ML061600452

License Amendment Request - Delete Reference to Banked Position Withdraw Sequence (BPWS)

ML061600452
Person / Time
Site:	Peach Bottom
Issue date:	06/08/2006
From:	Cowan P Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML061600452 (50)

Text

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Nuclear Exelon Nuclear www.exeloncorp.com 2 0 0 Exelon Way Kennett Square, PA 19348 10 CFR 50.90 June 8,2006 US. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278

Subject:

License Amendment Request - Delete Reference to Banked Position Withdrawal Sequence (BPWS)

Pursuant to 10 CFR 50.90, Exelon Generation Company, LLC (Exelon) hereby requests an amendment to Appendix A, Technical Specifications, of the Renewed Facility Operating Licenses DPR-44 and DPR-56. The proposed change modifies Technical Specifications (TS) 3.1.3, Control Rod OPERABILTY; TS 3.1.6, Rod Pattern Control; TS 3.3.2.1, Control Rod Block Instrumentation; TS 3.10.7, Control Rod Testing -

Operating, and; TS 3.1 0.8, SHUTDOWN MARGIN (SDM) Test - Refueling. The proposed change would replace the current references to Banked Position Withdrawal Sequence (BPWS) with references to the analyzed rod position sequence.

Exelon requests approval of the proposed changes by June 8,2007. Once approved, the amendment shall be implemented within 60 days. The proposed changes have been reviewed by the Plant Operations Review Committee and approved by the Nuclear Safety Review Board. No new regulatory commitments are established by this submittal.

We are notifying the Commonwealth of Pennsylvania of this application for changes to the Technical Specifications by transmitting a copy of this letter and its attachments to the designated State Official.

PBAPS Unit 2 & 3 LAR Delete Reference to Banked Position Withdrawal Sequence (BPWS)

June 8,2006 Page 2 If any additional information is needed, please contact Tom Loomis at (610) 765-5510.

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

Respectfully, Executed On Pamela B. cowan Director, Licensing & Regulatory Affairs Exelon Generation Company, LLC

Enclosures:

(1) Evaluation of Proposed Change (2) Markup of Proposed Technical Specification Page Changes (3) Markup of Proposed Technical Specification Bases Page Changes cc: S. J. Collins, Administrator, USNRC Region I J. Kim, Project Manager, USNRC F. Bowers, USNRC Senior Resident Inspector, Peach Bottom Atomic Power Station

ENCLOSURE 1 EVALUATION OF PROPOSED CHANGE

ENCLOSURE 1 EVALUATION OF PROPOSED CHANGES CONTENTS

SUBJECT:

DELETE REFERENCE TO BANKED POSITION WITHDRAWAL SEQUENCE (BPWS)

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria

6.0 ENVIRONMENTAL CONSIDERATION

7.0 PRECENDENT

8.0 REFERENCES

PBAPS Units 2 & 3 LAR Delete Reference to Banked Position Withdrawal Sequence (BPWS)

Evaluation of Proposed Changes ENCLOSURE 1

1.0 DESCRIPTION

This letter is a request to amend Renewed Facility Operating Licenses Nos. DPR-44 and DPR-56. The proposed change would replace the current references to Banked Position Withdrawal Sequence (BPWS) with references to the analyzed rod position sequence.

Exelon Generation Company, LLC (Exelon) requests approval of the proposed changes by June 8, 2007. Once approved, the amendment shall be implemented within 60 days.

2.0 PROPOSED CHANGE

The proposed change modifies:

1) Technical Specifications (TS) 3.1.3, Control Rod OPERABILTY, a) Condition D, b) Required Action D.1.

2) TS 3.1.6, Rod Pattern Control, a) Limiting Condition for Operation (LCO) 3.1.6, b) Conditions A and B, c) Surveillance Requirement 3.1.6.1.

3) TS 3.3.2.1; Control Rod Block Instrumentation a) Required Action C.2.2, b) Required Action D.1, c) Surveillance Requirement 3.3.2.1.8.

4) TS 3.10.7, Control Rod Testing - Operating, a) LCO 3.10.7.a.

5) TS 3.10.8, SHUTDOWN MARGIN (SDM) Test - Refueling, a) Limiting Condition for Operation 3.10.8.b.1.

The proposed change would replace the current references to Banked Position Withdrawal Sequence (BPWS) with reference to the analyzed rod position sequence.

provides the marked up TS pages. Enclosure 3 provides the marked up Bases pages for your information only. Final typed pages will be supplied prior to approval.

3.0 BACKGROUND

The proposed change is to replace the current references to Banked Position Withdrawal Sequence (BPWS) with reference to the analyzed rod position sequence.

As currently required in the identified TS sections, all control rod manipulations must comply with the requirements of the BPWS. These BPWS requirements are identified in NEDO-21231, Banked Position Withdrawal Sequence, dated January 1977.

PBAPS Unit 2 & 3 LAR - Delete Reference to Banked Position Withdrawal Sequence (BPWS)

Evaluation of Proposed Changes Page 2 Utilizing the words the analyzed rod position sequence in lieu of reference to only BPWS will allow greater flexibility in control rod startup and shutdown sequences that were not anticipated with the conversion to the Improved Technical Specifications, which occurred in 1995 (Reference 1) for Peach Bottom Atomic Power Station, Units 2 and 3.

The conversion to the Improved Technical Specifications incorporated reference to BPWS only. Utilizing the words the analyzed rod position sequence will provide greater flexibility in cycle-specific control rod patterns for cases when it is desirable to maintain a control rod fully inserted. This would include situations in which failed fuel suppression rods or suspected channel bow locations requiring rod insertion do not conform to BPWS requirements. In lieu of the use of only the BPWS, other analyses will be performed to develop modified startup/shutdown sequences and control rod patterns.

These sequences will be developed to minimize incremental control rod reactivity worth in accordance with the General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-15 (GESTAR-II), and U. S. Supplement, NEDE-24011-P-A-15-US, September, 2005, which incorporates NRC-approved methodology, and reviewed and approved in accordance with the 10 CFR 50.59 process. This change will allow startup/shutdown sequence modifications beyond those allowed by the general requirements of the BPWS and results in an overall reduction in unnecessary reactivity manipulations and associated operational challenges. This change will allow failed fuel to remain suppressed during plant startup/shutdown preventing further potential fuel damage and allows control rods to remain inserted in fuel cells with identified channel deformation. The change will also allow optimization of cycle-specific control rod startup and shutdown sequences that conform to the GESTAR-II requirements.

The revised TS wording was reviewed and approved as part of the Improved Technical Specifications (ITS) conversion for Dresden Nuclear Power Station, Units 2 and 3, LaSalle County Station, Units 1 and 2, and Quad Cities Nuclear Power Station, Units 1 and 2 (References 2, 3, and 4). Additionally, use of the words analyzed rod position sequence was justified in a response to a request for additional information to the U. S.

NRC (Reference 5) as part of the ITS conversion for these plants.

Bases changes are provided for your information in Enclosure 3.

