ML18039A365

Submits Response to RAI by NRC Re TS Change TS-384 Request for License Amend to Operate at Uprated Power Level of 3458 Mwt.Response to 980422,RAI for 971001 Proposed Change,Encl

ML18039A365
Person / Time
Site:	Browns Ferry
Issue date:	05/22/1998
From:	Abney T TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9806030179
Download: ML18039A365 (247)

Text

CATEGORY REGULA'x RY INFORMATION DXSTRIBUTI A SYSTEM (RIDS)

ACCESSION NBR:9806030179 DOC.DATE: 98/05/22 NOTARIZED: NO DOCKET FACIL:50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 50-296 Browns Ferry Nuclear Power Station, Unit 3, Tennessee 05000296 AUTH. NAME ABNEY,T.E.

AUTHOR AFFXLXATION Tennessee Valley Authority ~~p~A RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk) ~

SUBJECT:

Submits response to RAI by NRC re TS change TS-384 request C for license amend to operate at uprated power level of 3458 MWt. A D1STR1BDT10N CODE: A001D CORTES RECEIVED:LTR ENCL S1EE:

TITLE: OR Submittal: General Distribution E

NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-3 1 1 PD2-3-PD 1 1 DEAGAZIO,A 1 1 R INTERNAL: ACRS 1 1 01 1 1 NRR/DE/ECGB/A 1 1 NRR/DE/EMCB 1 1 NRR/DRCH/HICB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 NUDOCS-ABSTRACT 1 1 OGC/HDS3 1 0 ERNAL: NOAC 1 1 NRC PDR 1 1 D

U E

N t NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD) ON EXTENSION 415-2083 TOTAL NUMBER OF COPIES REQUIRED: LTTR 14 ENCL 13

"I Il 0

Tennessee Va! ley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 May 22, 1998 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D AC. 20555 Gentlemen:

In the Matter of .) Docket Nos. 50-260 Tennessee Valley Authority ) 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING UNITS 2 AND 3 TECHNICAL SPECIFICATION (TS) CHANGE TS - 384 REQUEST FOR LICENSE AMENDMENT FOR POWER UPRATE OPERATION (TAC NOS. M99711 AND M99712)

This letter provides additional information requested by NRC in support of TS-384. On October 1, 1997, TVA provided TS-384, an amendment to Operating Licenses DPR-52 and DPR-68 that will allow Units 2 and 3 to operate at an uprated power level of 3458 MWt.

The enclosure provides TVA's response to the April 22, 1998, NRC RAI for the October 1, 1997, proposed TS change. This letter includes replies to each of the NRC requests.

'P806030i79 'ti80522 PDR ADQCK 050002h0 P PDR  !

0 U. S. Nuclear Regulatory Commission Page 2 May 22, 1998 If you have any questions, please telephone me at (256) 729-2636.

e I T. E. Abney Manager of @censing and Indust y Affairs cc: See Page

4 E 'I

~ '

U.S. Nuclear Regulatory Commission Page 3 May 22, 1998 REFERENCES TVA letter to NRC dated, October 1, 1997, Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 Technical Specification (TS) Change TS-384 - Request For License Amendment for Power Uprate Operation

2. TVA letter to NRC dated March 16, 1998, Browns Ferry Nuclear Plant (BFN) Units 2 and 3 Technical Specification (TS) No. 384 Supplement 1 - Request for License Amendment for Power Uprate Operation

3. NRC letter to TVA dated April 22, 1998, Browns Ferry Plant Units 2, and 3 Request for Additional Information Regarding Technical Specification Change TS-384 Request for License Amendment for Power Uprate Operation (TAC Nos. M99711 and M99712)

~ '

U.S. Nuclear Regulatory Commission Page 4 May 22, 1998 Enclosure cc (Enclosure):

Albert W. De Agazio, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 Mr. Harold O. Christensen, Branch Chief U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 NRC Resident Inspector BFN Nuclear Plant 10833 Shaw Road Athens, Alabama 35611 L. Raghavan, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852

0 ENCLOSURE TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 BROWNS FERRY NUCLEAR PLANT (BFN) - RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING UNITS 2 AND 3 TECHNICAL SPECIFICATION (TS) CHANGE TS - 384 REQUEST FOR LICENSE AMENDMENT FOR POWER UPRATE OPERATION (TAC NOS. M99711 AND M99712)

This enclosure provides the TVA response to the April 22, 1998, NRC request for additional information (Reference 4) .

The following questions are related to the General Electric (GE),

Licensing Topical Report NEDC-32751P, "Power Uprate Safety Analysis for the Browns Ferry Nuclear Plant (BFN), Units 2 and 3," dated September 1997 which is Enclosure 5 to the TVA letter dated October 1, 1997 (Reference 1).

NRC RE VEST 1 a With respect to revised stress and fatigue analyses of mechanical components as a result of the power uprate, please:

(a) Describe the analytical methodology, assumptions, loading combinations and allowable limits used for evaluating piping stresses and cumulative fatigue usage factors (CUF), pipe supports, load limits at nozzles, penetrations, guides, valves, pumps, heat exchangers and anchors. Please address all mechanical components affected by the power uprate, including but not limited to, reactor vessel and internal components, control rod drive mechanisms, balance-of-plant (BOP) piping systems.

