ML18039A878

TS Change 399 to Licenses DPR-52 & DPR-68,increasing Allowable Leakage Rate Criteria for MSIVs & Requesting Exemption to Portions of 10CFR50,App J.Rev 0 to Rept 2100918-R-002 Encl

ML18039A878
Person / Time
Site:	Browns Ferry
Issue date:	09/28/1999
From:	Abney T TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18039A879	List: ML18039A878 ML18039A881 ML18039A882
References
TVA-BFN-TS-399, NUDOCS 9910120239
Download: ML18039A878 (39)

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TEQORY REGULAT RY INFORMATION DISTRIBUTIO ~SYSTEM (RIDS)

ACCESSION NBR:9910120239 DOC.DATE: 99/09/28 FACIL:50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee NOTARIZED: YES DOCKET ¹ 05000260 50-,296 Browns Ferry Nuclear Power Station, Unit 3, Tennessee 05000296 AUTH. NAME AUTHOR AFFILIATION Tennessee Valley Authority

'BNEY,T.E.

RECIP.NAME RECIPIENT AFFILIATION Records Management Branch (Document Control Desk)gippxm+s

SUBJECT:

TS Change 399 to licenses DPR-52 Ec DPR-68,increasing allowable leakage rate criteria for MSIVs.Rev 0 to 2100918-R-002 rept encl.

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Tennessee Valley Authority, Post Office Box 2000, Decatur, Afabama 35609-2000 TVA-BFN-TS-399 September 28, 1999 10 CFR 50.4 10 CFR 50.12 10 CFR 50.90 10 CFR 50.91 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:

In the Matter of Docket Nos. 50-260 Tennessee Valley Authority 50-296 A

BROWNS FERRY NUCLEAR PLANT (BFN) UNITS 2 AND 3 TECHNICAL SPECIFICATIONS (TS) CHANGE 399 INCREASED, MAIN STEAM ISOLATION VALVE (MSIV) LEAKAGE RATE LIMITS AND EXEMPTION FROM 10 CFR 50 APPENDIX J In accordance with the provisions of 10 CFR 50.4 and 50.90, TVA is submitting a request for a TS amendment (TS-399) to licenses DPR-52 and DPR-68 to increase the allowable leakage rate criteria for the MSIVs. In addition, in accordance with 10 CFR 50.12, TVA is requesting exemption to specific portions of 10 CFR 50, Appendix J to allow the exclusion of MSIV leakage from the summation of containment leak rate test results.

This change request is based on the utilization of the Boiling Water Reactor Owners'roup (BWROG) methodology described in NEDC-31858P, Revision 2, BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems. NRC has previously determined that the subject BWROG report is acceptable for direct reference

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L U.S. Nuclear Regulatory Commission Page 2 September 28, 1999 for individual licensee applications as documented in the associated NRC Safety Evaluation Report dated March 3, 1999.

BFN plant design does not include a MSIV leakage control system as do later vintage BWRs, therefore, there are no changes required in this area.

Specifically, TVA proposes to utilize the main steam drain lines to preferentially direct MSIV leakage to the main condenser. This drain path takes advantage of the large volume of the steam lines and condenser to provide holdup and plate-out of fission products that may leak through the closed MSIVs. In this approach, the main steam lines, steam drain piping, and the main condenser are used to mitigate the consequences of an accident to limit potential off-site exposures below those specified in 10 CFR 100 and in 10 CFR 50, Appendix A, General Design Criteria (GDC) 19 for control room dose limits.

The primary objective of this proposed TS change is to reduce the frequency of MSIV rebuilds during outages required to achieve the leakage rates specified in current TS (11.5 standard cubic feet per hour). This would extend the service life of the MSIVs as well as reduce the radiation exposure to personnel involved in MSIV maintenance activities. to this letter provides the description and justification for the proposed TS change, and the significant hazards and environmental impact considerations. Enclosure 2 contains mark-up copies of the appropriate pages from the current Unit 2 and 3 TS showing the proposed revisions.

The components and piping systems used in this treatment path have been evaluated to be capable of performing their function following a Design Basis Earthquake in accordance the technical methodology in the BWROG report, NEDC-31858P.

Refer to Enclosure 1,Section III.C and Enclosure 4 for details concerning the seismic evaluation of the piping systems and other components.

U. S. Nuclear Regulatory Commission Page 3 September 28, 1999 A plant-specific radiological analysis has been performed to assess the effects of the proposed increase in MSIV leakage criteria in terms of off-site doses and main control room dose. This analysis uses the holdup and plate-out factors described in NEDC-31858P. The analysis shows the dose contribution from the proposed increase in leakage criteria is acceptable compared to doses limits prescribed in 10 CFR 100 and 10 CFR 50 Appendix A, GDC 19. The results of the revised dose calculations are provided in Enclosure 1, Section III.B.

In addition, as provided by 10 CFR 50.12, an exemption to 10 CFR 50, Appendix J containment leakage requirements is also being requested which would allow exclusion of the MSIV leakage from the summation of containment leak rate test results. This exemption request is in Enclosure 3 and supports the TS change to increase the MSIV leakage criteria.

Regarding precedent, several other BWRs (Hatch 2, Duane Arnold, Limerick, Susquehanna, and LaSalle) have previously provided justifications for increased MSIV leakage TS using a similar approach based on NEDC-31858P, Revision 2. These applications were approved by NRC.