4.0 TECHNICAL ANALYSIS

The design basis accident that results in a positive reactivity insertion is the Control Rod Drop Accident (Updated Final Safety Analysis Report, Peach Bottom Atomic Power Station, Units 2 and 3, Section 14.6.2, Control Rod Drop Accident). The BPWS, as currently implemented, limits the potential reactivity increase from a postulated Control Rod Drop Accident (CRDA) during reactor startups and shutdowns below the Low Power Setpoint (LPSP) of 10% of Rated Thermal Power. CRDA analyses assume that the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analysis.

In order to limit the impact of a CRDA, the BPWS is applied to both reactor startup and shutdown processes. Utilizing rod pattern control systems, such as the Rod Worth Minimizer (RWM), the BPWS reduces the maximum control rod worth during the startup and shutdown process. The Rod Worth Minimizer or plant operators are functioning within the constraints of the banked position withdrawal sequences for control rod

PBAPS Unit 2 & 3 LAR - Delete Reference to Banked Position Withdrawal Sequence (BPWS)

Evaluation of Proposed Changes Page 3 manipulations and to limit reactivity worth. The RWM (LCO 3.3.2.1) provides backup to operator control of the withdrawal sequences to ensure that the initial conditions of the CRDA analysis are not violated.

Cycle-specific control rod patterns during startup and shut down conditions will continue to be controlled by the operator and the Rod Worth Minimizer (LCO 3.3.2.1, "Control Rod Block Instrumentation"), so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods inserted to 10% of Rated Thermal Power. As a result of this proposed change, these sequences will continue to limit the potential amount of reactivity addition that could occur in the event of a Control Rod Drop Accident (CRDA).

This proposed change will allow startup/shutdown sequence modifications beyond those allowed by the general requirements of the BPWS and results in an overall reduction in unnecessary reactivity manipulations and associated operational challenges. This proposed change will allow failed fuel to remain suppressed during plant startup/shutdown preventing further potential fuel damage and allows control rods to remain inserted in fuel cells with identified channel deformation. The proposed change will also allow optimization of cycle-specific control rod startup and shutdown sequences that conform to the GESTAR-II requirements.

These sequences will be developed to minimize incremental control rod reactivity worth in accordance with the General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-15 (GESTAR-II), and U. S. Supplement, NEDE-24011-P-A-15-US, September, 2005, which incorporates NRC-approved methodology, and reviewed and approved in accordance with the 10 CFR 50.59 process.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration Exelon has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

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

Response: No The proposed change modifies Technical Specifications (TS) 3.1.3, Control Rod OPERABILTY; TS 3.1.6, Rod Pattern Control; TS 3.3.2.1, Control Rod Block Instrumentation; TS 3.10.7, Control Rod Testing - Operating, and; TS 3.10.8, SHUTDOWN MARGIN (SDM) Test - Refueling. The proposed change would replace the current references to Banked Position Withdrawal Sequence (BPWS) with references to the analyzed rod position sequence. The use of the the analyzed rod position sequence will continue to minimize the consequences of an accident previously evaluated including the Control Rod Drop Accident (CRDA).

Additionally, the use of the words the analyzed rod position sequence will provide an equivalent level of protection during plant startups and shutdowns and therefore will not increase the consequences of an accident previously evaluated.

PBAPS Unit 2 & 3 LAR - Delete Reference to Banked Position Withdrawal Sequence (BPWS)

Evaluation of Proposed Changes Page 4 Control rod patterns during startup and shut down conditions will continue to be controlled by the operator and the Rod Worth Minimizer (RWM) (LCO 3.3.2.1, "Control Rod Block Instrumentation"), so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods inserted to 10% of Rated Thermal Power. As a result of this change, these sequences will continue to limit the potential amount of reactivity addition that could occur in the event of a Control Rod Drop Accident (CRDA).

Accidents are initiated by the malfunction of plant equipment, or the failure of plant structures, systems, or components. The proposed change will ensure that analyzed rod position sequences are developed to minimize incremental control rod reactivity worth in accordance with the General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-15 (GESTAR-II), and U. S. Supplement, NEDE-24011-P-A-15-US, September, 2005, NRC approved methodology, and reviewed and approved in accordance with the 10 CFR 50.59 process. These analyzed rod position sequences will limit the potential reactivity increase for a postulated CRDA during reactor startups and shutdowns below the Low Power Setpoint of 10% of Rated Thermal Power.

The proposed change will continue to ensure that systems, structures and components are capable of performing their intended safety functions.

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

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

Response: No The proposed change does not affect the assumed accident performance of the control rods, nor any plant structure, system, or component previously evaluated.

The proposed change does not involve the installation of new equipment, and installed equipment is not being operated in a new or different manner. The change ensures that control rods remain capable of performing their safety functions. No set points are being changed which would alter the dynamic response of plant equipment. Accordingly, no new failure modes are introduced.

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

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

Response: No The proposed change will ensure that analyzed rod position sequences are developed to minimize incremental control rod reactivity worth in accordance with the General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-15

PBAPS Unit 2 & 3 LAR - Delete Reference to Banked Position Withdrawal Sequence (BPWS)

Evaluation of Proposed Changes Page 5 (GESTAR-II), and U. S. Supplement, NEDE-24011-P-A-15-US, September, 2005, NRC approved methodology, and reviewed and approved in accordance with the 10 CFR 50.59 process. The proposed change will not adversely impact the plants response to an accident or transient. All current safety margins will be maintained.

There are no changes proposed which alter the set points at which protective actions are initiated, and there is no change to the operability requirements for equipment assumed to operate for accident mitigation.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

5.2 Applicable Regulatory Requirements/Criteria 10 CFR 50.36, "Technical specifications," provides the regulatory requirements for the content required in a licensee's TS. Criterion 3 of 10 CFR 50.36(c)(2)(ii) requires a limiting condition for operation to be established for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. 10 CFR 50.36 paragraph (c)(3) specifies that surveillance requirements should ensure that limiting conditions for operation are met.

Limiting Conditions for Operation, Conditions, Requirements, and Surveillance Requirements have been established to ensure that analyzed control rod positions are maintained and controlled to ensure the protection of systems, structures and components, and to minimize the impact of accidents and transients. The proposed change will ensure that analyzed rod position sequences are developed to minimize incremental control rod reactivity worth in accordance with the General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-15 (GESTAR-II), and U. S.

Supplement, NEDE-24011-P-A-15-US, September, 2005, NRC approved methodology, and reviewed and approved in accordance with the 10 CFR 50.59 process. Criterion 3 of 10 CFR 50.36(c)(2)(ii) and paragraph (c)(3) of 10 CFR 50.36 will continue to be met since full functionality will continue to be demonstrated.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation," or would change an inspection or surveillance requirement. However, the proposed amendment

PBAPS Unit 2 & 3 LAR - Delete Reference to Banked Position Withdrawal Sequence (BPWS)

Evaluation of Proposed Changes Page 6 does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

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

7.0 PRECEDENT The NRC has granted similar changes for the Dresden Nuclear Power Station, Units 2 and 3, LaSalle County Station, Units 1 and 2, and Quad Cities Nuclear Power Station, Units 1 and 2 Technical Specifications (Reference 2, 3 and 4).