TVA REPLY 1 a The analytical methodology, assumptions, loading combinations and allowable limits are different based on the mechanical .

component/system under evaluation. Thus, this response is divided into the following sections: Reactor Vessel Internal Components, Control Rod Drive System, Reactor Pressure Vessel, and Plant Piping Systems (nuclear steam supply system (NSSS) and BOP) .

19806030179 I

REACTOR VESSEL INTERNAL COMPONENTS For BFN, structural evaluation for a component consists of the determination of the stress value at the maximum stress location for power uprate loads in conjunction with a comparison of stress results to the allowable limits. Stress determination methods depend on the nature of the prior analysis of record for the component involved. The methods and assumptions used in the BFN power uprate project are based on the generic guidelines as documented in Section 5.5.1.1 and Appendix I of NEDC-31897P-A "Generic Guidelines for GE BWR Power Uprate," May 1992.

Based on calculations performed for the original structural evaluations, specific revised stress values were determined for power uprate loads'f structural evaluations for a component were originally based on detailed stress analyses, such as finite element modeling, the power uprate loads were compared with the load combinations for the prior analysis.

If the loads for power uprate were found to be less than or equal to the previous analysis loads the stress determination was concluded with stresses documented as remaining less than the previous analysis.

If power uprate loads exceeded the prior analysis load basis for detailed analyses, a revised stress was derived from the previously calculated stress by scaling the pre-uprate stresses by the ratio of power uprate load values to comparable load values in the prior analysis.

The loading combinations and allowable limits are based on the BFN current design bases which require load combinations and stress limits for reactor internal structures summarized in Table 1 (a) -1.

CONTROL ROD DRIVE SYSTEM For power uprate operating with a reactor dome pressure of 1050 psia, the increase in reactor dome pressure produces an operating pressure of 1085 psia at the reactor bottom head.

The control rod drive (CRD) mechanism has been designed for 1265 psia which is higher than the bottom head pressure for power uprate. The components of the CRD mechanism, which form part of the primary pressure boundary, were originally designed in accordance with the applicable American Society of E-2

Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)

Code,Section III. Since the original stress and fatigue analyses assumed a pressure of 1265 psia, these analyses were not revised for power uprate, and no new analytical computer codes were utilized. The power uprate engineering evaluations show that stresses and fatigue usage factors will remain within their original design allowables.

The limiting component of the CRD mechanism is the indicator tube which has a calculated primary membrane plus bending stresses of 20,790 psi. The allowable stress is conservatively specified as 26,060 psi (i.e., 1.5 Sm). The maximum stress on this component results from the maximum CRD internal hydraulic pressures of 1750 psig caused by an abnormal operating condition. Therefore, the stresses for the limiting CRD component will be maintained within the allowable stress criteria for the maximum power uprate operating pressure.

The CRD mechanism has been subjected to intensive testing at 1265 psia, which is higher than the maximum power uprate pressure. Based on the demonstrated performance of the CRD mechanism at these high pressures, it is concluded that deformations resulting from the power uprate pressure increase are of no significant consequence.

The small increase of approximately 4'ahrenheit (F) in reactor bottom head coolant temperature for power uprate conditions has an insignificant effect on the CRD mechanism.

The CRDs have been designed for temperatures up to 575',

which is higher than the maximum bottom head temperature of 532' for power uprate conditions'herefore, the existing design of the CRD mechanism is adequate for power uprate conditions.

The original analysis for cyclic operation of the CRD was conservatively evaluated in accordance with applicable requirements specified in the ASME BE PV Code,Section I II.

For example, when considering the loadings resulting from scram with a leaking scram discharge valve, scram with a failed buffer, and scram without CRD cooling water flow, the limiting component was found to be the CRD main flange. The fatigue usage factor is 0.15 .which is less than the allowable limit of 1.0. All requirements are satisfied even when E-3

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3)

Reactor Protection Systes Instruaentati on APPLICABLE COND IT IONS NODES OR RNUI RED REFERENCED OTHER CHANNELS FROH SPECIFIED PER TRIP REQUIRED SURVEILLANCE .ALLQIABLE FUNCTION COND I T I ONS SYSTEH ACTION D.1 REQUIREHENTS VALUE

1. Intermediate Range Honitors

~ a. Neutron F lux--High SR 3.3.1.1.1 S 120/125 SR 3.3.1 ~ 1.3 divisions of SR 3.3.1.1.5 full scale SR 3.3.1.1.6 SR 3.3.1 1.9

~

SR 3.3.1.1.14 5(a) SR 3.3.1.1.1 5 120/125 SR 3.3.1.1.4 divisions of SR 3.3.1.1.9 full scale SR 3.3.1.1.14

b. Inop SR 3.3.1.1.3 NA SR 3.3.1.1 '4 5(a) SR 3.3.1.1.4 NA SR 3.3.1 ' '4

2. Average Pouer Range Honitors 3.3.1.1.1 < 15X RTP

a. Neutron Flux -High, SR Setdoun SR 3.3.1.1.3 SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.9 SR 3.3.1.1.14

b. F lost Biased S inatated SR 3.3.1.1.1 < 0.58 M Theraet Pter -High SR 3.3.1.1 2 ~ + 66X RTP and SR 3.3.1.1.7 S 120X RTP SR 3.3.1 1.8

~

SR 3.3.1.1.9 SR 3.3.1.1.11 SR 3.3 1.1.14

~

3.3.1.1.1 < 120X RI'P

c. Neutron Flux -High SR SR 3.3.1.1.2 SR 3.3.1.1.7 SR 3.3.1.1.8 SR 3.3 '.1.9 SR 3.3.1.1 14~

(cont inued)

(a) Uith any control rod uithdrain from a core cell containing one or nere fuel asserhiies.

