Text

License Amendment Request for Proposed Revision of Technical Specification (TS) 3.4.17, Steam Generator (SG) Tube Integrity; TS 3.7.18, Steam Generator Level; TS 5.5.8, Steam Generator Program; and TS 5.6.6
	ML13018A350
Person / Time
Site:	Davis Besse
Issue date:	01/18/2013
From:	Lieb R; FirstEnergy Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	L-13-040
	Download: ML13018A350 (49)
	v • d • e

FENOC*

fr rstEnepy Nuclear

@rating Campany 55Ot North State Routa 2 OakHarbot Ahto 43449 flaymondA, LIeb Vice President, Nuclear 4t9-321-7678 Fwc:4tS32t-7582 January 18, 2013 L-13-040 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 10 cFR 50.90

Subject:

Davis-Besse Nuclear Power Station Docket No, 50-346, License No. NPF-3 License Amendment Request for Proposed Revision oJ "Tschnical

$pecification ffS).

3.4.:17, "Stqam Generator (SG\\ T-uhg Integrity":

T$ 3.7.18. "Steam,.fiBnerator Level"l Ts Q*-$*.9.,j'SteAm,gqnerator (SG) Program":

and TS 5.6.,S,,

"Steam Generator Tube lnspection Report" FirstEnergy Nuclear Operating Company (FENOC) hereby requests arnendment of Operating License NPF-3 for the Davis-Besse Nuclear Power Station (DBNPS). The proposed amendment would revise the following Technical Specifications (TS) associated with steam generators (SGs):

TS 3.4.17, "Steam Generator (SG) Tube Integrity" TS 3.7.18, "Steam Generator Level"

. TS 5.5.8, "Steam Generator (SG) Program"

. TS 5.6.6, "Steam Generator Tube Inspection Report" Revision of these TSs is required to support plant operation following replacement of the original SGs during a refueling outage scheduled to be completed at DBNP$ in April zAM. The propo-sed changes are necessary because of dimensional and material differences between the original SGs and the replacement SGs. Additionally, proposed changes are needed to address implementation issues associated with inspection

periods, and to address other administrative changes and clarifications, consistent with the guidance provided in Technical Specification Task Force Traveler TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,"

approved by the Nuclear Regulatory Commission (NRC) on October 27,2011 {76 FR 60763).

The FENOC evaluation of the proposed amendment is enclosed. Approval of the proposed license amendment is requested by February 1,2014, to be implemented prior to startup following the refueling outage during which the SG replacement is to occur.

a a

Davis-Besse Nuclear Power Station L-13-040 Page 2 There are no regulatory commitments contained in this letter. lf there are any questions, or if additional information is required, please contact Mr. Thomas A. Lentz, Manager

- Fleet Licensing, at 330-315-6810.

I declare under penalty of perjury that the foregoing is true and correct. Executed on January 18, 2013.

WM

Enclosure:

Evaluation of the Proposed Amendment cc: NRC Region lll Administrator NRC Project Manager NRC Resident Inspector Executive

Direetor, Ohio Emergency Management

Agency, State of Ohio (NRC Liaison)

Utility Radiological Safety Board

EVALUATION OF THE PROPOSED AMENDMENT

Subject:

Proposed Revision of Technical Specification (TS) 3.4.17, "Steam Generator (SG) Tube Integrity";

TS 3J,18, 'Steam Generator Level"; TS 5.5.8, "Steam Generator (SG) Program";

and TS 5.6.6, "Steam Generator Tube Inspection Report'for the Davis-Besse Nuclear Power Station, Unit Number 1 (DBNP$) Operating License Number NPF-3 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

4.1 No Significant Hazards Consideration Analysis 4.2 Applicabte Regulatory Requirements/Criteria 4.3 Precedent 4.4 Conclusions 5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

Attachments

1. Proposed changes to Technical Specifications, Annotated Copy

2. Proposed changes to Technical Specifications, Retyped Copy

3. Proposed changes to Technical Specification Bases, Annotated Copy

Davis-Besse Nuclear Power Station Evaluation of Proposed Amendment Page 2 of 11 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend Operating License Number NPF-3 for the Davis-Besse Nuclear Power Station, Unit Number 1 (DBNPS). The proposed change would revise four Technical Specifications (TS) associated with the steam generators (SGs). Replacement of the SGs is being perforrned as a design modification in accordance with the provisions of 10 CFR 50.59, "Changes, tests and experiments."

Nuclear Regulatory Commission (NRC) review and approval of the modification is not being requested herein.

Revision of the T$s is required to support plant operation following replacement of the original SGs, which is scheduled to be completed at DBNPS in April zAM. The proposed changes are necessary because of dimensional and material differences between the original SGs and the replacement SGs. Additionally, proposed changes are needed to address implementation issues associated with inspection periods, and to address other administrative changes and clarifications, consistent with the guidance provided in Technical Specification Task Force Traveler TSTF-5l0, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,"

approved by the Nuclear Regulatory Comrnission (NRC) on October 27,zOfi.

The TSs that would be revised by the proposed amendment are:

. TS 3.4.17, "Steam Generator (SG) Tube Integrity"

. TS 3.7.18, "Steam Generator Level"

. TS 5.5.8, "Steam Generator (SG) Program"

. TS 5.6.6, "Steam Generator Tube lnspection Report

Affected pages of the current TSs, annotated to show the proposed changes, are provided in Attachrnent

1. Re{yped TS pages are provided in Attachment

2. Proposed changes to the TS Bases are identified by annotation in Attachrnent

3. TS Bases are not part of the Technical Specifications, are not submitted for NRC approval, and are provided for information only.

TSs 3.4.17, 5.5,8, and 5.6.6 impose monitoring, inspection, repair and reporting requirements that ensure SG tube integrity is maintained consistent with DBNPS accident analysis assumptions and regulatory requirements.

The requirements currently imposed by these TSs are based on the analyses and tube materials of the original SGs. The proposed changes would impose requirements that reflect the analyses and tube materials of the replacement SGs, consistent with the guidance provided in TSTF-5l0, Revision 2.

TS 3.7.18 imposes SG secondary side inventory restrictions based on analyses specific to the original SG physical deslgn characteristics, to ensure that plant operation remains bounded by the values used in ine Main Steam Line Break (MSLB) analyses presented in the DBNPS Updated Safety Analysis Report (USAR). The proposed changes would impose inventory restrictions that are appropriate for the physical characteristics of the replacement SGs.

Davis-Besse Nuclear Power Station Evaluation of Proposed Amendment Page 3 of 1 2.0 DETAILED DESCRIPTION Gurrent TS 5.5.8, "Steam Generator (SG) Program,' provide$ requirements to establish and irnplement a program to ensure that SG tube integrity is maintained. Program req uirements include the following:

r Provisions for SG condition monitoring (TS 5,5,8.a)

. Performance criteria for SG tube integrity (TS 5.5.8.b)

. SG tube repair criteria (TS 5.5.8.c)

I Provisions for SG tube inspections (TS 5.5.8.d) o Provisions for monitoring operational primary to secondary LEAI(AGE (TS 5.5.8.e)

. Provisions for SG tube repair methods (TS 5.5.8.0 r Requirements for special visual inspections (TS 5.5.8.9)

Consistent with T$TF-510, Revision 2, approved tube repair methods for SGs are to be listed in TS 5.5.S. Because there are currently no approved repair methods for the replacement SGs to be installed at DBNPS, TS 5,5.8.a, c, and d would be revised to eliminate references to repairoptions and TS 5.5.8.f would be deleted in its entirety.

The requirements for special visual inspections of the internal auxiliary feedwater header, header-to-shroud attachment welds and external header thermal sleeves are unique to DBNPS, and were necessary due to operational events that damaged the internal auxiliary feedwater header in the original SGs. The design of the replacement SGs does not incorporate an internal auxiliary feedwater header. Because these requirements are not applicable to the replacement SGs, TS 5.5.8.9 would be deleted in its entirety. Minor editorial corrections are proposed for TS 5.5.8.b and there are no changes proposed for TS 5.5.8.e. Additional proposed changes to T$ 5,5.8 are consistent with TSTF-510, Revision 2.

Consistent with the proposed changes to TS 5.5.8 described above, Limiting Condition for Operation

{LCO) 3,4,17, and its associated ACTIONS and SURVEILLANCE REQUIREMENTS, would be revised to detete any reference to tube repair. The revised specification would only allow tube plugging, and directly reflects the wording provided in TSTF-510, Revision 2.

TS 5.6.6, "Steam Generator Tube lnspection Report,"

requires a report to be submitted within 180 days afterthe initial entry into MODE 4following completion of SG inspections required by TS 5.5.8 and specifies the required content for such report.