8.0 REFERENCES

1. Letter from U. S. NRC to G. A. Hunger (PECO Energy Company), Issuance of Improved Technical Specifications, Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, (TAC NOS. M90746 and M90747), dated August 30, 1995.

2. Letter from S. N. Bailey (U. S. NRC) to O. D. Kingsley (Exelon Generation Company, LLC), issuance of amendments associated with the Improved Technical Specifications for Dresden Nuclear Power Station, Units 2 and 3 (TAC. NOS.

MA8382 AND MA8383), dated March 30, 2001.

3. Letter from S. N. Bailey (U. S. NRC) to O. D. Kingsley (Exelon Generation Company, LLC), issuance of amendments associated with the Improved Technical Specifications for LaSalle County Station, Units 1 and 2 (TAC NOS. MA8388 AND MA8390), dated March 30, 2001.

4. Letter from S. N. Bailey (U. S. NRC) to O. D. Kingsley (Exelon Generation Company, LLC), issuance of amendments associated with the Improved Technical Specifications for Quad Cities Nuclear Power Station, Units 1 and 2 (TAC NOS.

MA8378 AND MA8379), dated March 30, 2001.

5. Letter from R. M. Krich (Commonwealth Edison Company) to U. S. Nuclear Regulatory Commission, Response to Request for Additional Information, dated October 9, 2000.

ENCLOSURE 2 PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 TECHNICAL SPECIFICATION PAGES MARKUP OF PROPOSED CHANGES Revised TS Pages Units 2 and 3 3.1-9 3.1-18 3.1-19 3.3-17 3.3-20 3.10-18 3.10-20

Control Rod OPERABILITY 3,1.3 ACTIONS (continued)

COND I T I ON Two or more inoperable control sods.

E.

Required Action and associ ated Completion Time o f Condition A, C, or 0 not met, Nine or more control rods inoperable.

PBAPS UNIT 2 REQUIRED ACTION OR 0.2 Restore control rod to OPERABLE status.

E. 1 Be i n MODE 3.

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COMPLETION TIME 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 4 hours 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Amendment No. 210

Rod Pattern Control 3.1.6 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Rod Pattern Control LCO 3.1.6 APPLICABILITY:

MODES 1 and 2 with THERMAL POWER s 10% RTP.

ACTIONS CON0 I T I ON A.

One or more control rods compl i ance w

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REQUIRED ACTION 1

(

1 1

NOTEIIIIIIIII A. 1 Rod worth minimizer (RWM) may be bypassed as allowed by LCO 3.3.2.1, "Contro 1 Rod Block Instrumentation.

Move associated control rod(s) to correct pos f t i on.

A.2 Declare associated control rod($)

inoperable.

PBAPS UNIT 2 3.1-18 COMPLETION TIME 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 8 hours

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(continued)

Amendment No. 210

Rod Pattern Control 3.1.6 S

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V

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I L

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A

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i FREQUENCY ACTIONS (continued)

CONDITION B.

Nine or more 0 control rods n REQUIRED ACTION Suspend wi t hdrawal o f control rods.

AND B.2 Place the reactor mode switch in the shutdown position.

COMPLETION TIME Immediately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SR 3.1.6.1 Ver RABLE control rods comply wit 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> PBAPS UNIT 2 3.1-19 Amendment No. 210

Control Rod Block Instrumentation 3.3.2.1 ACTIONS CONDITION C. (cont i nued)

0.

RWM inoperabie during reactor shutdown.

REQUIRED ACTION C.2.1.1 Verifyr 12 rods w i t hdram.

C.2.1.2 Verify by adini n i st r a t i ve methods that startup with RWH inoperable has not been performed' i n the last calendar year.

AND

\\1.2.2 Verify movement o f control rods i s i n licensed operator or alified f the 1 staff.

1.1 COMPLEJ'ION TIME Immed i a tel y Immediately luring control

.od movement During control rod movement (continued)

PBAPS UNIT 2 Amendment No. 210

Control Rod Block Instrumentation 3.3,2.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE Perform CHANNEL CALIBRATION.

SR 3.3.2.1.6 Verify the RWM i s not bypassed when THERMAL POWER i s 5 10% RTP.

Perform CHANNEL FUNCTIONAL TEST.

SR 3.3.2.1.8 Verify control rod sequenc RWM are i n conformance w i t FREQUENCY 24 months 24 months 24 months P r i o r t o declaring RWM OPERABLE f o l l o w i n g loading o f sequence i n t o RWM PBAPS UNIT 2 3 3-20 Amendment No, 232

Control Rod Testing-Operating 3.10.7 A.

Requirements o f the LCO not met.

3.10 SPECIAL OPERATIONS A. 1 Suspend performance o f the test and exception to LCO 3.1.6.

3.10.7 Control Rod Testing-Operating LCO 3.10.7 The requirements of LCO 3.1.6, "Rod Pattern Control," may be suspended to a1 1 ow performance of SDM demonstrations control rod scram time testing, control rod friction and the Startup Test Program, provided:

b.

The R19M is bypassed; the requirements o f LCO 3.3.2.1, "Control Rod Block Instrumentation,' Function 2 are suspended; and conformance to the approved control rod sequence for the specified test is verified by a second licensed operator or other qua1 ified member of the technical staff.

APPLICABILITY:

MODES 1 and 2 with LCO 3.1-6 not met.

ACTIONS COMPLETION TIME Immediately PBAPS UNIT 2 3.10-18 Amendment No. 210

SDM Test-Refueling 3.10.8 3.10 SPECIAL OPERATIONS 3.10.8 SHUTDOWN MARGIN (SDM) Test -Refueling LCO 3.10.8 The reactor mode switch position specified i n Table 1.1-1 f o r MODE 5 may be changed t o include the startup/hot standby position, and operation considered not t o be i n MODE 2, t o allow SDM testing, provided the following requirements are met:

a.

LCO 3.3.1.1, Reactor Protection System Instrumentation, MODE 2 requirements f o r Functions 2.a, 2.d and 2.e o f Table 3.3.1.1-1;

2.

Conformance t o the approved control rod sequence f o r the SDM t e s t i s v e r i f i e d by a second licensed operator o r other qua1 i f i e d member o f the technical staff;

c.

Each withdrawn control rod shall be coupled t o the associated CRD;

d.

A l l control rod withdrawals during out o f sequence control rod moves shall be made i n notch out mode;

e.

No other CORE ALTERATIONS are i n progress; and

f.

CRD charging water header pressure L 940 psig.

APPLICABILITY:

MODE 5 with the reactor mode switch i n startup/hot standby position.

PBAPS UNIT 2 3. 10-20 Amendment No. 232

Control Rod OPERABILITY 3.1.3 ACTIONS (continued)

CONDITION

0...-------- NOTE---------

Not appl i cab1 e when THERMAL POWER

> 10% RTP, Two or more inoperable control rods no E.

Required Action and associated Completion Time o f Condition A, C, or D not met.