BFN-UNIT 3 3.3-6 Amendment

considering the increased power uprate vessel bottom head pressure, thereby satisfying the peak stress intensity limits governed by fatigue. Fatigue usage factors based on actual plant loading cycles were not calculated since TVA is not required to collect such data and, therefore, does not have this information available'ther components of the CRD system will also be subjected to the dome pressure increase since the system must supply water at an increased pressure in order to maintain all of the CRD system functions. Since these components are not in direct contact with reactor coolant, they are not subject to the small temperature increase. The design pressure of these components is based on the maximum pressure that could be supplied by the CRD pumps. Since the maximum pressure to which these components will be subjected is enveloped by the maximum pressure capable of being supplied by the pumps, the existing design pressure is adequate for power uprate.

Components connected to the scram discharge volume will be subjected to the dome pressure increase and the small increase in the coolant temperature. Since the components are designed for 1250 psig, the maximum pressure to which these components will be subjected is enveloped by the design pressure. The small increase of approximately O' in reactor bottom head coolant temperature for power uprate conditions will have no effect on the components, which are designed for 280'.

REACTOR PRESSURE VESSEL A. General Procedure of Evaluation The procedure for reactor pressure vessel (RPV) structural evaluation used in the BFN power uprate project is based on the generic guidelines, as documented in Section 5.5.1.1 and Appendix I of NEDC-31897P-A "Generic Guidelines for GE BWR Power Uprate," May 1992. An ASME BRPV Code Section III evaluation is performed to assess the effects of changes in design basis operating conditions due to power uprate operation on limiting components of the BFN Units 2 and 3 RPVs. GE reviewed the currently applicable reactor vessel stress reports (References 5 through 11, 15 and 16) and reconciled any differences in reactor operating pressure, temperature, and flow at power uprate conditions.

For increases in design, emergency, and faulted conditions, primary stresses are recalculated for the new conditions.

For increases in normal and upset conditions, a screening criterion is used to determine which vessel components are analyzed as bounding components. Stresses for these components are scaled up based on the uprated conditions, and usage factors calculated and compared with allowable usage.

The screening and selection criterion for power uprate stress analysis is discussed in this section. This approach is consistent with NEDC-31897P-A, Appendix I. Table l(a)-2 lists the CUFs for RPV components for the pre-uprate conditions. RPV components which have original fatigue usage factors greater than 0.5 are selected for the power uprate stress analysis. The code allowable of 1.0 for CUF is still applied for the power uprate analyses. These selected components are considered to be the bounding components of the RPV. However, if operating and design conditions for an RPV component do not change, the original stress report is still valid and the component need not be reanalyzed. The following components have usage greater than 0.5 (see Table 1(a)-2) and are, therefore, selected for power uprate primary plus secondary (P+Q) and primary plus secondary plus peak (P+Q+F) stress analysis:

Feedwater Nozzle Recirculation Outlet Nozzle Support Skirt Main Closure Stud Additionally, since design, emergency, and faulted condition loads increase for the recirculation inlet nozzles, primary stresses were recalculated for these components. Effects of increased transient pressure loading (acoustic loads) on shroud support are also evaluated.

B. Method of Evaluation B.l Power Uprate Stress Analysis for Non-Bolting Materials The power uprate stress analysis uses the guidelines and procedures of the ASME BEPV Code,Section III (Code).

For the components under consideration, the 1965 Code with Addenda to and including Summer 1965 for Unit 2 E-5

(Reference 12) and Summer 1966 for Unit 3 (Reference 13) is the code of construction and the governing code.

However, if a component has undergone a design modification, the governing code is the code used in the stress analysis of the modified component. Feedwater nozzle and recirculation outlet nozzle were the only components that fall under such category.

1 ~ Design Conditions Design conditions (for example, design temperature and pressure) are typically defined as more severe than normal and upset operating conditions. Therefore, there are no changes in the design conditions due to power uprate, design stresses (general primary membrane, primary membrane plus primary bending) remain unchanged, and the Code requirements of Paragraphs N-414.1 through N-414.3 of References 12 and 13 and Section NB-3221 of Reference 14 (the governing code for the feedwater nozzle safe-end replacement) continue to be met for the RPV components analyzed.

2. Normal (Level A) and Upset (Level B) Conditions Effects of chan es in ressure tern erature and nozzle flow rates In general, changes in normal operating pressures, temperatures, and nozzle flow rates will increase the P+Q stress intensity ranges and the P+Q+F stress intensity ranges at a particular location on the RPV component.

The stress components [3 normal (cz o a0 ) and 3 shear (x , x,o, x0,)] of the P+Q stresses and the P+Q+F stresses consist of pressure stress components, thermal cycling stress components, and mechanical stress components (resulting from reaction loads from attached piping and/or a thermal sleeve, or from seismic loads) .

The normal stress due to pressure is directly proportional to the reactor coolant pressure, and the normal stress due to thermal cycling can be conservatively treated as proportional to the temperature change during a thermal transient (final E-6

transient temperature minus initial transient temperature), for small temperature variations as in the case of power uprate.

