Text

Technical Specifications, Issuance of Amendments Regarding Improved Control Rod Insertion Process, TSTF 476, Revision 1
	ML090540222
Person / Time
Site:	Hatch
Issue date:	10/01/2008
From:	; Plant Licensing Branch II
To:	;
	Martin R, NRR/DORL, 415-1493
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ML073230130	List: ML090540222;
References
	TAC MD6922, TAC MD6923
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-4 fOi sample analysts or \\nstrumefli calibration, or 6ssociated with radioactive apparatus or components; (S)

Southern Nuclear. pursuant 10 the Act and 10 CFR Parts 3D and 70, to possess, but not separate. such byproduct and special nucfear materials as may be produced by the operation of the facility.

C.

Thts renewed Rcense shan be deemed to contaIn end Is &ubJect to the conditions

,.-(i'pl,-clneoln*th'e"oIJowingCommlsslo,negulatlons'"ltf10CFR'Chapter I: -'Part20~-

Section 30.34 of Part 30, Section 40.4101 Part 40. Section 50.54 of Part 50, and Sedlon 70.32 of Part 70; all appRcabla previsions of the Act end the rules, regulaUoflS. and orders of Ute Commission ncM: or hereafter In effec~ end the additIOnal conditions speclfled or Incorporated below; (1)

Maximum Power Level Southern Nuclear Is authorfzed to operate the 'acIDty at steady state'

.reactor core power lavels not In excess of 2804 megaWBtts thermal (2)

TechnIcal BpeclficsUons The TechnlcaJ SpecfffcatJons (Appendix A) and the Enviro.nmental Protection Plan (Appendlx B), as rev1sed through Amendment No.

are hereby. Incorporated In.the renewed license. Southem Nuclear shaD operate the facUlty In accordance with the Technical Specifications end the Environmental Protection Plen.

The Survel1Jence Requirement (SR) contained In the Technical SpeclficatfOlis and Usted bilow, Is not required 10 be performed*

Irrvnedlalely upon Implementation of Amendment No. 195*. The SR Dsted below*shaD be successfully demonstrated prior to the time and conditIon specmed:

SR 3.8.1.18 &rui( be successfully demonstrated at 116 next reguJal1y scheduled performance

.(3).

EIre pr01ectlon Southem Nuclear shafllmplement and maintain In effeet aD provisions of the flre protection program, whIch Is referenced In the Updated FInal Safety AnaI)'Sfs Report for the facfflty. as contained In the updated Fii6 Hazards AnalysIs and Fire Protection Program for Edwtn I. Hatch Nuclear Plant Units 1 and 2, which was originally submlUed by letter.

daled July 22. 1986; SouthQm Nuclear may make changes to the fire

._--._- _.- -'. -- *'*protecUon program-w11hout-prfor-Commlssfonapprowlonlyffthechanges Renewed Ucense No. DPR*57 Amendment* No. 25R

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

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

1.

Rod Siock Monitor

a.

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

b.

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

c.

High Power Range* Upscale (c) 2 SR 3.3.2.1.1 SR 3.3.2.1.4 SR 3.32.1.7 S 105.9/125 divisions of full scale

d.

Inop (d) 2 SR 3.3.2.1.1 NA

e.

Downscale (d) 2 SR 3.3.2.1.1 SR 3.3.2.1.7

~ 93/125 divisions of full scale

2.

Rod Worth Minimizer lle)*2(e)

SR 3.3.2.1.2 SR 3.3.2.1.3 SR 3.3.2.1.5 SR 3.3.2.1.8 NA

3.

Reactor Mode Switch - Shutdown Position (I) 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 01 each withdrawn control rod has been confirmed.

(f)

Reactor mode switch in the shutdown position.

HATCH UNIT 1 3.3-19 Amendment No. 25R

Rod Pattern Control 83.1.6 BASES APPLICABLE SAFETY ANALYSES (continued) banked positions are established to minimize the maximum incremental control rod worth without being overly restrictive during 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 (Le., 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 coupied. 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 LeO 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, *Centre!

Rod OPERABILITY," consistent with the allowances for inoperable control rods in the BPWS.

(continued)

HATCH UNIT 1 B3.1-31

Rod Pattern Control B 3.1.6 BASES (continued)

APPLICABI L1TY In MODES 1 and 2, when THERMAL POWER is s; 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 callgm 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, II drifting from a control rod drive cooling water transient, leaking scram valves, or a power reduction to s 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)

?')A HATCH UNIT 1 B 3.1-32 Amendment No.

