Text

License Amendment Request 264, Revision 2: Application to Modify Improved Technical Specifications Regarding Steam Generator Tube Integrity and Response to Request for Additional Information
	ML070750085
Person / Time
Site:	Crystal River
Issue date:	03/14/2007
From:	Roderick D; Progress Energy Co
To:	; Document Control Desk, NRC/NRR/ADRO
References
	3F0307-07, TAC MD2054
	Download: ML070750085 (90)
	v • d • e

Progress Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 Ref: 10 CFR 50.90 March 14, 2007 3F0307-07 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

References:

Crystal River Unit 3 -

License Amendment Request #264, Revision 2:

Application to Modify Improved Technical Specifications Regarding Steam Generator Tube Integrity and Response to Request for Additional Information (TAC No. MD2054)

1. Crystal River Unit 3 to NRC Letter dated May 25, 2006, "Crystal River Unit 3

- License Amendment Request #264, Revision 0: Application to Modify Improved Technical Specifications Regarding Steam Generator Tube Integrity"

2. Crystal River Unit 3 to NRC Letter dated December 21, 2006, "Crystal River Unit 3 - License Amendment Request #264, Revision 1: Application to Modify Improved Technical Specifications Regarding Steam Generator Tube Integrity and Response to Request for Additional Information"

3. NRC to Crystal River Unit 3 Letter dated February 23, 2007, "Request for Additional Information RE: License Amendment Request to Implement TSTF-449, Crystal River Nuclear Plant, Unit 3 (TAC No. MD2054)"

Dear Sir:

On February 23, 2007, the Nuclear Regulatory Commission (NRC) issued a Request for Additional Information (RAI) regarding the license amendment request to modify the Crystal River Unit 3 (CR-3) Improved Technical Specifications (ITS) regarding steam generator tube integrity (Reference 3). In accordance with the provisions of 10 CFR 50.90, Florida Power Corporation (FPC), doing business as Progress Energy Florida, Inc., hereby provides Revision 2 to License Amendment Request #264 and the response to the RAI. The proposed amendment is consistent with NRC-approved Revision 4 to Technical Specification Task Force (TSTF)

Standard Technical Specification Change Traveler TSTF-449, "Steam Generator Tube Integrity." The availability of this ITS improvement was announced in the Federal Register on May 6, 2005 (70 FR 24126) as part of the consolidated line item improvement process. This submittal replaces Reference 2, License Amendment Request #264, Revision 1, in its entirety.

Progress Energy Florida, Inc.

Crystal River Nuclear Plant 15760 W. Powerline Street Crystal River, FL 34428

U.S. Nuclear Regulatory Commission 3F0307-07 Page 2 of 3 Attachment A provides the FPC response to the RAI. Attachment B provides a description of the proposed change and confirmation of applicability. Attachment C provides the existing ITS pages marked-up to show the proposed change, and Attachment D provides those same changes presented more formally with revision bars. Attachments E and F provide similar formats for the related Bases sections.

FPC requests approval of the proposed license amendment by May 1, 2007, with the amendment to be implemented within ninety days of issuance.

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

This letter establishes no new regulatory commitments.

The CR-3 Plant Nuclear Safety Committee has reviewed this request and recommended it for approval.

If you have any questions regarding this submittal, please contact Mr. Paul Infanger, Supervisor, Licensing and Regulatory Programs at (352) 563-4796.

Daniel L. Roderick Director Site Operations Crystal River Nuclear Plant DLR/dar Attachments:

A.

B.

C.

D.

E.

F.

Request for Additional Information Response Description and Assessment Proposed Improved Technical Specification Changes (Mark-up)

Proposed Improved Technical Specification Changes (Revision Bar Format)

Proposed Improved Technical Specification Bases Pages (Mark-up)

Proposed Improved Technical Specification Bases Pages (Revision Bar Format) xc:

NRR Project Manager Regional Administrator, Region II Senior Resident Inspector State Contact

U.S. Nuclear Regulatory Commission 3F0307-07 Page 3 of 3 STATE OF FLORIDA COUNTY OF CITRUS Daniel L. Roderick states that he is the Director Site Operations, Crystal River Nuclear Plant for Florida Power Corporation, doing business as Progress Energy Florida, Inc.; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the bel-ciis knowledge, information, and belief.

Daniel L. Roderick Director Site Operations Crystal River Nuclear Plant The foregoing document was acknowledged before me this I

day of 2007, by Daniel L. Roderick.

Signature of Notary Publi....

EU.. DPP State of Florida MY COMMISSION #_OD (Print, type, or stamp Commissioned Name of Notary Public)

Personally Known Produced

-OR-Identification

PRORESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #264, REVISION 2 Application to Modify Improved Technical Specifications Regarding Steam Generator Tube Integrity ATTACHMENT A Request for Additional Information Response

U.S. Nuclear Regulatory Commission Attachment A 3F0307-07 Page 1 of 5 Request for Additional Information Response On February 23, 2007, the Nuclear Regulatory Commission (NRC) issued a Request for Additional Information (RAI, Reference 3) concerning Revision 1 of License Amendment Request (LAR) #264 to modify the Crystal River Unit 3 (CR-3) Improved Technical Specifications (ITS) regarding steam generator tube integrity (Reference 1).

Florida Power Corporation (FPC), doing business as Progress Energy Florida, Inc., hereby provides the following response to this RAI. Since these responses involved changing some of the proposed text in Attachments C through F, this reply to the RAI constitutes Revision 2 to LAR #264, and replaces Revision 1, submitted December 21, 2006, in its entirety.

NRC Request

1. In your response to Question 8, you indicated that the second sentence of TS 5.6.2.10.4.a.1l.b was deleted since it refers to inspection categories that were deleted.

Although the inspection categories were deleted, this sentence contained expansion criteria for your inspection that were not captured in your proposal. Please discuss your plans to incorporate this expansion criteria into your proposed TSs (without reference to the inspection categories).

In addition, with respect to proposed TS 5.6.2.10.d.8, please discuss your plans to replace "degradation of a repair roll" with "flaw or flaws in a repair roll" since TSTF replaced the term "degradation" with the term "flaws."

Florida Power Corporation (FPC) Response

1. In LAR #264, Revision 1, CR-3 ITS 5.6.2.10.4.a.11.b (discussed in the question above) was moved to ITS 5.6.2.10.d.8. FPC agrees that when the inspection categories were deleted, a sentence requiring the examination of both Once Through Steam Generators (OTSGs) under certain circumstances was also deleted. FPC also agrees that this expansion criterion should be retained in ITS 5.6.2.10.d.8 even though the inspection category is being deleted.

To reflect this, the text for ITS 5.6.2.10.d.8 in Attachments C and D has been changed to require both OTSGs to be examined when a repair roll flaw(s) is the cause of a plant shutdown. This is now consistent with current requirements. FPC also agrees to replace the term "degradation of" in ITS 5.6.2.10.d.8 with "a flaw in" which is equivalent to the term, "a flaw or flaws in." The requested revision to Paragraph 5.6.2.10.d.8 has been changed to read as follows (the line-outs indicate deleted text and the bold text indicates added text):

"If the plant is required to shut down due to primary-to-secondary leakage and the cause is determined to be o

-ef a flaw in a repair roll, 100% of the repair rolls in AW both OTSGs shall be examined."

FPC feels this text revision remains technically consistent with TSTF-449, Revision 4, and therefore it does not change the description or the assessment presented in Attachment B to this amendment request.

U.S. Nuclear Regulatory Commission Attachment A 3F0307-07 Page 2 of 5 NRC Request

2. In your response to Question 1, you indicated that you would remove the exception to the accident-induced-leakage performance criterion.

Please discuss your plans to remove similar text from your Bases (refer to page B 3.4-78).

FPC Response

2. FPC agrees it is appropriate to remove the exception to the accident induced leakage performance criterion in the Bases on page B 3.4-78. The requested revision to the third paragraph on page B 3.4-78 in Attachments E and F has been changed to read as follows:

"The accident induced leakage performance criterion ensures that the primary to secondary LEAKAGE caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions. The accident analysis assumes that accident

-induced leakage does not exceed one gallon per minute per OTSG, except for specific "yes of degradation at specofi locations where the NRC hasaproe greater accident ind.ed leak-age. The accident induced leakage rate includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident."

FPC feels this text revision remains technically consistent with TSTF-449, Revision 4, and therefore it does not change the description or the assessment presented in Attachment B to this amendment request.

NRC Request

3. In your response to Question 8, you indicated that an additional sentence was added to clarify the repair criteria for intergranular attack (IGA). In reviewing the proposed text, it does not appear to address the repair criteria for all IGA indications in the first span of Steam Generator B. For example, it only addresses the acceptance criteria for those indications in the Inservice Inspection Surveillance Procedure (and there is no requirement to record the indications in this document). In addition, it does not specify the actions to take when these indications satisfy the repair criteria (i.e., the tubes must be plugged or repaired). Please discuss your plans to clarify the repair criteria for all first span IGA indications. The staff notes that the proposed repair criteria appear identical to the standard 40-percent depth based tube repair criteria and could possibly be deleted.

FPC Response

3. Text has been added to ITS 5.6.2.10.c.1 to reflect that action must be taken when indications satisfy the repair criterion. Also, reference made to the OTSG Inservice Inspection Surveillance Procedure has been removed since there is no requirement to record the indications in this document. Similar text exists in the first paragraph of ITS 5.6.2.1O.d.4, so deletion of the same reference in this section will be made, as well. Based on the above, the requested revision to ITS 5.6.2.1O.c.1 in Attachments C and D has changed to read as follows (the line-outs indicate deleted text and the bold text indicates added text):

U.S. Nuclear Regulatory Commission Attachment A 3F0307-07 Page 3 of 5 "Pit-like Intergranular Attack (IGA) indication means a bobbin coil indication confirmed by Motorized Rotating Pancake Coil (MRPC) or other qualified inspection techniques to have a volumetric, pit-like morphology characteristic of IGA. Inservice tubes with pit-like IGA indications in the first span of the B OTSG, idenf*Wed inh*teo OSG.isei..ee Inspe.ti. n Ste.

l e Pro..

dure are acceptable to rema... n in seriee provided the depth of the indication is less than 40% of the nominal tube wall thickness. Inservice tubes with pit-like IGA indications in the first span of the B OTSG with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged."

Similarly, the first paragraph of ITS 5.6.2.1O.d.4 in Attachments C and D has been changed to read as follows:

"Inservice tubes with pit-like IGA indications in the first span of the B OTSGr identified in the OTSG I.se...iee In.pection Surveillance Pr.ed.re must be inspected with bobbin and Motorized Rotating Pancake Coil (MRPC) eddy current techniques from the lower tube sheet secondary face to the bottom of the first tube support plate during each inservice inspection of the B OTSG."

FPC feels this text revision remains technically consistent with TSTF-449, Revision 4, and therefore it does not change the description or the assessment presented in Attachment B to this amendment request.

NRC Request

4. The proposed structure for TS 5.6.2.10.c could result in misinterpreting the requirements since the first sentence indicates that alternate repair criteria are discussed below and the next sentence is not an alternate tube repair criteria. Please discuss your plans to clarify the requirements.

Clarity could be achieved by: (1) numbering the first sentence (as number 1) and replacing "below" with "in Technical Specification 5.6.2.10.c.3"; (2) number the second sentence (as number 2); (3) numbering the third clause (as number 3) and adding "in Technical Specification 5.6.2.10.c.a" at the end of the clause; and (4) renumbering "1" and "2" as "a" and "b."

FPC Response

4. To address this concern, text in ITS 5.6.2.10.c was restructured, but not as described above.

The second paragraph (starting with "Tubes shall be plugged..." and ending with

"...sleeve/tube assembly. ") was moved to the first paragraph after the first sentence of that paragraph. The remaining of what had been the first paragraph (the text starting with "The non-sleeved region of a tube..." and ending with "...an alternate tube repair criteria discussed below. ") has become the second paragraph. No text has been added or deleted.

Reflecting the changes described above, the requested revision to ITS 5.6.2.10.c in Attachments C and D has been changed to read as follows:

"Provisions for OTSG tube repair criteria. Tubes shall be plugged if the sleeved region of a tube is found by inservice inspection to contain flaws in the (a) sleeve or (b) the pressure boundary portion of the original tube wall in the sleeve/tube assembly.

U.S. Nuclear Regulatory Commission Attachment A 3F0307-07 Page 4 of 5 "The non-sleeved region of a tube found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged or repaired except if the flaws are permitted to remain in service through application of an alternate tube repair criteria discussed below.

"The following alternate tube repair criteria may be applied as an alternative to the 40% depth based criteria..."

FPC feels this text revision remains technically unchanged and consistent with TSTF-449, Revision 4. Therefore, this revision does not change the description or the assessment presented in Attachment B to this amendment request.

NRC Request

5. In the second sentence of proposed TS 5.6.2.10.c.2, it would appear that the sentence should read, "provided the combined projected leakage from all sources of primary to secondary leakage." Please discuss your plans to clarify this requirement.

FPC Response

5. ITS 5.6.2.10.c.2 in Attachments C and D has been changed to include the words "sources of" in the second sentence of the first paragraph. The revised sentence now reads as follows:

"Tubes with axially oriented TEC may be left in-service using the method described in Topical Report BAW-2346P, Revision 0, provided the combined projected leakage from all sources of primary-to-secondary leakage, including axial TEC indications left in-service, does not exceed the Main Steam Line Break (MSLB) accident leakage limit of one gallon per minute, minus 150 gallons per day, per OTSG. "

FPC feels this text revision remains technically unchanged and consistent with TSTF-449, Revision 4.