Consistent with the proposed changes to TS 5.5.8 as discussed above, TS 5.6,6 would be revised to eliminate references to SG tube repairs. The proposed changes to TS 5.6.6 are consistent with TSTF-510, Revision 2.

TS 3.7.1B, "Steam Generator Levet,"

establishes limits on SG level based on operating MODE and plant operating conditions.

The four restrictions currently established in LCO 3.7.18 are based on the specific physical design characteristics and dimensions of the original SGs. Because of differences in design characteristics and dimensions of the replacement SGs, TS 3.7.18 would be revised to ensure appropriate secondary-side inventory limits are imposed.

Davis-Besse Nuclear Power Station Evaluation of Proposed Amendment Page 4 ot 11 Specifically, LCO 3.7.18.a provides the maximum SG level allowed in MODE 1 or 2, by reference to TS Figure 3.7.18-1

, "Maximum Allowable Steam Generator Level."

Figure 3J,18-1 identifies acceptable Steam Generator Operate Range level indication (in percent) as a function of steam superheat (in degrees F). Because of secondary-side dimensional and thermal performance differences between the original and replacement SGs, current TS Figure3,7.18-1 is not appropriate for use with the repfacement SGs and is to be replaced. Replacement Figure 3.7.18-1 was developed using the same methodology used for the original, with the supporting analyses based on the dimensions and thermal performance of the replacement SGs.

TS 3.7.18.b, c, and d provide the SG inventory limitations for MODE 3, based on plant operating conditions.

No changes are proposed to LCO 3.7.18.b, which establlshes the maximum $G water level with Steam and Feedwater Rupture Control System ($FRC$)

Instrumentation, Main Steam Line Pressure

- Low (LCO 3.3.11, Function 1), not bypassed. LCO 3J.18.c would be revised to reduce the maximum SG water level with LCO 3.3.11 Function 1 bypassed and both main feedwater pumps not capable of supplying feedwater to the SGs. LCO 3,7.18.d would be revised to reduce the maximum SG water level with LCO 3.3.11 Function 1 bypassed and one main feedwater pump capable of supplying feedwater to the SGs.

3.0 TECHNICAL EVALUATION The SGs in pressurized water reactor designs remove heat from the reactor coolant system (RCS) and produce stearn to operate the main generator and other balance-of-plant equipment. A principat function of the steam generators at the DBNP$ is to provide superheated steam at a constant pressure over the full range of power operation. Steam generator secondary side water inventory is maintained large enough to provide adequate primary to secondary heat transfer, and that inventory increases with increasing plant power. Inventory is controlled by the feedwater control system to maintain appropriate SG level and primary system temperature at various operating power levels.

SG tubes constitute the heat transfer surface between the primary and secondary systems, and, as such, are relied on to maintain the primary system's pressure and inventory, As an integral part of the reactor coolant pressure boundary (RCPB), the SG tubes isolate the radioactive fission products in the primary coolant from the secondary system in the SGs. Maintaining tube integrity ensures that the tubes are capable of performing their intended safety functions consistent with the plant licensing basis and applicab le regulatory requirements.

The licensing basis for the DBNPS includes the postulation of a SG tube rupture (SGTR) accident. In the event of a SGTR, primary coolant is released into the secondary side of the SG, and subsequently can be released to the environment through main steam safety valves or leak paths in the secondary system. A SGTR is a Class 3 design basis accident for which analyses are summarized in the Updated Safety Analysis Report (USAR), section 15.4,2, "Steam Generator Tube Rupture,"

ln order to ensure that the probability of a SGTR does not increase above that assumed in the accident analysis and that no other design basls accidents or transients result in tube failure, it is nece$sary to maintain SG tube integrity. For that purpose, TS 5.5.8,

Davis-Besse Nuclear Power Station Evaluation of Proposed Amendment Page5of11 "Steam Generator

($G) Program,"

imposes reguirements for monitoring, inspection, and maintenance to ensure SG tube integrity remains consistent with licensing basis assumptions related to SGTR and other design basis accidents and transients.

Current TS 5,5.8 requirements are based on many years of industry experience in operaling and maintaining the original SGs. Information developed throughout the commercial nuclear industry associated with SG tube degradation mechanisms and related inspection and detection techniques for SG tubing manufactured from Alloy 600 provides the basis for the specific inspections and inspection frequencies currently stipufated in TS 5.5.8. TSTF-510, Revision2, provides the NRC-approved revisions to TS 5.5.8 as provided in the Standard Technical Specifications for Babcock and \\Mlcox Plants (NUREG-1430, Revision 4). The revisions stipulate the SG tube inspections and inspection frequencies that have been determined to be appropriate for replacement SGs. TSTF-510, Revision 2, includes separate sections for the different types of tubing materials that may be used in replacement SGs. One of the tubing materials addressed is thermally treated Alloy 690, the material that is being used for the tubes in the DBNPS replacement SGs, The proposed changes to TS 5.5.8 directly reflect the wording provided in TSTF-S10, Revision 2,for thermally treated Alloy 090 tubing.

Consistent with TSTF-510, TS 5.5.8.c specifies a plant specific value for the maxirnum allowed tube flaw $ize, as the flaw depth in percent of tube wall thickness. By conservative analyses for the DBNPS replacement SGs, FENOC has determined that continued use of a maximum flaw depth of 40% through wall is appropriate.

Although unchanged from the current specified value for maximum flaw depth, this value is based on conservative calcutations of tube loads and flaw geometry appropriate for the DBNPS Replacement SGs, based on design and industry experience.

Current TS 3.4.17, "Steam Generator Tube lntegrity,"

establishes the requirement to maintain SG tube integrity in MODEs 1, 2, 3, and 4 and establishes the requirements for addressing SG tubes that satis! tube repair criteria through either tube plugging or repair in accordance with the Steam Generator Program. TSTF-510, Revision 2, specifies that reference to repair options be deleted from TS 3.4.17 if no repair methods are approved. Consistent with the proposed changes to TS 5.5.8 described

above, Limiting Condition for Operation (LCO) 3.4.17, and its associated ACTIONS and SURVEILLANCE REQUIREMENTS, would be revised to delete any reference to tube repair. The revised specification would only allow tube plugging, and directly reflects the wording provided in TSTF-510, Revision 2.

TS 5.6.6, "Steam Generator Tube Inspection Report,"

specifies the required content for a report to be submitted within 180 days after the inilial entry into MODE 4 following completion of SG inspections required by TS 5.5.8. The proposed changes to TS 5.6'6 reflectthe changes specified in TSTF-510, Revision 2, to clarify the reporting requirements and to ensure consistency with the proposed changes to TS 5.5.8. These changes have no'impact on plant operation or SG inspection requirements.

The licensing basis for DBNPS also includes analyses of postulated main steam line breaks (MSLB), as discussed in the USAR. SG operation must remain bounded by the inputs and assumptions of the MSLB accident analyses at all times. In the event of an MSLB, there is a rapid release of SG secondary side ma$s and energy to the area in which the break is located. A MSLB is a Class 3 Design Basis Accident for which

Davis-Besse Nuclear Power Station Evaluation of Proposed Amendment Page 6 of 11 analyses are summarized in USAR section 6.2, 'Containment Systems,"

and in section 15.4.4, "Steam Line Break."

TS 3.7.18, "Steam Generator Level,'establishes limits on SG level based on operating MODE and plant operating conditions.

The SG level limits imposed by this TS ensure that the combination of SG secondary-side inventory and energy available for release in the event of a MSLB do not exceed the mass and energy releases that have been analyzed for this event. Therefore, operation within the limits of LCO 3,7,18 ensures that the MSLB accident analyses remain bounding.

There are four restrictions stated in LCO 3.7,18, based in part on the specific physical design characteristics and dimensions of the original SGs. Because the replacement SGs differ in some of the design characteristics and dimensions that are the bases for these reslrictions, TS 3.7,18 would be revised to ensure appropriate secondary-side inventory limits are imposed on the replacement SGs.

Specifically, LCO 3.7.18.a provides the maximum SG water level allowed in MODE 1 or 2, by reference to TS Figure 3J.18-1, "Maximum Allowable Steam Generator Level."

Figure 3.7.18-1 is a curve of degrees of main steam superheat on the horizontal axis and SG level, as indicated on the Steam Generator Operate Range Level instrumentation, on the vertical axis. Because of secondary-side dimensional and thermal performance differences between the original and replacement SGs, the existing curve is not appropriate for use with the replacement SGs. The proposed new curve was developed using the same melhodology that was used for developing the original curue, but the supporting analyses are based on the dimensions and thermal performance of the replacement SGs. The existing USAR MSLB analysis remains bounding.