TVA has determined that there are no significant hazards considerations associated with the proposed changes and that the TS change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9) . The BFN Plant Operations Review Committee and the, Nuclear Safety Review Board have reviewed this proposed change, and determined that operation of BFN Units 2 and 3 in accordance with the proposed changes will not endanger the health and safety of the public. Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and enclosures to the Alabama State Department of Public Health.

A high priority is requested for the review of this proposed TS change to reduce the potential for MSIV refurbishment maintenance during the upcoming refuel outages. The next scheduled refuel outage begins in April 2000 on Unit 3, and

U.S. Nuclear Regulatory Commission Page 4 Sept:ember 28, 1999 TVA requests that the proposed TS be approved prior to that outage. The revised TS should be made effective within 60 days of NRC approval for both BFN units. contains provides a listing of commitments made in this submittal. If you have any questions, please contact me at (256) 729-2636.

Si cerely, T. E. Ab ey Manager Licensi and Industry Affai s Subscribed a to before me on is day of Sept:. y999 Notary Public y Commission Exp'ires 9/22/2002 Enclosures cc: see page 5

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U.S. Nuclear Regulatory Commission Page 5 September 28, 1999 Enclosures cc (Enclosures):

Chairman Limestone County Commission 310 West Washington Street Athens, Alabama 35611 Mr. Paul Frederickson, Branch Chief U.S. Nuclear Regulatory Commission Region II 61 Forsyth Street, S.W.

Suite 23T85 Atlanta, Georgia 30303 Mr. William O. Long, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611 State Health Officer Alabama State Department of Public Health 434 Monroe Street Montgomery, Alabama 36130-3017

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 and 3 PROPOSED TECHNICAL SPECIFICATIONS (TS) CHANGE TS-399 INCREASED MAIN STEAM ISOLATION VALVE (MSIV) LEAKAGE DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGE INDEX OF ENCLOSURES ENCLOSURE 1 DESCRIPTION OF PROPOSED CHANGE AND JUSTIFICATION I. DESCRIPTION OF THE PROPOSED TS CHANGE...... El-1 II. REASON FOR THE PROPOSED CHANGE El 1 III. SAFETY ANALYSZS

~ ~ ~ ~

....El-2 A.l DESCRIPTION OF ALTERNATE (ALT) PATH TO TO CONDENSER ~ ~ ~ ~ ~ ~ ~ ~ ~ El 4 A.2 RELIABILITY OF ALT PATH INCLUDING BOUNDARY VALVES......... ....El-5 B. RADIOLOGICAL DOSE ASSESSMENT.......... ~ ....El-7 C. STRUCTURAL INTEGRITY. ....El-7 C.l SEISMIC WALKDOWNS...................... ....El-8 C.2 VERIFICATION OF EARTHQUAKE DATABASES... ...El-10 C.3 SEISMIC ANALYSES FOR ALT DRAIN PATH.... ...El-11 C.4 SEISMIC DYNAMIC ANALYSIS OF TURBINE BUILDING..................... El-12 C.5 SEISMIC ANALYSIS FOR CONDENSER ...El-14 C.6

SUMMARY

OF SEISMIC CONSIDERATIONS...... ...El-16 IV. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION ...El-16 V. ENVIRONMENTAL IMPACT CONSIDERATION ...El-18 VI. REFERENCES ...El-19 ENCLOSURE 2 MARKED-UP TS/BASES CHANGES ENCLOSURE 3 APPENDIX J EXEMPTION I. APPLICABLE RULE.............................. .E3-1 II. REQUESTED EXEMPTION. .E3-1 III. JUSTIFICATION AND ASSESSMENT OF SAFETY IMPACT .E3-1 IV. AUTHORIZED BY LAW........ .E3-4 V. NO UNDUE RISK TO PUBLXC HEALTH AND SAFETY.... .E3-5 VI. BE CONSISTENT WITH HEALTH AND SAFETY......... .E3-5 VII. MUST ENTAIL SPECIAL CIRCUMSTANCES............ .E3-5 VIII. ENVIRONMENTAL IMPACT......................... .E3-9 ENCLOSURE 4 EQE SEISMIC EVALUATION REPORT ENCLOSURE 5 COMMITMENT LISTING

ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 and 3 PROPOSED TECHNICAL SPECIFICATIONS (TS) CHANGE TS-399 INCREASED MAIN STEAM ISOLATION VALVE (MSIV) LEAKAGE DESCRIPTION OF PROPOSED CHANGE AND JUSTIFICATION DESCRIPTION OF THE PROPOSED TS CHANGE TVA is requesting changes to the Units 2 and 3 TS Surveillance Requirement (SR) 3.6.1.3.10 to increase the allowed MSIV leakage from 11.5 standard cubic feet per hour (scfh) per valve to 200 scfh for individual MSIVs with a 400 scfh combined maximum pathway leakage for all four MSIV lines. The TS Bases are likewise being modified to match the proposed TS changes. A mark-up copy showing the exact TS and Bases changes is provided in Enclosure 2. Also, an Appendix J exemption is required to support the TS change.

It is provided in Enclosure 3.

REASON FOR THE PROPOSED CHANGE There are four steam lines installed on the BFN units, each provided with dual quick-closing MSIVs. These valves serve to isolate the reactor coolant system in the event of steam line breaks outside primary containment, a design basis loss-of-coolant accident (LOCA), or other events requiring containment isolation. For a detailed discussion of the system components and operating characteristics, refer to the descript'ion in Chapter 4.6.3 in the BFN Final Safety Analysis Report (FSAR).