Therefore, this revision does not change the description or the assessment presented in Attachment B to this amendment request.

NRC Request

6. In the first sentence of the second paragraph of proposed TS 5.6.2.10.c.2, you indicate that tubes identified with tube end cracking that meet the alternate repair criteria will be added to the existing list of tubes in the surveillance procedure. In TSTF-449, the phrase "satisfy the tube repair criteria" is used to indicate that an indication exceeds the tube repair criteria (i.e., the tube must be plugged or repaired). However, in the sentence in TS 5.6.2.10.c.2, the phrase "meet the alternate repair criteria" is used to indicate that if a tube is allowed to remain in service (i.e., it does not exceed the alternate repair criteria), then-it should be added to the list of tubes in the surveillance procedure. To ensure consistent terminology throughout this specification, discuss your plans to clarify TS 5.6.2.10.c.2 by indicating that tubes identified with tube end cracking that are allowed to remain in service under the alternate repair criteria will be added to the list.

U.S. Nuclear Regulatory Commission Attachment A 3F0307-07 Page 5 of 5 FPC Response

6. FPC agrees that text should be added to ITS 5.6.2.10.c.2 to clarify that the subject tubes do not just meet the alternate repair criteria, but have been allowed to remain in service under that criteria. The first sentence in the second paragraph of ITS 5.6.2.10.c.2 in Attachments C and D has been changed to read as follows (the line-out indicates deleted text and the bold text indicates added text):

"Tubes identified with TEC that meet are allowed to remain in service under the alternate repair criteria will be added to the existing list of tubes in the OTSG Inservice Inspection Surveillance Procedure."

FPC feels this text revision remains technically unchanged and consistent with TSTF-449, Revision 4. Therefore, this revision does not change the description or the assessment presented in Attachment B to this amendment request.

References:

1. Crystal River Unit 3 to NRC Letter dated December 21, 2006, "Crystal River Unit 3 -

License Amendment #264, Revision 1, Application to Modify Improved Technical Specifications Regarding Steam Generator Tube Integrity and Response to Request for Additional Information"

2. NRC to Crystal River Unit 3 Letter dated February 23, 2007, "Request for Additional Information RE: License Amendment Request to Implement TSTF-449, Crystal River Nuclear Plant, Unit 3 (TAC No. MD2054)"

PRORESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #264, REVISION 2 Application to Modify Improved Technical Specifications Regarding Steam Generator Tube Integrity ATTACHMENT B Description and Assessment

U.S. Nuclear Regulatory Commission Attachment B 3F0307-07 Page 1 of 4 Description and Assessment

1.0 INTRODUCTION

The proposed license amendment revises the requirements in the Crystal River Unit 3 (CR-3) Improved Technical Specification (ITS) related to steam generator tube integrity.

The changes are consistent with NRC approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity," Revision 4. The availability of this technical specification improvement was announced in the Federal Register on May 6, 2005 as part of the consolidated line item improvement process (CLIIP). This request replaces License Amendment Request

264, Revision 1, submitted December 21, 2006, in its entirety.

2.0 DESCRIPTION

OF PROPOSED AMENDMENT Consistent with the NRC-approved Revision 4 of TSTF-449, the proposed ITS changes include:

" Revised ITS definition of LEAKAGE

" Revised ITS 3.4.12, RCS [Reactor Coolant System] Operational Leakage

" New ITS 3.4.16, Steam Generator (OTSG) Tube Integrity

" Revised ITS 5.6.2.10, Steam Generator (OTSG) Program

Revised ITS 5.7.2.c, d, and e, Steam Generator Tube Inspection Report(s)

Proposed revisions to the ITS Bases are also included in this application. As discussed in the NRC's model safety evaluation, adoption of the revised ITS Bases associated with TSTF-449, Revision 4 is an integral part of implementing this ITS improvement. The changes to the affected ITS Bases pages will be incorporated in accordance with the ITS Bases Control Program.

3.0 BACKGROUND

The background for this application is adequately addressed by the NRC Notice of Availability published on May 6, 2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.

4.0 REGULATORY REQUIREMENTS AND GUIDANCE The applicable regulatory requirements and guidance associated with this application are adequately addressed by the NRC Notice of Availability published on May 6, 2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298),

and TSTF-449, Revision 4.

5.0 TECHNICAL ANALYSIS

Florida Power Corporation (FPC) has reviewed the safety evaluation (SE) published on March 2, 2005 (70 FR 10298) as part of the CLIIP Notice for Comment. This included the NRC staff's SE, the supporting information provided to support TSTF-449, and the changes associated with Revision 4 to TSTF-449.

FPC has concluded that the

U.S. Nuclear Regulatory Commission 3F0307-07 Attachment B Page 2 of 4 justifications presented in the TSTF proposal and the SE prepared by the NRC staff are applicable to CR-3 and justify this amendment for the incorporation of the changes to the CR-3 ITS.

6.0 REGULATORY ANALYSIS

A description of this proposed change and its relationship to applicable regulatory requirements and guidance was provided in the NRC Notice of Availability published on May 6, 2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.

6.1 Verification and Commitments The following information is provided to support the NRC staff's review of this amendment application:

Plant Name, Unit No.

Crystal River Unit 3 Steam Generator Model(s):

177FA Effective Full Power Years (EFPY) of service for currently Approximately 19.2 as of Refuel 14 (Nov. 05, 2005) installed OTSGs Tubing Material Alloy 600 Stress Relieved Number of tubes per OTSG 15,531 Number and percentage of tubes OTSG A OTSG B plugged in each OTSG 351 (2.3%)

862 (5.6%)

Number of Tubes repaired in each OTSG OTSG A OTSG B Tubes wisleeves (Inservice) 159 156 Tubes wi repair rolls (Inservice) 948 1401

- Primary Water Stress Corrosion Cracking Degradation mechanism(s)

(PWSCC) identified Outside Diameter Intergranular Attack/Stress Corrosion Cracking (OD IGA/SCC)

Wear / Fretting / Thinning Per SG: 150 gallons per day (gpd) per LCO 3.4.12.d Cur-rent primary-to-secondary 341.

leakage limits:

Total:

No total limit specified in ITS Temperature condition leakage is evaluated at: room temperature

U.S. Nuclear Regulatory Commission 3F0307-07 Attachment B Page 3 of 4 Approved Alternate Tube Repair Criteria (ARC):

1. First Span IGA

2. Tube End Cracks (TECs)

Approved by: Amendment 158 dated 10/28/97 Applicability: Inservice tubes with pit-like IGA indications in the first span of OTSG B Any special limits on allowable accident leakage:

None Any exceptions or clarifications to the structural performance criteria that apply to the ARC: None Approved by: Amendments 188 dated 10/01/99 and 222 dated 10/31/05 Applicability: Inservice tubes with axially-oriented TECs in either OTSG Any special limits on allowable accident leakage:

1 gallon per minute (gpm) minus 150 gpdfor TECs combined with all other postulated accident leakage Any exceptions or clarifications to the structural performance criteria that apply to the ARC: None Approved OTSG Tube Repair Methods

1. Sleeves

2. Repair Rolls

-f Approved by: Amendment 136 dated 09/11/91 Applicability limits, if any: (ITS 5.6.2.10.4.a.1l.a)

No more than five thousand sleeves may be installed in each OTSG.

Repair criteria: 40% of the sleeve wall thickness Approved by: Amendment 198 dated 09/10/01 Applicability limits, if any: None.

Repair criteria: 40% of the initial wall thickness Performance criteria for accident Primary to secondary leak rate values assumed in leakage licensing basis accident analysis, including assumed temperature conditions: 1 gpm at room temperature assumed in the CR-3 Final Safety Analysis Report 7.0 NO SIGNIFICANT HAZARDS CONSIDERATION FPC has reviewed the proposed no significant hazards consideration determination published on March 2, 2005 (70 FR 10298) as part of the CLIIP. FPC has concluded that the proposed determination presented in the notice is applicable to CR-3 and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.9 1(a).

U.S. Nuclear Regulatory Commission Attachment B 3F0307-07 Page 4 of 4 8.0 ENVIRONMENTAL EVALUATION FPC has reviewed the environmental evaluation included in the model SE published on March 2, 2005 (70 FR 10298) as part of the CLIIP. FPC has concluded that the staff's findings presented in that evaluation are applicable to CR-3 and the evaluation is hereby incorporated by reference for this application.

9.0 PRECEDENT This application is being made in accordance with the CLIIP. FPC is not proposing variations or deviations from the ITS changes described in TSTF-449, Revision 4, or the NRC staff's model SE published on March 2, 2005 (70 FR 10298). Several clarifications have been requested by the NRC Staff as described in Attachment A. These proposed changes are consistent with the CLIIP model application.

10.0 REFERENCES

Federal Register Notices:

Notice for Comment published on March 2, 2005 (70 FR 10298)

Notice of Availability published on May 6, 2005 (70 FR 24126)

PRORESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #264, REVISION 2 Application to Modify Improved Technical Specifications Regarding Steam Generator Tube Integrity ATTACHMENT C Proposed Improved Technical Specification Changes (Mark-up)

&iikeeitte indicates deleted text.

ifghlightedNx indicates added text.

TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.11 Emergency Feedwater Initiation and Control (EFIC)

System Instrumentation................

3.3-26 3.3.12 Emergency Feedwater Initiation and Control (EFIC)

Manual Initiation......................

3.3-30 3.3.13 Emergency Feedwater Initiation and Control (EFIC)

Automatic Actuation Logic.............

3.3-32 3.3.14 Emergency Feedwater Initiation and Control (EFIC)-Emergency Feedwater (EFW)-Vector Valve Logic...................................

3.3-34 3.3.15 Reactor Building (RB)

Purge Isolation-High Radiation.....................................

3.3-35 3.3.16 Control Room Isolation-High Radiation..........

3.3-36 3.3.17 Post Accident Monitoring (PAM)

Instrumentation.. 3.3-38 3.3.18 Remote Shutdown System..........................

3.3-42 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4-1 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits...........

3.4-1 3.4.2 RCS Minimum Temperature for Criticality.........

3.4-3 3.4.3 RCS Pressure and Temperature (P/T) Limits.......

3.4-4 3.4.4 RCS Loops-MODE 3...............................

3.4-6 3.4.5 RCS Loops-MODE 4...............................

3.4-8 3.4.6 RCS Loops-MODE 5, Loops Filled.................

3.4-10 3.4.7 RCS Loops-MODE 5, Loops Not Filled.............

3.4-13 3.4.8 Pressurizer.....................................

3.4-15 3.4.9 Pressurizer Safety Valves.......................

3.4-17 3.4.10 Pressurizer Power Operated Relief Valve (PORV)..

3.4-19 3.4.11 Low Temperature Overpressure Protection (LTOP)

System...................................

3.4-21 3.4.12 RCS Operational LEAKAGE.........................

3.4-22 3.4.13 RCS Pressure Isolation Valve (PIV)

Leakage......

3.4-24 3.4.14 RCS Leakage Detection Instrumentation...........

3.4-27 3.4.15 RCS Specific Activity...........................

3.4-30 13.4.16 Steam Generator (OTSG)

Tube Integrity...........

3.4-34 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5-1 3.5.1 Core Flood Tanks (CFTs).........................

3.5-1 3.5.2 ECCS-Operating.................................

3.5-4 3.5.3 ECCS-Shutdown..................................

3.5-7 3.5.4 Borated Water Storage Tank (BWST)...............

3.5-9 3.6 CONTAINMENT SYSTEMS................................

3.6-1 3.6.1 Containment.....................................

3.6-1 3.6.2 Containment Air Locks...........................

3.6-3 3.6.3 Containment Isolation Valves....................

3.6-8 3.6.4 Containment Pressure............................

3.6-15 3.6.5 Containment Air Temperature.....................

3.6-16 (continued)

Crystal River Unit 3 ii Amendment No. 16-1

TABLE OF CONTENTS B 3.3 B 3.3.12 B 3.3.13 B 3.3.14 B 3.3.15 INSTRUMENTATION (continued)

Emergency Feedwater Initiation and Control (EFIC)

Manual Initiation...................

Emergency Feedwater Initiation and Control (EFIC)

Automatic Actuation Logic...........

Emergency Feedwater Initiation and Control (EFIC)-Emergency Feedwater (EFW)-Vector Valve Logic................................

Reactor Building (RB)

Purge Isolation-High Radiation..................................

Control Room Isolation-High Radiation........

Post Accident Monitoring (PAM)

Instrumentation Remote Shutdown System........................

B 3.3-100 B 3.3-105 B 3.3-110 B

B B

3.3.16 3.3.17 3.3.18 B

B B

B 3.3-114 3.3-119 3.3-124 3.3-145 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4-1 B 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits..........

B 3.4-1 B 3.4.2 RCS Minimum Temperature for Criticality........

B 3.4-6 B 3.4.3 RCS Pressure and Temperature (P/T) Limits......

B 3.4-9 B 3.4.4 RCS Loops-MODE 3..............................

B 3.4-17 B 3.4.5 RCS Loops-MODE 4..............................

B 3.4-22 B 3.4.6 RCS Loops-MODE 5, Loops Filled................

B 3.4-27 B 3.4.7 RCS Loops-MODE 5, Loops Not Filled............

B 3.4-33 B 3.4.8 Pressurizer....................................