The other three TS 3.7.18 restrictions, 3.7.18.b, c, and d, provide the SG inventory limitations for MODE 3. The totat mass and energy available for release in MODE 3 is a function of both the initiaf SG level and the available feedwater capacity, which is assumed in the MSLB analyses to continue to feed the SG for some length of time following the steain line break. The associated MODE 3 analyses address the various feedwater capacities that can exist and the associated maximum allowed SG level. The MODE 3 LCO restrictions limit lhe combination of the initial SG level and the available feedwater capability to ensure that the current MODE 1 MSLB analysis remains bounding. To provide the technical bases for the proposed LCO 3.7.18.b, c, and d requirements, calculations were performed using the same gnethodology as that used in the original calculations.

These calculations utilize the current MODE 1 MSLB mass and energy release assumptions and the design characteristics of the replacement SGs. As a result, adjustments to limits on maximum water level and Main Feedwater Pump capability are proposed for LCOs 3.7.18 c and d, as appropriate.

No changes to the existing USAR MSLB analyses are required to support the proposed MODE 3 TS changes.

4.0 REGULATORYEVALUATION

FirstEnergy Nuclear Operating Co is requesting amendmenl of Operating License Number NPF-3 for the Davis-Besse Nuclear Power Station, Unit Number 1 (DBNPS),

to

Davis-Besse Nuclear Power Station Evaluation of Proposed Amendment Page 7 of 11 revise four Technical $pecifications (TS) associated with the steam generators (SG).

The TSs that would be revised by the proposed amendment are:

. TS 3.4.17, "Sleam Generator (SG) Tube Integrity"

. TS 3.7.18, "Steam Generator Level" I TS 5.5.8, "Steam Generator (SG) Program"

. TS 5.6.6, "Steam Generator Tube Inspection Report" Revision of the TSs is required to support plant operation following replacement of the original SGs, which is scheduled to be completed at DBNPS in April zAM. Proposed changes are necessary because of dimensional and material differences between the original SGs and the replacement SGs. Additionally, proposed changes are needed to address implementation issues associated with the inspection periods, and address other administrative changes and clarifications, consistent with the guidance provided in Technical Specification Task Force Traveler TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,n approved by the Nucf ear Regulatory Commission (NRC) on October 27,2011.

4,1 No Significant Hazards Consideration Analysis Pursuant to 10 CFR 50.91(a),

"Notice for public comment,'FENOG hereby provides the required analysis about the issue of no significant hazards consideration, using the three standards set forth in 10 CFR 50.92, lssuance of amendment."

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

Respoqse: No.

For T$ 3.4.17, "Steam Generator (SG) Tube Integrity,"

a steam generator tube rupture (SGTR) event is the relevant design basis accident analyzed in the licensing basis for DBNPS. TS 3.4.17 and TS 5.5.8, "Steam Generator (SG) Program,"

impose monitoring and inspection requirements that ensure tube integrity is rnaintained.

The proposed changes to these TSs would implement monitoring and inspection requirements appropriate for the design and materials of the replacement SGs. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the $G tubes are inspected such that that the integrity of the SG tubes is verified to be maintained at a tevel that prevents an increase in the probability of a SGTR. Therefore the proposed changes to these TSs will not increase the probability of a SGTR.

The radiological consequences of a SGTR are bounded by using conservative assumptions in the design basis accident analysis, and are dependent upon the pre-existing primary-to-secondary leak rate, the flow rate through the ruptured tube, the radiological isotopic content of the RCS and the release paths. The monitoring and inspection requirements imposed by TS 3.4.17 and TS 5.5.8 are intended to ensure that SG tube integrity is maintained.

The proposed changes to these TSs would irnplement monitoring and inspection requirements appropriate for the design and

Davis-Besse Nuclear Power Station Evaluation of Proposed Amendment Page8of11 materials of the replacement SGs and would not affect radiological releases in the event of an SGTR. The radiological isotopic content of the RC$ and the release paths are not affected by any of the requirements in the current TS 3.4.17 or TS 5.5.8 or proposed revisions thereto. Therefore, the proposed changes to these TSs will not increase the consequences of a SGTR.

TS 5,6.b, "Steam Generator Tube tnspection Report," specifies information that is to be reported to the NRC following SG inspections performed in accordance with the Steam Generator Program requirements contained in TS 5.5.8. The requirement to provide this report is administrative in nature and the content of this report can have no effect on the probability or the consequences of an accident previously evaluated.

LCO 3.7.18, "Steam Generator Level," ensures that the plant is operated within the SG inventory limits that were used as initiat conditions in the current accident anatysis for a Main Steam Line Break (MSLB). The SG inventory is not an accident initiator and does not atfect any accident initiator. Therefore, the proposed changes in SG inventory limits will not increase the probability of a MSLB accident.

The radiological consequences of a MSLB are dependent upon the total SG inventory released, the SG primary-to-secondary leakage rale, the radiological isotopic content of the RCS, and the release paths. The revision to LCO 3,7.18 will ensure that the total inventory released remains bounded by the existing analysis. None of the other factors listed above are affected by the revised operating limits on SG inventory that are proposed in the revisions to LCO 3.7.18. Therefore, the proposed changes in SG inventory limits will not increase the consequences of a MSLB.

Based on the above, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluaied.

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

Response

No.

The proposed changes support replacement of the SGs at the DBNPS.

Replacement of the SGs is being performed as a design modification in accordance with the provisions of 10 CFR 50.59, "Changes, tests and experiments,"

The proposed changes to TS 3.4.17, TS 5.5.8 and TS 5.6.6 would implement monitoring and inspection requirements appropriate for the design and materials of the replacement SGs, and establish appropriate reporting requirements.

These changes would not affect the method of operation of the SGs. The proposed changes to TS 3.7.18 would ensure that the replacement SGs will be operated in accordance with existing analyses.

None of the proposed changes would introduce any changes to the plant design. ln addition, the proposed changes would not impact any other plant system,or component.

Davis-Besse Nuclear Power Station Evaluation of Proposed Amendrnent Page 9 of 11 The proposed changes would continue to prevent loss of SG tube integrity, and would ensure operation within the bounds of existing accident analyses.

There are no accident initiators created or affected by these changes.

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

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

Response: No.

The SG tubes in pressurized water reactors are an integral part of the reactor coolant'system (RCS) pressure boundary and, as such, are relied upol !q maintain ihe primary system's pressure and inventory. As part of the RCS pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the

$G tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In sumrnary, the safety function of a SG is maintained by ensuring the integrity of its tubes and the ability to remove residual heat from the primary system.

The proposed changes will ensure that the existing margins of safety are. _

maintained following tne replacement of SGs. The changes to LCO 3.4.17 and TSs 5.5.8 and 5.6.6 impose requirements for SG tube integrity monitoring, inspection, and reporting that will ensure that there is no reduction in the ability of the tubes to perform their RCS pressure boundary and heat transfer functions. The changes to LGO 3.7.18 ensure the MSLB accident analyses remain bounding. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above responses, FENOC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50,92(c),

and, accordingly, a finding of "no significant hazards consid eratio n" is j ustified.

4.2 Applicable Regulatory Requirements/Criteria The following lists the regulatory requirements and plant-specific bases related to the proposed changes.

. The regulatory basis for TS 3.4,17, "Steam Generator Tube Integrity,'

TS 5.5.S, "Steam Generator (SG) Program,"

and TS 5.5.6, "Steam Generator Tube Inspection Report," is to ensure that the integrity of the reactor coolant pressure boundary is maintained consistent with design requirements and accident analyses, and to ensure that appropriate information relative to SG inspection activities is reported to the NRC.

. The regulatory basis for TS 9J,18, "Steam Generator Level" is to ensure that operation is bbunded by the initial condition assumptions for the mass and energy released to containment in the MSLB accident analysis. Failure to mainiiin these initial conditions could result in an increase in the maximurn

Davis-Besse Nuclear Power $tation Evaluation of Proposed Amendment Page fi ol 11 temperature and pressure in containment, exces$ive cooling of the RCS, and related core reactivity effects.

The proposed amendment has been evaluated against the following NUREG-0800 Standard Review Plan sections to determine whether applicable regulations and requirements would continue to be met.

. Section 5.4, Reactor Goolant System Component and $ubsystem Design

. Section 5,4,2.1, $team Generator Materials o Section 5.4.2.2, Steam Generator (SG) Program

. Section 10.3, Main Steam Supply System

. Section 10.4.9, Auxiliary Feedwater System (PWR)

FENOC has determined that the proposed amendment does not require any exemptions or relief from regulatory requirements, and does not affect conformance to any 10 CFR 50 Appendix A General Design Criteria as described in the NUREG sections or in the DBNPS USAR.

