1.7 scale

d. Inop (d) 2 SR 3.3.2.1.1 NA

e. Downscale (d) 2 SR 3.3.2.1.1 > 93/125 SR 3.3.2.1.7 divisions of full scale

2. Rod Worth Minimizer 1(e), 2(e) 1 SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.5 SR 3.3.2.1.8

3. Reactor Mode Switch - Shutdown Position (f) 2 SR 3.3.2.1.6 NA (a) THERMAL POWER > 29% and < 64% RTP.

(b) THERMAL POWER ->64% and < 84% RTP.

THERMA Pexcept during the reactor shutdown process if (c) THERMAL POWER > 84%. Jthe coupling of each withdrawn control rod has (d) THERMAL POWER > 29%. been confirmed.

(e) With THERMAL POWER < 10% RTIF'.

(f) Reactor mode switch in the shutdown position.

HATCH UNIT 2 3.3-19 Amendment No. M,

Rod Pattern Control B 3.1.6 BASES APPLICABLE banked positions are established to minimize the maximum SAFETY ANALYSES incremental control rod worth without be g overly restrictive during (continued) normal plant operation. Generic analysi of the BPWS (Ref. 1) has demonstrated that the 280 cal/gm fuel da a e limit will not be violated during a CRDA while following the BPWS of operation. The evaluation provided by the generic BPWS analysis (Ref. 8) allows a limited number (i.e., eight) and corresponding distribution of fully inserted, inoperable control rods that are not in compliance with the Add insert 1 B here. sequence. This analysis may be modified by plant specific evaluations.

Rod pattern control satisfies Criterion 3 of the NRC Policy Statement (Ref. 9).

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 BPWS. This LCO only applies to OPERABLE control rods. For inoperable control rods required to be inserted, separate requirements are specified in LCO 3.1.3, "Control Rod OPERABILITY," consistent with the allowances for inoperable control rods in the BPWS.

APPLICABILITY In MODES 1 and 2, when THERMAL POWER is < 10% RTP, the CRDA is a Design Basis Accident and, therefore, compliance with the assumptions of the safety analysis is required. When THERMAL POWER is > 10% RTP, there is no credible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Ref. 2). In MODES 3, 4, and 5, since the reactor is shut down and only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will remain subcritical with a single control rod withdrawn.

ACTIONS A.1 and A.2 With one or more OPERABLE control rods not in compliance with the prescribed control rod sequence, actions may be taken to either correct the control rod pattern or declare 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 may be the result of "double notching," drifting from a (continued)

HATCH UNIT 2 B 3.1-31

Rod Pattern Control B 3.1.6 BASES ACTIONS B.1 and B.2 (continued) 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 verified to be in compliance with the BPWS at a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency to ensure the assumptions of the CRDA analyses are met. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency was developed considering that the primary check on compliance with the BPWS is performed by the RWM (LCO 3.3.2.1), which provides control rod blocks to enforce the required sequence and is required to be OPERABLE vAhen operating at < 10% RTP.

REFERENCES 1. NEDE-24011-P-A-US, "General Electric Standard Application for Reactor Fuel, Supplement for United States," (revision specified in the COLR).

2. 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, 1988.

3. NUREG-0979, Section 4.2.1.3.2, April 1983.

4. NUREG-0800, Section 15.4.9, Revision 2, July 1981.

5. 10 CFR 100.11.

6. NEDO-21778-A, "Transient Pressure Rises Affected Fracture Toughness Requirements for Boiling Water Reactors,"

December 1978.

10. NEDO 33091-A, Revision 2, h 7. ASME, Boiler and Pressure Vessel Code.

"Improved BPWS NEDO-21231, "Banked Position Withdrawal Sequence,"

Control Rod Insertion January 1977.

Process," July 2004.

18.

9. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

HATCH UNIT 2 B 3.1-33

INSERT IB When performing a shutdown of the plant, an optional BPWS control rod sequence (Ref. 10) may be used provided 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 Reference 10 control rod sequence for shutdown, the rod worth minimizer may be reprogrammed to enforce the requirements of the improved BPWS control rod insertion process, or bypassed in accordance with the allowance provided in the Applicability Note for the Rod Worth Minimizer in Table 3.3.2.1-1.

In order to use the Reference 10 BPWS 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 10. If the requirements for use of the BPWS control rod insertion process contained in Reference 10 are followed, the plant is considered to be in compliance with BPWS requirements, as required by LCO 3.1.6.

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 1. Rod Block Monitor (continued)

SAFETY ANALYSES, LCO, and Two channels of the RBM are required to be OPERABLE, with their APPLICABILITY setpoints within the appropriate Allowable Values, to ensure that no single instrument failure can preclude a rod block from this Function.