B 3.4-37 B 3.4.9 Pressurizer Safety Valves......................

B 3.4-43 B 3.4.10 Pressurizer Power Operated Relief Valve (PORV)

. B 3.4-47 B 3.4.11 Low Temperature Overpressure Protection (LTOP)

System..................................

B 3.4-52 B 3.4.12 RCS Operational LEAKAGE........................

B 3.4-53 B 3.4.13 RCS Pressure Isolation Valve (PIV)

Leakage.....

B 3.4-58 B 3.4.14 RCS Leakage Detection Instrumentation..........

B 3.4-65 B 3.4.15 RCS Specific Activity..........................

B 3.4-71 3.4.16 Steam Generator (OTSG)

Tube Integrity..........

B 3.4-75 B

B B

B 3.5 3.5.1 3.5.2 3.5.3 3.5.4 3.6 3.6.1 3.6.2 3.6.3 3.6.4 3.6.5 3.6.6 EMERGENCY CORE COOLING SYSTEMS (ECCS)

Core Flood Tanks (CFTs)........................

ECCS-Operating................................

ECCS-Shutdown.................................

Borated Water Storage Tank (BWST)..............

B B

B 3.5-1 3.5-1 3.5-9 3.5-20 3.5-24 CONTAINMENT SYSTEMS................................

B 3.6-1 Containment....................................

B 3.6-1 Containment Air Locks..........................

B 3.6-6 Containment Isolation Valves...................

B 3.6-15 Containment Pressure...........................

B 3.6-29 Containment Air Temperature....................

B 3.6-32 Reactor Building Spray and Containment Cooling Systems..............................

B 3.6-35 (conti nued)

Crystal River Unit 3 vi Amendment No. 1-8--Z

Definitions 1.1 1.1 Definitions LEAKAGE (conti nued)

3.

Reactor Coolant System (RCS)

LEAKAGE through a steam generator (O:+/-SG-tu-Ie to the secondary system [(primary to secondar7F ILEAKAGE)].

b.

Unidentified LEAKAGE All LEAKAGE that is not identified LEAKAGE.

c.

Pressure Boundary LEAKAGE MODE NUCLEAR HEAT FLUX HOT CHANNEL FACTOR (FQ(Z))

NUCLEAR ENTHALPY RIS1EN HOT CHANNEL FACTOR SF)

OPERABLE -OPERABILITY PHYSICS TESTS LEAKAGE (except OTSG tube leakage

.i-rmar__to secnodary LEAKAGE) through a non-isolable fault in an RCS component body, pipe wall, or vessel wall.

A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1.

F,(Z) shall be the maximum local linear power density in the core divided by the core average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions.

F*H shall be the ratio of the integral of linear power along the fuel rod on which minimum departure from nucleate boiling ratio occurs to the average fuel rod power.

A system, subsystem, train, component, or device shall be OPERABLE when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

(continued)

Crystal River Unit 3 1.1-5 Amendment No. +4-9 Crystal River Unit 3 1.1-5 Amendment No. 1-49

RCS Operational LEAKAGE 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 RCS Operational LEAKAGE LCO 3.4.12 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE;

b.

1 gpm unidentified LEAKAGE;

c.

10 gpm identified LEAKAGE; and

d.

150 gpd of primary to secondary through any one steam generator LEAKAGE (OTSG).

Two TS~ sall beOPERABLrE-.

APPLICABILITY:

MODES 1, 2,

3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

RCS opeerational A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months LEAKAGE not within within limits.

limits for reasons other than pressure boundary LEAKAGE 4L_

primary_' to secondary ILEAKAGE.

B.

Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

AND OR B.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Pressure boundary LEAKAGE exists.

Primary to secondary

]LEAKAGE not withinr

ýlimit.I Crystal River Unit 3 3.4-22 Amendment No. +%

I

RCS Operational LEAKAGE 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 NOTES----------------------

'LiNot required to be performed in MODE 4.

Not required in MODE 3 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation.

12.

Not applicable to primary to secondar

!LEAKAGE.

Verify RCS operational LEAKAGE is within' Ilimits by performance of Perform RCS water inventory balance during steady state 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months i

SR 3.4.12.2 NOTE----------------------

,Not required to be performed until 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months si after establishment of steady state operation.

Verify steam generater tube I I ILC, I i

ity i i n accordance with the Steam Generator Tube Su..... "11....-

-n..

a rý....

urveillane Program,. iVerify primary to Isecondary LEAKAGE is 150 gallons per da-*

ithrough any one steam generator.-

Ift accordanc-e with the Steam Generator Tube Surveillance 72rogram.

Crystal River Unit 3 3.4-23 Amendment No. 1-4-9

O)TSG Tube Integrity 13.4.16 13.4 REACTOR COOLANT SYSTEM (RCS)]

13.4.16 Steam Generator (OTSG)

Tube Integrit2]

[LCO 3.4.16 OTSG tube integrity shall be maintained.1 FAh0 All OTSG tubes satisfying the tube repair criteria shall be Iplugged or repaired in accordance with the Steam Generator7 Program.

APPLICABILITY:

MODES 1, 2, 3, and 4.1 A-CTIONS NOTE--------------------------------

!Separate Condition entry is allowed for each OTSG tube.

CONDITION RQUIRED ACTION COMPLETION TIME FA.

One or more OTSG A.1 Verify tube inteegrityj 7da tubes satisfyingth-

'of the affected tube repair criterial tube(s) is maintained and not pluggeddor' until the next, repai red in___

refueling outage or' accordance with the PTSG tube inspection.

Steam Generator' P rog ram.

_N_

A.2 Plug or repair the' Prior to entering' affected tube(s) in MODE 4 following laccordance with the' the next refueling,

!Steam Generator, outage or OTSG Program.

tube inspection B.

Required Action and B.1 Be in MODE 3.

hours 7associ ated Completion Time of Condition Ar AND not met._

B.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OTSG tube integrlt*

not maintained.*-

Crystal River Unit 3 3.4-34 Amendment No. XX

QTSG Tube Integrity 1

13.4.16 SURVEILLANCE REQUIREMENTS_

,WRVEI LLANC CFREQUENCY SR 3.4.16.1 Verify OTSG tube integrity in accordance' In accordance with',

with the Steam Generator Program.

the Steamn_

,Generator Progra'

i

-3.4.16.2 Verify that each inspected OTSG tube that Prior to entering' Isatisfies the tube repair criteria is-MODE 4 following a]

plugged or repaired in accordance with the

.TSG tube-

'Steam Generator Program..

jinspectioQ Crystal River Unit 3 3.4-35 Amendment No. XX

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals (continued) 5.6.2.10 Steam Generator (OTSG) Tube Surveillance Program Each OTSG shall be demonstrated OPERABLE by performance of th following aumnted inservice inspection program.

1.

Eaeh OTSG shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of OTFSGs5 specified in Table 5.6.2 1.

2.

The OTSG tube minimum sample size, inspection result celassification, and the corresponding action required shl be as specified in Table 5.6.2 2. The inservice inspection of OTSG tubes shall -be per formed at the frequenie.s

-specified in Specification 5.6.2.10.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 5.6.2.10.4. The tubes selece for each isrceinspection shall include at least 3%o the total number of tubes in all OTSGs. The tubes selected for these npcin shall be selected on a random basis exeeptt

a. Where experience in similar plants with similar water chemistry indicates criticeal areas to be inspected, then at least 50% of the tubes inspected shall be fro these critical areas.

b.

The first inservic inspection (subsequent to the preserIIce inspection) of each OTSG shall include:

1. All nonplugged tubes that previously ha detectable wall penetrations (>20%), and

2. Tubes in those areas where exprinceha i ndiceated potenti a--l prblms e.The second and third isrceinspections may be less than a full tube Inspection by concentrating (select ing at least 50% of the tubes to be inspected) the inspHein on those areas of the tube sheet array antd on hos potions of the tubes where tubes with imperfections w.ere previously found-.

d.

Tubes Hin speceific limited areas which are distinguished by unique operating conditions or physical constructo may be excluded from random samples if all such tubesy (continued)

Crystal River Unit 3 5.0-13 Amendment No. 149

Procedures, Programs and Manuals 5.6 Procedures, Programs and Manuals 5.6.2.1O OTSG Tube Surveillance Program (continued) in the specific area of an OTSG are inspected with the inspection result classification and the corresponding action.reuired as specified in Table 5.6.2 3. No credit will be taken for these tubes in meem.....

sample s iz reurements. Degraded or defective tube found in ths ra ilnot be considered in determining the inpcinresults category as long as the mode of degradation is unique to that area andno random in nature.

t-Inservice tubes with pit like.GA indications in the first span of the B OTSG, identified in the OTSG Inservice Inspection Surveillance Procedure, must be ins-pected with bobbin and Motorized Rotating Pancake Coil11 ('MRPC) eddy current techniques from the lower tb se

... secondary face to the bottom of the first tube support plate during each inservice inspection of thB OTSG.

No credit is to be taken for this inspect-,ionl inl meeting iiu sample size requirements for, theradom inspection.

Defective tubes found during this inspection are to be plugged or sleeved. Degraded or defective tubes found during this inspection are not to be considered in determining the inspctin results category for the random inspection,' u-nlesstthe degradation mechanism identified is a mechanism other than pi t i ke GA-.

f.-Tubes in service with axially oriented tube end cras (TEC) are identified in the OTSG Inservice Inspectio Surveillance proceedure.

The portion of the tube with the axial TEC must be inspected using the motorized rotating coil eddy current technique during each subsequent inspection.

No credit is to be taken for this inspection for meeting the minimum sample size requirement for random sample inspection.

Tubes identified with TEC that meet the alternate rpi criterCia will be added to the existing list of tubjes in the OTSG Inservicle Inspection Surveillance procedure-.

Tubles identified with TEC during the previous inspection which meet the criteria to remain in service will not be included when calcuilating the inspectionI categolry of the (conti nued)

Crystal River Unit 3 5.0-14 Amendment No. +&8

rrocedures, Prog rams and Manuals 5-.6 56--

,rocedures, Programs and Manuals The inspection data for tubes with xill oriented TEC inudica.ttions shall be comare totepeius inspection data tomnto h

ndiatlilon for growth.

Tubes with axiallynoriented TEC mybletin service.

usingthemethod decIITe*b ed in I

Top Re port AW 2346...,

Revsio 0,prvided the combined poectedleýakag J.U /-.ILU UI.)U UU

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L NOTE For

the, inspýecti on conducted in accordance with 5.6.2.1O.2.f only tubeýs w it h T[EC indications identified after the 19,7 inspetion will be i-ncluided i'n the be pe rcentage-calcutlations.

(contii nued) 55.0 1:4A Amendment No.

222 trt...

.-... I..-;

r Q*ý.*

It ve r n *:

Procedures, Programs and Manuals 5.6 5--6-rrocedures, rrograms and Manualsy

&-rG-rP:-AGOTSG Tube Surveillance rrogram (continued)

Category Inspection Results C 1 Less than 5% of the total tubes.inspected areqdegrded tubes and none of theinpce tube~s are defective.

C 2 One or more tubes, but not more than 1-%

of the total tubes inspce are defecti+ve,'o-r between 5% and 10% of theimr total tubes inspctedaredegraded tubes.

C 3 More than 10% of the total tubes. inspected' are degraded tubes or more tha_1loite isected tubes are defective.

3. The above re~quired inservice inspectons of OT-SG tubes shl be perfm ed M

U.

Ia te following feuencie

a. -1.s e-..Lfl L.

.4-b!-

m-Lat inftervals after th!epreviu isetion. if two consectiv insectonsfolowig ervice under all volatile treatmen t' 6 INJ co-n' diti ons, no:t i ncludi ngth p

esul

'V nr d1? --!

ir~"'

SI t-i.

W ilU III UI LC Iu 1 4

,fl

.,e 4

IIIU i Ie

./I V l

insecios emonst rate tha preioul obse!rved derdtion ha4otc ntinued an d no additional degýj~aradaFtion has occurred,the inpecto inevlmay 1II "MC I(I w IIJ E

1 1

.. E L

I a

I a; I-I U.L I V IE,

b x ed to a maximum f once..pr

..40. onhs

b. if the inservic isetion of aOTconducted i accordance w.it Table 5.6.2 2 or TabIL e 5.6.2 3 require 1ItoLE at.

leas one1.-EILL~

nspection freqencysha..l.

l applytunti a subsequent l

inspection demonstrates th Ct I

thirdsa*mpleinspe.cti is not required. if the C)iseto rsus classification. is due to inluin newh TEC indications that meet the critera to reman in service, no reduction i n i nspecti on frequency. i equi red.

e. Additional unscheduled inservic inpecti--ons shall be perfor-mned on each OTSG i n accordiancewthte is sapeInpcin rspecified in Table 5.6.2 2 or Tablme 5.6.2 3 duing the s utdwn subsequent to any of the followin odtos U

AI

=II=I=;I-I I

I IhI UIUC1 LOk l UL.

n eL y U I I U

M I

3 3 W I I I.I I

.IICA I-I IIU II]I Ij

-. w" Z

ItU LM

I I 0l 1 L

M Vnc I

Il UIIL n exessof the limits of Specification 3.4.12-,

2. A seismic occuirrence greater than the Operating Basis Earthquake,

3. A loss of eclant accident requirin actuation Of the engineered safeguards, or

4. A main' st-eam line or feedwater line break.

(conti nued)

Crystal River Unit 3 5.0-15 Amendment No. 2-2-Z

Procedures, Programs and Manuals 5.6,6-Procedures, Programs and ManualsI 5.6.2.10 OTSG Tube Surveillance Program (continued)ý

4.