4.3 Precedent The proposed changes to TS 3.4,17, "Steam Generator Tube Integrity,"

TS 5.5.8, "Steam Generator (SG) Program,"

and TS 5.6.6, "Steam Generator Tube Inspection Report,"

are consistent with sections 3.4.17, "$team Generator Tube Integrity,u 5,5.9, "Steam Generator (SG) Program,"

and 5.6.7, "Steam Generator Tube Inspeclion Report" of the Standard Technical Specifications

- Babcock and Wilcox Piants (NUREG-1430, Revision

4) and TSTF-S10, Revision 2, approved by the NRG on October 27,2011.

4.4 Gonclusions The proposed changes are intended and structured to maintain compliance with the appiicable regulitory requirernents and criteria identified in section 4.2 and with the guidance provided in NR0-approved TSTF-510, Revision 2.

The proposed changes to TS 3.4.17, TS 5.5.8, and TS 5.6.6 implement NRC-approved guidance that provides reasonable assurance that tube integrity will be maintained in the interyal between SG inspections.

These proposed changes result in continued assurance of the function and integrity of $G tubes.

The change$

to LCO 3,4.17, "Steam Generator Tube Integrity,"

and TS 5.5.8, "Steam Gbnerator (SG) Program,"

will ensure that SG tube integrity is maintained at a high level, consistent with the assumptions and probabilities previously applied to SG iube rupture events and other ptant events that challenge SG tube integrity' These changes also implement the NRC-approved guidance disseminated in TSTF-sl0, Revision 2.

The change$

to TS 5.6.6, "Steam Generator Tube lnspection Report,"

clarifiJ the required contents of this report and irnplement the NRC-approved guidance disseminated in TSTF-510, Revision 2.

The proposed changes to TS 9J.18, nsteam Generator Level," implement the same type of controls that exist for the original SGs. The changes would only

Davis-Besse Nuclear Power $tation Evaluation of Proposed Amendrnent Page 11 of 11 revise the specific values to be appropriate for the design characteristics and dimensional ditferences in the replacement SGs. The proposed changes provide the same asburance of conformance to the current licensing basis.

The changes to LCO 3.7.18, "Steam Generator Level,"

will ensure that the secondary side inventory in the replacement SGs is maintained within the limits previously analyzed for an MSLB event.

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

5.0 ENVIRONMENTAL CONSIDERATION FENOC has determined that the proposed license amendment would change a requirement with respect to the installation or use of a facility component located within the restricted area, as delined in 10 CFR 20, or would change an inspection or surveillance requirement.

However, the proposed amendrnent does not involve:

(i) a significant hazard's consideration; (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in t0 CFR 51.22(cXg).

Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. DBNPS License No. NPF-3, Amendment No. 286, Dated October 2,2012.

2.

Technical Specification Task Force, TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,"

October 27,2011

3.

DBNPS Updated Safety Analysis Report

. Section 15.4,2, "Steam Generator Tube Rupture"

. Section 15.4.4, "Steam Line Break"

4.

Code of Federal Regulations, Title 10, Energy r 10 CFR 51, "ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC LICENSING AND RELATED REGULATORY FUNCTION$-

Proposed Revlslon of Technlcal Specification (TS) 3.4.17, e'Steam Generator (SG)

Tube Integrlty";

T$ 3.7.18, "Steam Generator Level"; TS 5.5.8, "$team Generator (SG) Program"l and TS 5.6.6, "Steam Generator Tube Inspectlon Report" for the Davis-Besse Nuclear Power Station, Unit Number 1 (DBNPS)

Operating License Number NPF-3 Attaqh!4.qnf 1,

, Proposed Ghanges to Technlcal Specifications Annotated Copy (12 Pages Follow)

SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17

$team Generator (SG) Tube Integrity LCO 3,4.17 APPLICABILITY:

ACTIONS SG tube integrity shall be maintained.

AI\\tA All SG tubes satistying the tube rcpaipplUggingcriteria shall be plugged er reBaire+ln accordance with the Steam Generator Program, MODES 1,2, 3, and 4.

NOTE----

Separate Condition entry is allowed for each SG tube.

CONDITION One or more $G tubes satisffing the tube reBair olugging criteria and not plugged er+epairedin accordance with the Steam Generator Program.

B. Required Action and associated Completion Time of Condition A not met.

oR SG tube integ;ity not maintained.

COMPLETION TIME 7 days Prior to entering MODE 4 following the next refueling outage or SG tube inspection 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months REQUIRED ACTION A.1 Verify tube integritY of the affected tube(s) is maintained until the next refueling outage or $G tube inspection.

AND A.2 Plug e+repatrthe affected tube(s) in accordance with the Steam Generator Program.

Be in MODE 3.

Be in MODE 5.

Davis-Besse 3.4.17-1 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

SG Tube lntegritY 3.4.17 sR 3.4.17.1 sR 3.4.17.2 SURVEILLANCI Veriff SG tube integrity in accordance with the Steam Generator Program.

Veriff that each inspected SG tube that satisfies the tube rpaiFplggging-criteria is plugged er repaired in accordance with the Steam Generator Program.

FREQUENCY In accordance with the Steam Generator Program Prior to entering MODE 4 following a $G tube inspection Davis-Besse 3.4.17-2

Stearn Generator Level 3,7.18 3.7 PLANT SYSTEMS 3.7.18 Steam Generator Level LCO 3.7.18 Water Level of each steam generator shall be:

a.

Less than or equal to the maximum water level shown in Figure 3.7.18-1 when in MODE 1 ar 2i

b.

s 96% Operate Range with LCO 3.3.11, "Steam and Feedwater Rupture Control System (SFRCS)

Instrumentation,"

Function 1 (Main Steam Llne Pressure Low) not bypassed when in MODE 3;

c.

396.]!0/o Operate Range with LCO 3.3.11, Function 1 bypassed and both main feedwater (MFW) pumps not capable of supplying feedwater to the steam generators when in MODE 3; and

d.

s 50 inches Startup Range with LCO 3.3.11, Function 1

bypassed and one er$ettrMFW pumpe capable of supplying feedwater to the steam generators when in MODE 3.

APPLICABILITY:

ACTIONS MODES 1, 2, and 3.

Enter applicable Conditlons and Required Actions of LCO 3.1.1, "SHUTDOWN MARGIN (SDM),'I wtren high steam generator water level resutts in exceeding the SDM limits.

CONDITION COMPLETION TIME B.

Water level in one or more stearn generators not within limits.

Required Action and associated Completion Time not met, 15 mlnutes REQUIRED ACTION A.1 Restore steam generator level to within limit.

8.1 Be in MODE 3.

ANq 8.2 Be in MODE 4.

Davis-Besse 3.7.18-1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

Stearn Generator Level FREQUENCY sR 3.7.18.1 SURVEI LLANCE REQUIREMENTS SURVEILLANCE Verify steam generator water level to be within limits.

Provided for context only. There are no changes proposed for this page.

Davis-Besse 3.7.18-2

Steam Generator Level 3.7.18 Unaccepteble Operrtlng Brglon trble Ope r\\lng Reglon paoe.

Figure 3.7.18-1 (page 1 of 1)

Maxlmum Allowable Steam Generator Level Davis-Besse 3.7.18-3

bs

-Lo oJ oct)

(Etr, o*,g a,eo o

q 100.0 95.0 90.0 85.0 80.0 75.0 70.0 65.0 60.0 55.0 50.0 45.0 40,0 35.0 30.0 25.0 20.0 15.0 10,0 5.0 0.0 0.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0 55.0 60.0 65.0 Maln Steam SuPerheat (deg. F) nsert 1, Figure 3.7,18-1 UNACCEPTABLE ACCEPTABLE (0,31.5)

Programs and Manuals 5.5 Prograrns and Manuals 5.5.8 a,

$teqm G eneratorJS,G)

Prqg.rarn A Steam Generator Program shall be established and irnplemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall incl ude the followi ng4revisiene:

Provisions for condition monitoring asses$ments.

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 SG Inspection

outage, as determined from the inservice inspection results or by other means, prior to the plugging

++epiFof tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspectedl pl pluggedr-er+epaireel to confirrn that the performance criteria are being met.

Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity,

. accident induced

leakage, and operational LEAKAGE.

Structural integrity performance criterion: AII in-service stearn generator tubel dfrblt retain structural integri$ over the full range of normal operating conditions (including staftup, operation ln the power range, hot standby, and cool downl-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 ihe 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 licenslng basis, shall atso be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the a$se$$ment of tube integrity' those loads that do significantly atfect burst or collapse shall be determined and assessed in combination with the loads due to pressure wlth a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assurned in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed l gprn psl $4 1.

2.

Davis-Besse

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (9G),Prggram (continued)

3.

The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

Provisions for SG tube repaieplUggingcriteriai.

Tu.bq$

f.ound by inse inspection to cbntain flaws with a denth go.u.al, to or exceedino 40Yo of the norninal tube wall thickness shall be qlggggd.