Xn the case of nozzles, increases in coolant flow through the nozzle will increase the forced convection heat transfer coefficients on the inside (fluid side) surface of the nozzle. These increases in heat transfer coefficients will change the temperature distribution through the nozzle, thus changing the thermal stresses in the nozzle slightly. However, it is judged that small changes in the heat transfer coefficient on the nozzle inside surface have negligible effects on the temperature distribution through the nozzle.

For the power uprate analysis, a scaling approach was used to conservatively scale up the original stresses to account for pressure and temperature increases due to power uprate. Xn,many pressure vessel calculations, the three stress directions of the orthogonal coordinate system (r, 9, and z) at a particular location on the component of interest are chosen because the shear stress components are mostly negligible and, thus, the normal stress components are the principal stresses. Therefore, the principal stress due to pressure is directly proportional to the coolant pressure and the principal stress due to thermal cycling is proportional to the temperature change during a thermal transient, provided that the normal stress directions and the principal stress directions coincide. Since there are usually no changes in the mechanical stresses due to power uprate, the new (power uprate) value for principal stress is:

<total, new +pressure,old* (pnew/pold) ++thermal,old* (~ new/~ old) ++nechanlcal or

+total, new +pressure,old* ( ) ++theaaal,old* ( ) ++mechanical where:

(SCF) Pressure scaling factor = (pn.w/p,ld)

(SCF) Thermal scaling factor = (dT,/bTold) b,T Final transient temp. - initial transient temp.

E-7

If the flow rate increase due to power uprate is large, a flow factor (SCF)p must also be applied. The (SCF) is required because an increase in the flow rate will increase the heat transfer and, therefore, the thermal stress. For this case, the (SCF)p is determined as a function of heat transfer coefficients. The overall thermal scaling factor becomes:

SCFT = (bTnew/bTold)

• SCFP For stress reports which list values for pressure stresses, thermal cycling stresses, and mechanical stresses separately, one may calculate power uprate principal stresses by scaling the original pressure stress by (SCF) and the original thermal cycling stress by (SCF) and combining them with the original P

mechanical stress. For stress reports which do not list these values separately, a conservative scaling technique was developed where the original principal stress components are scaled up by the larger of (SCF) P and (SCF) T . This is conservative because (a) the larger scaling factor (SCF) is used to scale up the stresses, and (b) the mechanical stress is scaled up as well. In addition, if the scaling factor is less than unity, the power uprate principal stresses are not scaled down. In that case, an SCF value of 1.0 is used (i.e., the original values are retained).

To further simplify the scaling technique, the larger scaling factor (SCF) can be applied directly to the original stress intensity values instead of to the original principal stresses. Stress intensity (or "stress difference") is determined by taking the algebraic difference between any pair of principal stresses. The following example illustrates why SCF can be applied directly to the stress intensities:

S12,new +1, new +2, new

(<1 old* SCF) - (<2 old* SCF)

(+l,old +2,old)

• SCF S12,old

• SCF E-8

ASME Code stress limits for normal and u set conditions According to Paragraph N-414.4 of References 12 and 13 and Sections NB-3222 and NB-3223 of Reference 14, the P+Q stress limit is met if the maximum primary plus secondary stress intensity range (S) at a location on the component is less than 3Sm of the material. If the 3Sm limit is not met, then plastic behavior is assumed and simplified elastic-plastic analysis can be used to determine structural adequacy.

For those components which do not meet the requirements of Paragraph N-415.1 of References 12 and 13 or Paragraph NB-3222.4(d) of References 14, a fatigue evaluation must be performed to assure that the component does not fail by material fatigue. For adequacy, the cumulative fatigue usage factor must be less than 1.0.

Procedure for calculatin ower u rate P+ stress in tensi t ran e The following procedure is used for calculating the power uprate P+Q stress intensity range (S,) for the limiting location on the RPV component of interest.

This power uprate value will then be compared with the ASME Code stress limit.

(a) Determine the pressure scaling factors [(SCF) ] P and the thermal scaling factors [(SCF) ] for the stress cycles originally analyzed using the appropriate power uprate operating condition changes. Determine the larger of the two scaling factors for each stress cycle. This value is SCF.

(b) Multiply SCF for each stress cycle by the original P+Q stress intensity values of the original governing stress report. The original governing stress report is the most recent stress report listed in Paragraph 2.1 of References 15 and 16.

(c) Determine the maximum absolute value of the extremes of the range through which the power E-9

0 uprate stress intensities (calculated in Step b) fluctuate over time. This value is S,. Note that there are two stress cycles associated with this value.

(d) Determine the allowable P+Q stress intensity range (3S ,) evaluated at the maximum temperature of the two limiting stress cycles of Step 3.

However, according to Note 1 of Figure N-414 of References 12 and 13, the average value of 3Sm for the highest and lowest temperatures of, the metal during the transient can be used in the analysis if it can be shown that the secondary stress is due to thermal loads and not mechanical loads.

(e) Compare S, with 3S ,. If S, ( 3S ,, the ASME Code P+Q stress limit is met. If S, )

3S ,, then simplified elastic-plastic analysis must be used, using power uprate values where applicable.

Procedure for ower u rate fati ue evaluation The following procedure is used for calculating the power uprate cumulative fatigue usage factor (U,) for the limiting location on the RPV component of interest.