Acceptance e

I I LeI aI:L

a. Vocabulary as used in this Specification:

1.

Tubing or Tube means that portion of the tubeo sleeve which forms the primary system to secondary system pressure boundary.

2.

Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-eurrent testing indications below 20% of the Iominal tube wall thickness, if detectable, may be Lonsidered as imperfectionsy.

3. Degradation means a srieinduced cracking, wastage, wear, or general corrosion ocrigo either inside or outside of a tube.

4. Degraded Tube means a tube containing degradatio

>20% through wall but < 40% through wall in the pressure boundary.

5. %6 Degradation/% Through wall means the percentage of the tube (pressure boundary) wall thicknessý affected or removed by degradation.

6.

Defective Tube means a tube containing degradato S40%

through wall in the pressure boundary. Any tube which does not permit the passage of the eddy current inspecton,.

probe shall be deemed defective tube.

7.

Pit like Intergranular Attack (IGA) indication means a bobbi n coil i ndi cati on confi rmed by Motorized Rotating Pancake Coil (MRPC) or othe qualified inspection techniques to have a volumetric, pit like morphology characteristic Of

!GA.

(conti nued)

Crystl Rivr Uni 3

5.0-I16 Amendment No. +80**

%/'

IIUILILUI Crystal River Unit 3 5.0-16 Amendment No. 1&0

Procedures, Programs and Manuals 5.6 5

I rrocedures, Programs and Manuails

-O+/-SC-Tube-Surveillance Program (continued)

8.

Plugging/Repair Limit means the extent of pressure boundary degradation beyond which the tube shal either be removed from service by installation of plugs or the area of degradation shall be removed from service (a new pressure boundary established3 usin anApproved Repair Technique. The plugging/repair limit is 40% through wall for all pressure boundary degradation.

9. Unserviceable describes the condition of a tubei it leaks o c

n a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss of coolant accident, or a main steam line or feedwater line break, as specified in 5.6.2.10.3.c, above.

10. Tube Inspection means an inspection of the OTSG tube pressure boundary.

-11. Approved Repair Technique means a technique, ote than plugging, that has been accepted by the NRC as a methodology to remove or repair degraded or defective portions of the pressure boundary and to establish a new pressure boundary.

Following are Approved Repair Techniques:

a)

Sleeve installation in accordance with the B&W process (or method) described in report BAW 2120P.

No more than five thousand sleeves may be installed in each OTSG.

b) installation of repair rolls in the upper and lower tubesheets in accordance with BAW 2303rP,-

Revision 4. The repair process (single, overlapping, or multiple roll) may-b-e pefred in each tube.

The repair roll are will be emI d using eddy urrent methods, following installation.

The repais r roll mut be free-of imprfections and degradation fo the repair t be considered acceptable.

(continued)

Crystal River Unit 3 5.0-17 Amendment No. 1%

Procedures, Programs and Manuals.6 5.-r-Procedures, Programs and Manuals 5-62.--I lOTSG Tube Surveillance Program (continued)

The repair roll in each tube will be inspece during each subsequent inservice inspection whil1e the tube wi th a repai r roll isi sevie The repair roll will be considered a specific limited area and will be excluded from the random sampling. No credit will be taken for meeting the minimu saple size.

if primary to secondary leakage results inta shutdown of the plant and the cause i determined to be degradation in a repair roll 100% of the repair rolls in that OTSG shall be exmnd. if that i nspecti',on r esul ts i n eneing Category C 2 or C 3 for specific limited area inspection, as detailed in Tablýe 5.6.2 3, 100% of the repair rolls shall be examined in the other OTS-G.

1:2.

Tube End Cracks (TEC) are those crack like eddy Lurrent indications, circumferentially and/or axially oriented, that are within the inconel clad region of the primary face of the upper and lower tubesheets, but do not extend into the carbont steel to Inconel clad interface.

b. The OTSG shall be determined OPERABLE after cmltn the corresponding acin plug or repair all tubes exceeding the p! ugin/eai r limit) requi red by Tablýe 5.6.2 2 (and Table-5,.6r.2 3 if theprvsosf Specification 5.6.2.1:0.2.d are uiie)

I.nservice tubes with pit like IGA indications in the "B" OTSG firtsa shall be monitored for growth of these indicat,.,I..

using a test probe equivalent to the hig.h frequency In "i probe used in the 1I997 inspection.

The indicated percentage throughwall value from the current insecion shall be compared to the indicated percentage th r ouwall value from the 1997 inspection.

kEcntinued) r-....

L SI V

l w ý P.

1 1 pq I

1 1.

No' 198

.l Y

.I I

I IV lI III M-Hll~ l!llll H

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals

[5.6.2.10 Steam Generator (OTSG)

Program A Steam Generator Program shall be established and implemented td ensure that OTSG tube integrity is maintained.

In addition, the'

,Steam Generator Program shall include the following provisions:

a._ 1Provisions for condition monitoring assessments.

Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced-Ileakage.

The "as found" condition refers to the condition of the tubing during an OTSG inspection outage, as,

'determined from the inservice inspection results or by other means, prior to the plugging or repair of tubes.

Condition monitoring assessments shall be conducted during each outag-during which the OTSG tubes are inspected, plugged, orL repaired to confirm that the performance criteria are being met.

b. Performance criteria for OTSG tube integrity.

OTSG tub integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.I---

11.

Structural integrity performance criterion:

All in-iservice steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power, range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents.

This includes retaining aF safety factor of 3.0 against burst under normal steadyL state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burstF applied to the design basis accident primary-to---.-

secondary pressure differentials.

Apart from the b-a-e Srequirements, additional loading conditions associated with the design basis accidents, or combination of

,accidents in accordance with the design and licensing' basis, shall also be evaluated to determine if the -

associated loads contribute significantly to burst orI collapse.

In the assessment of tube integrity, those, loads that do significantly affect burst or collapse' shall be determined and assessed in combination with thý loads due to pressure with a safety factor of 1.2 on the' combined primary loads and 1.0 on axial secondary loads.

(continued)

Crystal River Unit 3 5.0-13 Amendment No. XX

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 15.6.2.10 OTSG Program (continued~j F2.

Accident induced leakage performance criterion : Th

primary to secondary accident induced leakage rate ftor lany design basis accident, other than an OTSG tuber-

ýrupture, shall not exceed the leakage rate assumed 1i'n the accident analysis in terms of total leakage rate Tfor all OTSGs and leakage rate for an individual 0TSG._

Leakage is not to exceed one gallon per minute per OTS.I

3. The operational LEAKAGE performance criterion isý specified in LCO 3.4.12, "RCS Operational LEAKAGE."

,c. Provisions for OTSG tube repair criteria.

Tubes shallTb plugged if the sleeved region of a tube is found by7L jinservice inspection to contain flaws in the (a) sleeve or' (b) the pressure boundary portion of the original tube wall

[In the sleeve/tube assembly.

T.he non-sleeved region of a tube found by inservic{

inspection to contain flaws with a depth equal to or, exceeding 40% of the nominal tube wall thickness shall i plugged or repaired except if the flaws are permitted to' rremain in service through application of an alternate tube repair criteria discussed below.[

The following alternate tube repair criteria may be applied

'as an alternative to the 40% depth based criteria:

,1.

Pit-like Intergranular Attack (IGA) indication means-Ibobbin coil indication confirmed by Motorized Rotatin' IPancake Coil (MRPC) or other qualified inspection'--"

techniques to have a volumetric, pit-like morphology characteristic of IGA.

Inservice tubes with pit-like IGA indications in the first span of the B OTSG are'._

acceptable provided the depth of the indication is lss than 40% of the nominal tube wall thickness.

InserviceI tubes with pit-like IGA indications in the first span ofI the B OTSG with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.*

(continued)

Crystal River Unit 3 5.0-14 Amendment No. XX

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals

[.6.2.10 OTSG Program (contic-nued)]

'2. Tube End Cracks (TEC) are those crack-like eddy current indications, circumferentially and/or axially oriented,,]

that are within the Inconel clad region of the primary face of the upper and lower tubesheets, but do not, extend into the carbon steel-to Inconel clad interface.

Tubes with axially oriented TEC may be left in-service using the method described in Topical Report BAW-2346P,*

Revision 0, provided the combined projected leakage from lall sources of primary-to-secondary leakage, including'-_

axial TEC indications left in-service, does not exceed' ithe Main Steam Line Break (MSLB) accident leakage limit

'of one gallon per minute, minus 150 gallons per day, peg 9TSG.

The contribution to MSLB leakage rates from TEC indications shall be determined utilizing the'____

methodology in Addendum B dated August 10, 205ýtdl rTopical Report BAW-2346P, Revision 0.

The projection of, TEC leakage that may develop during the next operatingý cycle shall be determined using the methodology in-ddendum C dated August 30, 2005 to Topical Report' BVWý 12346P, Revision 0.F

'Tubes identified with TEC that are allowed to remain i'n service under the alternate repair criteria will be',

added to the existing list of tubes in the OTSG*

iInservice Inspection Surveillance Procedure.

The Inspection data for tubes with axially oriented TEC_

indications shall be compared to the previous inspect6in data to monitor the indications for growth.

Tubes with crack-like indications within the ca steel portion of the tubesheet, circumferentiallyL oriented TEC, or volumetric indications within the Inconel clad region of the tubesheet shall be repaired

ýusing the appropriate method from 5.6.2.10.f or removed from service by plugging the tube.

(continued)

Crystal River Unit 3 5.0-15 Amendment No. XX

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.10 OTSG Program (continued)

,'d.____Provisions for OTSG tube inspections.

Periodic 0TSG inspections shall be performed.

The number and portions the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks)K that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-*

tubesheet weld at the tube outlet, and that may satisfy the" applicable tube repair criteria.

The tube-to-tubesheet weld is not part of the tube.

In tubes repaired by sleeving, the portion of the original tube wall between the sleeve'sl-ljoints is not an area requiring re-inspection.

In addition' to meeting the requirements of d.1 through d.8 below, the-inspection scope, inspection methods, and inspectionr intervals shall be such as to ensure that OTSG tubs_

aintegrity is maintained until the next OTSG inspection.

assessment of degradation shall be performed to determine*

the type and location of flaws to which the tubes may be Isusceptible and, based on this assessment, to determines which inspection methods need to be employed and at what, locations.1

,i.

Inspect 100% of the tubes in each OTSG during the first

[refueling outage following OTSG replacement.l V2.

Inspect 100% of the tubes at sequential periods of 60

'effective full power months.

The first sequentialF period shall be considered to begin after the first

ýinservice inspection of the OTSGs.

No OTSG shall_

ioperate for more than 24 effective full power months -so

!one refueling outage (whichever is less) without being'

,inspected.

3.

If crack indications are found in any OTSG tube, thenI the next inspection for each OTSG for the degradation' mechanism that caused the crack indication shall nots_

exceed 24 effective full power months or one refueling outage (whichever is less).

If definitive information*

such as from examination of a pulled tube, diagnostic7 non-destructive testing, or engineering evaluation,

,indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.[---

(continued)

Crystal River Unit 3 5.0-16 Amendment No. XX

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 15.6.2.10 OTSG Program (continued)

Inservice tubes with pit-like IGA indications in the' 7first span of the B OTSG must be inspected with bobbin and Motorized Rotating Pancake Coil (MRPC) eddy current techniques from the lower tube sheet secondary face to the bottom of the first tube support plate during each Linservice inspection of the B OTSG.

fInservice tubes with pit-like IGA indications in the "B"1 OTSG first span shall be monitored for growth of these e' indications by using a test probe equivalent to the high frequency bobbin probe used in the 1997 inspection.

The'

,Indicated percentage through-wall value from the current

,inspection shall be compared to the indicated percentagA through-wall value from the 1997 inspection.-

5.

Tubes in-service with axially oriented tube end crack

-(TEC) are identified in the OTSG Inservice Inspection, Surveillance Procedure.

The portion of the tube with the axial TEC must be inspected using the motorized_

rotating coil eddy current technique every 24 effective' Ifull power months or one refueling outage, whichever is l1ess.

16.

If the plant is required to shut down due to primary-to-isecondary leakage and the cause is determined to be degradation of the TEC portion of the tubes, 100% of the tubes with TEC in that OTSG shall be examined in the,_

location of the TEC.

If more than 1% of the examined tubes satisfy the tube repair criteria, 100% of the,

,tubes with TEC in the other OTSG shall be examined -in

,the location of the TEC.J

7.

The repair roll in each tube will be inspected every24

,effective full power months or one refueling outage 9 _

(whichever is less) while the tube with a repair roll is in service.i

18. If the plant is required to shut down due to primary-to-'

'secondary leakage and the cause is determined to be a flaw in a repair roll, 100% of the repair rolls in botTf OTSGs shall be examined.i (continued)

Crystal River Unit 3 5.0-17 Amendment No. XX

Procedures, Programs and Manuals 5.6 F5.6 Procedures, Programs and Ma"nual-s 15.6.2.10 OTSG Proqram (continued)]

Provisions for monitoring op-erational primary to secondarý LEAKAGE.F if.

Provisions for OTSG tube repair methods.

Steam generatoi tube repair methods shall provide the means to reestablisF the RCS pressure boundary integrity of OTSG tubes without removing the tube from service.