The setpoints are calibrated consistent with applicable setpoint methodology (nominal trip setpoint).

Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable.

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 Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50 49) are accounted for.

The RBM is assumed to mitigate the consequences of an RWE event when operating > 29% RTP. Below this power level, the consequences of an RWE event will not violate the MCPR SL or the 1% plastic strain design limit; therefore, the RBM is not required to be OPERABLE (Ref. 3).

1standard . and 14

2. Rod Worth Minimizer The RWM enforces the banked osition withdrawal seq ence (BPWS) to ensure that the initial condition of the CRDA analysi are not violated. The analytical methods nd assumptions usell in evaluating the CRDA are summarized in Re rences 4, 5, 6, * . In addition, the Reference 6 analysis (Generi BPWS analysis) may be modified by plant specific evaluations. The BPWS requires that control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. Requirements that (continued)

HATCH UNIT 2 B 3.3-44

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

SAFETY ANALYSES, LCO, and the control rod sequence is in compliance with the BPWS are APPLICABILITY specified in LCO 3.1.6, "Rod Pattern Control."

The RWM Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 10).

2B here.

Add insert 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 available and required to be OPERABLE (Ref. 7). Special circumstances provided 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 inoperable control rods, or to allow correction of a control rod pattern not in compliance with the BPWS. 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 BPWS, and therefore OPERABILITY of the RWM, is required in MODES 1 and 2 when THERMAL POWER is

< 10% RTP. When THERMAL POWER is > 10% RTP, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Refs. 5 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 SDM ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.

3. Reactor Mode Switch - Shutdown Position During MODES 3 and 4, and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical; therefore, no positive reactivity insertion events are analyzed. The Reactor Mode Switch - Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.

The Reactor Mode Switch - Shutdown Position Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 10).

(continued)

HATCH UNIT 2 B 3.3-45

INSERT 2B When performing a shutdown of the plant, an optional BPWS control rod sequence (Ref. 14) 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 14 control rod insertion sequence for shutdown, the rod worth minimizer may be reprogrammed to enforce the requirements of the improved BPWS control rod insertion process, or it can be bypassed if it is not programmed to reflect the optional BPWS shutdown sequence, as permitted by the Applicability Note for the RWM in Table 3.3.2.1-1.

Control Rod Block Instrumentation B 3.3.2.1 BASES REFERENCES 7. NRC SER, "Acceptance of Referencing of Licensing Topical (continued) Report NEDE-2401 1-P-A," "General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17,"

December 27, 1987.

8. NEDC-30851-P-A, "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation,"

October 1988.

9. GENE-770-06-1, "Bases for Changes To Surveillance Test Intervals And Allowed Out-Of-Service Times For Selected Instrumentation Technical Specifications," February 1991.

10. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

11. NEDC-3241 OP-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)

Retrofit Plus Option III Stability Trip Function," October 1995.

12. NRC Safety Evaluation Report for Amendment 174.

13. NRC Safety Evaluation Report for Amendment 176, Quarterly Surveillance Extension.

14. NEDO 33091-A, Revision 2, "Improved BPWS Control Rod Insertion Process," July 2004.

HATCH UNIT 2 B 3.3-52

Edwin I. Hatch Nuclear Plant Enclosure 3 Technical Specifications and Bases Clean Typed Pages

Control Rod Block Instrumentation 3.3.2.1 Table 3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE

1. Rod Block Monitor

a. Low Power Range - Upscale (a) 2 SR 3.3.2.1.1 s 115.5/125 SR 3.3.2.1.4 divisions of full SR 3.3.2.1.7 scale

b. Intermediate Power Range - Upscale (b) 2 SR 3.3.2.1.1 < 109.7/125 SR 3.3.2.1.4 divisions of full SR 3.3.2.1.7 scale

c. High Power Range - Upscale (c) 2 SR 3.3.2.1.1 5 105.9/125 SR 3.3.2.1.4 divisions of full SR 3.3.2.1.7 scale

d. Inop (d) 2 SR 3.3.2.1.1 NA

e. Downscale (d) 2 SR 3.3.2.1.1 >-93/125 SR 3.3.2.1.7 divisions of full scale

2. Rod Worth Minimizer lI(e). 2(e) 1 SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.5 SR 3.3.2.1.8

3. Reactor Mode Switch - Shutdown Position (fM 2 SR 3.3.2.1.6 NA a) THERMAL POWER > 29% and < 64% RTP.

(b) THERMAL POWER > 64% and < 84% RTP.

(c) THERMAL POWER > 84%.