1, Tubee feund by ineeruie+inspeetien t eentain flawsrin a region of the tube that eenteine ne-reBairt with s depthqual te e'exeeeding the eleeve that ent 3

Tubee with a fla$hin-either Barent tuhe er the sleeve; within-a'sleeve 4,.' Tubee with a flaw ln a repair relfhalf b+pfugged'

d.

Provisions for SG tube inspections.

Periodic SG tube inspections shall be perforrned. 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 ffaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlot to the tube-to-tubesheet weld at the tube outlet, and that may s atisfy the a pplicab I e tube repaieBlgggingcriteria. Th e tube-to-tubesheet weld is not part of the tube.

in$

the tub+and-tube rell, eutbeard ef the new rellarea in'thetube sheet an inspeetion beeause it ie ne lengeFparfef the pressure beundaty eHe+tho eteirv+l+instette+ln additlon to meeting the requirements of d.1' d.? and dj3.tnreugh45 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. *+A degladatiqn assessment ef Sgmdatio+shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to

' deterrnine which inspection methods need to be employed and at what locations.

1.

tnspect 100% of the tubes in each SG during the first refueling outagefollowingSG@.

Davis-Besse

Programs and Manuals 5.5 5.5 Prograrns and Manuals 5.5.8 Steam Generator (SG) Program (continued) pewer menths, Th lng

2.

Afreithe firgtlefgeilng ouiage followina $G installation.

inspect e4c.h q-9 at least every 72 effective full power month$ or at leest ?very third refueling outaqe {whichever results in mo[gJrequent iffidition.

the rUinimum number ottuhes insnpqted at each scneduied Insnection snallbg thg number of tubes in ell $Gq

@er pjgg,inspection outages scneduled in eqgh inspectiorulqrig.d.Fs defined in a. b. c and d below. lf a d.egr3datiQn nsqessrnent lnOicates the notential for a tvpe of deqradqtign tgoqcur at alocation not breviously inspected with a technique.capable--qf.

Cetecting tfris tyoe of degr-adation at ttris locelip-n,and that may satisfy the applicable iube plugaino.g-titpria.

the minimum number of locations inspeCted witn sug.

h q. capable insnection technique 9urilg t-re rem.ai,rlder d

tion period mav be prorated. The-Jrflct'pn oi Iocations to be inspected for this potentiel,-typ"e of dporadation-at tfris location at lhe end of-thp inspection period shall be no less thelt tfre ratio of tfre nr,mher of times the $G is scheduled to be lnspecled i n @

aner tn e Uetermin qtion-t nat.a rlgUv-f-o n!..oI.

Oe iv ne ogqr..rf.ripq,at thiq location divided,Ptt the total nunlbgl.o.f-t the SG is scheduled to be inspected,in thq Each inspection period defined bejow $Ay be exlerrded up t,o 3 effective full power months to include a $G inspection outage in an insnection period And the su,bg,eqH9nl irrspection period begins at the conctusion of the include-L SQ inspection outage.

g) Aflet lhe first refueling outage following SG installation.

lnsDec!

100%_of

,th-e_lubes durinq the next 144 etfective full Power monlhs. This constitutes the flrst Insnectlon period:

b) During the next 120.effeQ!i.V.,g full power months. inspect 1Q0% of tho tubes. Thls constitutes thg qpcond insp-ection period:

c) During the next 96 effective full oower m,onthg.

insqect.J-0,.0,.%

of tfre iubes. fnis c d) purinq the remalnino life of the SGs. insoect 10070 of the tube-$

every 72 effectivejull power mgnlhq,-.This g,p.nstitutes the fourth and su bseouent inspectionl?gllqd$..

3.

lf crack indications are found in any SG tube, then the next inspection for each affected-and potentially affected SG for the Davls-Besse

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Stearn GeneJertor (SG) Prooram (continued) degradatlon mechanism that caused the crack indication shall not exceed 24 etfectlve full power months or one refuellng outage (whlchever

+s-lessresults in more frequ igES)' lf definitive information, such as from examination of a pulled tube, diagnostic non-deslructive testing, or englneering evaluation indicates that a Grack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

4-During eael,r Beriedie SG tube inspeetienr inspeet'l00Yo ef the tHbes thaffreve been repeired by the-repair-rell proes, This speeial e++ne+etl-reBain 6-InsBeet BeriBheral tubes in the vioinig ef the eeeured internal t,

the,NEth tube uppert ptate during eeh pesiediesG tube i

the entire eireurnferenee eFthe eteam generater and shall tetal

e.

Provisions for rnonitoring operational prirnary to secondary LEAIGGE' ir the purBbses ef these Speifieatienst.'tube plugging ie net a repalpAll 1,

Sleeving-in.aeeerdanee with TeBieal Repert BAW,*I20P' nei gaps"tass than i/4 ineh' or hai wi eleves shall b+perfermed en the affeeted S thresgh{heauxlliary iene' 5.5.9 Secondary Water -Ch.emistrv Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation.

The program shall include:

Davis-Besse

Programs and Manuals 5,5 ldentification of a sampling schedule for the critical variables and control points for these variables; ldentification of the procedures used to measure the values of the critical

. variables; ldentification of process sampling points; Procedures for the recording and management of data; Procedures defining corrective actions for all off control point chemistry conditions; and A procedure ldentifying the authority responsible for the interpretation of the data and the sequence and timing of admlnistrative events, which is required to Initiate corrective action.

5.5.10 V"-e,.ntilation Filter Testing Program (VFTP)

A program shalt be established to implement the following required testing of safety related filter ventilation systems in accordance with Regulatory Guide 1.52, Revision 2, ANSI/ASME N510-1980, and ASTM D 3803-1989.

a., Demonstrate for each of the safety related systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 1.lo/owhen tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI/ASME N510-19S0 at the system flowrate specified below.

Safety Related Ventilation System Station Emergency Ventilation System (EVS)

Control Roorn Emergency Ventllation System (cREVS) a.

b.

c.

d.

e.

Flowrate (cfrn)

> 7200 and 5 8800 e 2970 and 5 3030 Davis-Besse

Reportin g Requirements 5.6 5.6 Reporting Requirements 5.6.6

$le.Ar:n Generator Tube Inspection Report A report shatl be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.8, "Steam Generator (SG) Program." The report shall include:

a.

The scope of inspections performed on each SG;

b.

ne*v+gDegradationmechanisrnsfound;

c.

Nondestructive examination techniques ulilized for each degradation mechanism;

d.

Location, orientation (if linear), and measured sizes {if avaitable) of service

, induced indications;

e.

Number of tubes plugged er+epair+during the inspection outage for each aetfu+deg radation mechanism

t.

+otal-Ihe-nurnber and percentage of tubes plugged er+epa+re+to date; and the effective plugging percentage in ea,ch SG:

g.

The results of condition monltorlng, including the results of tube pulls and

. in-situ testing; SG-anC i,

ReBair methed utili=ed end th numbepef+ubee repaired by eaeh reBair 5.6.7 Remote Shutdown System Renort

\\A/hen a report is required by Condition C of LCO 3.3.18, "Remote Shutdown System," a report shall be submitted within the following 30 days. The report shall outline the action taken, the cause of the inoperability, and the plans and schedule for restoring the control circuit or transfer switch of the Function to OPERABLE status.

Davis-Besse 5.&1

Proposed Revision of Technlcal Specification (TS) 3.4.17, "Steam Generator (SG)

Tube lntegrity"; T$ 3.7.18, "$team Generator Level"; TS 5.5.8, "Steam Generator (SG) Program"; and TS 5.6.6, "Steam Generator Tube Inspectlon Report" for the Davis-Besse Nuclear Power Station, Unit Number 1 (DBNPS)

Operating License Number NPF-3 Proposed Ghanges to Technlcal Specifications Retyped Gopy The following is a list of the affected Technical Specification page$.

3.4,17-1 3.4.17-2 3,7.18-1 3.7.18-2*

3.7.18-3 5.5-5 5.5-6 5.5-7 5.5-8*

5.6-1

No Ghange. Page provided for context only.

SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 APPLICABILITY:

ACTIONS SG tube integrity shall be maintained.

AN.D-All SG tubes satisfying the tube plugging criteria shall be plugged in accordance with the Steam Generator Program.

MODES 1,2,3, and 4.

NOTE----

Separate Condition entry is allowed for each SG tube.

CONDITION COMPLETION TIME One or more SG tubes satisfying the tube plugging criteria and not plugged in accordance with the Steam Generator Program.

Prior to entering MODE 4 following the next refueling outage or SG tube A.

Required Action and associated Completion Time of Condition A not met.

OR SG tube integrity not maintained.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours REQUIRED ACTION A.1 Veriff tube integritY of the affeeted tube(s) is maintained until the next refueling outage or SG tube inspection.