For a steam line break, the MSIV isolation function terminates the blowdown of reactor coolant in sufficient time to prevent an uncontrolled release of radioactivity from the reactor to the environment by limiting the loss of coolant inventory. For the LOCA event, the valves isolate the reactor from the environment and prevent the direct release of fission products. Although the MSIVs are designed to provide a low leakage barrier, industry and BFN experience indicates that maintain the valves to it meet is difficult to always the current TS SR 3.6.1.3.10 performance criteria of 11.5 scfh.

Section 3.0 of NEDC-31858P, Revision 2, BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems, (Reference 1) summarizes industry experience with reactor MSIV leakage, and shows that many plants have difficulty in consistently meeting the low leakage rate criteria. BFN has observed similar experience with an average of two MSIVs exceeding the leakage limit each operating cycle. Rework of a MSIV to meet the 11.5 scfh criteria requires approximately 1000 to 2000 man-hours depending on the level of effort, and accumulates a radiation dose of approximately 4.5 man-rem for each complete refurbishment.

Multiple rebuilds of the MSIVs also have a potential negative effect on valve integrity in the long term. With a 200 scfh limit for individual MSIVs and a 400 scfh combined maximum pathway leakage allowance as proposed in this TS change, only one MSIV rework would have been necessary for the last three BFN Unit 2 cycles and the two most recent Unit 3 operating cycles. It is obvious that a revised TS would markedly reduce the number of MSIV rebuild operations.

Approval of this proposed TS change would be an economic benefit to TVA in terms of direct costs and a reduction in outage activities. The change would also lower personnel radiation exposure and improve the performance integrity of the MSIVs by reducing the number of maintenance activities associated with restoring the leakage to a overly strict lower limit. As discussed in the next section, these benefits can be achieved by establishing an alternate MSIV leakage path to the condenser through the inherently rugged main steam line drain system.

SAFETY ANALYSIS TVA proposes to utilize the main steam and main steam drain lines to preferentially direct MSIV leakage to the unit's main condenser as described in BWROG Topical Report, NEDC-31858P, Revision 2. This path takes advantage of the large volume of the steam lines, drain lines, and condenser to provide holdup and plate-out of fission products that may leak through the closed MSIVs following a postulated LOCA. In using this approach, the main steam lines, steam drain piping, and the main condenser are used to mitigate the consequences of an accident to limit potential off-site exposures below those e E1-2

specified in 10 CFR 100 and control room doses below 10 CFR 50 Appendix A, General Design Criteria (GDC) 19 values. NRC has previously determined that the BWROG methodology is acceptable as documented in the March 3, 1999, NRC Safety Evaluation Report (SER) (Reference 2) for NEDC-31858P, Revision 2, and that the subject BWROG report can be used for direct reference for individual licensee applications.

Off-site dose calculations and control room dose calculations have been revised as summarized in Enclosure 1 Section III.B using the alternate leak treatment (ALT) path and an increased MSIV leakage term in accordance with the methodology described in NEDC-31858P, Revision 2. The revised calculations show that 10 CFR 100 and GDC 19 limits are maintained with the proposed relaxed MSIV leakage criteria.

Because the original BFN design basis of the Turbine Building main steam piping and components was not Seismic Class I, seismic verification walkdowns and evaluations of representative piping/supports have been performed to demonstrate the main steam line piping and components that comprise the ALT path to the condenser are rugged and able to perform the safety function of MSIV leakage control and treatment following a Design Basis Earthquake (DBE). In addition, BFN retained the services of EQE International (EQE), Oakland, California, as a consultant to conduct a review of the earthquake experience data on the performance of nuclear power plants and condensers.

According to EQE, based on past earthquake experience, welded steel piping and condensers constructed to normal industrial practices have been found to be seismically rugged and not susceptible to a primary collapse mode of failure as a result of the seismic vibration modes experienced at sites examined in the earthquake database.

Using the earthquake experience-based methodology, supplemented by the walkdowns and analytic evaluations, has been concluded the primary components in the alternate it MSIV leakage path can be relied upon to maintain structural integrity during earthquakes. The seismic evaluation results are summarized in Section III.C of this Enclosure and a copy of the EQE report for Units 2 and 3 is provided in Enclosure 4.

Regarding precedent, several other BWRs (Hatch 2, Duane Arnold, Limerick, Susquehanna, and LaSalle) have previously provided justifications for increased MSIV leakage TS using a similar approach based on NEDC-31858P, Revision 2. These TS change requests were approved by NRC.

A.1 DESCRIPTION OF ALTERNATE (ALT) FLOW PATH TO CONDENSER Figure 3.1, MSIV Seismic Verification Boundary, of the EQE report (Enclosure 4) provides a flow diagram schematic which shows the ALT pathway from the MSIVs to the condenser and the boundary valves associated with the ALT pathway for BFN Units 2 and 3. As discussed in NEDC-31858P, Revision 2, the ALT pathway establishes a seismically equivalent route to contain and direct leakage from the MSIVs to the condenser following a design basis LOCA.

The ALT path is from outboard side of the MSIVs through four 3-inch lines which join a 4-inch drain line path to

'the condenser. The flow path is through normally open Flow Control Valves (FCVs) 1-168, 169, 170, and 171, and continues through FCV 1-S7, FCV 1-58, and FCV 1-59 to the main condenser. FCV 1-57 is normally open and FCVs 1-58 and 1-59 are normally closed valves. FCV 1-S9 has a 4" bypass line which also routes to the condenser. The bypass around FCV 1-59 is free of valves and orifices.