(d) THERMAL POWER > 29%.

(e) With THERMAL POWER < 10% RTP, except during the reactor shutdown process if the coupling of each withdrawn control rod has been confirmed.

(i) Reactor mode switch in the shutdown position.

HATCH UNIT 1 3.3-19 Amendment No.

Rod Pattern Control B 3.1.6 BASES APPLICABLE banked positions are established to minimize the maximum SAFETY ANALYSES incremental control rod worth without being overly restrictive during (continued) normal plant operation. Generic analysis of the BPWS (Ref. 1) has demonstrated that the 280 cal/gm fuel damage limit will not be violated during a CRDA while following the BPWS mode of operation. The evaluation provided by the generic BPWS analysis (Ref. 8) allows a limited number (i.e., eight) and corresponding distribution of fully inserted, inoperable control rods that are not in compliance with the sequence. This analysis may be modified by plant specific evaluations.

When performing a shutdown of the plant, an optional BPWS control rod sequence (Ref. 10) may be used provided 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 Reference 10 control rod sequence for shutdown, the rod worth minimizer may be reprogrammed to enforce the requirements of the improved BPWS control rod insertion process, or bypassed in accordance with the allowance provided in the Applicability Note for the Rod Worth Minimizer in Table 3.3.2.1-1.

In order to use the Reference 10 BPWS 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 10. If the requirements for use of the BPWS control rod insertion process contained in Reference 10 are followed, the plant is considered to be in compliance with BPWS requirements, as required by LCO 3.1.6.

Rod pattern control satisfies Criterion 3 of the NRC Policy Statement (Ref. 9).

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 BPWS. This LCO only applies to OPERABLE control rods. For inoperable control rods required to be inserted, separate requirements are specified in LCO 3.1.3, "Control Rod OPERABILITY," consistent with the allowances for inoperable control rods in the BPWS.

(continued)

HATCH UNIT 1 B 3.1-31

Rod Pattern Control B 3.1.6 BASES (continued)

APPLICABILITY In MODES 1 and 2, when THERMAL POWER is 5 10% RTP, the CRDA is a Design Basis Accident and, therefore, compliance with the assumptions of the safety analysis is required. When THERMAL POWER is > 10% RTP, there is no credible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Ref. 2). In MODES 3, 4, and 5, since the reactor is shutdown and only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will remain subcritical with a single control rod withdrawn.

ACTIONS A.1 and A.2 With one or more OPERABLE control rods not in compliance with the prescribed control rod sequence, actions may be taken to either correct the control rod pattern or declare 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 may be the result of "double notching," drifting from a control rod drive cooling water transient, leaking scram valves, or a power reduction to < 10% RTP before establishing the correct control rod pattern. The number of OPERABLE 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. When the control rod pattern is not in compliance with the prescribed sequence, all control rod movement must be stopped except for moves needed to correct the rod pattern, or scram if warranted.

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 or other qualified member of the technical staff. This ensures that the control rods will be moved to the correct position. A control rod not in compliance with the prescribed sequence is not considered inoperable except as required by Required 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 reasonable, 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.

B.1 and B.2 If nine or more OPERABLE control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence.

(continued)

HATCH UNIT 1 B 3.1-32

Rod Pattern Control B 3.1.6 BASES ACTIONS B.1 and B.2 (continued)

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 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 or other qualified member of the technical staff.

When nine or more OPERABLE control rods are not in compliance with BPWS, 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 mode switch in shutdown, the reactor is shut down, and as such, does not meet the applicability 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 verified to be in compliance with the BPWS at a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency to ensure the assumptions of the CRDA analyses are met. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency was developed considering that the primary check on compliance with the BPWS is performed by the RWM (LCO 3.3.2.1), which provides control rod blocks to enforce the required sequence and is required to be OPERABLE when operating at < 10% RTP.

REFERENCES 1. NEDE-24011-P-A-US, "General Electric Standard Application for Reactor Fuel, Supplement for United States," (revision specified in the COLR).

2. 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, 1988.

3. NUREG-0979, Section 4.2.1.3.2, April 1983.

4. NUREG-0800, Section 15.4.9, Revision 2, July 1981.

(continued)

HATCH UNIT 1 B 3.1-33

Rod Pattern Control B 3.1.6 BASES REFERENCES 5. 10 CFR 100.11.

(continued)

6. NEDO-21778-A, "Transient Pressure Rises Affected Fracture Toughness Requirements for Boiling Water Reactors,"

December 1978.

7. ASME, Boiler and Pressure Vessel Code.

8. NEDO-21231, "Banked Position Withdrawal Sequence,"

January 1977.

9. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

10. NEDO-33091 -A, Revision 2, "Improved BPWS Control Rod Insertion Process," July 2004.

HATCH UNIT 1B B 3.1-34

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 1. Rod Block Monitor (continued)

SAFETY ANALYSES, LCO, and Two channels of the RBM are required to be OPERABLE, with their APPLICABILITY setpoints within the appropriate Allowable Values, to ensure that no single instrument failure can preclude a rod block from this Function.

The setpoints are calibrated consistent with applicable setpoint methodology (nominal trip setpoint).

Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable.

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 Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.

The RBM is assumed to mitigate the consequences of an RWE event when operating > 29% RTP. Below this power level, the consequences of an RWE event will not exceed the MCPR SL or the 1% plastic strain design limit; therefore, the RBM is not required to be OPERABLE (Ref. 3).

2. Rod Worth Minimizer The .RWM enforces the banked position withdrawal sequence (BPWS) 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 4, 5, 6, 7, and 14. In addition, the Reference 6 analysis (Generic BPWS analysis) may be modified by plant specific evaluations. The standard BPWS requires that control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions.

Requirements that (continued)

HATCH UNIT 1 B 3.3-44

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

SAFETY ANALYSES, LCO, and the control rod sequence is in compliance with the BPWS are APPLICABILITY specified in LCO 3.1.6, "Rod Pattern Control."

When performing a shutdown of the plant, an optional BPWS control rod sequence (Ref. 14) 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 14 control rod insertion sequence for shutdown, the rod worth minimizer may be reprogrammed to enforce the requirements of the improved BPWS control rod insertion process, or it can be bypassed if it is not programmed to reflect the optional BPWS shutdown sequence, as permitted by the Applicability Note for the RWM in Table 3.3.2.1-1.

The RWM Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 10).

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 available and required to be OPERABLE (Ref. 7). Special circumstances provided 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 inoperable control rods, or to allow correction of a control rod pattern not in compliance with the BPWS. 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 BPWS, and, therefore, OPERABILITY of the RWM, is required in MODES 1 and 2 when THERMAL POWER is

< 10% RTP. When THERMAL POWER is > 10% RTP, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Refs. 5 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 SDM ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.

3. Reactor Mode Switch - Shutdown Position During MODES 3 and 4, and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed (continued)

HATCH UNIT 1 [] 3.3-45

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 3. Reactor Mode Switch - Shutdown Position (continued)

SAFETY ANALYSES, LCO, and to be subcritical; therefore, no positive reactivity insertion events are APPLICABILITY analyzed. The Reactor Mode Switch - Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.

The Reactor Mode Switch - Shutdown Position Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 10).

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 (MODE 3, 4, or 5), 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 withdrawal blockis required to be OPERABLE. 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 Interlock") provides the required control rod withdrawal blocks.

ACTIONS A.1 With one RBM channel inoperable, the remaining 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 Completion 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 the event occurring coincident with a failure in the remaining OPERABLE channel.

B.1 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.

(continued)

HATCH UNIT 1 B 3.3-46

Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS B.1 (continued)

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.1, 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 reliability 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 rods, was not performed in the last calendar year (i.e., in the last 12 months). These requirements minimize the number of reactor startups initiated with 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 0.2.1.2 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 qualified member of the technical staff (e.g., a qualified shift technical advisor or reactor engineer). The RWM may be bypassed under these conditions to allow continued operations. In addition, Required Actions of LCO 3.1.3 and LCO 3.1.6 may require bypassing the RWM, during which time the RWM must be considered inoperable with Condition C entered and its Required Actions taken.

D.1 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 qualified member of the technical staff. The RWM may be bypassed under these conditions to allow the reactor shutdown to continue.

(continued)

HATCH UNIT 1 B 3.3-47

Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS E.1 and E.2 (continued)

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 cells 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 Control REQUIREMENTS Rod Block instrumentation Function are found in the SRs column of Table 3.3.2.1-1.

The Surveillances are modified by a second Note to indicate that when an RBM channel is placed in an inoperable status 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. 9) 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 System input.

(continued)

HATCH UNIT 1 B 3.3-48

Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.1 (continued)

REQUIREMENTS Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 184 days is based on reliability analyses (Ref. 11).

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. 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 block occurs. This test is performed as soon as possible after the applicable conditions are entered. 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

< 10% RTP 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 < 10% RTP in MODE 1. This allows entry into MODE 2 (and if entered during a shutdown, concurrent power reduction to < 10% RTP) for SR 3.3.2.1.2 and THERMAL POWER reduction to < 10% RTP in MODE 1 for SR 3.3.2.1.3 to perform the required Surveillances ifthe 92 day on an ALTERNATE TEST BASIS 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 92 day on an ALTERNATE TEST BASIS Frequency is based on a review of the surveillance test history and Reference 13.