_s OR Nine or more control rods i noperabl e, REQUIRED ACTION D. 2 Restore control rod to OPERABLE status.

E. 1 Be in MODE 3.

COMPLETION TIME 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 4 hours 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> PBAPS UNIT 3 3.1-9 Amendment No. 214

Rod Pattern Control 3.1.6 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Rod Pattern Control LCO 3.1.6 0

uirements o f t

APPLICABILITY:

MODES 1 and 2 with THERMAL POWER s 10% RTP.

ACTIONS

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CONDITION A.

One o r more OP control rods n PBAPS UNIT 3 REQUIRED ACTION

.----.I..- NOTE..----.--

A. 1 Rod worth minimizer (RWM) may be bypassed as allowed by LCO 3.3.2.1, "Control Rod Block I ns t rumen t a t i on.

Move associated control rod(s) t o correct p o s i t i on.

OR A.2 Declare associated control rod( s) inoperable.

3.1-18 COMPLETION TIME 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 8 hours (continued)

Amendment No. 214

Rod Pattern Control 3.1.6 ACTIONS (continued)

CONDITION B.

Nine or more OPERABLE control rods no compl i ance wit REQUIRED ACTION Suspend withdrawal o f control rods.

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

COMPLETION TIME Immediately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SURVEILLANCE REQUIREMENTS SR *3.1.6.1 Ver OPERABLE control rods comply 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> wit PBAPS UNIT 3 3.1-19 Amendment No. 214

Control Rod Block Instrumentation 3.3.2.1 i

ACTIONS CONDITION C.

(continued)

D.

RWM inoperable during reactor shutdown.

PBAPS UNIT 3 REQUIRED ACTION C.2.1.1 Verify zr 12 rods withdrawn, C.2.1.2 Verify by admi n i s t r a t i ve met hods that st a r t up w i t h RWM inoperable has not been performed i n the l a s t cal endar year.

AND C.2.2 Verify movement o f control rods is in' c

other qualified member o f the technical. staff.

Verify movement of control rods is accordance wi t h by a second 1 ic operator o r other qua1 i fied member of the technical staff.

COMPLETION TIME Immediately Immediately During control rod movement During control rod movement (continued) 3.3-17 Amendment No. 214

Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE I

Perform CHANNEL CALIBRATION.

SR 3.3.2.1.6 V e r i f y the RWM i s not bypassed when THERMAL POWER i s s 10% RTP.

SR 3,3.2.1.7 NOTE-------------------

Not required t o be performed u n t i l 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> a f t e r reactor mode switch i s i n the shutdown posi ti on.

Perform CHANNEL FUNCTIONAL TEST.

SR 3.3.2.1.8 V e r i f y control rod sequenc RWM are i n conformance w i t FREQUENCY 24 months 24 months 24 months P r i o r t o decl a r i ng RWM 0 PE RAB L E fol 1 owing loading o f sequence i n t o RWM PBAPS UNIT 3 3.3-20 Amendment No. 234

Control Rod Testing-Operating 3.10.7 3. 10 SPECIAL OPERATIONS 3.10.7 Control Rod Testing-Operating LCO 3.10.7 The requirements o f LCO 3.1.6, "Rod Pabbern Control, may suspended to a1 low performance of SDM demonstrations, control rod scram time testing, control rod friction testing, and the Startup lest Program, provided:

b.

The RWM is bypassed; the requirements of LCO 3.3.2.1, "Control Rod Block Instrumentation, " Function 2 are suspended; and conformance to the approved control rod sequence for the specified test is verified by a second 1 icensed operator or other Qua1 i f i ed member of the technical staff.

APPLICABILITY:

MODES 1 and 2 with LCO 3.1.6 not met.

ACTIONS COND I TI ON A.

Requirements o f the LCO not met.

REQUIRED ACTION A. 1 Suspend performance of the test and exception to LCO 3.1.6.

COMPLETION TIME Immediately PBAPS UNIT 3 3.10-18 Amendment No. 214

SDM Test -Refuel i n g 3.10.8 I

3.10 SPECIAL OPERATIONS 3.10.8 SHUTDOWN MARGIN (SDM) Test -Refuel i n g LCO 3.10.8 The reactor mode switch p o s i t i o n specified i n Table 1.1-1 f o r MODE 5 may be changed t o include the startup/hot standby position, and operation considered not t o be i n MODE 2, t o allow SDM testing, provided the following requirements are met:

a.

LCO 3.3.1.1, "Reactor Protection System Instrumentation," MODE 2 requirements f o r Functions 2.a, 2.d and 2.e o f Table 3.3.1.1-1; requirements o f he control rod s

/-

OR

2.

Conformance t o the approved control rod sequence f o r the SDM t e s t i s v e r i f i e d by a second licensed operator o r other qua1 i f i e d member o f the technical s t a f f ;

c.

Each withdrawn control rod shall be coupled t o the associated CRO;

d.

A l l control rod withdrawals during out o f sequence control rod moves shall be made i n notch out mode;

e.

f.

No other CORE ALTERATIONS are i n progress; and CRD charging water header pressure 2 940 psig.

APPLICABILITY:

MODE 5 with the reactor mode switch i n startuplhot standby posit ion.

PBAPS UNIT 3 3.10-20 Amendment No. 234

ENCLOSURE 3 PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 TECHNICAL SPECIFICATION BASES PAGES MARKUP OF PROPOSED CHANGES Revised Bases Pages Unit 2 -

B 3.1-17 B 3.1-18 B 3.1 -1 9 B 3.1-21 B 3.1-35 B 3.1-36 B 3.1-37 B 3.1-38 B 3.3-48 B 3.3-49 B 3.3-56 B 3.10-33 Unit 3 -

B 3.1 -1 7 B 3.1 -1 8 B 3.1-19 B 3.1-21 B 3.1 -35 B 3.1 -36 B 3.1-37 B 3.1 -38 B 3.3-49 B 3.3-50 B 3.3-57 B 3.1 0-33

i i

Control Rod OPERABILITY B 3,1,3 BASES ACTIONS b.1. A.2. A.3, (continued) stuck position and the highest worth OPERABLE control rod assumed to be fully withdrawn, The allowed Goupletion Time o f 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to verify SM i s adequate, considering that with a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable o f providing the required scram and shutdown reactivity. Failure to reach MODE 4 is only likely if an additional control rod adjacent to the stuck control rod also fails ta insert during a required scram, Even with the postulated additional single failure o f an adjacent cuntrol rod to insert, sufficient reactivity reach and maintain lrtODE 3 conditions LL 01 remains to Giith two or mre withdrawn control rods stuck, the plant must be brought to WE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The occurrence o f more than one control rod stuck at a withdrawn position increases the probabi 1 i ty that the reactor cannot be shut dawn if required. Insertion o f all insertable control rods eliminates the possibility o f an additional failure o f a letion Tim o f With one or more control rods inoperable for reasons other than being stuck tn the withdrawn position, (including a control rod which is stuck in the fully inserted position) operatfon may continue, provided the control rods are fully inserted within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and disarmed (electrically or hydraulically) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be hydraulically disamed by closing the drive water and exhaust water isolation valves. The control rods can be el ectri cat 1 y di sarmed by di scannect i ng power from all four directional control valve solenoids. Required Action C.1 i s modified by a Note, which allows the RWM to be bypassed Of required to allow insertton o f the inoperable k o n t inued ).