For the purposes of these Specifications, tube plugging is not a repair.

All acceptable tube repair methods are listed below.I

'1. Sleeve installation in accordance with the B&W process!

(or method) described in report BAW-2120P.

No more thai five thousand sleeves may be installed in each OTSG.

t2.

Installation of repair rolls in the upper and lower'

-tubesheets in accordance with BAW-2303P, Revision 4.

The repair process (single, overlapping, or multiple roll) may be performed in each tube.

The repair roll,

'area will be examined using eddy-current methods,,-

following installation.

The repair roll must be free of flaws for the repair to be considered acceptable.

If the repair roll is unacceptable, the tube must be' repaired or plugged*.

(continued)

Amendment No. XX Crystal River Unit 3 5.0-18

Procedures, Programs and Manuals 5.6 Anl I~ r r-r' n

-1 r-..

-1 At-i1*

RI ~ ~ ~ ~ ~ ~

In

-ri

-Crn rr rirm frn L'ýJI%.1I.I1J

.I.IJL.LIXV.L...L.

JLld.41JI L-A, I.LJI'4 Preservice inspect ion@--

Number of OT^SGs T

40 First Inservice Inspection GnIe SeTnd and Subsequent inseriS The finse Vice inspection may be limited to one OTFS.G

'JI I 111i i

E L

V.

L I I

.. 1 1

.LU l-III1JL*

I4 I I I

C

/

LU I

-a

%J I I I Vj GL

"ýj tA ez-LII~

LUL~~3 I I A

rne li e

su t Or TneL T.-I r t

wIe ta Io UI

'.I

ý I

-IA I

.E.

I

-U J

IILIIII I

,~I L

I

.IU LIIC;L I41 I I 3L IIJI I ~VIUU IIL H;LE

,LIJIIE indicate that both lTSGs are perfori.ngin a like manner. Note that under soe ircumstances, the operating conditions in one OT-SG may be found to be I

more severe -than those in the other UiSG.

under suc L"

" 1 I.

2 r _

A C]eCunisaees th:-e smpi e

sequenee 3sill 1 De modilried "to I ImIIsI LtI III3 L

s everI e

LUl I

e lullS.

IL Crystal River Unit 3 5.0-24 Amendment No. 1-4-9

Procedures, Programs and Manuals 5.6 TABLE 5.6.2 2 (page 1 of 1)

UJI

)J*

I UUL

+/-I'd..rL'.i

'I'UI 1st Sample inspeetiein I

2-rd Sam~ple i-s-pe-ct-i-r I 3rd Sample inspeetio Sample -5z-e I[Resttt I AeiJ Resul t

~

Acto ftes~uý I

Aje A minimum of S-tubes-per e:FSG E-4 N/'A N7/A N/A N7/A 4

4 4

Pl uj wlr de F-I ve 2S tubes 1/2 th4UI.is O iG.i Nome N/A N/A Plug or

?epei-r def eeti Ve tubes-and 4 th i.....

4S tube q,

defeet+i'e Per#*r acti on for C 3--resvtil of-firs-t Sample.-

G-3 E-3--res-i-le of first*i s-amol-Te-N/A N-/A 4

4 4

E-3

+lPeet a! 1 thi... E)G, pl ug or repai r defeetiv'e

~tubes,

-nveet 2S each-other

?i-fi-NPr per

...... and

.mVtif I\\RE M! other eTSs a-re E-1 Noei~

N/A N/A Some GTSGs PerN/ANfor/A addq-trema4 C.2.resti.

G~s a-re of-seeentd E--3 s-ample.-

.TS.

is C 3 inspeet-allI eaeh-OTSG(7 plug or defeetive nIUtI iy i,RiE, per

.OCFRO *.72.

N7/A N/A to -

.2 Ii! II nU wmer NE isI tn IIJIumve ofxI.J idii I

in tC unit an nI EIis tIII nuLJverC ar u Ib,.Si nspeetea auri Crystal River Unit 3 5.0-25 Amendment No.

1-&0

Procedures, Programs and Manuals 5.6 TA r r-r r

-l1

")

f 1

1,,

%I"

-A1 LA-L Iz JA.. LI I LU) FRLA INtnCLL I 1tjN 1st Sample Inspection of a Sample Inspection of a "Specific Limited Area J

"Specific Limited Area" Sample-Si5 Result Aeim Result AegE4 on 100%6 of area E-4 None N/1AN/

im-both E-2 Pl-gor N4A N/1A d-Fefective tubt G-3 Plug or N/A N

defetv 100% of area E-4 None N/A N/A in one OT-SC

(-2 Plug or E-4 Nome de-fectv tubes-and E-2 Plug or 1:0eet96O

-repai r Of defective eorrespending tbes-5-.

area-tn

--3 Plug or other-OTSGE repair defective tube-ST E-4 Plug or E-4 None repair__

defective E-2 Plug or tubes-and rpi ilnspect-+ee96 defecive

+f tubes.

eorresponding area in E-3 Plug or other OTSG.

rpi defective

_t u

b e

s-.

Crystal River Unit 3 5.0-26 Amendment No.

+&0

Reporting Requirements 5.7 5.7 Reporting Requirements 5.7.1.2 Not Used 5.7.2 Special Reports Special Reports shall be submitted in accordance with 10 CFR 50.4 within the time period specified for each report.

The following Special Reports shall be submitted:

a.

When a Special Report is required by Condition B or F of LCO 3.3.17, "Post Accident Monitoring (PAM)

Instrumentation," a report shall be submitted within the following 14 days.

The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

b.

Any abnormal degradation of the containment structure found during the inspection performed in accordance with ITS 5.6.2.8 shall be reported to the NRC within 30 days of the current surveillance completion.

The abnormal degradation shall be defined as findings such as delamination of the dome concrete, widespread corrosion of the liner plate, corrosion of prestressing elements (wires, strands, bars) or anchorage components extending to more than two tendons and group tendons force trends not meeting the requirements of 10CFR50.55a(b)(2)(ix)(B).

The report shall include the description of degradation, operability determination, root cause determination and the corrective actions.

c.A report shall be submitted within 180 days after the" 7initial entry into MODE 4 following completion of an__

inspection performed in accordance with the Specificat-io 5.6.2.10, Steam Generator (OTSG)

Program.

The report shaTl]

i ncl ude:

11.

The scope ofinspections performed on each OTSG,

-2. Active degradation mechanisms found-3.[

Nondestructive examination techini ques utilized for eachi deegradati on mechanism,*

14.

Location, orientation (if linear), and measured sijzes L(if available) of service induced indications,j (continued)

Crystal River Unit 3 5.0-28 Amendment No. 22-2

Reporting Requirements 5.7 5.7 Reporting Requirements 5.7.2 Special Reports (continued)

5.

K Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism, 6

Total number and percentage of tubes plugged o repai red to datej

7.

The results of condition m6nitoring-,including th-Iresults of tube pulls and in-situ testing_,]

87 The effective plugging percentage for all plugging and

'tube repairs in each OTSG,*

1. _Repaihrmethod utilized and the number of tubes repaired by each repair method.j

e. Following each inservice inspection of steam generator prir to L l

asension into MODE 4.

1.

Number of tubes plugged and repaired,

2.

Craek like indications and assessment of growth fr indications in the first span; S-J--

Results of in situ pressure testing, if performed, and 4-

"'Number of tubes and axially oriented TEC indieations left in.,

the projected accident leakage, and an assessment of growth for TEC indication..

(conti nued)

Crystal River Unit 3 5.0-28 Amendment No. 2-2-Z

Reporting Requirements 5.7 5.7 Reporting Requirment 5.7.2 Special Reports Eeontinued)ý

d. Results of OTSG tube inspections that fall into Category C 3 shall be reported to the NRC in accordance with I1OCFR5O.72.

e. The complete results of the OTSG tube inservice inspection snaii De suomittea o

11e IKL witnin I Iays1arter Irea.e closure f.llowing restart. The report shall include:

1.

Number and extent of tubes inspeeted,

2.

Location and percent of wall thickness penetration for each indication of an imperfectioni,-

3-EO.]Location, bobbin coil amplitude, and axial and circumferential extent (if determined) for each first span IGA indication, and 'an assessment of growth forej lindications in the first span of OTSG B, and

4.

Identification of tubes plugged or repied and specification of the repair methodology implemented for eaeh-tube-j.LNumber of as-found and as-left tubes with TEC indications, number of as-found and as-left TEC indications, the number of as-found and as-left TEC indications as a function of tubesheet radius, the as-found, as-left, probability of detection and new TEC leakage for upper and lower tubesheet indications. The projected accident leakage and an assessment of growth

ýfor TEC indications will be provided.

An assessment of the adequacy of the predictive methodology in Addendum C to Topical Report BAW-2346P, Revision 0, including assessing the distribution of indications found in each OTSG to ensure the assumption regarding the similarity of the distribution of indications remain consistent from one cycle to the next and that the assumption of a linear increase in leak rate remain valid.

Corrective actions in the event that the assessment indicates the assumptions can not be fully supported.

Crystal River Unit 3 5.0-29 Amendment No. 2-2-Z

PRORESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #264, REVISION 2 Application to Modify Improved Technical Specifications Regarding Steam Generator Tube Integrity ATTACHMENT D Proposed Improved Technical Specification Changes (Revision Bar Format)

TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.11 Emergency Feedwater Initiation and Control (EFIC)

System Instrumentation................

3.3-26 3.3.12 Emergency Feedwater Initiation and Control (EFIC)

Manual Initiation......................

3.3-30 3.3.13 Emergency Feedwater Initiation and Control (EFIC) Automatic Actuation Logic.............

3.3-32 3.3.14 Emergency Feedwater Initiation and Control (EFIC)-Emergency Feedwater (EFW)-Vector Valve Logic...................................

3.3-34 3.3.15 Reactor Building (RB)

Purge Isolation-High Radiation.....................................

3.3-35 3.3.16 Control Room Isolation-High Radiation..........

3.3-36 3.3.17 Post Accident Monitoring (PAM)

Instrumentation.. 3.3-38 3.3.18 Remote Shutdown System..........................

3.3-42 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4-1 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits...........

3.4-1 3.4.2 RCS Minimum Temperature for Criticality.........

3.4-3 3.4.3 RCS Pressure and Temperature (P/T) Limits.......

3.4-4 3.4.4 RCS Loops-MODE 3...............................

3.4-6 3.4.5 RCS Loops-MODE 4...............................

3.4-8 3.4.6 RCS Loops-MODE 5, Loops Filled.................

3.4-10 3.4.7 RCS Loops-MODE 5, Loops Not Filled.............

3.4-13 3.4.8 Pressurizer.....................................

3.4-15 3.4.9 Pressurizer Safety Valves.......................

3.4-17 3.4.10 Pressurizer Power Operated Relief Valve (PORV)..

3.4-19 3.4.11 Low Temperature Overpressure Protection (LTOP)

System...................................

3.4-21 3.4.12 RCS Operational LEAKAGE.........................

3.4-22 3.4.13 RCS Pressure Isolation Valve (PIV)

Leakage......

3.4-24 3.4.14 RCS Leakage Detection Instrumentation...........

3.4-27 3.4.15 RCS Specific Activity...........................

3.4-30 3.4.16 Steam Generator (OTSG)

Tube Integrity............

3.4-34 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5-1 3.5.1 Core Flood Tanks (CFTs).........................

3.5-1 3.5.2 ECCS-Operating.................................

3.5-4 3.5.3 ECCS-Shutdown..................................

3.5-7 3.5.4 Borated Water Storage Tank (BWST)...............

3.5-9 3.6 CONTAINMENT SYSTEMS................................

3.6-1 3.6.1 Containment.....................................

3.6-1 3.6.2 Containment Air Locks...........................

3.6-3 3.6.3 Containment Isolation Valves....................

3.6-8 3.6.4 Containment Pressure............................

3.6-15 3.6.5 Containment Air Temperature.....................

3.6-16 (continued)

Crystal River Unit 3 ii Amendment No.

TABLE OF CONTENTS B 3.3 B 3.3.12 B 3.3.13 B 3.3.14 B 3.3.15 INSTRUMENTATION (continued)

Emergency Feedwater Initiation and Control (EFIC)

Manual Initiation...................

Emergency Feedwater Initiation and Control (EFIC)

Automatic Actuation Logic...........

Emergency Feedwater Initiation and Control (EFIC)-Emergency Feedwater (EFW) -Vector Valve Logic................................

Reactor Building (RB)

Purge Isolation-High Radiation..................................

Control Room Isolation-High Radiation........

Post Accident Monitoring (PAM)

Instrumentation Remote Shutdown System........................

B 3.3-100 B 3.3-105 B 3.3-110 B

B B

3.3.16 3.3.17 3.3.18 B

B B

B 3.3-114 3.3-119 3.3-124 3.3-145 B 3.4 B 3.4.1 3.4.2 3.4.3 3.4.4 3.4.5 3.4.6 3.4.7 3.4.8 3.4.9 3.4.10 3.4.11 3.4.12 3.4.13 3.4.14 3.4.15 3.4.16 3.5 3.5.1 3.5.2 3.5.3 3.5.4 3.6 3.6.1 3.6.2 3.6.3 3.6.4 3.6.5 3.6.6 REACTOR COOLANT SYSTEM (RCS)

B 3.4-1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits..........

B 3.4-1 RCS Minimum Temperature for Criticality........

B 3.4-6 RCS Pressure and Temperature (P/T) Limits......

B 3.4-9 RCS Loops-MODE 3..............................