AND A.2 Plug the affected tube(s) in accordance with the $team Generator Program.

Be in MODE 3.

Be in MODE 5.

Davis-Besse 3.4.17-1

SG Tube Integri$

3.4.17 SURVEILLANCE REQUI REMENTS sR 3.4.17.1 sR 3,4.17.2 SURVEILIANCE Verify SG tube lntegrig in accordance with the Steam Generator Program.

Veriff that each inspected SG tube that satisfies the tube plugging criteria is pluggod in accordance with the Steam Generator Prograrn.

FREQUENCY ln accordance with the Steam Generator Program Prior to entering MODE 4 following a SG tubo inspection Davis-Besse 3.4.17-2

Steam Generator Level 3.7.18 3.7 PLANT SYSTEMS 9.7.18 Steam Generator Level LCO 3.7.18 Water Level of each steam generator shall be:

a.

Less than or equal to the maximum water level shown in Figure 3.7.18-1 when in MODE 1 or 2;

b.

< 90% Operate Range with LCO 3.3.11,

"$team and Feedwater Rupture Control System (SFRCS)

Instrumenlation,"

Function 1 (Main Steam Line Pressure Low) not bypassed when in MODE 3;

c.

s74% Operate Range with LGO 3.3.11, Function 1

bypassed and both main feedwater (MFW) purnps not capable of supplying feedwater to the steam generators when in MODE 3; and

d.

s 50 inches Startup Range with LCO 3.3.11, Function 1

bypassed and one MFW pump capable of supplying feedwater to the steam generators when in MODE 3.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS NOTE---'

Enter applicable Conditions and Required Actions of LGO 3,1,1, "SHUTDOWN MARGTN (SDM)," when high steam generator water level results in exceeding the SDM limits.

CONDITION COMPLETION TIME Water level in one or more steam generators not within limits.

15 minutes Required Action and associated Completion Time not met.'

A.

B.

REQUIRED ACTION A.1 Restore steam generator level to within limit.

8.1 AND 8.2 Be in MODE 3.

Be in MODE 4.

Davis-Besse 3.7.18-1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

SURVEI LLANCE REQUI REM ENTS sR 3.7.18.1 Stearn Generator Level 3.7.18 FREQUENCY SURVEILLANCE Verify steam generator water level to be within limits.

Provided for context only. There are no changes proposed for this page.

Davis-Besse 3.7.18-2

Steam Generator Level

].7,18 100.0 95.0 90.0 85.q 80.0 75.0 70.0 65.0 60.0 55.0 50.0 45.0 40.0 35.0 30.0 25.0 20.0 15.0 10.0 s.0 0.0 (60.3,96)

UNACCEPTABLE

e

..t 0) d)

J ocrl tr G,u oe d)ctooa ACGEPTABLE (0, 31.5) 0.0 5,0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0 55.0 60.0 65.0 Ta.a Maln Steam

$uPerheat (deg. F)

Figure 3,7.1S-1 (page 1 of 1)

Maximum Allowable Stearn Generator Level Davis-Besse 3.7.18-3

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 a.

Stea.m Generator (SG) Prosram A Steam Generator Program shall be established and irnplemented to ensure that $G tube integrity is maintained. In addition, the Steam Generator Program shall include the following:

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 SG inspection outage, as detennined from the inservice inspection results or by other means, prior to the plugging of tubes.

Conditlon monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the perforrnance criteria are being met.

' Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAI(AGE.

1.

Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (includlng startup, operation in the power range, hot standby, and cool down), 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 prlmary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primaryto-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the deslgn basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to deterrnine 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 cornbination with the loads due to pressure with a safety factor of 1.2 on the combined prlmary loads and 1,0 on axial secondary loads.

2.

Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm per SG.

b.

Davis-Besse 5,5-5

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator.(9G)

Program (continued)

' 3.

The operatlonal LEAMGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

c.

Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 4A% of the nominal tube wall thickness shall be plugged.

d.

Provisions for SG tube inspections.

Periodic SG tube inspections shall be pefformed. The number and portions of the tubes Inspected and methods of inspection shall be performed with the objective of detecting flaws o{ any type ie.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-totubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not pirt of ttii'tube. tn aOciition to meeting the requirements of d.1, d.2 and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed

' to determine the type and location of flaws to which the tubes may be susceptible and, based on this a$sessment, to determine which inspection methods need to be employed and at what locations.

1.

Inspect 100% of the tubes in each $G during the first refueling outage following SG installation.

Davis-Besse

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Stearn Generator (SG) Program (continued) 2, After the first refueling outage following SG installation, inspect each SG at least every 72 etfective full powel months or at least every third refueling outage (whichever results in rnore frequent inspections).

In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the nurnber of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. lf a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. Tha fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the deterrnination that a new form of degradation could potentiatly be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power rnonths to include a SG inspection iutage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

a) After the first refueling outage following $G installation, inspect 100o/o of the tubes during the next 144 effective full power months. This constitutes the first inspection period; b) During the next 120 ettective full power months, inspect 100% of the tubes. This constltutes the second inspection period; c) During the next 96 effective full power months, inspect 1009o of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subseguent inspection periods.

3.

If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the Davis-Besse 5.5-7

Programs and Manuals 5.5 5.5 Programs and Manuals 5,5.8 5.5.9 5.5.10 Steam Generator (SG) Program (continued) degradation mechanism that caused the crack indication shall not exceed 24 etfective full power months or one refueling outage (whichever results in more frequent inspections).

lf definitive information, such as frorn examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication ls not associated with a crack(s),

then the indicatlon need not be treated as a crack.

e.

Provisions for monitoring operational primary to secondary LEAKAGE.

Secgnd a nl$/.gter Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation.

The program shall include:

a.

ldentification of a sampling schedule for the criticalvariables and control points for these variables;

b.

tdentification of the procedures used to measure the values of the critical

, variables;

c.

ldentlficatlon of process sampling points;

d.

Procedures for the recording and management of data;

e.

Procedures defining corrective actions for all off control polnt chemistry conditions; and

f.

A procedure identiffing the authority responsible for the interpretation of tho data and the sequence and timlng of administrative events, which is required to initiate correctlve action, Ventilation Filter Testino Proqr-qm ffFTP)

A program shall be established to lmplement the following required testing of safety related filter ventilation systems in accordance with Regulatory Guide 1.52, Revlsion 2, ANSI/ASME N510-1980, and ASTM D 3803-1989.

a, ' Demonstrate for each of the safety related systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 1,Oo/owhen tested in accordance with Regulatory duide 1.52, Revlsion 2, and ANSI/A$ME NSt0-1980 at the system flowrate specified below.

Safety Related Ventilation Svslqm.

Station Emergency Ventilatlon System {EVS)

. Control Room Emetgency Ventilation System (cREVS)

Flowrate,(qfm).

> 7200 and s 8800

> 2970 and s 3630 Davis-Besse 5.5-8

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 Steqm Genera-tor Tube lnspection Report 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 Specificatlon 5.5.8, "Steam Generator (SG) Program." The report shall include:

The scope of inspections performed on each SG; Degradation mechanisms found; Nondestructive examination techn iques utilized for each degradation mechanism;

Location, orientation (if linear), and measured sizes (if available) of service induced indications; Number of tubes plugged during the inspection outage for each degradation mechanism; The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG; The results of condition monitoring, including the results of tube pulls and In-situ testlng; 5.6.7 BCmojp.$.hU!*our,s-$ystem Bepo$

When a report is required by Condition C of LCO 3.3.18, "Remote Shutdown System,"

a report shall be submitted within the following 30 days. The report shall outline the action taken, the cause of the inoperability, and the plans and schedute for restoring the control circuit or transfer switch of the Function to OPERABLE StAtUS.

a.

b.

d, e.

f.

Davis-Besse 5.6-1

Proposed Revision of Technlcal Specification (TSl 3.4,17, "Steam Generator (SG)

Tube Integrity"; TS 3.7.18, "Steam Generator Level"; TS 5.5.8, "$team Generator (SG) Program"; and TS 5,6.6, "Steam Generator Tube Inspectlon Reporf' for the Davis-Besse Nuclear Power Statlon, Unit Number { (DBNPS}

Operatlng License Number NPF-3.

Proposed Ghanges to Technlcal Specification Bases Annotated Copy (11 Pages Follow)

Technical Specification Bases are not part of the Technical Specifications, are not submitted for Nuclear Regulatory Commission Approval and are provided for information

only,

ProvjC.gd for Information Only B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.17 Steam Generator (SG) Tube Integrity BASES No changes to this page.

Provided for context onlY.

SG Tube Integrity g 3.4.17 BACKGROUND Steam generator (SG) tubes are smalldiarneter, thin walled tubes that carry prirnary coolant through the primary to secondary heat exchangers.