PBAPS UNIT 2 I& 3.1-17 Revision No. 2

Control Rod OPERABILITY 8 3.1.3 BASES ACTIOMS c.1 and C.2 (continued) control rods and contfnued operation. LCO 3.3.2, I provides additional requirements when the RW i s bypassed to ensure compliance with the CROA analysis, The allowed Completion TJmes w e reamable, considering t h e small number o f allowed inoperable control rods, and provide time to Insert and disarm the control rods i n an orderly manner and without chal 1 enging pl ant systems w

i acceptable, considering the low probability o f a CRDA occurring.

If any Required Action and associated Completion T h e o f Condition A, C, or D are not Met, or there are nine or more inoperable control robs, the piant must be brought to a MODE i n which the LCO does not apply. To achieve thfs status, the plant must be brought to MORE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. ThSs ensures all insertable control rods are inserted and places the reactor in a condition that does not require the active function (i.e.s scram) o f the control rods. The number o f control rods permitted to be inoperable when operating above 10% KTP (e.ga9 no CRDA considerations) could be more than the value specified, but the occurrence o f a large number o f k o n t i nuedl PBAPS UNIT 2 B 3.1-18 Revision ffo. 2

Control Rod OPERABILITY 6 3.1.3 BASES ACTIONS (continued)

I f

inoperable control rods could be indicat9ve o f a generic problem, and investigation and resolution o f the potential problera should be undertaken. The allowed Completion Time o f 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> i s reasonable, based on operating experience, to reach WDE 3 froa full power in an orderly manner and without chal 1 enging plant systems.

SURVEI LLARlCE SR 3.1.3.1 REQUIREMEMS The position o f each control rod amst be determined t o ensure adequate information on control rod position i s availablsr to the operator for deteninCng control rod OPEWILITY and conroll ing rod patterns.

Control rod posftion m y be detemined by the use of OPERABLE position indicators, by moving control rods to a positlon with an OPERABLE indicator, or by the use o f other appropriate methods. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> frequency o f this SR is based on opwatlng experience related to expected changes IOn control rod position and the availability o f control rod posttion indicatjons i n the control room.

Control rod insertion capabitity t s demnstrated by fnserting each partially or fully withdrawn control rod at hast one notch and observing that the contral rod mves.

The control rod may then be returned to i t s origlnal position. This ensures the control rod i s not stuck and i s movement and considering the large testing sample o f SR 3.1.3.2.

Furthemre, the 31 day Frequency takes into account operating experfence related to changes i n CRD performance. A t any time, f f a control rod is imtnovable, a (continued)

PBAPS UNIT 2 B 3.1-19 Revision No. 0

Control Rod OPERABILITY 8 3.1.3 BASES i'

SURVEILLANCE SR 3.1.34 (continued)

REQUIREMENTS t o the "full out" position during the performance o f SR 3,1,3.2.

This Frequency i s acceptable, consldering the low probabilfty that a contriol rod will become uncoupled when it is not being to uncoupl ing events ved and operating experience re1 ated REFERENCES 1,

UFSAR, Secttons l. 5 J. l and 1.5.2-2.

2.

UFSAR, Section 14.6.2.

3.

UFSAR, Appendix K,Section VI.

UFSAR, Chapter 14.

PRAPS UNIT 2 B 3.1-21 Revision No

Rod Pattern Control 8 3.1.6 RodOpattem control satisffes Criterion 3 of the NRC Policy Statement.

\\

APPLICABILITY In MHIES 1 and 2, when THERMAL POWER is s 10% RTP, the CROA Js a Design Basis Accldent and, therefore, compliance with the assumptions o f the safety analysts i s required. When THERMAL WER i s > 10% RTP, there i s no credfble control rod configuration that results in a controll rod worth that could exceed the 280 cat/@ fuel danrage l i m i t during a CRDA (Ref. 2).

I n MOES 3, 4, and 5, sJnce the reactor 9s shut down and only a single control rod can be withdrawn from a core cell containing fuel asserabl ies, adequate SOH ensures that the consequences o f a CWA are acceptable, since the reactor wSll remain subcritical with a slngle control rod w i t hdrawn.

(continued)

PBAPS UNIT 2 8 3.1-35 Revision No. 0

i BASES (continued)

Rod Pattern Control B 3.1.6 ACTIONS cooling water transient, leaklng scram valves, or a pouer reduction to s 1oX RTP before establtshing the correct control rod pattern.

The number o f OPERABLE control rods not I n compliance wfth the prescrfbed sequence i s lfraited t o elght, to prevent the operator from attempting to correct a control rod pattern that signifkantly deviates frola the prescribed sequence.

When the control rod pattern i s not in cow1 i m e with the prescribed sequence, a1 1 control rod movement must be stopped except for wves needed t o correct the rod pattern, or stram if warranted, Required ActSon A.1 i s modtfied by a Note which allows the RUM t o be bypassed t o atlow the affected control rods to be returned to their correct positim.

LCQ 3.32.1 requires verification o f control rod movement by a second licensed operator or a qualified member o f the technical staff (Le.,

personnel traJned i n accordance wtth an approved training by Required Action A.2, The allowed o f 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> i s reasonable, considering the restrfctions on the number of allawed out o f sequence control rods and the low probabilfty o f a CROA occurring during the tiore the control JikLmuu If nfne or mre OPERABLE control rodsare the control rod pattern significantly dew prescribed sequence. Control rod withdrawal should be suspended 4mediately to prevent the potential for further deviation from the prescribed sequence.

Control rod insertion to correct control rods withdrawn beyand thefr:

allowed position is allowed since, i n general, insertion o f i

PBAPS UNIT 2 B 3.1-36 Reviston No. 0

Rod Pattern Control B 3.1.6 MSES ACTIONS B.1 and 6.2 (eantfnued) control rods has less inpact on control rod worth than wfthdrawals have, Required Actlon B.1 i s modlfied by a ?dote which allows the RWM t o be bypassed t o allow the affected contra1 rods to be returned to thejr correct position.

LCO 3.3.2.1 requires verfficatton o f control rltd movement by a second licensed operator or a qualffled aien&er o f the technical staff.

control Fods are not i n reactor mode swftch must "be placed or i s shut dam, and as such, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is reasonable to allow stwe cantpl f ance, and f s on withtn 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, tilth the made requtremnts of this LCO, appropriate relativs to the lorr probabjlity of a CROA occurring wfth the control rods out o f sequence, i

SURVE f LLANCE REQUIREHENTS pattern i s verifi

3.

LCFSAR, Sectlon 14,6.2.3.

4.

5.

10 CFR 100.11.

~ U R E 6 - 0 8 ~ ~,

Section 15.4-9, R w ~ s I o ~ 2, July 1981, cont f w e d l B 3.1137 RevDslon No. 0 PBAPS UNIT 2

Rod Pattern Control B 3.1.6 BASES REFERENCES

6.