Establishment of the ALT path is based on valves FCV 1-58 and FCV 1-59 (although FCV 1-59 is not strictly required because of the bypass line) being opened by operators in response to those events taking credit for the availability of the ALT path to the condenser. On approval of the proposed TS, operating procedures will be revised to provide procedural requirements to establish the ALT path to the condenser.

FCVs 1-168, 169, 170, 171, and 57 are normally open motor operated valves which would remain open on loss of off-site power. FCV 1-58 and FCV 1-59 are normally closed valves which would require operator action to align the ALT path to the condenser. As noted above, these two valves are powered from essential power buses with emergency diesel generator back-up. To further ensure valve reliability, FCV 1-58 and FCV 1-S9 will be added to the In-Service Test (IST) program and will be periodically stroke tested. However, as stated above, FCV 1-59 has an open full bypass line, therefore, its operation is not essential to align the ALT path.

The ALT pathway includes the main steam and main steam drain piping, and branch lines, which act as a boundary volume to contain the MSIV leakage and direct ALT pathway to the condenser.

it via the The boundary valves have been reviewed and documented in Reference 4. All of the boundary valves fall into one of the following categories:

1) manual isolation valves that are normally closed;

2) Motor Operated Valves that are normally closed; 3) Air Operated Valves that are normally open, but fail closed on loss of power, loss of air, or loss of control signal;

4) Air Operated Valves that are normally open which will require an operator action to close; and 5) valves isolated by a spring assisted in-line check valve (new) which has a opening pressure in excess of the ALT drain path differential pressure once the MSIVs have closed and the line has depressurized.

The following changes to boundary valves are being made:

1) Pressure Control Valve (PCV) 1-147 is currently an air operated valve which fails open. Therefore, PCV-1-147 is being modified by changing the operator to an operator that will fail closed on loss of power, air, or control signal; and 2) the steam line supplies to the Offgas Preheaters currently do not have adequate boundary valves.

Therefore, in-line check valves are being added to these lines to serve as the boundary valves, Refer to Figure 3-1 of the EQE Report for the location of these valves.

A secondary passive flow path also exists from the MSIVs to the condenser. This flow path consists of; 1) four 2-inch bypass lines, which contain 0.25" orifices around FCV 1-168, 169, 170 and 171; 2) a 1" bypass around the FCV 1-58 valve which has a 0.1875" orifice and a normally open manual valve, HCV-1-525 and; 3) a 4" bypass line around FCV 1-59.

A.2 RELIABILITY OF ALT PATH INCLUDING BOUNDARY VALVES As noted previously, Figure 3.1 of the EQE report provides a flow diagram schematic which shows the ALT leakage pathway from the MSIVs to the condenser, and the boundary valves associated with the ALT path for Units 2 and 3. The boundary valves and material properties of the drain line piping and branch lines are documented in a TVA design calculation (Reference 4) and summarized in Table 4-3 of the EQE report. All boundary valves are either closed

during normal system operation or fail closed upon loss of power, or loss of control air or hydraulic pressure. The rationale for these valves being acceptable boundary valves is documented in Reference 4.

To establish the primary ALT path to the condenser flow path to the condenser, FCV 1-S8 and FCV 1-59 will be opened using hand switches in the main control room. FCV 1-58 will also auto-open for certain combinations of the MSIVs being closed and turbine speed conditions. As noted above, the opening of FCV 1-59 is not, essential since it inch non-orificed bypass line. Both FCV 1-58 and FCV 1-59 has a 4-are powered from essential power buses with emergency diesel backed power. Therefore, it is likely that they will be available during and after a LOCA event concurrent with a loss of off-site power.

The most limiting single active failure would be the failure of FCV 1-58 valve to open. In this condition, MSIV leakage flow would be diverted through the 1" orificed bypass line around FCV 1-58 and through normally open manual valve HCV 1-S25. With the 0.1875" orifice, calculated that the majority of MSIV leakage would it is still be directed to the condenser with a smaller remainder through the closed Main Steam Stop/Control Valves to the high pressure turbine. Therefore, even in the unlikely event of this single active failure, the bulk of the MSIV leakage will still be routed to the condenser.

As part of the ALT path isolation boundary, two in-line check valves are being added to the steam supply lines for the Offgas Preheaters as shown in Figure 3-1 of the EQE report. These check valves will be spring assisted to remain closed against less than a 5 pound per square inch (psi) differential pressure. This is based on the pressure downstream of the MSIVs being 1 or 2 psi once, the steam lines have depressurized and the condenser has lost vacuum.

The check valves will be ASME Section III, Class 2 valves.

Spring assisted in-line check valves have been used in the nuclear and non-nuclear industry to resolve issues with other types of check valves. This type of in-line check valve has been very reliable, and provides leak tight closure. Therefore, a spring assisted check valve is well suited for this application.

The References 6a and 6b drawings show those portions of the ALT path that fall under the BFN In-Service Inspection (ISI) program. These drawings show that the majority of the ALT drain path and boundary paths fall within the ISI ASME Section III Class 2 Program. Inclusion of this piping in the ISI program further ensures the reliability of this piping through periodic inspections.

B. RADIOLOGICAL DOSE ASSESSMENT TVA has previously submitted off-site and control room dose calculation results to NRC in conjunction with the Power Uprate Project on March 30, 1999 (Reference 7).