B 3.4-17 RCS Loops-MODE 4..............................

B 3.4-22 RCS Loops-MODE 5, Loops Filled................

B 3.4-27 RCS Loops-MODE 5, Loops Not Filled............

B 3.4-33 Pressurizer....................................

B 3.4-37 Pressurizer Safety Valves......................

B 3.4-43 Pressurizer Power Operated Relief Valve (PORV).

B 3.4-47 Low Temperature Overpressure Protection (LTOP)

System..................................

B 3.4-52 RCS Operational LEAKAGE........................

B 3.4-53 RCS Pressure Isolation Valve (PIV)

Leakage.....

B 3.4-58 RCS Leakage Detection Instrumentation..........

B 3.4-65 RCS Specific Activity..........................

B 3.4-71 Steam Generator (OTSG)

Tube Integrity...........

B 3.4-75 B

B B

EMERGENCY CORE COOLING SYSTEMS (ECCS)

Core Flood Tanks (CFTs)........................

ECCS-Operating................................

ECCS - Shutdown.................................

Borated Water Storage Tank (BWST)

CONTAINMENT SYSTEMS................................

Containment....................................

Containment Air Locks..........................

Containment Isolation Valves...................

Containment Pressure...........................

Containment Air Temperature....................

Reactor Building Spray and Containment Cooling Systems.............................

B B

B 3.5-1 3.5-1 3.5-9 3.5-20 3.5-24 3.6-1 3.6-1 3.6-6 3.6-15 3.6-29 3.6-32 B 3.6-35 (continued)

Crystal River Unit 3 vi Amendment No.

Definitions 1.1 1.1 Definitions LEAKAGE (conti nued)

3.

Reactor Coolant System (RCS)

LEAKAGE through a steam generator to the secondary system (primary to secondary LEAKAGE).

b.

Unidentified LEAKAGE All LEAKAGE that is not identified LEAKAGE.

c.

Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a non-isolable fault in an RCS component body, pipe wall, or vessel wall.

MODE NUCLEAR HEAT FLUX HOT CHANNEL FACTOR (FQ(Z))

NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (F*H)

OPERABLE-OPERABILITY PHYSICS TESTS A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1.

FQ(Z) shall be the maximum local linear power density in the core divided by the core average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions.

FNH shall be the ratio of the integral of linear power along the fuel rod on which minimum departure from nucleate boiling ratio occurs to the average fuel rod power.

A system, subsystem, train, component, or device shall be OPERABLE when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

(continued)

Crystal River Unit 3 1.1-5 Amendment No.

RCS Operational LEAKAGE 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 RCS Operational LEAKAGE LCO 3.4.12 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE;

b.

1 gpm unidentified LEAKAGE;

c.

10 gpm identified LEAKAGE; and

d.

150 gpd of primary to secondary LEAKAGE through any one steam generator (OTSG).

APPLICABILITY:

MODES 1, 2,

3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

RCS operational A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months LEAKAGE not within within limits.

limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

B.

Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

AND OR B.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

Crystal River Unit 3 3.4-22 Amendment No.

RCS Operational LEAKAGE 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 NOTES----------------------

1. Not required to be performed in MODE 4.

Not required in MODE 3 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.

Verify RCS operational LEAKAGE is within limits by performance of RCS water inventory balance.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months SR 3.4.12.2 ---------------

NOTE-----------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify primary to secondary LEAKAGE is

150 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months gallons per day through any one steam generator.

Crystal River Unit 3 3.4-23 Amendment No.

OTSG Tube Integrity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 Steam Generator (OTSG)

Tube Integrity LCO 3.4.16 OTSG tube integrity shall be maintained.

AND All OTSG tubes satisfying the tube repair criteria shall be plugged or repaired in accordance with the Steam Generator Program.

APPLICABILITY:

MODES 1, 2,

3, and 4.

ACTIONS

NOTE-----------------------------------

Separate Condition entry is allowed for each OTSG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more OTSG A.1 Verify tube integrity 7 days tubes satisfying the of the affected tube repair criteria tube(s) is maintained and not plugged or until the next repaired in refueling outage or accordance with the OTSG tube inspection.

Steam Generator Program.

AND A.2 Plug or repair the Prior to entering affected tube(s) in MODE 4 following accordance with the the next refueling Steam Generator outage or OTSG Program.

tube inspection B.

Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Compl eti on Time of Condition A AND not met.

B.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR OTSG tube integrity not maintained.

Crystal River Unit 3 3.4-34 Amendment No.

OTSG Tube Integrity 3.4.16 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify OTSG tube integrity in accordance In accordance with with the Steam Generator Program.

the Steam Generator Program SR 3.4.16.2 Verify that each inspected OTSG tube that Prior to entering satisfies the tube repair criteria is MODE 4 following a plugged or repaired in accordance with the OTSG tube Steam Generator Program.

inspection Crystal River Unit 3 3.4-35 Amendment No.

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.10 Steam Generator (OTSG)

Program A Steam Generator Program shall be established and implemented to ensure that OTSG tube integrity is maintained.

In addition, the Steam Generator Program shall include the following provisions:

a.

Provisions for condition monitoring assessments.

Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage.

The "as found" condition refers to the condition of the tubing during an OTSG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging or repair of tubes.

Condition monitoring assessments shall be conducted during each outage during which the OTSG tubes are inspected, plugged, or repaired to confirm that the performance criteria are being met.

b.

Performance criteria for OTSG tube integrity.

OTSG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1.

Structural integrity performance criterion:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents.

This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials.

Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse.

In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

(continued)

Crystal River Unit 3 5.0-13 Amendment No.

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.10 OTSG Program (continued)

2.

Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than an OTSG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all OTSGs and leakage rate for an individual OTSG.

Leakage is not to exceed one gallon per minute per OTSG.

3.

The operational LEAKAGE performance criterion is specified in LCO 3.4.12, "RCS Operational LEAKAGE."

c.

Provisions for OTSG tube repair criteria.

Tubes shall be plugged if the sleeved region of a tube is found by inservice inspection to contain flaws in the (a) sleeve or (b) the pressure boundary portion of the original tube wall in the sleeve/tube assembly.

The non-sleeved region of a tube found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged or repaired except if the flaws are permitted to remain in service through application of an alternate tube repair criteria discussed below.

The following alternate tube repair criteria may be applied as an alternative to the 40% depth based criteria:

1.

Pit-like Intergranular Attack (IGA) indication means a bobbin coil indication confirmed by Motorized Rotating Pancake Coil (MRPC) or other qualified inspection techniques to have a volumetric, pit-like morphology characteristic of IGA.

Inservice tubes with pit-like IGA indications in the first span of the B OTSG are acceptable provided the depth of the indication is less than 40% of the nominal tube wall thickness.

Inservice tubes with pit-like IGA indications in the first span of the B OTSG with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

(continued)

Crystal River Unit 3 5.0-14 Amendment No.

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.10 OTSG Program (continued)

2.

Tube End Cracks (TEC) are those crack-like eddy current indications, circumferentially and/or axially oriented, that are within the Inconel clad region of the primary face of the upper and lower tubesheets, but do not extend into the carbon steel-to Inconel clad interface.

Tubes with axially oriented TEC may be left in-service using the method described in Topical Report BAW-2346P, Revision 0, provided the combined projected leakage from all sources of primary-to-secondary leakage, including axial TEC indications left in-service, does not exceed the Main Steam Line Break (MSLB) accident leakage limit of one gallon per minute, minus 150 gallons per day, per OTSG.

The contribution to MSLB leakage rates from TEC indications shall be determined utilizing the methodology in Addendum B dated August 10, 2005 to Topical Report BAW-2346P, Revision 0. The projection of TEC leakage that may develop during the next operating cycle shall be determined using the methodology in Addendum C dated August 30, 2005 to Topical Report BAW-2346P, Revision 0.

Tubes identified with TEC that are allowed to remain in service under the alternate repair criteria will be added to the existing list of tubes in the OTSG Inservice Inspection Surveillance Procedure.

The inspection data for tubes with axially oriented TEC indications shall be compared to the previous inspection data to monitor the indications for growth.

Tubes with crack-like indications within the carbon steel portion of the tubesheet, circumferentially oriented TEC, or volumetric indications within the Inconel clad region of the tubesheet shall be repaired using the appropriate method from 5.6.2.10.f or removed from service by plugging the tube.

(continued)

Crystal River Unit 3 5.0-1S Amendment No.

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.10 OTSG Program (continued)

d.

Provisions for OTSG tube inspections.

Periodic OTSG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria.

The tube-to-tubesheet weld is not part of the tube.

In tubes repaired by sleeving, the portion of the original tube wall between the sleeve's joints is not an area requiring re-inspection.

In addition to meeting the requirements of d.1 through d.8 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that OTSG tube integrity is maintained until the next OTSG inspection.

An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1.

Inspect 100% of the tubes in each OTSG during the first refueling outage following OTSG replacement.

2.

Inspect 100% of the tubes at sequential periods of 60 effective full power months.

The first sequential period shall be considered to begin after the first inservice inspection of the OTSGs.

No OTSG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.

3.

If crack indications are found in any OTSG tube, then the next inspection for each OTSG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less).

If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

(continued)

Crystal River Unit 3 5.0-16 Amendment No.

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.10 OTSG Program (continued)

4.

Inservice tubes with pit-like IGA indications in the first span of the B OTSG must be inspected with bobbin and Motorized Rotating Pancake Coil (MRPC) eddy current techniques from the lower tube sheet secondary face to the bottom of the first tube support plate during each inservice inspection of the B OTSG.

Inservice tubes with pit-like IGA indications in the "B" OTSG first span shall be monitored for growth of these indications by using a test probe equivalent to the high frequency bobbin probe used in the 1997 inspection.

The indicated percentage through-wall value from the current inspection shall be compared to the indicated percentage through-wall value from the 1997 inspection.

5.

Tubes in-service with axially oriented tube end cracks (TEC) are identified in the OTSG Inservice Inspection Surveillance Procedure.

The portion of the tube with the axial TEC must be inspected using the motorized rotating coil eddy current technique every 24 effective full power months or one refueling outage, whichever is less.

6.

If the plant is required to shut down due to primary-to-secondary leakage and the cause is determined to be degradation of the TEC portion of the tubes, 100% of the tubes with TEC in that OTSG shall be examined in the location of the TEC.

If more than 1% of the examined tubes satisfy the tube repair criteria, 100% of the tubes with TEC in the other OTSG shall be examined in the location of the TEC.

7.

The repair roll in each tube will be inspected every 24 effective full power months or one refueling outage (whichever is less) while the tube with a repair roll is in service.

8.

If the plant is required to shut down due to primary-to-secondary leakage and the cause is determined to be a flaw in a repair roll, 100% of the repair rolls in both OTSGs shall be examined.

(continued)

Crystal River Unit 3 5.0-17 Amendment No.

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.10 OTSG Program (continued)

e.

Provisions for monitoring operational primary to secondary LEAKAGE.

f.

Provisions for OTSG tube repair methods.

Steam generator tube repair methods shall provide the means to reestablish the RCS-pressure boundary integrity of OTSG tubes without removing the tube from service.

For the purposes of these Specifications, tube plugging is not a repair.

All acceptable tube repair methods are listed below.

1.

Sleeve installation in accordance with the B&W process (or method) described in report BAW-2120P.

No more than five thousand sleeves may be installed in each OTSG.

2.

Installation of repair rolls in the upper and lower tubesheets in accordance with BAW-2303P, Revision 4.

The repair process (single, overlapping, or multiple roll) may be performed in each tube.

The repair roll area will be examined using eddy-current methods following installation.

The repair roll must be free of flaws for the repair to be considered acceptable.

If the repair roll is unacceptable, the tube must be repaired or plugged.

(continued)

Crystal River Unit 3 5.0-18 Amendment No.

Procedures, Programs and Manuals 5.6 THIS PAGE INTENTIONALLY LEFT BLANK Crystal River Unit 3 5.0-24 Amendment No.

Procedures, Programs and Manuals 5.6 THIS PAGE INTENTIONALLY LEFT BLANK Crystal River Unit 3 5.0-2S Amendment No.

Procedures, Programs and Manuals 5.6 THIS PAGE INTENTIONALLY LEFT BLANK Crystal River Unit 3 5.0-26 Amendment No.

Reporting Requirements 5.7 5.7 Reporting Requirements 5.7.1.2 Not Used 5.7.2 Special Reports Special Reports shall be submitted in accordance with 10 CFR 50.4 within the time period specified for each report.

The following Special Reports shall be submitted:

a.

When a Special Report is required by Condition B or F of LCO 3.3.17, "Post Accident Monitoring (PAM)

Instrumentation," a report shall be submitted within the following 14 days.

The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

b.

Any abnormal degradation of the containment structure found during the inspection performed in accordance with ITS 5.6.2.8 shall be reported to the NRC within 30 days of the current surveillance completion.

The abnormal degradation shall be defined as findings such as delamination of the dome concrete, widespread corrosion of the liner plate, corrosion of prestressing elements (wires, strands, bars) or anchorage components extending to more than two tendons and group tendons force trends not meeting the requirements of 10CFR50.55a(b)(2)(ix)(B).

The report shall include the description of degradation, operability determination, root cause determination and the corrective actions.

c.

A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.6.2.10, Steam Generator (OTSG)

Program.

The report shall include:

1.

The scope of inspections performed on each OTSG,

2.

Active degradation mechanisms found,

3.