Rod Pattern Control B 3.1.6 BASES ACTIONS B.1and 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 REQUIREMENTS SR 3.1.6.1 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 s 10% ATP.

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, Ju~y 1981.

(continued)

HATCH UNIT 1 83.1-33 Amendment No. 258

Rod Pattern Control B 3.1.6 BASES REFERENCES (continued)

5.

10 CFR 100.11.

6.

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

December 1978.

7.

ASME, Boiler and Pressure Vesser 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 1 B 3.1-34 Amendment No. 258

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY

1. Rod Block Monitor (continued)

Two channels of the RBM are required to be OPERABLE, with their 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 AWE event will not exceed the MCPA 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 Amendment No. 258

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE SAFETY ANALYSES, LCD, and APPLICABILITY

2. Rod Worth Minimizer (continued) the control rod sequence is in compliance with the BPWS are specified in LCD 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 LCD 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 GRDA (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 B 3.3-45 Amendment No. 258

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY

3. Reactor Mode Switch - Shutdown Position (continued) 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).

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 wilhdrawal blocks.

ACTIONS 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.

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 Amenoment No. 25A

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 (Le., 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 Leo 3.1.3 and LeO 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.

With the RWM inoperable during a reactor shutdown, the operator is still capable of enforcing the prescribed control rod sequence.

Required Action 0.1 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 Amendment No. 25R

Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS (continued)

E.1 and E.2 With one Reactor Mode Switch - Shutdown Position control rod withdrawal block channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod withdrawal block function. However, since the Required Actions are consistent with the normal action of an OPERABLE Reactor Mode Switch - Shutdown Position Function (Le., 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 SOM 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 REQUIREMENTS As noted at the beginning of the SRs, ~he SRs for each Control 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 Amendment No. 258

Control Rod Block Instrumentation B 3.3.2.1 BASES SU RVEI LLANCE REQUIREMENTS SR 3.3.2.1.1 (continued)

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 1 B 3.3-49 A~endment Nn

?~R

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

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 REQUIREMENTS SR 3.3.2.1.7 (continued) 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 iQto 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 1. 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.

(continue.s!l HATCH UNIT 1 B 3.3-51 Amendment No. 258

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-32410P-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 Amend~ent No. 25R

4*

(6)

Southern Nuclear, pursuant \\0 the Act and 10 CFR Parts 30 and 70, to.

possess, but not separate, such byproduct and special nuclear materials as may be produced by the operaUon of the facDlly.

c.

This renewed ncense shaD be deemed to contain, and Is subJect to, the conditions speclfled In the following Commission regulations In 10 OFA

.Chap'er I: Part 20. SecUon 30.34 01 Part 30, Section 40.41 of Part 40, Sectlon 50.64 of Part 50, and Section 70.32 of Part 70; aft applicab'e provisions of the Act and the rules, regu'aUons, and orders of the Commfsslon now or hereafter In

eftect; and the additional condltlons!-speclfiedor Incorporated-below:.

(1)

Maximum power Leye' Southern Nuclear Is authorized to operate the faclilly Bl steady stale reactor core power 'evels not fn excess of 2.804 megawatts therma'. In accordance *wlth the condlUons specIDed herein.

(2)

Iechn~ISPedmC8tions The Technfcal SpecfflcaUons (Appendix A) and the EnvIrorunental Protection PIS" (Appendix Bl, as reVIsed through Amendment No. 202 are hereby Incorporated In the renewed license. Southem 'Nuclear anau operate thefaclily In accordance with the TeetvllcaJ Specfficatlons and Ihe EnvIronmenta' Protection Plan.

(3)

AddltlgnaJ CondltJons The matlers specified In the following condJUons shan be completed to the saUsfacUon cf'the Commlsslon wtthln the stated time periods following the Issuance of the renewed license or within the operaltonal restrk:tlons lnd1cated. The removal of these condltJons shall be made by an amendment to the license supported by a favorable evaluation by the Commission.

'(8) fire Protect1oQ SoUthern Nuclear shaR Implement and maintain In effect aft proVIsIons of the fire protection program, which fs'referenced In the the Updated FInal safety Analysis Report for Ihe facfJlty, as Contained I

'The original Doensee authorized to possess, use, and operate the faCIlity was GeorgIa. Power I

Company (GPO). Consequenfly, certaIn hlstorfcal references fo GPe remain In certain license

,conditions. -

Renewed Ucense No. NPF*5 Amendmenl No. 7.07.