NaW)-21778-A, "Transient Pressure Rfses Affected Fracture Toughness Requirements far Boil Ing Water Reactors, December 1978.

(continued)

PBAPS UNIT 2 B 3.1-38 Revision No. 0

Control Rod B1 ock Instrumentat 3 on B 3.3*2*1 BASES APPL I CABLE SAFETY ANALYSES, (cont i nued)

LCO, and APPLICABILITY The RM i s assumed to mitigate the consequences o f an RWE event when operatfng ;k 30% RTP.

Below this power level, the consequences of an RUE event will not exceed the KPR SL and, therefore, the RSM is not regutred to be OPERABLE (Ref. 1). When operating < 90% RTP, analyses (Ref. I) have shown that with an initial MCPR 2: 1.70, nu RUE event will result in exceeding the KPR SL.

Also, the analyses demonstrate that when operating at z 90% RTP wSth WPR L 1.40, no WE event will result fn exceeding the MCPR Sl. (Ref. 1). Therefore, under these conditions, the RW i s also not required to be OPERABLE.

f The RWM Function satisfies Criterion 3 of the NRC Policy S t a temn t.

Since the Rid! is a hardwfred system desSgned t o act as a backup to operator contrml o f the rod sequences, only one channel o f the RW is avatlable and requfred to be OPERABLE ircwtances provided f o r i n the CO 3.1.3, Controt Rod OPERABfLITY, and sitate bypassfng the RWM to allow n with inoperable control rods, or to f a control rod pattern not i n compliance RW may be ttypassed as required by these St must be considered inoperable and o f this CCO followed.

(cant 5 nued 1.

PBAPS UNIT 2 B 3.3-48 Revision No. 0

Control Rod Block Instrumentatton B 3.3,2,1 BASES i

APPlICABLE SAFETY ANALYSE therefore OPERABILITY o f the is > 10% RTP, there i s no ion that results i n a control 286 cal/gm fuel damage limit In MOOES 3 and 4, all inserted into the are; In MOOE 5, sfnte only a awn fran a care cell ate S M ensures that the able, since the reactor 1 and 2 when fHERM&L POWER 3s During MODES 3 and 4, and during MODE 5 when the reactor mode swjtch is required to be in the shutdown position, the core is assumed to be subcritfcal; therefore, no positfve reactivity insertion events are analyzed. The Reactor Mode Swi tch-Shutdown Position contra1 rod WS thdrawa7 block ensures that the reactor remains subcritical by blocking cointrol rod uithdrauat? thereby preserving the assumptions o f the safety analysis satisfies Criterion 3 o f the RRC Polfcy Statement.

ctor M e Swi tch-Shutdown Posit Two channds are rsqufred t o be OPERABLE to ensure that no single channel failure will preclude a rad block when required. There Js no Altwable Value for this Function since the channels are mchanically actuated based solely on reactor mode switch posWon.

During shutdm condWons (HOD 3, 4, or 51, no posltive reactivity insertion events are analyzed because assumptions are that control rod wfthdrawal blocks are provided to prevent criticality. Therefore, when the reactor mode switch i s in the shutdown position, the control rod wfthdrawal block i s required t o be 0PERkBt.E.

During MODE 5 with the reactor W e switch i n the refueling posttion, the refuel positfan one-rod-out Snterlack (LCO 3.9.2? "Refuel Position One-Rod-Out Interlock") provfdes the required control rod withdrawal blocks.

(continued)

\\

PSAPS UNIT 2 8 3.3-49 Revisfan No. 0

I i Control Rod Block Instrumentation B 3.3e2.1 BASES SURVEI LLAMCE SR 3.3.2.1.?

(continued)

REQUIREMENTS The 24 month Frequency is based on the need to perfom this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance w e r e performed with the lceactor at powe!r.

Operating experience has shown these components w i l l pass the Surveillance when perfomed at the 24 month Frequency.

SR 3.3.2.1.8 The RWM will only enforce the pLtoper control rod sequence if the rod sequence i s 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, Suxveilfance is performed once prior to declaring FWM OPERABLE following loading of sequence into RWM, since this is when rod sequence input errors are possible.

The REFERENCES 1,

NEDC-32162-P, "Maximum Extended Load Line L i m i t and ARTS Implrovemrsnt Program A n a l y s i s for Peach Bottom Atomic Power Station, Units 2 and 3," Revision 1, February 1993,

(

3.20.3.4.8 and 7.

nModifications to the Requirements for Control Rod Drop Accident Mitigating SyStGRiS," BWR Owners' Group,

6.

NRC SER, "Acceptance of Referencing of Licensing Topical Report NEDE-24013-P-A, @'

"General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17," December 27, 1987.

(continued)

PBAPS UNIT 2 B 3,3-56 Revision No. 0

BASES LCO second licensed o (continued) technical staff.

an Inadvertent cri not confom t o the specified i n CCO 3 sequance control rod withdrawals] must be made i n the individual notched withdrawal der t o minlmire the potential reactlvlty insertion associated w i t h each mvetmnt.

Coupling integrity o f withdrawn control rods is required t o rsintrrize the probability o f a CRDA and ensure proper functioning o f the wfthdrawn control rods, if they are required to scram. Because the reactor vessel head may be removed during these tests, no other CORE ALTERATI'ONS may be In progress, Furthemre, since the control rod scram funetion with the RCS at atmospheric pressure relies solely on the CR_D acemfator, i t is essential that the CRD charging water header remain pressurized.

fhfs Special Operatfans LCO then allows changing the Table 1.1-1 reactor mode swltch position rsqutremnts t o include the startup/hot standby posftiorr, such that the SW tests may be performed whtler i n WOOE 5.

t APPLICABILITY These SLIM test Special Operations requirements are only applicable if the SDM tests we t o be perfomred while i n MOO 5 with the reactor vessel head renroved or tbe head With one or mre contra1 rods dfscovered uncaupfed during thfs S e d a l Operation, a centralled fnsertion o f each uncoup r ed control rod 1s required; efther t o atteapt recoupling, or t o preclude a control rod drop, Thls controlled insertion i s preferred since, if the control rod fails t o follow the drive as lt is withdrawn ( f a., 1s "stuck" i n an inserted positlon), placing the reactor mode switch i n the shutdawn posltion per Requlred Action B.1 could cause substantial secondary damage.

If recoup1 ing i s not accorrplished, operation may continue, provided the control rods are fully fnserted withtn 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and disarmed (electrically OF hydraulically) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Inserting a

[cont 1 ntredl PBAPS UNIT 2 8 3.10-33 RevCsion No, 0

Control Rod OPERABILITY B 3.1.3 BASES ACTIONS A.1.

A.2.