This power uprate value will then be compared with the ASME Code stress limit.

(a) Multiply SCF for each stress cycle by the original P+Q+F stress intensity values of the original governing stress report.

(b) For each of the limiting stress cycle pairs used in the fatigue analysis of the original governing stress report, determine the absolute value of the difference of the power uprate P+Q+F stress intensities (calculated in Step 1). This value is Sp+Qip, new E-10

(c) Determine the power uprate alternating stress intensity, S, ,, for each of the original limiting stress cycle pairs as follows:

Salt, new (1/2)

• Ke ,* (E,/Ea)

• Sp+g+p new where eg new Simplified elastic-plastic factor 1.0, for S, c 3S

1. 0 + [ (1-n) /n (m-1) ] * [ (S,/3S ,) - 1],

for 3S ,< S,< 3mS (E,/E ) Elastic modulus correction factor (both factors are described in Paragraphs NB-3228.3 and NB-3222.4 of Reference 14).

(d) Use S,, as the value of the ordinate when entering the applicable design fatigue curve in References 12 and 13 or Reference 14 to find the corresponding allowable number of cycles (N,,)

for each of the limiting stress cycle pairs.

(e) Calculate the power uprate incremental fatigue usage factor (U, , = n,/N, ,) for each of the limiting stress cycle pairs, where n, is the lesser of the actual number of design cycles for each pair. The lesser number is used because the value of the P+Q+F stress intensity range for the limiting stress cycle pair is only experienced by the component over the lesser number of cycles.

(f) Calculate the power uprate cumulative fatigue usage factor (U, = ZU, ,) . If U, < 1.0, the ASME Code limit is met.

(g) Satisfy special stress limits of Paragraph NB-3228 of Reference 14 if applicable.

3. Emergency (Level C) and Faulted (Level D) Conditions Primary stresses due to emergency and faulted conditions are recalculated for those RPV components whose emergency and faulted loads increase and the stresses are evaluated against the Code requirements of Sections NB-3224 and NB-3225 of Reference 14.

B.2 Power Uprate Stress Analysis for Bolting Materials

1. Design Conditions Since there are no changes in the design conditions due to power uprate, the bolt design stresses remain unchanged and the ASME Code requirements of Section N-416 of References 12 and 13 and Section NB-3231 of Reference 14 continue to be met.

2. Normal (Level A), Upset (Level B), and Emergency (Level C) Conditions Effects of chan es in ressure and tern erature ln general, changes in normal operating pressures and temperatures will increase both bolt service stresses (the stresses averaged across the bolt cross section and at the periphery of the bolt cross section) and the peak bolt stresses.

ASME Code stress limits for normal u set and emer enc condi ti ons According to Paragraph N-416 ' of References 12 and 13 and Paragraph NB-3232.1 of Reference 14, structural adequacy is met if the maximum value of service stress, averaged across the bolt cross section and neglecting stress concentrations, is less than 2Sm for the bolting material.

According to Paragraph N-416.1 of References 12 and 13 and Paragraph NB-3232.2 of Reference 14, structural adequacy is met if the maximum value of service stress at the periphery of the bolt cross section is less than 3Sm of the bolting material.

For those components which do not meet the requirements of Paragraph N-415.1 of References 12 and 13 or Paragraph NB-3222.4(d) of Reference 14, a fatigue evaluation must be performed to assure that the bolts do not fail by material fatigue. For adequacy, the cumulative fatigue usage factor must be less than 1.0.

Procedure for calculatin ower u rate service stresses The general procedure for calculating the power uprate service stresses for the limiting location on the bolt, and comparing it to the ASME Code stress, is as follows:

(a) Determine the pressure scaling factors [(SCF),]

and the thermal scaling factors [(SCF)~1 for all stress cycles originally analyzed using the appropriate power uprate operating condition changes. Determine the larger of the two scaling factors for each stress cycle. This value is SCF.

(b) Multiply SCF for each stress cycle by the original service stress values from the original governing stress report, both averaged across the bolt cross section and at the bolt periphery.

(c) Determine the allowable service stress values (2S , for the service stress averaged across the bolt cross section, and 3S , for the service stress at the bolt periphery), evaluated at the maximum temperature for the limiting stress cycles.

(d) Compare the power uprate service stresses with their allowable values. The ASME Code stress limits are met if the allowable values are not exceeded.

Procedure for ower u ra te fa ti ue eval ua ti on The general procedure for calculating the power uprate cumulative fatigue usage factor (U,) for the limiting location on the bolt, and comparing it to the ASME Code stress limit, is as follows:

(a) Multiply SCF for each stress cycle by the original peak bolt stress values of the original governing stress report.

(b) For each of the limiting stress cycle pairs used in the fatigue analysis of the original governing stress report, determine the absolute value of the

difference of the power uprate peak bolt stresses (calculated in Step 1). This value is Sp Determine the power uprate alternating stress intensity (S,,) for each of the original limiting stress cycle pairs as follows:

Salt:,new (1/2) * (E /E )

• S...,

where (E./E.) = same factor as in Section B.1.2 (c) Use S, , as the value of the ordinate when entering the design fatigue curve to find the corresponding allowable number of cycles (N, ,)

for each of the limiting stress cycle pairs.