The SG tubes have a number of important safety functions. $team generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to rnaintaln the primary system's pressure and inventory. The SG tubes isolate the radioactive fisslon products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systerns to remove heat from the primary system. Thls Specification addresses only the RGPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.4.4, "RCS Loops - MODES 1 and 2," LCO 3.4.5' "RCS Loops - MODE 3," LCO 3.4.6, "RCS Loops - MODE 4," and LCO 3.4.7,.'RCS Loops - MODE 5, Loops Filled."

SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing

basis, including applicable regulatory requirements.

$team 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 phenornena such as denting and wear. These degradation mechaniims can irnpair tube integrity if they are not managed effectively.

The SG performance criteria are used to manage SG tube degradation.

Specification 5.5.8, "Steam Generator (SG) Program,"

requires that a program be established and lmplemented to ensure that SG tube integrity is rnaintained.

Pursuant to Specification 5.5.8, tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operationat LEAKAGE, The SG performance criteria are described in Specification 5.5.8. Meeting the SG performance criteria provides reasonable assurance of maintaining tube Integrity at normal and accident conditions.

The processes used to meet the SG performance criteria are defined by the Steam Generator Program Guidelines (Ref. 1).

Davis-Besse

Provided for Inforrnatlon Only SG Tube IntegritY B 3.4.17 BASES APPLICABLE SAFETY ANALYSES The steam generator tube rupture (SGTR) accident is the llmiting design basis event for SG tubes and avoiding an SGTR is the basis for this Specification.

The analysis of a SGTR event assumes a bounding primary to secondary LEAI(AGE rate equal to the operational LEATGGE rate limits in LCO 3,4.13, URCS Operational LEAKAGE,"

plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysls for a SGTR assumes the contaminated secondary fluid is released to the atmosphere vla maln steam safety valves.

The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their 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 secondary LEAIGGE from all

$Gs oi 1 gallon per minute. DOSE EQUIVALENT l-131 is assumed to be equivalent to 1% failed fuel in the accident analysis.

The dose consequences of these events are within the limits of GDC 19 (Ref. 2)'

10 CFR 100 (Ref. 3) or the NRC approved licensing basis (e.9., a srnall fraction of these limits).

Steam generator tube integrity satisfies Criterion 2 of t0 cFR 50.36(cX2xii).

LCO The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfrT the reBaiFplgggigcriteria be plugged r+pair+in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam G e ne iato r Prog ram seBaieplUggj3gcriteria is reBaired+eremoved from service by plugging. lf a tube was determined to satisfy the repair plUgginq priteria but was not plugged+r+pai+x*, the tube may still have tube integrity.

In the context of this Specification, ag SG tube is defined as the entire length of the tub between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. includins the tub

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

A SG tube has tube integrity when it satisfles the SG performance criteria.

The SG performance criteria are defined in $pecification 5.5.8, and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criterla.

There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAIGGE. Fallure to meet any one of these criteria is considered failure to meet the LCO.

Davis-Besse

Provided for Information Onlv No changes to this page.

Provided for context only.

SG Tube lntegrity I B 3.4.17 LCO (continued)

The structural Integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structurat integrity of the SG 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.9., opening area increased in response to constant pressure) accompanied by ductite (plastic) tearing of the tube material at the ends of the degradation."

Tube collapse ls deflned 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 curye becomes zeroJ The structural integrity performance criterlon provides guidance on assessing loads that have a significant etfect on burst or collapse. In that context, the term "significant" is defined as'An accident loading condition other than diffbrential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting bursUcollapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis.

The division between prirnary and secondary classifications will be based on detailed analysis and/or testing.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code, Section lll,

$ervice Level A (normal operating conditions) and Service Level B (upset or abnormat conditions) transients included in the design specification.

This includes safety factors and applicable design basis loads based on ASME Code, Section lll, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121(Ref.

5).

The accident induced leakage performance criterion ensurs that the primary to secondary LEAKAGE caused by a design basls accldent, other than a'SGTR, is wlthln the accident analysis assumptions.

The accident analysis assumes that accldent induced leakage does not exceed 1 gpm per SG. The accident induced leakage rate includes any primary to secondary LEAI$GE existing prior to the accident In addition to primary to secondary LEAI(AGE induced during the accident, The operational LEAMGE performance criterion provides an obsorvable indication of SG tube conditions during plant operation. The limit on operational LEATGGE is contained in LCO 3.4.13, "RCS Operational LEAlfiGE," and lirnits prirnary to secondary LEAI(AGE through any one Davis-Besse

Provided for lnformation Onlv SG Tube Integrity B 3.4.17 BASES LCO (continued)

$G to 150 gallons per day. This lirnit is based on the assurnption that a single crack leaking this amount would not propagate to a SGTR under the stress conditlons of a LOCA or a main steam line break. lf this amount of LEAMGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1,2,3, or 4, RCS conditlons are far less challenging in MODES 5 and 6 than during MODES 1,2,9, and 4. In MODES 5 and 6, prirnary to secondary differential pressure is low, resulting in lower stregses and reduced potential for LEAIGGE.

ACTIONS The ACTIONS are modified by a Note clariffing that the Conditions may be entered independently for each SG tube. This ls acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube, Complying with the Required Actions may allow for continued operatlon, and subsequent affected SG tubes are governed by subsequent Condition entry and application of associated Required Actions.

A.1 andA.2 Condition A applies if it is discovered that one or more SG tubes examined In an inservice inspection satisfy the tube rBaiFplwg!ru criteria but were not plugged er+epai+e*in accordance with the Stearn Generator Program as required by SR 3.4.17.2.

An eveluatien of gg Steam generator tube integrity is based on meeting the $G performance criteria described in the Steam Generator Program. The SG fBeiFpigggiry-criteria define SG tube degradation.

The mea,sured value rnust be adjusted for measurement uncertaintlr and-thaLallo$t4or predicted flaw growth between inspections ptovide assurance that the SG performance criteria will continue to be rnet. In order to determine lf a SG tube that should have been plugged er reBaire+has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met untit the next refueling outage or SG tuba inspection.

The tube integrity determination is based on the measured flaw gize ffi th+tsbe-at the time the situation is discovered qnd adjgstments fg.[

meagqrenent qnpgf-tFintypnd the estimated growth of the degradation prior to ahe next SG tuna inspection.

lf it is determined that tube Integrlty is not being maintained, Gondition B applies.

Davis-Besse

P.[o..vj.ded for lnformation Only SG Tube IntegritY B 3,4.17 BASES ACTIONS A.1 and A.2 (continued)

A Completlon Tlme of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a $G tube that may not have tube integrity.

lf the evaluation determines that the affected tube(s) have tube integrity, Required Action A.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged-+r rpair+prior to entering MODE 4 following the next refueling outage or -

SG inspection.

This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.

8.1 an{-F.?

lf the Requtred Actions and associated Completion Times of Condition A

are not met or if SG tube integrity is not being maintained, the reactor must be brought to 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 .

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

SURVEILLANCE REQUIREMENTS sR 3.4.17.1 During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-0G, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes-is performed.

The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpo$e of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previou$

operating period.

The Steam Generator Program deterrnines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfuing the tube repair-BlUgghtrcriteria, lnspection scope (i.e., which tubegor areas of tubing witnin tne SG are to be inspected) is a function of existing and potential degradation locations.

The Steam Generator Prograrn also Davis-Besse

Provi4ed,.for lnformation Only SG Tube Integrity B 3.4,17 BASES SURVEILLANCE REQUIREMENTS SR 3.4.17.1 (continued) specifies the inspection rnethods to be used to find potential degradattbn.

Inspection methods are a function of degradation morphology, non-destructive examinatlon (N DE) technique capabilities, and inspection locations.

The Stearn Generator Program defines the Frequency of SR 3.4,17.1.

The Frequency is determlned by the operational assessment and other Iimits in the SG examination guidelines (Ref. 6). The $team Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 5.5.8 contains prescriptive requirements concernlng inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

lf crack indicatiglg are found in any $G tubeotFq maximum insnection interval fof all, affected and potentiallv affgcted SQs is restriCied bv Specification 5.5.8 until subseauent inspections sunPort extending the inspection interval.

qR..3.4.17.?

During an SG inspectlon, any inspected tube that satisfies the $team Ge neiator Prog ram reBaie.BlUgging-crite ria i s repaireekreremoved fro rn service by plugging. The tube pluooing repai+criteria delineated in Specificatlon 5.5.8 are intended to ensure that tubes accepted for continued service satisff the SG performance criteria with allowance for error In the flaw size measurement and for future flaw growth. In addition, the tube repai+BlUggingcriteria, in conjunction with other elernents of the Stearn GeneraloiPiogram, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s).