A,3, and A.4 (continued) stuck position and the highest worth OPERABLE control rod assumed to be fully withdrawn, The allowed Completion T i m o f 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to verify SDM i s adequate, considering that with a single control rod stuck in a withdrawn position, the remaining OPERAIBLE: control rods are capable of providing the required scrai and shutdown reactivity. Failure to reach WOE 4 is only likely if an additional control rod adjacent to the stuck control rod also fails to insert during a required scram. Even with the postulated additional single failure o f an adgacent control rod to insert, sufficient reactivity reach and maintain MDE 3 conditions u

With two or more withdrawn control rods stuck,. the plant must be brought to MODE: 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, The occurrence o f more than one control rod stuck at a wlthdrawn position increases the probability that the reactor cannot be shut down if required, Insertion o f all insertable control rods eliminates the possibility o f an addftional failure o f a control rod to insert. The allowed Completion Time o f sonable, based on operatfn om full power conditions i 1 1 eng i ng plant sys terns.

C.1 and C.2 With one or more control rods inoperable for reasons other than being stuck in the withdrawn position (including a control rod which is stuck in the fully inserted position) operation may continue, provided the control rods are fully inserted within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and disarmed (electricalfy or hydraulically) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected. The control rod i s disarmed to prevent inadvertent withdrawal during subsequent operations, The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves. The control rods can be electrical Jy di sarmed by di sconnect i ng power from all four directional control valve solenoids. Required Action C.1 is modified by a Note, which allows the RWM to be bypassed if required to allow insertion o f the inoperable kont inuedl PBAPS UNIT 3 B 3.1-17 Revision No, 2

Control Rod OPERABILITY 8 3.1.3 BASES ACTIONS

-2 (continued) control rods and cont i nued operat 5 on additional requirements when the RklFTa is bypassed to ensure compljance with the CRDA analysis.

The allowed Completion Times are reasonable, considering the small number o f allowed inoperable control rods, and provide time to insert and disarm the control rods i n an orderly manner and without chal 1 eng i ng pl ant sys terns.

J 2 % L m u u LCO 3.3.2.1 provides restore the control rods to acceptable, considering the low probability o f a CRDA occurring.

If any Required Action and associated Completion Time o f Condition A, C, or D are not met, or there are nine o r more inoperable control rods, the plant must be brought t o a MODE in which the LCO does not apply, To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, This ensures all insertable control rods are inserted and places the reactor in a condition that does not require the active function (Le., scram) o f the control rods. The number o f control rods permitted to be inoperable when operating above 10% RTP (e.g., no CRDA considerations) could be more than the value specified, but the occurrence o f a large number o f kont hued1 i

PBAPS UNIT 3 8 3.1-18 Revision No. 2

Control Rod OPERABILXTY B 3.1.3 BASES ACTIOMS (continued)

Inoperable control rods could be indicative o f a generic probtem, and investigation and resolution o f the potential problem should be undertaken, The allowed Completion T.frae o f 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging plant system, SURVEl LLANCE SR 3.1.3.1

~ E Q U ~ ~ E ~ E N T S The position o f each control rod must be determined to ensure adequate information on control rod position is available to the operator for determining control rod OPERABfLffY and cantrolling rod patterns. Contra1 rod position may be determined by the use o f OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use o f other appropriate methods.

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency o f this SR is based on operating experience related to expected changes in control rod position and the availability o f control rod position Indications in the control room.

apabil i ty 5s demonstrated by ially or fully withdrawn control rod at least one notch and observing that the control rod moves.

The control rod may then be returned to its original position. This ensures the control rod i s not stuck and i s m o v ~ ~ n t and consfdering the large testtng sample o f SR 3.1.3.2.

Furthemore, the 31 day Frequency takes into account operating experience related to changes in CRD performance. A t any tim, if a control rod i s imovable, a.

f con t I nued 1 B 3,149 Revision No. 0 PBAPS UNIT 3

Control Rod OPERABILITY B 3.1.3 BASES SURVEILLANCE SR 3.1.3.3 (continued)

REQUIREMENTS t o the "full out" position during the performance o f SR 3.1.3.2.

This Frequency i s acceptable, considerlng the law probability that a control rod will become uncoupled when it is not being nowed and operating experience related to uncoupt ing events.

REFERENCES

1.

UFSAR, Sections 1.5-1.1 and 1.5.2.2,

2.

UFSAR, Section 14.6.2.

3.

W A R, Appendtx K,Section VI.

I PBAPS UNIT 3 B 3.1-21 Revision No. 0

Rod Pattern Control B 3.1.6 BASES APPLICABLE SAFnY ANALYSES P

has demonstrated that ot be violated d rods not in compliance Roo pattern control satisf-ies Criterian 3 o f the NRC Policy Statement.

LCO Compl i ance wS th the prescrf bed control rod sequences APPLICABI L ITY In WDES f and 2, when THERMAL POKR i s s 10% RTP, the CRDA i s a Design Basis Accident and, therefore, compliance with the assmptions o f the safety analysis 4s required.

When THERMAL POWER 0s > foX #fP, there f s no credible control rod conffguratfon that results i n a control rod worth that could exceed the 280 cal/gm fuel damage IWt during a CRaA (Ref. 2).

In HOOES 3, 4, and 5, since the reactor i s shut down and only a single control rod can be withdrawn from a core cell containing fuel assembliirs, adequate SOM ensures that the consequences of a CROA are acceptable, since the reactor will renrafn subcritical with a single control rod withdrawn.

(continued)

PBAPS UNIT 3 8 3.1-35 Revision No. 0

I ACTIONS With one compl i ance with the ens-may be taken to n or declare the associated control rods inoperable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

N6ncowpliance wjth the prescribed sequence niay be the result o f double notching, drifting frm a control rod drive coaling water transient, leaking scraar valves, or a power reduction t o s 10% RTP before establishing the correct control rod pattern.

The nurarber o f OPERABLE contra1 rods not fn compliance with tbe prescribed sequence f s limited to eight, t o prevent the operator from attempting to correct a control rod pattern that significantly deviates from the prescrtbed sequence.

Yhen the contra1 rod pattern i s not i n compl iance with the preseri bed sequence, a1 1 control rod movement must be stopped except for moves needed t o correct the rod pattern, or scraa if warranted.

Required Action A.1 4s modified by a Note which allows the Rldn t o be bypassed t o allw the affected control rods t o be returned to their correct position.

LCO 3.3.2.1 requires verificatlon o f control red nrovement by a second licensed operator or a qualified member o f the technical staff (i.e.,

A.2.

The allowed ble, considering the restrictions on the nusrber o f allowed out o f sequence control rods and the low probability o f a CROA occurrlng durtng the time the control rods are out o f sequence.

B.1 and 8 4 If nine or more OPERABLE cont the control rod pattern significantly devi prescribed sequence. Control rod withdrawal should be suspended inrmediately ta prevent the! potential for further deviation from the prescribed sequence.

Control rod insertion t o correct control rods withdrawn beyond their allowed position i s allowed since, i n general, insertion o f I cant.i nued 1 PBAPS UNIT 3 8 3.1-36 Revision No. 0

Rod Pattern Control B 3.1.6 BASES ACTIONS

@.I and S.2 (continued) control rods has less impact on control rod worth than withdrawals have. Required ActIon B.1 i s modffied by a Note which allows the RWM to be bypassed t o allow the affected control rods to be returned to their correct position.