SR 3.3.2.1.4 The RBM setpoints are automatically varied as a function of power.

Three Allowable Values are specified in Table 3.3.2.1-1, each within a specific power range. The power at which the control rod block Allowable Values automatically change are based on the APRM signal's input to each RBM channel. Below the minimum power setpoint, the RBM is automatically bypassed. These power Allowable Values must be verified periodically to be less than or equal to the specified values. If any power range setpoint is nonconservative, then the affected RBM channel is considered inoperable. Alternatively, the power range channel can be placed in the conservative condition (i.e.,

enabling the proper RBM setpoint). 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 (continued)

HATCH UNIT 1 B 3.3-49

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

REQUIREMENTS tested in SR 3.3.1.1.2 and SR 3.3.1.1.8. The 24 month Frequency is based on a review of the surveillance test history and Reference 12.

SR 3.3.2.1.5 The RWM is automatically bypassed when power is above a specified value. The power level is determined from APRM power signals. The automatic bypass setpoint must be verified periodically to be

> 10% 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 24 month Frequency is based on Reference 12.

SR 3.3.2.1.6 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. 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 18 month 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 SR.

The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 24 month Frequency is based on a review of the surveillance test history and Reference 12.

SR 3.3.2.1.7 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the (continued)

HATCH UNIT 1 B 3.3-50

Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.7 (continued)

REQUIREMENTS 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 3.3.1.1.8.

The 24 month Frequency is based on a review of the surveillance test history and Reference 12.

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.

REFERENCES 1. FSAR, Section 7.5.8.2.3.

2. FSAR, Section 7.2.2.4.

3. NEDC-30474-P, "Average Power Range Monitor, Rod Block Monitor, and Technical Specification Improvements (ARTS)

Program for Edwin I. Hatch Nuclear Plants," December 1983.

4. NEDE-24011-P-A-US, "General Electrical Standard Application for Reload Fuel," Supplement for United States, (revision specified in the COLR).

5. 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.

6. NEDO-21231, "Banked Position Withdrawal Sequence,"

January 1977.

(continued)

HATCH UNIT 1 B 3.3-51

Control Rod Block Instrumentation B 3.3.2.1 BASES REFERENCES 7. NRC SER, "Acceptance of Referencing of Licensing Topical (continued) Report NEDE-2401 1-P-A," "General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17,"

December 27, 1987.

8. NEDC-30851-P-A, "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation,"

October 1988.

9. GENE-770-06-1, "Bases For Changes To Surveillance Test Intervals and Allowed Out-Of-Service Times For Selected Instrumentation Technical Specifications," February 1991.

10. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

11. NEDC-3241 OP-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)

Retrofit Plus Option III Stability Trip Function," October 1995.

12. NRC Safety Evaluation Report for Amendment 232.

13. NRC Safety Evaluation Report for Amendment 234, Quarterly Surveillance Extension.

14. NEDO-33091-A, Revision 2, "Improved BPWS Control Rod Insertion Process," July 2004.

HATCH UNIT 1 B 3.3-52

Control Rod Block Instrumentation 3.3.2.1 Table 3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE 1 Rod Block Monitor

a. Low Power Range - Upscale (a) 2 SR 3.3.2.1.1 S 115.5/125 SR 3.3.2.1.4 divisions of full SR 3.3.2.1.7 scale

b. Intermediate Power Range - Upscale (b) 2 SR 3.3.2.1.1 < 109.7/125 SR 3.3.2.1.4 divisions of full SR 3.3.2.1.7 scale

c. High Power Range - Upscale (c) 2 SR 3.3.2.1.1 < 105.9/125 SR 3.3.2.1.4 divisions of full SR 3.3.2.1.7 scale

d. Inop (d) 2 SR 3.3.2.1.1 NA

e. Downscale (d) 2 SR 3.3.2.1.1 2:93/125 SR 3.3.2.1.7 divisions of full scale

2. Rod Worth Minimizer 1(e), 2(e) 1 SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.5 SR 3.3.2.1.8

3. Reactor Mode Switch - Shutdown Position (M) 2 SR 3.3.2.1.6 NA (a) THERMAL POWER - 29% and < 64% RTP.

(b) THERMAL POWER 2 64% and < 84% RTP.

(c) THERMAL POWER - 84%.

(d) THERMAL POWER > 29%.

(e) With THERMAL POWER < 10% RTP, except during the reactor shutdown process if the coupling of each withdrawn control rod has been confirmed.