(d) Calculate the power uprate incremental fatigue usage factor (U, , = n,/N, ,) for each of the limiting stress cycle pairs, where n, is the lesser of the actual number of cycles for each pair. The lesser number is used because the value of the peak bolt stress range for the limiting stress cycle pair is only experienced by the bolt over the lesser number of cycles.

(e) Calculate the power uprate cumulative fatigue usage factor (U, = ZU, ,) . If U, < 1.0, the ASME Code limit is met.

3 ~ Faulted (Level D) Conditions The stresses due to faulted conditions remain unchanged and the Code requirements of Section NB-3235 of Reference 14 continue to be met for the bolts.

C. Assumptions C.1 Design Condition Changes As noted in Paragraph 4.3 of the Design Specifications for the BFN power uprate program (References 15 and 16),

the power uprate design requirements (Design Pressure =

1250 psig, Design Temperature = 575o F) are unchanged from the original design requirements specified in the

RPV purchase documents, except for the recirculation inlet nozzle loads described in Section C.2.

C.2 Operating Condition Changes Paragraph 4.4.1 of References 15 and 16 lists the changes to the reactor vessel thermal cycles. Paragraph 4.4.2 of References 15 and 16 lists the changes to the Reactor Vessel Nozzles Thermal Cycles. Per Paragraphs 4.4 '.3 of References 15 and 16, the recirculation inlet nozzle thermal sleeve hydraulic loads, Fy, Fz, and Mx are revised from 8. 7 kips, -21. 5 kips, and 84. 6 in-kips to 10.0 kips, 20.2 kips, and 93.0 in-kips, respectively.

D. Loading Combinations Loading combinations used in the original or modified stress analyses are unchanged for power uprate conditions.

E. Allowable Stress Limits Allowable stresses used in the original or modified stress analyses are unchanged for power uprate conditions.

PLANT PIPING SYSTEMS NSSS and BOP Plant Piping Systems (Safety Related) 1 The scope of evaluation and the evaluation method for the safety related piping systems used in the BFN power uprate project are described in Section 5.5.2 and in Appendix K of NEDC 31897P-A, "Generic Guidelines for GE BWR Power Uprate,"

May 1992. Existing BFN design basis documents (e.g., design specifications and piping stress reports) were reviewed to determine the design and analytical basis for safety related piping systems. The power uprate parameters of safety related piping systems were compared with the existing analytical basis to determine effect in temperature and pressure due to power uprate conditions. The power uprate temperatures and pressures are bounded by the temperatures and pressures used in the pre-uprate piping analyses and therefore, no new calculations were made.

The results of these evaluations demonstrate that the requirements of the existing code of record are satisfied for

all safety related piping systems at the power uprate conditions. Interface loads on system components (pipe supports, load limits at nozzles, penetrations, guides, valves, pumps, heat exchangers and anchors) have not increased due to power uprate conditions, and do not exceed component acceptance criteria.

Plant Piping Systems (Non-Safety Related)

In addition to the safety-related piping systems, the non-safety related piping systems expected to be affected by power uprate conditions, such as condensate and feedwater heaters piping, were evaluated against predicted power uprate conditions and confirmed to be within the existing design envelope. The approach is also in accordance with the generic guideline for power uprate (NEDC-31897P-A). The pressure and temperature increases for non-safety related piping systems and components attached to safety related piping systems were also evaluated. Based on previous power uprate engineering experiences and plant's operating histories, the non-safety related BOP piping and components performance are considered acceptable for power uprate conditions.

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TABLE 1(a) -1 LICENSING BASIS LOADING COMBINATIONS AND ALLOWABLE STRESS Loading Combination Stress Allowable Buckling Category I Pm Pm+Pb Criteria Upset Upset RIPD Normal RIPD 1.0 Sm 1.5 Sm < 0.40 cr plus OBE Emergency Normal RIPD LOCA 1.5 Sm 2.25 Sm 5 0.60 cr plus DBE Faulted LOCA plus 2.0 Sm 3.0 Sm 0. 80 DBE 0'IPD

- Reactor Internal Pressure Differential Loads DBE - Design Basis Event seismic loads OBE - Operating Basis seismic loads LOCA - Loss of Coolant Accident Buckling Criteria - Based on the allowable ratio of permissible loads/critical load where o' is the stress determined by buckling stability analysis.

Pm - general primary membrane stress Pb - primary bending stress E-17

TABLE 1(a) -2

SUMMARY

OF CUMULATIVE FATIGUE USAGE FACTORS BEFORE POWER UPRATE Pre-Power Uprate Cumulative RPV Component Fatigue Usage Unit 2 Unit 3 Feedwater Nozzle (rapid plus system cycling) 0dgc 1 0 Recirculation Inlet Nozzle 0.425 '.425 Recirculation Outlet Nozzle 0.717"'. 0 717 "

Core Spray Nozzle 073 0.073 CRD Hydr. Sys. Ret. Nozzle 0. 363 0.363 CRD Penetration 0.005 0.005 2" Instrumentation Nozzle 0.06 0.06 Support Skirt 0.55 0.

Refueling containment 55'.328 skirt 0.328 Shroud Support 0.17 0.17 Main Closure:

Closure Shell/Flange 0.000 0.000 Stud 0.762 0. 762'.032 Vessel Shell 0.032 Selected for usage calculation in power uprate evaluation.