These were reviewed by NRC in a supplemental SER dated August 3, 1999 (Reference 8). The Reference 7 calculations were performed using an 11.5 scfh per MSIV line leakage term (46 scfh total) via the ALT path and modeled in accordance with NEDC-31858P, Revision 2.

For this TS change, the off-site dose calculations and control room dose calculations have been revised using a total net MSIV leakage of 400 scfh for all four MSIV lines (Reference 5). This is consistant with the proposed change to the TS 3.6.1.3.10. A comparison of the results between 46 scfh and 400 scfh leakage is provided in the table below. All doses are in man-rem units.

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Iii"';r'aVd';i'::,::I,';': 4 ~ 059 22.36 5.837 5.838 50.92 71.65

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':Seti;:::;:.;:;::-".;:;::,:';;:;:-;:;', 0.1084 0.1005 0.1005 0.3502 0.4173 The revised calculations show that margin is maintained to 10 CFR 100 and GDC 19 dose limits with the proposed increased MSIV leakage criteria.

C. STRUCTURAL INTEGRITY Portions of the main steam piping and components in the

Turbine Building at BFN were not originally designed as Seismic Class I. Therefore, seismic verification walkdowns and evaluations of representative piping/supports were performed to demonstrate the main E1-7

steam line piping and components that comprise the ALT path were rugged, and would be able to perform the safety function of MSIV leakage control following an DBE.

In addition, BFN retained the services of EQE as a consultant to conduct a review of the earthquake experience data on the performance of nuclear power plants and condensers. Using the experience-based methodology, supplemented by the walkdowns and analytic evaluations, it has been concluded the primary components in the ALT leakage path can be relied upon to maintain structural integrity. The seismic evaluation results are summarized in this section and a copy of the EQE report is provided in Enclosure 4.

C. 1 SEISMIC WALKDOWNS The seismic walkdown scope included the ALT path from the outboard MSIVs to the main condenser and includes the piping, instrumentation, valves, and equipment that would be required to maintain the primary ALT drain pathway and secondary path boundaries. The seismic verification boundary was established to envelop the ALT pathway shown on Figure 3-1 of the EQE report.

Various design attributes of the as-installed equipment, piping, and tubing were reviewed and evaluated by a Seismic Walkdown Team to ensure that BFN installations were representative of the earthquake comparison database design practices and that components were free of known seismic vulnerabilities. Earthquake experience had previously identified that certain specific conditions were more prone to failure of piping and tubing systems and components during seismic events. Some of the conditions evaluated in the seismic walkdown reviews included:

~ Piping, pipe supports and equipment design attributes

~ Seismic anchor movements

~ Seismic interaction issues (Class II/I issues and proximity issues)

~ Valve design attributes The above design attributes and conditions are discussed in more detail in Sections 3.1.1, 3.1.2, 3.1.3, and 3.1.4 of the EQE Report. Conditions which did not conform with the walkdown screening guidelines or which were judged by

the Seismic Walkdown Team to require further review were documented as open items and listed as "outliers".

BFN Unit 3 walkdowns were conducted on the ALT pathway system piping, equipment and components consistent with the scope of the ALT path as shown in Figure 3-1 of the EQE report and as documented in "Browns Ferry Unit 3 NSIV Seismic Verification Summary Report" which is contained in a BFN design calculation (Reference 3). The original scope of the seismic verification walkdown boundary (Reference 9) for Unit 2 was somewhat less than that performed for Unit 3 based on differences in planned modifications at the time. For consistency, the Unit 2 seismic verification boundary will be expanded to match the Unit 3 scope. Therefore, additional Unit 2 seismic verification walkdowns will be performed during or prior to the next Unit 2 refueling outage (Spring 2001) and any new outliers will be resolved prior to completion of the Unit: 2 outage.

As a result of the walkdowns by the Seismic Walkdown Team, 34 outliers were identified for Unit 3 and 26 for the smaller scope Unit 2 walkdown discussed above. A brief summary of the identified outliers is listed on Table 3-3 for Unit 3 and Table 3-4 for Unit 2 in the EQE Report.

Some addit:ional outliers are likely to be identified as a result of t: he planned expansion of the Unit 2 walkdown boundary. Examples of some of the outliers identified by the Seismic Walkdown Team are:

~ short rod hangers

~ spans exceeding USAS B31.1 recommendations

~ differential displacements of main pipe and branch lines

~ equipment anchorage deficiencies

~ valve performance evaluation

~ condenser and condenser anchorage evaluation

~ proximity and potential impact of piping with equipment, structural features, and other piping All outliers, except for the qualification of isolation boundary valve PCV-1-147 to the steam seal system and outliers associated with main steam supply to the Offgas Preheaters (Table 4-2 of EQE report), were remedied on Unit 3 during the 1998 Unit 3 Cycle 8 outage (Reference 3 and Reference 11). The above remaining Unit 3 outliers will be addressed during the next Unit 3 outage scheduled

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for the Spring of 2000. The Unit 2 outliers and any new outliers resulting from the additional Unit 2 seismic verification walkdowns will be addressed during the Unit 2 Cycle 11 outage in the Spring of 2001. Tables 4-1 and 4-2 of the EQE Report provide a status for resolution of the existing outliers for Unit 2 and Unit 3 outliers, respectively.