Nondestructive examination techniques utilized for each degradation mechanism,

4.

Location, orientation (if linear), and measured sizes (if available) of service induced indications, (continued)

Crystal River Unit 3 5.0-28 Amendment No.

Reporting Requirements 5.7 5.7 Reporting Requirements 5.7.2 Special Reports (continued)

5.

Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism,

6.

Total number and percentage of tubes plugged or repaired to date,

7.

The results of condition monitoring, including the results of tube pulls and in-situ testing,

8.

The effective plugging percentage for all plugging and tube repairs in each OTSG,

9.

Repair method utilized and the number of tubes repaired by each repair method,

10.

Location, bobbin coil amplitude, and axial and circumferential extent (if determined) for each first span IGA indication, and an assessment of growth for indications in the first span of OTSG B, and

11.

Number of as-found and as-left tubes with TEC indications, number of as-found and as-left TEC indications, the number of as-found and as-left TEC indications as a function of tubesheet radius, the as-found, as-left, probability of detection and new TEC leakage for upper and lower tubesheet indications. The projected accident leakage and an assessment of growth for TEC indications will be provided.

An assessment of the adequacy of the predictive methodology in Addendum C to Topical Report BAW-2346P, Revision 0, including assessing the distribution of indications found in each OTSG to ensure the assumption regarding the similarity of the distribution of indications remain consistent from one cycle to the next and that the assumption of a linear increase in leak rate remain valid.

Corrective actions in the event that the assessment indicates the assumptions can not be fully supported.

Crystal River Unit 3 S.0-29 Amendment No.

PRORESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #264, REVISION 2 Application to Modify Improved Technical Specifications Regarding Steam Generator Tube Integrity ATTACHMENT E Proposed Improved Technical Specification Bases Pages (Mark-up) s44ke ei4 indicates deleted text.

ifghlighted text indicates added text.

RCS Operational LEAKAGE B 3.4.12 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.12 RCS Operational LEAKAGE BASES BACKGROUND During the life of the plant, the joint and valve interfaces contained in the RCS can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration.

The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety.

This LCO specifies the types and amounts of LEAKAGE.

10 CFR 50, Appendix A, GDC 30 (Ref.

1),

requires means for detecting and, to the extent practical, identifying the source of reactor coolant LEAKAGE.

Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

OPERABILITY of the leakage detection systems is addressed in LCO 3.4.14, "RCS Leakage Detection Instrumentation."

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration.

Therefore, detecting, monitoring, and quantifying reactor coolant LEAKAGE is critical.

Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leaktight.

Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

APPLICABLE SAFETY ANALYSES Except for primary to secondary LEAKAGE, the safety analyses do not address operational LEAKAGE.

However, other operational LEAKAGE is related to the safety analyses for a LOCA in that the amount of leakage can affect the probability of such an event.

The safety analysis for an event resulting in steam discharge to the atmosphere assumes 1: gpI primary to seeondary LEAKAGE as the initial ondition.

(continued)

Crystal River Unit 3 B 3.4-53 Revision No.

RCS Operational LEAKAGE B 3.4.12 BASES APPLICABLE that primary to secondary LEAKAGE from all steam generators]

SAFETY ANALYSES (OTSGs) is one gallon per minute or increases to one gallon (continued) per minute as a result of accident induced conditions.

The ILCO requirement to limit primary to secondary LEAKAGE, through any one OTSG to less than or equal to 150 gall6oins per day is significantly less than the conditions assumed in the safety analysis.l The FSAR (Ref.

3) analysis for steam generator tube rupture (SGTR) assumes the contaminated secondary fluid is only briefly released via safety valves and the majority is steamed to the condenser.

The 1 gpm primary to secondary LEAKAGE sfet-ynalysis assumption is relatively inconsequential in terms of offsite dose.

The safety analysis for the Steam Line Break (SLB) accident assumes the enti-r 1 gpm primary to secondary LEAKAGE +n one [is through the affected generator as an initial condition (Ref.

4).

The dose consequences resulting from the SLB accident meet the acceptance criteria defined in 10 CFR 50.67.

RCS operational LEAKAGE satisfies Criterion 2 of the NRC Policy Statement.

LCO RCS operational LEAKAGE shall be limited to:

a.

Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration.

LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE.

Violation of this LCO could result in continued degradation of the reactor coolant pressure boundary (RCPB).

LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

b.

Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment atmosphere and sump level monitoring equipment can detect within a reasonable time period.

Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

(continued)

Crystal River Unit 3 B 3.4-54 Revision No. 7

RCS Operational LEAKAGE B 3.4.12 BASES LCO

c.

Identified LEAKAGE

[(conti nu-d)]

to 10 gpm of identified LEAKAGE is considered Ullow able because LEAKAGE is from known sources that do not interfere with the detection of unidentified LEAKAGE and is well within the capability of the RCS makeup system.

Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE).

Violation of this LCO could result in continued degradation of a component or system.

J...

IAr

.L....

L A-This LEAKAGE limit is established to ensure that tube initially leaking during normal peaindo not a

cnservative limit which is cosistent wth the Criteria.

CR 3has, 2r-te to

olutariy adopt ti conservative limit to ensure plant shutdown in a timely manner in rspose to detection of primary to beLL i

t ude I

the totIr all A I I IIIJI 4

IIIIIEIL I*

/-l I I l-L M

i identified LEAKAG[.

Two OTSGs are also required to be OPERABLE.

This reqirment is met by satifyn teag nted inservice inspection

-e)rmet f h team Generator Tube Surveillance Program (Specification

,d

-Pi-r-ikýirr t Sco-[ýEAKAkGE-thr-u-cih A-nyOn-e O~

Th leliI*t Iof7L*II Igallons per day per-OTS`GII-sI IJsl-dlnIU

'the operational LEAKAGE performance criterion in NEI, j97-06, Steam Generator Program Guidelines (Ref. 5).-

IThe Steam Generator Program operational LEAKAGE

,performance criterion in NEI 97-06 states, "TheRCs.'

operational primary to secondary leakage through ang5-o

,ne SG shall be limited to 150 gallons per day."

Thi Ilimit is based on operating experience with OTSG tube

'degradation mechanisms that result in tube leakage:[

IThe operational leakage rate criterion in conjuncthion,

,with the implementation of the Steam Generator Program

'is an effective measure for minimizing the frequencymof

!steam generator tube ruptures.

Lrconti nued)j Crystal River Unit 3 B 3.4-55 Revision No.

c 5i 4

RCS Operational LEAKAGE B 3.4.12 BASES ACTIONS A.1 If unidentified LEAKAGE- 'or identified LEAKAGE, or priar*

to secndary L[AK."G are in excess of the LCO limits, the LEAKAGE must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down.

This action is necessary to prevent further deterioration of the RCPB.

B.1 and B.2 If any pressure boundary LEAKAGE exists o*primary_*

ssecondary LEAKAGE is not within limits,] or if unidentified-o

deified, orV, primary to s@eondary LEAKAGE cannot be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the reactor must be placed in a lower pressure condition to reduce the severity of the LEAKAGE and its potential consequences.

The reactor must be placed in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

This action reduces the LEAKAGE and also reduces the stresses that tend to degrade the pressure boundary.

The Completion Times allowed are reasonable, based on operating experience, to reach the required conditions from full power conditions in an orderly manner and without challenging plant systems.

In MODE 5, the pressure stresses acting on the RCPB are much lower and further deterioration is much less likely.

SURVEILLANCE SR 3.4.12.1 REQUIREMENTS Verifying RCS LEAKAGE within the LCO limits ensures that the integrity of the RCPB is maintained.

Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection.

Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance.

Primary to secondary LEAK(AG[ is also measured by performance of an RES water, invntory balance in conjunction with efflue monitoring within the secondary steam and condensate l(conti nued)l Crystal River Unit 3 B 3.4-56 Revision No. 10

RCS Operational LEAKAGE B 3.4.12 BASES SURVEILLANCE SR 3.4.12.1 (continued)

REQUIREMENTS The RCS water inventory balance must be performed with the reactor at steady state operating conditions and with RCS temperature greater than 400'F.

The test must be performed prior to entry into MODE 2 if it has not been performed within the past 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months near normal operating pressure.

This surveillance is imoidified by two notes.

Note 1 states!

~i*

is not required to be performed for entry into MODE 4 er-MODE 3 or for non-steady state conditions in MODE 3, but must be performed in MODE 3 above 400'F if 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation are achieved.

If the test is not performed prior to all other requirements for entry into MODE 2 being satisfied, entry into MODE 2 must be delayed until steady state operation is established and the requirements of SR 3.0.4 are satisfied.

Steady state operation is required to perform a meaningful water inventory balance; calculations during maneuvering are not useful.

For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP pump seal injection and return flows.

Note 2 states that this SR is not applicable to primary t Isecondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accuratelyby an RCS water inventory balance.[

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is reasonable to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents.

SR 3.4.12.2 This SR provides the means necessary to determine OTSG.

OPERABILITY in an operational MODE. The requirement to demontrat OTSG% tube integrity in accordance with the Steam Generator Tube Surveillance Program emphasizes the impotane of OTSG tube integrity, even though this Survellane cannot be performed at normal operating eondition-s-.

(conti nued) rytl Rvr UnIt II 3

L* UI**JLLC IILJ j

I

.*l~IUll.

B34-5 ReIsI onI No. 3ll Crystal River Unit 3 B 3.4-57 Revision No. M

RCS Operational LEAKAGE B 3.4.12 BASES S -URVEILLANCE SR 3.4.12.2 (continued)]

REQUIREMENTS This SR verifies that primary to secondary LEAKAGE is

]ess than or equal to 150 gallons per day through any one OTSG.

Satisfying the primary to secondary LEAKAGE limit ensures, that the operational LEAKAGE performance criterion in the

,Steam Generator Program is met.

If this SR is not met,]

compliance with LCO 3.4.16, "Steam Generator Tube-

!Integrity," should be evaluated.

The 150 gallons per day Ilimit is measured at room temperature as described iný Reference 6.

The operational LEAKAGE rate limit applies to LEAKAGE through any one OTSG.

If it is not practical to H assign the LEAKAGE to an individual OTSG, all the primary, to secondary LEAKAGE should be conservatively assumed to bý from one OTSG.F-The Surveillance is modified by a Note which states tha

,the Surveillance is not required to be performed until 122 hours0.00141 days 0.0339 hours 2.017196e-4 weeks 4.6421e-5 months after establishment of steady state operation.

Foq RCS primary to secondary LEAKAGE determination, steady' state is defined as stable RCS pressure, temperature, powe Ilevel, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.l-T.he Surveillance Frequency of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is a reasonable interval to trend primary to secondary LEAKAGE and___

recognizes the importance of early leakage detection in the I

prevention of accidents.

The primary to secondary LEAKAGE

' Is determined using continuous process radiation monitorsl or radiochemical grab sampling in accordance with the EPRIj

,guidelines (Ref.

6).J-REFERENCES

1.

10 CFR 50, Appendix A, GDC 30.

2.

Regulatory Guide 1.45, May 1973.

3.

FSAR, Section 14.2.2.2.

4.

FSAR, Section 14.2.2.1.

15.

NEI 97-06, "Steam Generator Program Guidelines."

EPRI, "Pressurized Water Reactor Primary-to-Secondary LLeak Guidelines."'

Crystal River Unit 3 B 3.4-58 Revision Amendment No.

149

OTSG Tube Integrity 1B 3.4.16 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.16 Steam Generator (OTSG)

Tube Integrity BASES JA-CKGROUND Steam generator (OTSG) tubes are small diameter, thin, Walled tubes that carry primary coolant through the primary to secondary heat exchanges.

The OTSG tubes have a number' of important safety functions.

Steam generator tubes arep an integral part of the reactor coolant pressure boundaryL (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory.

The OTSG tubes isolate' the radioactive fission products in the primary coolantr from the secondary system.

In addition, as part of the' RCPB, the OTSG tubes are unique in that they act as the' heat transfer surface between the primary and secondary

,systems to remove heat from the primary system.

ThisL,__-

Specification addresses only the RCPB integrity function of the OTSG.

The OTSG heat removal function is addressed by,

ýLCO 3.4.4, "RCS Loops -

MODE 3,"

LCO 3.4.5, "RCS Loops MODE 4," LCO 3.4.6, "RCS Loops -

MODE 5, Loops Filled," and

ýis implicitly required in MODES 1 and 2 in order to prevent.

'a Reactor Protection System actuation (LCO 3.3.1).-

TSG tube integrity means that the tubes are capable ofL performing their intended RCPB safety function consistent-with the licensing basis, includingaapplicable regulatory requi rements.--

ISteam generator tubing is subject to a variety of, degradation mechanisms.

Steam generator tubes may__

experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress

corrosion cracking, along with other mechanically induced phenomena such as denting and wear.

These degradation

'mechanisms can impair tube integrity if they are not-

'managed effectively.

The OTSG performance criteria aFý

,used to manage OTSG tube degradation.1 Spcification 5.6.2.10, "Steam Generator (OTSG)

Program,"

requires that a program be established and implemented to ensure that OTSG tube integrity is maintained.

Pursuant t*o

'Specification 5.6.2.10, tube integrity is maintained when-the OTSG performance criteria are met.

There are threeF-OTSG performance criteria: structural integrity, acciden-t iinduced leakage, and operational LEAKAGE.