Reference 1 provides guidance for performing operational assessments to veriff that the tubes remaining in servlce will continue to meet the SG performance criteria.

The Frequency of prior to entering MODE 4 following a SG inspection ensures ihat ttre Surveillance has been completed and all tubes meeting the repaiFpluggi$Lcriteria are plugged er+epaire+prior to subiecting the SG tubes to significant prirnary to secondary pre$sure differential.

Davis-Besse

Provided for Information Only BASES No changes to this page.

Provided for context only.

SG Tube Integrity B 3.4.17 REFERENCES 1.

2.

3.

4.

5.

6.

NEI 97-06, "$team Generator Program Guidelines."

10 CFR 50 Appendix A, GDC 19.

10 cFR 100.

ASME Boiler and Pressure Vessel Code, Section lll, $ubsection NB.

Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes,"

August 1976.

EPRI, "Pressurized Water Reactor Steam Generator Exarnination Guidelines."

Davis-Besse

Provided for Information Only Steam Generator Level B 3.7.18 B 3.7 PLANT SYSTEMS B 3.7.18 Steam Generator Level BASE$

BACKGROUND A principal function of the steam generators is to provide superheated steam at a constant pressure (935 psia) over the power range. Steam generator water inventory is maintained large enough to provide adequate primary to secondary heat transfer. Mass inventory and indicated water level in the stearn generator increases with load as the length of the four heat transfer regions within the steam generator vary.

Inventory is controlled indirectly as a function of power and maintenance of a constant average primary system temperature by the feedwater controls in the Integrated Control System.

The maximum operating steam generator levet is based primarily on preservlng the initial condition assumptions for the steam generator inventory used in the main steam line break (MSLB) accident analysis (Ref. 1). The mass and energy release data thatare input into the pealc pres$ure analysis of the containment vessel were generated with the ifEnpS/MOD2-B&W computer code. The analysis was performed with the bounding plant conditions to maximize heat generated in the Reactor Coolant System (RCS), heat transfer from the primary to secondary

systems, ind maxirnum inventory in the steam generators.

Each of these cbnditions maxirnizes the mass and energy release from the MSLB. The analysis includes evaluation of the reactivi$

transient due to the MSLB.

gen{t+1er+pel'ating at 100% Bo r

55s0o1b, As a steam generator becomes fouled and the operating level approaches ine limit of 96%, the mass inventory in the downcomer region increases. In matching unit data of startup level versus power, the steam generator performance codes have shown that fouling of the bwer tube support plates does not significantly change the heat transfer characteristics of the steam generator.

Thus, the steam temperature, or superheat, is not degraded due to the fouling of the tube support

plates, and mass inventory changes are malnly due to the added level in the downcomer.

However, increasing the fouling of the steam generator tube surfaces gg ingrgasing levels of tube plugqinq degrades the heat transfer capabllity of ne steam generator, increases the mass inventory, and decreaSes the steam superheat at 1000/o RTP. The resulF were presented as the ameunt el mees inventery in eeeh eteam generater vereue eBrating Davis-Besse

Provided for Information Only Steam Generator Level B 3.7.18 BASE$

BACKGROU ND (continued)

The lirniting curve, which was determined from several steam generator performance code runsr conservatively bounds lhg_steam generator mass inventory value, when operating at power levels < 1000/6.

The curve presents the limit on each steam Senerator'$

mass inventory (in Onerate Range level ifrdication) as a function of ste.SFl superheat.

The points displayed in Figure 3.7,18-1 represent a 56#4e56J00Ib mass value, at the @

Operate Range levelx and steam superheat values based on a conservatively calculated mass inventory..The v.elue used for mass invgEtpJyjs conslstent.with the 56.000 lbm value assumed in calculations s u p po rt i nq.lhg_[4

$"tB a n-a I yqlq, The etea m generater perfermanee analyete alse+ndieated theLstaFtuFand ful I range level i nstflJmenb are inadeguabinC ieator ef-etea m ge neratr lf the mixture water level should rise above the 9G7o upper limit, the steam superheat would tend to decrease due to reduced feedwater heating through the aspirator ports. Normally, a

reduction in water tevel is manually initiated to maintain steam flow through the aspirator port by reducing the power level. Thus, the superheat versus level llmitation also tends to ensure that, in normal operation, water level will remain clear of the aspirator ports.

Feedwater nozzle flooding would impair feedwater heatlng, and could result in excessive tube to shell temperature dlfferentials, excessive tubesheet temperature differentials, and large variations in pressurizer level.

APPLICABLE SAFETY ANALYSES The most limiting Design Basis Accident that would be affected by stearn generator operaiing level is a main steam line failure, This accident is ivaluated in Reference

1. The pararneter of interest is the rnas$ of water, or inventory, contained in the steam generator due to its role in towering Reactor Coolant System (RCS) temperature (return to criticality concern),

and in ralsing containment pressure during an MSLB accident.

A higher inventory causes the effects of the accident to be rnore severe.

Figure 3.7.18-1 is based upon maintaining inventory

< 5#4,S50J0@lbnU Uthigh.j-E qgnsistent VUi![

lhg valgg assurned (56.000.1bm) in calculations diipporting ihe -lgure 3,7,1 MSLB analysis.

' whieh assurned epp Ire plant response when operatlng at the limit of the Figure is be+mdingt ferbpunded_by the MSLB analysis.

considering all plant effects (e.9.'

steam superheat and downcomer voiding).

Davis-Besse

Provided for Information Only No changes to this page.

Provided for context only.

Stearn Generator Level B 3.7.18 The steam generator level satisfies Criterion 2 of 10 CFR 50.36(cX2XiD.

BASES LCO This LCO is required to preserue the initial condition assumptions of the accident anatyses. Failure to meet the rnaximum steam generator level LCO requirements can result in additional mass and energy released to contalnment, and excessive cooling (and related core reactivity effects) following an MSLB. In addition, feedwater nozzle flooding would impair feedwater heating, and could result in excessive tube to shell temperature d iffere ntials and excessive tu b esheet ternpe rature grad i e nts.

APPLICABILITY ln MODES 1, 2, and 3, a maximum steam generator water fevel is required to preserve the initial condition assumption for steam generator inventory used in the main steam line failure accident analysis (Ref. 1). In MODE 3, limits on steam generator water level (in conjunction with meeting the requirements of LCO 3.1.1, 'SHUTDOWN MARGIN (SDM)")

will also prevent a return to criticality in the event of an MSLB.

In MODES 4, 5, and 6, the water in the steam generator has a low specific enthalpy; therefore, there is no need to limit the steam generator inventory when the unit is in this condition.

ACTIONS ln the event a high steam generator water level results in exceeding

!he-SDM limits of LCO 3.1,1, "SHUTDOWN MARGIN (SDM),"

the ACTIONS Note directs entry into the applicable Conditions and Required Actions of LCO 3.1.1. This is an exception to LCO 3.0,6 and ensures the proper actions are iaken for SDM not within the required limits.

AJ Wth the steam generator level in excess of the maximurn limit, action must be taken to restore the level to within the bounds assumed in the analysis. To achieve this status, the water level is restored to within the limit. The 15 minute Completion Time is considered to be a reasonable time to perform this evolution.

8.1 lf the water level in one or more steam generators cannot be restored to withln the limits, the unit must be placed in a MODE that minimizes the accident risk. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and In MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operatlng experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

Davis-Besse

Provided for Information Only No changes to this page.

Provided for context only.

Steam Generator Level B 3.7.18 BASES SURVEILLANCE SR 3.7.18.1 REQUIREMENTS This SR verifies the steam generator level to be within acceptable limits.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is adequate because the operator will be aware of unit evolutions that can affect the steam generator level between checks. Furthermore, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications available in the control roorn, including alarms, to

.alert the operator to steam generator level status.

REFERENCES

1. UFSAR, Section 15.4.4, Davis-Besse

v t e Letter Sequence Request
TAC:MF0536, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection (Approved, Closed)
Initiation Request, Request, Request, Request, Request Acceptance Supplement Administration Questions 29 August 2013 14 November 2013 Answers 27 September 2013 13 December 2013 Results Approval Other:
MONTHYEAR2013-02-13 [Table View]

L-13-040, License Amendment Request for Proposed Revision of Technical Specification (TS) 3.4.17, Steam Generator (SG) Tube Integrity; TS 3.7.18, Steam Generator Level; TS 5.5.8, Steam Generator Program; and TS 5.6.6

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L-13-040, License Amendment Request for Proposed Revision of Technical Specification (TS) 3.4.17, Steam Generator (SG) Tube Integrity; TS 3.7.18, Steam Generator Level; TS 5.5.8, Steam Generator Program; and TS 5.6.6

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