Leo 3.3.2.1 requires vertficatfon o f control rod movemnt by or a qua1 i fied member a f the control rods are not i n e reactor mode switch must be placed ctor i s shut down, and as such, lity requirements o f this LCO.

wfthin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

With the mode o f 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> i s reasonable to allow insertion o f control rods to restore coarpliance, and i s appropriate relative to the low probability o f a CRDA occurring w4th the control rods out o f sequence.

SURVEILLANCE SR 3.1.6.1

~ E Q U ~ R E ~ N T S.

3.

UFSAR, Sectfon 14.6.2.3.

4.

NUREG-0800, SectIan 15.4.9, Reviston 2, July 1981.

-%L..+u---

5.

10 CFR 100.11, lcont i nued ),

PBAPS UNIT 3 B 3.1-37 Revision No. 0

Rod Pattern Control B 3.1.6 BASES REFERENCES

6.

NEDO-21778-A, "Transient Pressure Rises Affected Fracture Teughness Requirements f o r Boil ing Water Reactorss December 1978, ASME, Boiler and Pressure Vessel Code, (continued)

PBAPS UNIT 3 B 3.1-38 Revision No. 0

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 1,

Rod Block Monitor (continued)

SAFEJY ANALYSES,

LCO, and APPLICABILIJY The RBM i s assumed to mitigate the consequences o f an RWE event when operating z 30% RTP. Below this power level, the consequences o f an RWE event will not exceed the flCPR SL and, therefore, the RBM i s not required to be OPERABLE (Ref. 1).

When operating < 90% RTP, analyses (Ref. 3) have shown that with an initial MCPR 2 1.70, no RWE event will result in exceeding the MCPR S1.

Also, the analyses demonstrate that when operating at L 90% RTP with MCPR L 1.40, no RWE event will result in exceeding the MCPR SC (Ref. 1). Therefore, under these conditions, the RBM is also not required to be OPE

2.

Rod Worth Minimizer ed.

The anal 1 methods and re summarized in e RWM Function satisfies Criterion 3 o f the NRC Policv U

S t at emen t.

f a controi rod pattern not in-compliance e RWM may be bypassed as required by these en it must be considered inoperable and the Required Actions o f this LCO followed.

Since the RWM is a hardwired system designed to act as a backup to operator control o f the rod sequences, only one channel of the RWM is available and required to be OPERABLE (Ref. 6).

Special circumstances provided f o r in the Required Action o f LCO 3.1.3, "Control Rod OPERABILITY," and LCO 3.1.6 may necessitate bypassing the ROJM to allow continued operation with inoperable control rods, or to (continued)

PBAPS u m 3 B 3.3-49 Revision No. 3

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE

2. Rod Worth Minimizer (cantinued)

SAFETY ANALYSES, and therefore OPERABILITY o f the 1 and 2 when THERMAC POWER is POWER 1s > 10% RTP, there is no In MODES 3 and 4, all ration that results i n a control the 280 cal/ga fuel damage 'tinit during a CRDA (Refs. 4 and 6).

control rods are required to be inserted into the core; therefore, a CRDA cannot occur.

In HODE 5, since only a stngle control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the cansequences o f a CRDA are acceptable, since the reactor will be subcritical.

3, Reactor Mode Switch-Shutdown Position

//

\\--".-:

___ r" During MODES 3 and 4, and during MODE 5 when the reactor mode swftch is requlred to be i n the shutdown position, the core is assumed to be subcr9tical; 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 analys i s.

The Reactor Mode Switch - Shutdown Positbn Function satisfies Criterion 3 o f the WRC Pol icy Statement.

Two channels are required to be OPERABLE to ensure that no singte channel failure wilt 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 critical ity. Therefore, when the reactor mode switch i s in the shutdown position, the control rod withdrawal block is required to be OPERABLE.

During M O E 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.

(cont.5 nued)

PBAPS UNIT 3 6 3-3-50 Revision No. 3

Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2..1.7 (continued)

REQU I REMENTS The 24 month Frequency is based on the need to perform this Surveillance under the condftfons that apply during a plant outage and the potential f o r an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience has shown these components will pass the Surveillance when performed at the 24 month Frequency.

The RWM will only enforce the proper control rod sequence if the rod sequence i s properly input into the RM computer.

This SR ensures that the proper sequence i s loaded into the RWM so that it can perform its intended function, The Surveillance i s performed once prior to declaring RWM OPERABLE following loading of sequence into RW, since this

' is when rod sequence fnput errors are! possible.

REFERENCES

1.

NEDC-32162-P, nMaximum Extended Load Line Limit and ARTS Improvement Program Analysis for Peach Bottom Atomic Power Station, Units 2 and 3," Revision 1, February 1993.

7.10.3.4.8 and 7.16.3.

-,..&&.U.+--.L-'-*s=l ir, ements for Control Rod terns, " BWR Owners' Group,

5.

NEM)-23231, "Banked Position Withdrawal Sequence, January 1977.

6.

EIRC SER, "Acceptance o f Referencing o f Licensing Topical Report NEDE-24011-P-A, " "General Electric Standard Application f o r Reactor" Fuel Revision 8,

~ n d m ~ n t 17," December 27, 1987.

t con t i nued 1 PBAPS UNIl 3 6 3.3-57 Revision No. 3

SDM Test --Refuel i ng B 3.10.8 BASES an inadvertent cri not conform to the specified in CCO 3 sequence control rod withdrawals) must be made in the individual notched withdrawal mode to miniinize the potential reactivity insertion associated with each movement.

Coupling integrity o f withdrawn control rods is required to minimize the probability o f a CROA and ensure proper functioning of the withdrawn control rods, if they are required to scram. Because the reactor vessel head may be removed during these tests, no other CORE ALTERATIONS may be in pragress, Furthemre, since the control rod scram function with the RCS at atmospheric pressure re1 ies solely on the CRD accumulator, it.is essential that the CRD charging water header remain pressurized. Jhis Special Operations LCO then allows changing the Table 1.1-1 reactor mode switch position requirements to include the startuplhot standby position, such that the SDEvl tests may be performed while in WOE 5.

APPLICABILITY These SDlvl test Special Operations requirements are only applicable if the SDW tests are to be performed while in MODE 5 with the reactor vessel head removed ar the head MODES are unaffected by this LCO.

ACTIONS A. l and A.2 With one or more control rods discovered uncoupled during this Special Operation, a controlled insertion o f each uncoupled control rod i s required; either to attempt recoupling, or to preclude a contra1 rod drop, This controlled insertion is preferred since, if the control rod fails to fallow the drive as it is withdrawn ( L e o s is "stuck" i n an inserted position), placing the reactor mode switch in the shuCdawn position per Required Action B.1 could cause substantfa1 secondary damage.

If recoup1 ing is not accompl ished, operation may continue, provided the control rods are fully inserted within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and disarmed (electrically or hydraulically) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Inserting a (continued)

PBAPS UNIT 3 8 3.10-33 Revision No. 0