Control Rod Block Instrumentation 3.3.2.1 Table 3.32.1-1 (page 1011)

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

1.

Rod Block Monitor

a.

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

b.

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

c.

High Power Range - Upscale (el 2

SR 3.3.2.1.1 SR 3.3.2.1.4 SR 3.3.2.1.7 S 105.9/125 divisions of full scale

d.

Inop (d) 2 SR 3.3.2.1.1 NA

e.

Downscale

[d) 2 SR 3.3.2.1.1 SA 3.3.2.1.7

~ 931125 divisions of full scale 2

Rod Worth Minimizer 1(e),2(e)

SA 3.3.2.1.2 SR 3.3.2.1.3 SR 3.3.2.1.5 SR 3.3.2.1.8 NA

3.

Reactor Mode Switch - Shutdown Position (I) 2 SR 3.3.2.1.6 NA (a)

THERMAL POWER ~ 29% and < 64% RTP.

(b)

THERMAL POWER ~ 64% and < 84% RTP.

(cl 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.202

Rod Pattern Control B 3.1.6 BASES APPLICABLE SAFETY ANALYSES (continued) banked positions are established to minimize the maximum incremental control rod worth without being overly restrictive during 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 (Le., 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 83.1-31 Amendment No. 202

Rod Pattern Control B3.1.6 BASES (continued)

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 GRDA (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 GRDA 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 ;s 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 GRDA 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 Amendment No. 202

Rod Pattern Control 83.1.6 BASES ACTIONS 8.1 and B.2 (continued)

Control rod withdrawal should be suspended immediately to prevent the potential for further deviation rrom 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 8PWS, 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 REQUIREMENTS SR 3.1.6.1 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 S 10% RTP.

REFERENCES

1.

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

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 2 83.1-33 A,TIennment No. 202

Rod Pattern Control B 3.1,6 BASES REFERENCES (continued)

5.

10 CFR 100.11.

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 Amendment No. 202

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY

1. Rod Block Monitor (continued)

Two channels of the RBM are required to be OPERABLE, with their 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 TheAWM 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 UNIT2 B 3.3-44 Amendment No. 202

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY HATCH UNIT 2

2. Rod Worth Minimizer (continued) the control rod sequence is in compliance with the BPWS are 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 GRDA 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 (continue~

B 3.3-45 Amendment No. 202

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and AP PLiCABI L1TY

3. Reactor Mode Switch - Shutdown Position (continued) 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).

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 contr:ol 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 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.

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 pertorming 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 Amendment No. 202

Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS fLL(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 (Le., 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.

With the RWM inoperable during a reactor shutdown, the operator is still capable of enforcing the prescribed control rod sequence.

Required Action 0.1 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 B 3.3-47 Amendment No. l02

Control Rod Block Instrumentation B 3,3.2,1 BASES ACTIONS (continued)

E.1 and E.2 With one Reactor Mode Switch - Shutdown Position control rod withdrawal block channel inoperable, the remaining OPERABLE channel is adequate to pertorm 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 (Le., 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 SOM 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 REQUIREMENTS As noted at the beginning of the SAs, the SAs for each Control Rod Block instrumentation Function are found in the SAs 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.

SA 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 ftmendment No. 202

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

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 SAs. 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 ;s 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 Amendment No. 202

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

SA 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 Amendment No. 202

Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE REQUIREMENTS SR 3.3.2.1.7 (continued) 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-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 2 83.3-51

Control Rod Block Instrumentation B 3.3.2.1 BASES REFERENCES (continued)

7.

NRC SER, "Acceptance of Referencing of Licensing Topical 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.

NEOC-32410P-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.

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

HATCH UNIT 2 B 3.3-52 AmendMent No. 202

v t e Letter Sequence Approval
TAC:MD6922, Improved BPWS Control Rod Insertion Process (NEDO-33091) (Approved, Closed) TAC:MD6923, Improved BPWS Control Rod Insertion Process (NEDO-33091) (Approved, Closed)
Initiation Acceptance... Administration Questions Answers Results Approval, Approval Other:
MONTHYEAR2008-10-01 [Table View]

ML090540222

Text

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