C.2 VERIFICATION OF EARTHQUAKE DATBASES To establish the applicability of NEDC-31858P, Revision 2, regarding usage of earthquake experience-based methodology for demonstrating the. seismic ruggedness of the main steam piping and associated components for BFN, comparisons of the ground response spectra of selected database facilities with BFN design basis ground spectrum were made. In general, the earthquake experience database sites have experienced strong ground motions that are in excess of the BFN DBE at the frequency range of interest (i.e., about 1 Hertz and above for piping and rigid range of equipment). Many of the database site ground motions envelope the conservatively estimated BFN DBE ground spectrum by large factors in various frequency bands within the 1 Hertz and above range. Of the 13 database facilities reviewed and approved by the NRC as documented in Reference 2, 10 were selected for comparison to BFN and are shown in Figures 2-2 through 2-11 of the EQE Report.

Based on the comparison, it is concluded that the BFN DBE ground spectrum is generally bounded by the earthquake experience database sites at the frequencies of interest.

Hence, the use of an earthquake experience-based approach for demonstrating the seismic ruggedness of non-seismically analyzed main steam piping and associated components at BFN, is consistent with NEDC-31858P, Revision 2 recommendations and with the limitations of the NRC staff's SER (Reference 2).

The main steam piping and condensers in the earthquake experience database exhibited substantial seismic ruggedness, even when they typically were not designed to resist earthquakes. This is a common conclusion in studies of this type on other plant items such as welded steel piping in general, anchored equipment such as motor control centers, pumps, valves, structures, and so forth.

That is, with limited exceptions, normal industrial construction and equipment typically have substantial inherent seismic ruggedness, even when they are not

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designed for earthquakes. No failures of the main steam piping were found. Anchored condensers have also performed well in past earthquakes with damage limited to minor internal tube leakage.

NEDC-31858P, Revision 2 contains detailed discussions and comparisons of main steam piping and condenser design in several earthquake experience database sites and example General Electric (GE) Mark I, II, and III plants in the U.S. The general conclusions of these comparisons are as follows:

GE plant designs are similar to or more rugged than those in the earthquake experience database that exhibited good earthquake performance The possibility of significant failure in GE BWR main steam piping or condensers in the event of an eastern United States DBE is highly unlikely; and that Any such failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented Additional detailed discussions of the seismic experience database comparisons and applicability to BFN are in Section 2 of the EQE Report.

C.3 SEISMIC ANALYSES FOR ALT DRAIN PATH The majority of the MSIV ALT path piping systems and related components at BFN (those segments downstream of the outboard MSIVs and the outboard drain isolation valve) are located in the Turbine Building are not designated as Seismic Class I systems. Thus, these piping systems were not typically seismically analyzed, but rather were designed and installed to the requirements of USAS B31.1-1967.

As part of the plant specific seismic verification of the non-seismic designed ALT piping, and related supports and components using the earthquake experience-based approach outlined in NEDC-31858P, Revision 2, the following reviews were performed to demonstrate that the piping and related supports fall within the bounds of the earthquake experience database:

~ Review of the design codes and standards, piping design parameters, and support configurations

~ Seismic verification walkdowns to identify potential piping concerns The BFN ALT piping systems consist of welded steel pipe and standard support components, and have support spacing that generally meet the USAS B31.1 recommended spans.

Design bases for the portions of piping associated with the ALT pathway are further tabulated in Table 4-3 of EQE Report. Table 4-4 of the EQE Report presents a general summary of the piping data that constitute the seismic experience data. Comparisons of BFN and selected database piping parameters are presented in Table 4-5 of the EQE Report. Overall, BFN piping design is similar to and well represented by those designs found in the experience database sites that have been shown to perform well in past earthquakes.

Since the BFN FSAR does not reference Appendix A to 10 CFR 100, the seismic adequacy of the ALT piping was addressed by performing seismic verification walkdowns to identify specific design attributes associated with poor seismic performance following the guidelines outlined in Section 3.1 of EQE Report. Additionally, bounding evaluations were performed for typical support configurations using evaluation criteria discussed in Section 4.1 of EQE Report. Table 4-6 of EQE Report summarizes the results of the support and anchorage evaluations for the selected bounding configurations as documented in Reference 10.

The seismic evaluations, consisting of verification walkdowns, bounding support evaluations, and resolution of the identified walkdown outliers, provide reasonable assurance that the ALT drain path piping, related supports, and components will remain functional in the event of a DBE at BFN.

C.4 SEISMIC DYNAMIC ANALYSIS OF TURBINE BUILDING The BFN Turbine Building is classified as a Class II structure. Class II structures are generally defined as structures which are important to reactor operation, but are not essential for preventing an accident which would endanger the public health and safety, and are not essential for the mitigation of the consequences of

accidents. Class II designated structures are also required to not degrade the integrity of any Class I structure.

The Turbine Building houses the main turbines, generators, condensers, other auxiliary systems and balance-of-plant systems, and related piping. It is a reinforced concrete structure below the operating floor, elevation 617 feet, and is supported on steel H-piles. The turbine foundation is separated from the rest of the Turbine Building foundations and frames by means of a one inch expansion joint. The steel super-structure above elevation 617 feet is framed by transverse welded steel rigid frames which span approximately 107 feet. An expansion joint is provided between a two-bay frame for the first two units and a single bay frame for Unit 3.