Unless otherwise stated, the values are taken from the original RFV vendor stress reports (Reference 5 for Unit 2 and Refeience 6 for Unit 3).

From feedwater nozzle modification stress reports (Reference 7 for rapid cycling and Reference 8 for system cycling).

d From recirculation inlet nozzle modification stress report (Reference 9).

'rom recirculation outlet nozzle modification stress report (Reference 10).

From core spray nozzle modification stress report (Reference 11).

J NRC RE UEST 1 b (b) Indicate whether the analytical computer codes used in the evaluation are different from those used in the original design-basis. If not, please justify their use including their qualification for such applications, and discuss how code differences are reconciled.

TVA REPLY 1 b The same computer codes used in the BFN original design basis evaluations for stress and fatigue analyses of the RPV and reactor internal mechanical components were used for the power uprate evaluation. No computer codes were used for the other structural evaluations. No new analyses were performed for safety and non-safety related piping systems and components for power uprate conditions, hence no analytical computer codes were used.

NRC RE VEST 1 c (c) Provide a comparison of the maximum calculated stress, and CUF for the components evaluated, at the design basis and power uprate conditions. Please indicate whether the revised analyses are performed to the Code of record; otherwise describe the Code and Code edition used for the revised analyses and justify their use. Discuss how the calculated CUFs compare to those resulting from the actual loading cycles based on the data recorded during plant operation.

TVA REPLY 1 c The reply to Request 1(c) is based on the mechanical component/system under evaluation. Therefore, the following sections are provided: Reactor Vessel Internal Components, Control Rod Drive System , Reactor Pressure Vessel, and Plant Piping Systems (NSSS and BOP).

REACTOR VESSEL INTERNAL COMPONENTS No specific ASME code edition was applicable to the reactor internals during the time of BFN Units 2 and 3 design and construction (1966 - 1977) . However, consistent with the original evaluation of the reactor internals and the BFN Final

Safety Analysis Report (FSAR), the power uprate evaluation was performed in accordance with the intent of the ASME BRPV Code Section III, Subsection NB, 1989 Edition for structural design criteria in conjunction with additional BFN FSAR structural criteria. Use of the ASME Code is justified as a conservative structural evaluation criteria.

Table 1(c) -1 provides a comparison of maximum stresses at power uprate and design basis (100. power) for selected reactor internal components. The stress values listed result from the same analysis method except for the input loads which are at 100% or 105% power accordingly.

It is noted .that in some cases, there is little or no change in maximum stress results due to 105% power, as the governing loading was seismic, which did not change. Also, for some components, the faulted LOCA RIPD loads were less at 105%

power.

CONTROL ROD DRIVE SYSTEM Please see Reply 1(a) for the discussion on the Code of record and Code edition used for the power uprate evaluation.

REACTOR PRESSURE VESSEL As stated in Reply 1(a), six components were evaluated for power uprate:

Feedwater Nozzle Recirculation Outlet Nozzle Support Skirt Main Closure Stud Recirculation inlet Nozzle Shroud Support Except for the feedwater nozzle and recirculation inlet nozzle, the other components were evaluated to the Codes of Record (see Reply 1(a)). Since both of these components were modified, the use of Code other than the Code of Record was justified in the respective modification design documentation.

Thus, during the power uprate evaluations, the use of the later Code editions and addenda (1974 Edition with addenda up to and including Summer 1976 for feedwater nozzle, 1980 Edition with addenda up to and including Winter 1981 for E-20

Recirculation inlet nozzle) were not required to be justified.

See Tables 1(c)-2 and 3 for results.

2'm act of ower u rate acoustic loads on the shroud su ort Primary stress analysis results from Reference 5, Report ll, are summarized in Table 1(c) -4. The methodology from Reference 5, Report ll, is used to calculate uprate stresses, based on the increased acoustic loads. For the purposes of this assessment, the acoustic load combination is considered equivalent to a faulted condition load combination. The applicable ASME Code editions (References 12 and 13) did not include design and analysis rules for faulted conditions.

Therefore, rules contained in a later edition of the Code (Reference 14) are used. Results are shown in Table 1(c)-4 and compared to the original results.

The calculated CUFs in both the original and power uprate fatigue analyses are conservatively based on the maximum design number of loading cycles, over the life of the plant, for each design basis transient. Since it is not a requirement of the applicable design Codes or the BFN licensing basis, actual loading cycle data (other than number of reactor startups and shutdowns) has not been systematically recorded at BFN. It is therefore not possible to compare the CUFs calculated using design basis loading cycles to those resulting from the loading cycles that have actually occurred.

PLANT PIPING SYSTEMS NSSS and BOP The power uprate temperatures and pressures are bounded by the pre-uprate temperatures and pressures used for piping analysis and, therefore, the calculated pipe stresses, support loads, equipment nozzle loads and containment penetrations loads from existing stress reports are bounded for power uprate conditions. No new calculations were made.

The power uprate parameters affecting fatigue are bounded by the parameters used in the pre-uprate evaluations and the CUF calculations are not affected by power uprate conditions.

For non-safety related piping systems and components, the evaluations were performed to the code of record. The pressure and temperature increases for non-safety related piping systems and components are within their design

requirements. Based on previous power uprate engineering experiences and plant's operating histories, the non-safety related BOP piping and components performance are considered acceptable for power uprate conditions.