(f) Reactor mode switch in the shutdown position.

HATCH UNIT 2 3.3-19 Amendment No.

Rod Pattern Control B 3.1.6 BASES APPLICABLE banked positions are established to minimize the maximum SAFETY ANALYSES incremental control rod worth without being overly restrictive during (continued) normal plant operation. Generic analysis of the BPWS (Ref. 1) has demonstrated that the 280 cal/gm fuel damage limit will not be violated during a CRDA while following the BPWS mode of operation. The evaluation provided by the generic BPWS analysis (Ref. 8) allows a limited number (i.e., eight) and corresponding distribution of fully inserted, inoperable control rods that are not in compliance with the sequence. This analysis may be modified by plant specific evaluations.

When performing a shutdown of the plant, an optional BPWS control rod sequence (Ref. 10) may be used provided 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 Reference 10 control rod sequence for shutdown, the rod worth minimizer may be reprogrammed to enforce the requirements of the improved BPWS control rod insertion process, or bypassed in accordance with the allowance provided in the Applicability Note for the Rod Worth Minimizer in Table 3.3.2.1-1.

In order to use the Reference 10 BPWS 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 10. If the requirements for use of the BPWS control rod insertion process contained in Reference 10 are followed, the plant is considered to be in compliance with BPWS requirements, as required by LCO 3.1.6.

Rod pattern control satisfies Criterion 3 of the NRC Policy Statement (Ref. 9).

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 BPWS. This LCO only applies to OPERABLE control rods. For inoperable control rods required to be inserted, separate requirements are specified in LCO 3.1.3, "Control Rod OPERABILITY," consistent with the allowances for inoperable control rods in the BPWS.

(continued)

HATCH UNIT 2 B 3.1-31

Rod Pattern Control B 3.1.6 BASES (continued)

APPLICABILITY In MODES 1 and 2, when THERMAL POWER is 5 10% RTP, the CRDA is a Design Basis Accident and, therefore, compliance with the assumptions of the safety analysis is required. When THERMAL POWER is > 10% RTP, there is no credible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Ref. 2). In MODES 3, 4, and 5, since the reactor is shut down and only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will remain subcritical with a single control rod withdrawn.

ACTIONS A.1 and A.2 With one or more OPERABLE control rods not in compliance with the prescribed control rod sequence, actions may be taken to either correct the control rod pattern or declare 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 may be the result of "double notching," drifting from a control rod drive cooling water transient, leaking scram valves, or a power reduction to < 10% RTP before establishing the correct control rod pattern. The number of OPERABLE 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. When the control rod pattern is not in compliance with the prescribed sequence, all control rod movement must be stopped except for moves needed to correct the rod pattern, or scram if warranted.

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 or other qualified member of the technical staff. This ensures that the control rods will be moved to the correct position. A control rod not in compliance with the prescribed sequence is not considered inoperable except as required by Required 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 reasonable, 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.

B.1 and B.2 If nine or more OPERABLE control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence.

(continued)

HATCH UNIT 2 B 3.1-32

Rod Pattern Control B 3.1.6 BASES ACTIONS B.1 and B.2 (continued)

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 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 or other qualified member of the technical staff.

When nine or more OPERABLE control rods are not in compliance with BPWS, 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 mode switch in shutdown, the reactor is shut down, and as such, does not meet the applicability 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 verified to be in compliance with the BPWS at a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency to ensure the assumptions of the CRDA analyses are met. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency was developed considering that the primary check on compliance with the BPWS is performed by the RWM (LCO 3.3.2.1), which provides control rod blocks to enforce the required sequence and is required to be OPERABLE when operating at

< 10% RTP.

REFERENCES 1. NEDE-24011-P-A-US, "General Electric Standard Application for Reactor Fuel, Supplement for United States," (revision specified in the COLR).

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

"Amendment 17 to General Electric Licensing Topical Report NEDE-2401 1-P-A," BWROG-8644, August 15, 1988.

3. NUREG-0979, Section 4.2.1.3.2, April 1983

4. NUREG-0800, Section 15.4.9, Revision 2, July 1981.

(continued)

HATCH UNIT 2 B 3.1-33

Rod Pattern Control B 3.1.6 BASES REFERENCES 5. 10CFR 100.11.

(continued)

6. NEDO-21778-A, "Transient Pressure Rises Affected Fracture Toughness Requirements for Boiling Water Reactors,"

December 1978.

7. ASME, Boiler and Pressure Vessel Code.

8. NEDO-21231, "Banked Position Withdrawal Sequence,"

January 1977.

9. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

10. NEDO-33091 -A, Revision 2, "Improved BPWS Control Rod Insertion Process," July 2004.

HATCH UNIT 2 B 3.1-34

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 1. Rod Block Monitor (continued)

SAFETY ANALYSES, LCO, and Two channels of the RBM are required to be OPERABLE, with their APPLICABILITY setpoints within the appropriate Allowable Values, to ensure that no single instrument failure can preclude a rod block from this Function.

The setpoints are calibrated consistent with applicable setpoint methodology (nominal trip setpoint).

Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable.

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 Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50 49) are accounted for.

The RBM is assumed to mitigate the consequences of an RWE event when operating > 29% RTP. Below this power level, the consequences of an RWE event will not violate the MCPR SL or the 1% plastic strain design limit; therefore, the RBM is not required to be OPERABLE (Ref. 3).

2. Rod Worth Minimizer The RWM enforces the banked position withdrawal sequence (BPWS) 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 4, 5, 6, 7, and 14. In addition, the Reference 6 analysis (Generic BPWS analysis) may be modified by plant specific evaluations. The standard BPWS requires that control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions.

(continued)

HATCH UNIT 2 B 3.3-44

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

SAFETY ANALYSES, LCO, and the control rod sequence is in compliance with the BPWS are APPLICABILITY specified in LCO 3.1.6, "Rod Pattern Control."

When performing a shutdown of the plant, an optional BPWS control rod sequence (Ref. 14) 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 14 control rod insertion sequence for shutdown, the rod worth minimizer may be reprogrammed to enforce the requirements of the improved BPWS control rod insertion process, or it can be bypassed if it is not programmed to reflect the optional BPWS shutdown sequence, as permitted by the Applicability Note for the RWM in Table 3.3.2.1-1.

The RWM Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 10).

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 available and required to be OPERABLE (Ref. 7). Special circumstances provided 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 inoperable control rods, or to allow correction of a control rod pattern not in compliance with the BPWS. 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 BPWS, and therefore OPERABILITY of the RWM, is required in MODES 1 and 2 when THERMAL POWER is

< 10% RTP. When THERMAL POWER is > 10% RTP, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Refs. 5 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 SDM ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.

3. Reactor Mode Switch - Shutdown Position During MODES 3 and 4, and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed (continued)

HATCH UNIT 2 B 3.3-45

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 3. Reactor Mode Switch - Shutdown Position (continued)

SAFETY ANALYSES, LCO, and to be subcritical; therefore, no positive reactivity insertion events are APPLICABILITY analyzed. The Reactor Mode Switch - Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.

The Reactor Mode Switch - Shutdown Position Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 10).

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 (MODE 3, 4, or 5), 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 withdrawal block is required to be OPERABLE. 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 Interlock") provides the required control rod withdrawal blocks.

ACTIONS A.1 With one RBM channel inoperable, the remaining 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 Completion 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 the event occurring coincident with a failure in the remaining OPERABLE channel.

B.1 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.

(continued)

HATCH UNIT 2 B 3.3-46

Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS B.1 (continued)

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.1, 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 reliability 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 rods was not performed in the last calendar year (i.e., in the last 12 months). These requirements minimize the number of reactor startups initiated with 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 qualified member of the technical staff (e.g., a qualified shift technical advisor or reactor engineer). The RWM may be bypassed under these conditions to allow continued operations. In addition, Required Actions of LCO 3.1.3 and LCO 3.1.6 may require bypassing the RWM, during which time the RWM must be considered inoperable with Condition C entered and its Required Actions taken.

D.1 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 qualified member of the technical staff. The RWM may be bypassed under these conditions to allow the reactor shutdown to continue.

(continued)

HATCH UNIT 2 [] 3.3-47

Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS E.1 and E.2 (continued)

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 cells 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 Control REQUIREMENTS Rod Block instrumentation Function are found in the SRs column of Table 3.3.2.1-1.

The Surveillances are modified by a second Note to indicate that when an RBM channel is placed in an inoperable status 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. 9) 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 System input.

(continued)

HATCH UNIT 2 B 3.3-48

Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.1 (continued)

REQUIREMENTS Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 184 days is based on reliability analyses (Ref. 11).

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. 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 block occurs. This test is performed as soon as possible after the applicable conditions are entered. 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

< 10% RTP 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 < 10% RTP in MODE 1. This allows entry into MODE 2 (and if entered during a shutdown, concurrent power reduction to < 10% RTP) for SR 3.3.2.1.2 and THERMAL POWER reduction to < 10% RTP in MODE 1 for SR 3.3.2.1.3 to perform the required Surveillances if the 92 day on an ALTERNATE TEST BASIS 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 92 day on an ALTERNATE TEST BASIS Frequency is based on a review of the surveillance test history and Reference 13.