Rolled shapes, plates, and bars used for structural steel are in accordance with ASTM A36. ASTM A325 bolts were used for field structural bolt connections. For welding, E70XX electrodes are used. The steel super-structure frames are braced to provide rigidity in the direction of the Reactor Building and provide support, for the turbine cranes as well as the elaborate girt system. Frames are designed with fixed bases to resist lateral forces from the overhead cranes and wind loads due to 100 mile per hour wind in addition to supporting the vertical dead and live loads. The design of the steel super-structure is based on AISC "Manual of Steel Construction", 6th Edition.

The compressive strength of the structural concrete (f',)

is 3000 psi at 28 days cure time except that turbine building columns are 4000 psi. For evaluation/reanalysis of the structure, long-term concrete strength gain may be used. Reinforcing steel used is in accordance with ASTM A432 Grade 60 or ASTM A615 with f~ = 60000 psi and E, = 29 x 10'si. Beams and slabs have been designed by ACI working stress methods and columns designed by working stress method, and checked by ACI ultimate strength design method using a load factor of 1.8.

Where masonry walls exist in the Turbine Building, they are generally used as removable shield walls or non-load bearing partition walls. Since non-reinforced masonry walls do not perform well during seismic events, masonry walls were specifically reviewed during the seismic

verification walkdowns. Resolution of any masonry wall outliers typically focus on relocating the ALT path components outside the potential zone of influence of the masonry wall. For instance, outlier 12-1 for Unit 3 (see Table 4-2 of EQE report) noted an interaction with a masonry wall. The proposed resolution of this outlier will be to reroute the main steam supply line and add new isolation boundary valves to the Offgas Preheaters which are outside the influence of the masonry wall.

Performance of the Turbine Building and other non-seismic structures during a seismic event is of interest to the MSIV leakage only to the extent that the building structure and its internal components survive and not degrade the capabilities of the ALT pathway and condenser.

The NEDC-31858P, Revision 2, survey of this type of industrial structure has, in general, confirmed that excellent past seismic performance exists. There are no known cases of structural collapse of either turbine buildings at power stations or structures of a similar construction.

Based on the above design bases for the BFN Turbine Building, and the excellent seismic performance of similar types of industrial structures in past strong-motion earthquakes as documented in NEDC-31858P, Revision 2, the BFN Turbine Building is expected to remain structurally intact following a DBE. Additional detailed discussions of the seismic assessment of the Turbine Building is in Section 4.3 of the EQE Report. Table 4-7 of the EQE Report summarizes the design basis of the BFN Turbine Building and the applicable design codes used.

C.5 SEISMIC ANALYSIS FOR CONDENSER The BFN condensers consist of three single-pass, single pressure, radial flow type surface condensers. Each condenser is located beneath each of the three low pressure turbines, and is structurally independent. Table 4-8 of the EQE Report lists the design data for the BFN condensers and an earthquake experience database site listed in the NEDC-31858P, Revision 2. In addition, design characteristic comparisons of the BFN condensers with the selected database condensers are shown in Figures 4-2 through 4-5 of the EQE Report. The BFN condenser design data is comparable to the data for the database

site. The BFN condensers were also evaluated for structural integrity subject to seismic DBE loads.

Results of the evaluation indicate that the condenser shell stresses are small. Maximum stress ratios, based on AISC allowables, are 0.12 for combined axial and bending and 0.10 for shear (Reference 10).

The condenser support anchorage consists of a center key and six support feet that are arranged as shown in Figure 4-6 of the EQE Report. The center support is a fixed anchor and consists of a built-up wide flange H section embedded 4 feet into the concrete pedestal, which is connected to the Turbine Building base mat and welded to the bottom plate of the condenser. The support plates consist of 2 to 3 anchors of 2- to 2-1/2- inch diameter bolts. Each anchor bolt has greater than 5 feet nominal length with approximately 48 inches of embedment into the concrete pedestal, which is connected to the Turbine Building base mat. These supports are designed to resist vertical operating loads and are slotted radially from the center key to allow for thermal growth. Shear forces are transferred to the wide flange shaped anchor in the center and to the anchor bolts and shear keys to the support feet and carried through the concrete pedestal to the Turbine Building base mat.

The anchorage for the BFN condenser is comparable with the performance of the anchorages for similar condensers in the earthquake experience database. The shear areas of the condenser anchorage, in the directions parallel and transverse to the turbine generator axis, divided by the seismic demand, were used to compare with those presented in NEDC-31858P, Revision 2, and are shown in Figures 4-7 and 4-8 of the EQE Report, respectively. The BFN condenser anchorage shear area to seismic demand is substant:ially greater t:han the selected database sites.

The condenser support anchorage was also evaluated and the results indicate that the combined seismic DBE and operational demand is less than the anchorage capacity based on the AISC allowables. Maximum stress ratios are 0.70 for bolt tension in the perimeter support feet, and 0.86 for shear in the center support built-up section (Reference 10).

The above comparisons of the condenser seismic experience data and the anchorage capacity evaluations demonstrate that the conclusions presented in the NEDC-31858P, Revision 2, can be applied to the BFN condensers. That

is, a significant failure of the condenser in the event of a DBE at BFN is highly unlikely and contrary to the large body of historical earthquake experience data.

C.6

SUMMARY

OF SEISMIC CONSIDERATIONS Seismic verification walkdowns and evaluations of representative piping/supports were performed to demonstrate the main steam line piping and components that comprise the ALT path were rugged, and would be able to perform the safety function of MSIV leakage control following an DBE. Therefore, it has been concluded the primary components in the ALT leakage path can be relied upon to maintain to structural integrity.

NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION TVA is submitting a request for an amendment to the Unit 2 and 3 Technical Specifications (TS) to increase the allowable leakage rate criteria for the Main Steam Isolation Valves (MSIVs). This change request is based on the utilization of the Boiling Water Reactor Owners'roup (BWROG) methodology described in NEDC-31858P, Revision 2, BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems. TVA proposes to utilize the inherently rugged main steam drain lines to preferentially direct MSIV leakage to the main condenser.

This drain path takes advantage of the large volume of the steam lines and condenser to provide holdup and plate-out of fission products that may leak through the closed MSIVs. In this approach, the main steam lines, steam drain piping, and the main condenser are used to mitigate the consequences of an accident to limit potential off-site exposures below those specified in 10 CFR 100 and 10 CFR 50 Appendix A, General Design Criteria (GDC) 19 for control room dose limits.

TVA has concluded that operation of BFN Units 2 and 3 in accordance with the proposed change to t: he TS does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91(a)(1), of the three standards set forth in 10 CFR 50.92(c).

~ -5 The ro osed amendment does not involve a si nificant increase in the robabilit or conse uences of an accident reviousl evaluated.

TVA proposes to utilize the main steam drain lines to preferentially direct MSIV leakage to the main condenser. This drain path takes advantage of the large volume of the steam lines and condenser to provide holdup and plate-out of fission products that may leak through the closed MSIVs. In this approach, the main steam lines, steam drain piping, and the main condenser are used to mitigate the consequences of an accident to limit potential off-site exposures below those specified in 10 CFR 100 and 10 CFR 50 Appendix A, GDC 19 for control room dose limits.

Seismic verification walkdowns and evaluations of representative piping/supports were performed to demonstrate the main steam line piping and components that comprise the ALT path were rugged, and able to perform the safety function of MSIV leakage control following an Design Basis Earthquake (DBE). Thus, has been concluded the primary components in the MSIV it alternate treatment flow path can be relied upon to maintain structural integrity.

Therefore, the proposed amendment does not involve changes to structures, components, or systems which would affect the probability of an accident previously evaluated in the Browns Ferry Final Safety Analysis Report (FSAR).

A plant-specific radiological analysis has been performed to assess the effects of the proposed increase in MSIV leakage criteria in terms of off-site doses and main control room dose. This analysis uses the holdup and plate-out factors described in NEDC-31858P, Revision 2. The analysis shows the dose contribution from the proposed increase in leakage criteria is acceptable compared to doses limits prescribed in 10 CFR 100 and 10 CFR 50, Appendix A, GDC 19. Therefore, the proposed changes do not significantly increase the consequences of an accident previously evaluated.

l B. The ro osed amendment does not create the ossibilit of a new or different kind of accident from an accident reviousl evaluated.

The proposed changes require the use of the main steam piping and the condenser to process MSIV leakage.

This additional function does not compromise the reliability of these systems. They will continue to function as intended and not be subject to a failure of a different kind than previously considered. In addition, MSIV functionality will not be adversely impacted by the increased leakage limit. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

C. The ro osed amendment does not involve a si nificant reduction in a mar in of safet The proposed change to TS Surveillance Requirement 3.6.1.3.10 to increase the allowable MSIV leakage does not involve a significant reduction in the margin of safety. The allowable leak rate specified for the MSIVs is used to quantify a maximum amount of leakage assumed to bypass containment. The results of the re-analysis supporting these changes were evaluated against the dose limits contained in 10 CFR 100 for off-site doses and 10 CFR 50, Appendix A, GDC 19 for control room doses. Sufficient margin relative to the regulatory limits is maintained even when conservative assumptions and methods are utilized. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

ENVIRONMENTAL IMPACT CONSIDERATION The proposed change does not involve a significant hazards consideration, a change in the types of, or increase in, the amounts of any effluents that may be released off-site, or a significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.

II VI . REFERENCES BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems, NEDC-31858P, Revision 2, September 1993

2. Letter from F. Akstulewicz, NRC, to T. A. Green, General Electric, March 3, 1999, Safety Evaluation for GE Topical Report, NEDC-31858P, Revision 2

3. BFN Calculation CD-Q0001-980038, Main Steam Seismic Ruggedness Evaluation, (Unit 3 Summary Report)

BFN Calculation MD-Q0001-960036, MSIV Leakage Containment System Boundaries, Physical Properties, System 001

5. BFN Calculation ND-Q0031-920075, Control Room and Off-site Doses Due to a LOCA TVA drawings:

a) 2I 3 47E801 1 ISII 2 ISI b) 2I 3 47E807 2 ISI

7. Letter from TVA to NRC dated March 30, 1999, BFN-Resolution of Control Room Emergency Ventilation (CREV) System Issues With Regard to License Condition Associated With Units 2 and 3 Power Uprate Operating License Amendments 254 and 214 (TAC Nos. M99711 and M99712)

8. Letter from NRC to TVA dated August 3, 1999, Safety Evaluation Supplement, BFN Units 2 and 3, Radiological Dose Calculations Associated with Power Uprate License Amendment Nos. 254 and 214 (TAC Nos. MA5991 and MA5992)

,9. BFN Calculation CD-N2001-990112, Main Steam Seismic Ruggedness Evaluation, (Unit 2 Summary Report) 10 BFN Calculation CD-N0001-990113, Seismic Evaluation Report, Unit 0 BFN Calculation CD-N0001-980039, Main Steam Seismic Ruggedness Verification