The OTSG (continued)

Crystal River Unit 3 B 3.4-75 Revision No. XX

OTSG Tube IntegrityI B 3.4.16 BASES BACKGROUND performance criteria are described in Specificatio-_

_(continued) 5.6.2.10.

Meeting the OTSG performance criteria provides

]reasonable assurance of maintaining tube integrity a normal and accident conditions.]-

The processes used to meet the OTSG performance criteria are defined by the Steam Generator Program Guidelines!

,(Ref. 1).

APPLICABLE The steam generator tube rupture (SGTR) accident is th#

,SAFETY ANALYSES limiting design basis event for OTSG tubes and avoiding h

SGTR is the basis for this Specification.

The analysis ofl a SGTR event assumes a bounding primary to secondarY LEAKAGE rate equal to the operational LEAKAGE rate limrits in LCO 3.4.12, "RCS Operational LEAKAGE,"

plus the leakag~e rate associated with a double-ended rupture of a singleý tube.

The accident analysis for a SGTR assumes theF contaminated secondary fluid is only briefly releasedto ithe atmosphere via safety valves and the majority is' discharged to the main condenser.j The analysis for design basis accidents and transientsl other than a SGTR assume the OTSG tubes retain theiý structural integrity (i.e., they are assumed not to__

rupture).

In these analyses, the steam discharge to the atmosphere is based on the total primary to secondar7,-

LEAKAGE from all OTSGs of one gallon per minute or is-assumed to increase to one gallon per minute as a result of]

accident induced conditions.

For accidents that do not'--

,involve fuel damage, the primary coolant activity level of]

DOSE EQUIVALENT 1-131 is assumed to be equal to the LCOd 13.4.15, "RCS Specific Activity," limits.

For accidentsi ithat assume fuel damage, the primary coolant activity is a

function of the amount of activity released from the' damaged fuel.

The dose consequences of these eventsa w ithin the limits of GDC 19 (Ref.

2),

10 CFR 50.67 (Ref.3)]

or the NRC approved licensinq bases (e.g., a small fraction of these li mits).

Siteam generator tube integrity satisfies Criterion 2 ofl1(d

,CFR 50.36 (c)_(2)(li)_

[LCO The [CO requires that OTSG tube integrity be mainta7ined.*_*

The LCO also requires that all OTSG tubes that satisfy the repair criteria be plugged or repaired in accordance with' the Steam Generator Program.

(continued)

Crystal River Unit 3 B 3.4-76 Revision No. XX

OTSG Tube Integrity

'B 3.4.16 BASES

'LCO During an OTSG inspection, any inspected tube thai (continued) satisfies the Steam Generator Program repair crite'ria i's repaired or removed from service by plugging.

If a tube was determined to satisfy the repair criteria but was not plugged or repaired, the tube may still have tubeý I ntegri ty.

TIn the context of this Specification, an OTSG tube ijs, defined as the entire length of the tube, including thi tube wall and any repairs made to it, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet.

The tube-to-tubesheet weld is not

!considered part of the tube.

An OTSG tube has tube integrity when it satisfies the OTSG performance criteria.

The OTSG performance criteria are-

'defined in Specification 5.6.2.10, "Steam Generator,'

Program," and describe acceptable OTSG tube performance.

The Steam Generator Program also provides the evaluation.

process for determining conformance with the OTSG performance criteria.1 IThere are three OTSG performance criteria: structural integrity, accident induced leakage, and operationalL ILEAKAGE.

Failure to meet any one of these criteria sI

ýconsidered failure to meet the LCO.J The structural integrity performance criterion provides a margin of safety against tube burst or collapse under*

normal and accident conditions, and ensures structural integrity of the OTSG tubes under all anticipated_

transients included in the design specification.

Tube burst is defined as, "The gross structural failure of the tube wall.

The condition typically corresponds to an'.

unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile' (plastic) tearing of the tube material at the ends of th*

degradation."

Tube collapse is defined as, "For the load

'displacement curve for a given structure, collapse occurs' at the top of the load versus displacement curve where the slope of the curve becomes zero."

The structural integrity performance criterion provides guidance on assessing loads' that have a significant effect on burst or collapse.

In'-'

that context, the term "significant" is defined as "An*

accident loading condition other than differential press (continued)

Crystal River Unit 3 B 3.4-77 Revision No. XX

ITSG Tube Integrity B 3.4.16:

BASES LCO is considered significant when the addition of such lod-(continued) in the assessment of the structural integrity performanceI criterion could cause a lower structural limit or limiting burst/collapse condition to be established."

For tube'-

integrity evaluations, except for circumferentialE degradation, axial thermal loads are classified as7 __

secondary loads.

For circumferential degradation, th classification of axial thermal loads as primary or-secondary loads will be evaluated on a case-by-case basis.

The division between primary and secondary classifications!

will be based on detailed analysis and/or testingj

,*-tructural integrity requires that the primary mieimbrane stress intensity in a tube not exceed the yield strength for all ASME Code,Section III, Service Level A (normal operating conditions) and Service Level B (upset or, abnormal conditions) transients included in the design__

specification.

This includes safety factors and applicabl1 design basis loads based on ASME Code,Section III,L Subsection NB (Ref.

4) and Draft Regulatory Guide 1.121 L(Ref. 5)-.

IThe accident induced leakage performance criterion ensures!

that the primary to secondary LEAKAGE caused by a design'-

basis accident, other than a SGTR, is within the accident, lanalysis assumptions. The accident analysis assumes that accident induced leakage does not exceed one gallon per, minute per OTSG.

The accident induced leakage rater

'includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE' jinduced during the accident.

The operational LEAKAGE performance criterion provides an' observable indication of OTSG tube conditions during plant operation.

The limit on operational LEAKAGE is contained in LCO 3.4.12, "RCS Operational LEAKAGE,"

and limitsl.

primary to secondary LEAKAGE through any one OTSG to 15 gallons per day.

This limit is based on the assumption that a single crack leaking this amount would not propagate o a SGTR under the stress conditions of a LOCA or a main_

steam line break.

If this amount of LEAKAGE is due to more

,than one crack, the cracks are very small, and the above

'assumption is conservative.[

L(conti nued)j Crystal River Unit 3 B 3.4-78 Revision No.

XX

6TSG Tube Integrity

ýB 3.4.16:

BASES FA-PPLICABILITY Steam generator tube integrity is challenged when the_

pressure differential across the tubes is large.

Large

'differential pressures across OTSG tubes can only bez

'experienced in MODE 1, 2,

3, or 4.

RCS conditions are far less challenging in MODES 5 and 6

,than during MODES 1, 2,

3, and 4.

In MODES 5 and 6,*

primary to secondary differential pressure is low,[

resulting in lower stresses and reduced potential fori LEAKAGE.

ACTIONS The ACTIONS are modified by a Note clarifying that the'

,Conditions may be entered independently for each OTSG ýtube.

This is acceptable because the Required Actions provide' appropriate compensatory actions for each affected OTSG tube.

Complying with the Required Actions may allow forL___

continued operation, and subsequent affected OTSG tubes are governed by subsequent Condition entry and application ofj associated Required Actions.)

A.1 and A.2

,Condition A applies if it is discovered that one or more OTSG tubes examined in an inservice inspection satisfy týj*

itube repair criteria but were not plugged or repaired in accordance with the Steam Generator Program as required by SR 3.4.16.2.

An evaluation of OTSG tube integrity of the' affected tube(s) must be made.

Steam generator tube_-

integrity is based on meeting the OTSG performance criter-e-ia described in the Steam Generator Program.

The OTSG repair, criteria define limits on OTSG tube degradation that allow for flaw growth between inspections while still providing*

assurance that the OTSG performance criteria will continue to be met.

In order to determine if an OTSG tube that-should have been plugged or repaired has tube integritya evaluation must be completed that demonstrates that ther

,OTSG performance criteria will continue to be met until'-thie next refueling outage or OTSG tube inspection.

The tube*

integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next OTSG tube inspection.

If it is determined that tube integrityY is not being maintained, Condition B ap ples.

(continued)

Crystal River Unit 3 B 3.4-79 Revision No. XX

)TSG Tube Integrity

'B 3.4.16:

BASES

'A-CTIONS A.1 and A.2 (continued)J A Completion Time of 7 days is suffi-cient to complete the evaluation while minimizing the risk of plant operation w ith an OTSG tube that may not have tube integrity.F

,If the evaluation determines that the affiect-ed tube(s) F-ha

ýtube integrity, Required Action A.2 allows plant operation*

,to continue until the next refueling outage or OTSG_____

inspection provided the inspection interval continues to be supported by an operational assessment that reflects theý affected tubes.

However, the affected tube(s) must be_

plugged or repaired prior to entering MODE 4 following th-e next refueling outage or OTSG inspection.

This Completion' ITime is acceptable since operation until the next--

inspection is supported by the operational assessmen.

B.1 and B.I

ýI-f the Required Actions and associated Completion Times of,

'Condition A are not met or if OTSG tube integrity is notý,

being maintained, the reactor must be brought to MODE 37 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .!

The allowed Completion Times are reasonable, based on operating experience, to reach the desired plant conditions' from full power conditions in an orderly manner and wLithout hallenging~

pant systems.

,URVEI LLANCE SR 3.4.16.-11

'REUIREMENTS During shutdown periods the OTSGs are inspected as required by this SR and the Steam Generator Program.

NEI 97-06,]

Steam Generator Program Guidelines (Ref.

1),

and itsý referenced EPRI Guidelines, establish the content of-the

,Steam Generator Program.

Use of the Steam Generator__

1Program ensures that the inspection is appropriate and consistent with accepted industry practices.1 During OTSG inspections a condition monitoring assessment of the OTSG tubes is performed.

The condition monitoring' assessment determines the "as found" condition of the OTSG tubes.

The purpose of the condition monitoring assessment'

'is to ensure that the OTSG performance criteria have been-met for the previous operating period.r (continued)

Crystal River Unit 3 B 3.4-80 Revision No. XX

,OTSG Tube Integrity

ýB 3.4.16 BASES

-URaVEILLANCE SR 3.4.16.1 (continued)j REQUIREMENTS,

[The Steam Generator Program determines the scope of th*

inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repair criteria.

!Inspection scope (i.e., which tubes or areas of tubing within the OTSG are to be inspected) is a function of'

,existing and potential degradation locations.

The Steam -

Generator Program also specifies the inspection methods to be used to find potential degradation.

Inspection methods'

'are a function of degradation morphology, non-destructive'

'examination (NDE) technique capabilities, and inspection 11 ocati ons.-

The Steam Generator Program defines the Frequency of S.

3.4.16.1.

The Frequency is determined by the operationil]

assessment and other limits in the OTSG examination guidelines (Ref.

6).

The Steam Generator Program usesl Sinformation on existing degradations and growth rates to, determine an inspection Frequency that provides reasonabl' assurance that the tubing will meet the OTSG performance4

ýcriteria at the next scheduled inspection.

In addition,l Specification 5.6.2.10 contains prescriptive requirements, concerning inspection intervals to provide added assurance

,that the OTSG performance criteria will be met between ischeduled inspections.]

SSR 3.4.16.2I During an OTSG inspection, any inspected tube thfat__

Isatisfies the Steam Generator Program repair criteria i*

repaired or removed from service by plugging.

The tube repair criteria delineated in Specification 5.6.2.10 ar4 intended to ensure that tubes accepted for continued lservice satisfy the OTSG performance criteria with' lallowance for error in the flaw size measurement and forL future flaw growth.

In addition, the tube repair criteria_,]

in conjunction with other elements of the Steam Generator Program, ensure that the OTSG performance criteria will continue to be met until the next inspection of the subject tube(s).

Reference 1 provides guidance for performing___

9operational assessments to verify that the tubes remaining in service will continue to meet the OTSG performance criteria.r Steam generator tube repairs are only performed using papproved repair methods as described in the Steam Generator Program.

(continued)

Crystal River Unit 3 B 3.4-81 Revision No. XX

IOTSG Tube Integrit 1

B 3.4.16:

BASES

'SURVEILLANCE SR 3.4.16.2 (continued)]

'REQUIREMENT$S___

jThe Frequency of prior to entering MODE 4 following a OTSG_

inspection ensures that the Surveillance has been completed and all tubes meeting the repair criteria are plugged or, repaired prior to subjecting the OTSG tubes to significant primary to secondary pressure differential.F REFERENCES

1.

NEI 97-06, "Steam Generator Program Guidelines."

,2.

10 CFR 50 Ap-pendix A, GDC 19.Q F3.

10 CFR 50.67.1 L_ ASME Boiler and Pressure Vessel Code,Section III,]

'Subsection NB.I

5. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," August 1976.

6.

EPRI, "Pressurized Water Reactor Steam Generatr 2 Exami nati on Gui delines. "1 Crystal River Unit 3 B 3.4-82 Revision No. XX

PRORESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #264, REVISION 2 Application to Modify Improved Technical Specifications Regarding Steam Generator Tube Integrity ATTACHMENT F Proposed Improved Technical Specification Bases Pages (Revision Bar Format)

RCS Operational LEAKAGE B 3.4.12 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.12 RCS Operational LEAKAGE BASES BACKGROUND During the life of the plant, the joint and valve interfaces contained in the RCS can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration.

The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety.

This LCO specifies the types and amounts of LEAKAGE.

10 CFR 50, Appendix A, GDC 30 (Ref.

1), requires means for detecting and, to the extent practical, identifying the source of reactor coolant LEAKAGE.

Regulatory Guide 1.45 (Ref.

2) describes acceptable methods for selecting leakage detection systems.