SR 3.3.2.1.4 The RBM setpoints are automatically varied as a function of power.

Three Allowable Values are specified in Table 3.3.2.1-1, each within a specific power range. The power at which the control rod block Allowable Values automatically change are based on the APRM signal's input to each RBM channel. Below the minimum power setpoint, the RBM is automatically bypassed. These power Allowable Values must be verified periodically to be less than or equal to the specified values. If any power range setpoint is nonconservative, then the affected RBM channel is considered inoperable. Alternatively, the power range channel can be placed in the conservative condition (i.e.,

enabling the proper RBM setpoint). 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 (continued)

HATCH UNIT 2 B 3.3-49

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

REQUIREMENTS tested in SR 3.3.1.1.2 and SR 3.3.1.1.8. The 24 month Frequency is based on a review of the surveillance test history and Reference 12.

SR 3.3.2.1.5 The RWM is automatically bypassed when power is above a specified value. The power level is determined from APRM power signals. The automatic bypass setpoint must be verified periodically to be

> 10% 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 24 month Frequency is based on Reference 12.

SR 3.3.2.1.6 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. -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 18 month 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 SR.

The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 24 month Frequency is based on a review of the surveillance test history and Reference 12.

SR 3.3.2.1.7 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the (continued)

HATCH UNIT 2 B 3.3-50

Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.7 (continued)

REQUIREMENTS 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 3.3.1.1.8.

The 24 month Frequency is based on a review of the surveillance test history and Reference 12.

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.

REFERENCES 1. FSAR, Section 7.6.2.2.5.

2. FSAR, Section 7.6.8.2.6.

3. NEDC-30474-P, "Average Power Range Monitor, Rod Block Monitor, and Technical Specification Improvements (ARTS)

Program for Edwin I. Hatch Nuclear Plants," December 1983.

4. NEDE-2401 1-P-A-US, "General Electrical Standard Application for Reload Fuel," Supplement for United States, (revision specified in the COLR).

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

"Amendment 17 to General Electric Licensing Topical Report NEDE-2401 1-P-A," BWROG-8644, August 15, 1986.

6. NEDO-21231, "Banked Position Withdrawal Sequence,"

January 1977.

(continued)

HATCH UNIT 2 B63.3-51

Control Rod Block Instrumentation B 3.3.2.1 BASES REFERENCES 7. NRC SER, "Acceptance of Referencing of Licensing Topical (continued) Report NEDE-24011-P-A," "General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17,"

December 27, 1987.

8. NEDC-30851-P-A, "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation,"

October 1988.

9. GENE-770-06-1, "Bases for Changes To Surveillance Test Intervals And Allowed Out-Of-Service Times For Selected Instrumentation Technical Specifications," February 1991.

10. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

11. NEDC-3241OP-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)

Retrofit Plus Option III Stability Trip Function," October 1995.

12. NRC Safety Evaluation Report for Amendment 174.

13. NRC Safety Evaluation Report for Amendment 176, Quarterly Surveillance Extension.

14. NEDO-33091-A, Revision 2, "Improved BPWS Control Rod Insertion Process," July 2004.

HATCH UNIT 2 B 3.3-52

Edwin I. Hatch Nuclear Plant Enclosure 4 List of Regulatory Commitments

Edwin I. Hatch Nuclear Plant Enclosure 4 List of Regulatory Commitments The following table identifies those actions committed by Southern Nuclear Operating Company in this document for Edwin I. Hatch Nuclear Plant. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

Type Scheduled One-Time Continuing Completion Date Commitment Action Compliance (If Required)

1. Changes will be made to operational procedures to ensure that before reducing power to the low power setpoint (LPSP), operators shall confirm control rod coupling integrity for all rods that are fully withdrawn.

Control rods that have not been confirmed coupled and X Before are in intermediate positions implementation must be fully inserted prior to power reduction to the LPSP.

If shutdown is required and all rods which are not confirmed coupled cannot be fully inserted prior to the power dropping below the LPSP, then the original standard BPWS must be adhered to.

2. Changes will be made to operational procedures to ensure that after power drops below the LPSP, rods may be inserted from notch position 48 to notch position 00 without stopping at intermediate positions. However, it is recommended that control rods be inserted in the same Before order as specified for the Xefor original/standard BPWS as much as possible. When in the process of shutting down following improved BPWS with the power below the LPSP, no control rod shall be withdrawn unless the control rod pattern is in compliance with standard BPWS requirements.

E4 - 1

NL-07-1709, Technical Specifications Revision Request to Adopt TSTF-476, Improved BPWS Control Rod Insertion Process

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