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TABLE 1 (c) -1 MAXIMUM STRESS VALUES Upset Conditions Stress Stress Reactor Stress Location & Result Result Allowable Component Type 100% power 105t power Stress Shroud Base Junction 2.18 ksi 2. 18 ksi 7.72 ksi (buckling)

Shroud Bolts* 33.93 ksi 33.99 ksi 34.95 ksi Head Core Buckling pressure 24.58 psid 25.24 psid 28.0 psid Plate differential Faulted Conditions Stress Stress Reactor Stress Location & Result Result Allowable Component Type 100> power 105% power Stress Shroud Base Junction 10.09 ksi 10.80 ksi 15.44 ksi (buckling)

Shroud Bolts* 41.27 ksi 41.43 ksi 69.9 ksi Head Core Buckling pressure 32.0 psid 30.0 psid 42.0 psid Plate differential Jet Pump Diffuser Base 24.02 ksi 26.69 ksi 38.0 ksi Assembly Junction* .

• primary membrane + primary bending E-23

TABLE 1(c) -2

SUMMARY

OF POWER UPRATE STRESS AND FATIGUE RESULTS Original Power Power Original Power Power P+Q SI Uprate Uprate Fatigue Uprate Uprate Component (ksi) P+Q SI Allowable Usage Usage Allowable (ksi) (ksi)

Feedwater Nozzle 45.2 47.2 80.1 1.0 0.984 1.0 Recirculation Outlet 73.2 75.5 80.1 0.717 '.779 1.0 Nozzle Main Closure Stud 47.1 49.2 73.4 98.9 103.3 110.1 0.762 0.762 1.0 Skirt 'upport 115.9 80.1 0.55 0.904 1.0 System plus rapid cycling, from Reference 7.

For all locations except the inner surface downstream of the secondary seal, which is qualified by the power uprate fracture mechanics evaluation in accordance with NUREG-0619.

First line of stress is averaged across bolt section; second line is maximum at periphery.

P+Q stress is acceptable by elastic-plastic analysis, consistent with NEDC-31897P-A, Appendix I.

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TABLE 1(c) -3 PRIMARY STRESS RESULTS FOR THE RECIRCULATION INLET NOZZLE Pre-Uprate Uprate Allowable Category Condition Location Stress, Stress, Stress, ksi ksi ksi Pm Design Nozzle 8.5 8.5 26.7 Safe End 12.3 12 ' 17.3 Level C Nozzle 10.1 10.1 44.0 Safe End 13.4 13.4 20.7 PL+PB Design Nozzle 16.0 16.4 40.1 Safe End 17.7 18.5 25.9 Level C Nozzle 22.2 22.5 66.0 Safe End 21.7 21.7 31.0 Note: Since CUF for the pre-uprate analysis was below screening criteria value of 0.5, the CUF for power uprate conditions for this components was not evaluated TABLE 1 (c) -4 SHROUD SUPPORT LOCAL PLUS BENDING (PL+PB) STRESS COMPARISON Loading Condition Calculated Stress, Allowable Stress, ksi ksi Design Operating 24.5 1.5 Sm = 34.95 Faulted (Power Uprate) 66.0 3.0 Sm = 69.9 E-25

NRC RE UEST 1 d (d) CUFs in Table 3-3 for the reactor vessel is provided at four locations; feedwater nozzle, recirculation outlet nozzle, main closure stud and the support skirt. Provide a comparison of CUFs for the limiting reactor vessel components between the design basis analysis and the power uprate condition reconciled. Explain why the CUF for the feedwater nozzle based on the pre-uprate power condition is greater than the calculated CUF that incorporates the power uprate condition.

TVA REPLY 1 d A comparison of CUFs for the limiting reactor components between the pre-uprate design basis analysis and the power uprate condition reconciled is shown in Table 1(c)-2. The CUF for feedwater nozzle for power uprate conditions was lower than pre-uprate conditions due to reasons outlined below:

(a) The high CUF blend radius region (portions past the sleeve seals) is qualified during power uprate conditions by inspections in accordance with NUREG-0619 rather than ASME Code fatigue analysis acceptance criteria. This eliminates the blend radius locations where CUF would have been beyond 1.0 due to power uprate conditions. Fracture mechanics

.assessment for feedwater nozzle in accordance with NUREG-0619 was performed for the power uprate condition and demonstrated the structural integrity of the feedwater nozzle blend radius.

(b) The seal refurbishment period is chosen to be 30 years such that all locations other than the blend radius region have system plus rapid CUF for power uprate conditions below 1.0 (highest value being 0.984 as reported in Table 1(c)-2).

NRC RE VEST 1 e (e) Please discuss whether you expect modifications to piping, equipment or their supports for the power uprate. Xf any, list the piping systems and pipe supports requiring modification and discuss the nature of these modifications.

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TVA REPLY 1 e As stated in Section 3.5 and 3.12 of Enclosure 5 to Reference 1, there is no adverse impact to the piping, equipment or their supports due to the power uprate implementation. Therefore, no modifications to these systems are required.

NRC RE UEST 2 It is stated (page 3-1) that main steam relief valves (MSRVs) operate in the safety (spring) mode only and the over pressure analysis assumes MSRV opening tolerance to be at least 3. above the nominal setpoints with one MSRV out of service. Table 5-1 indicates that the assumed setpoints include a +3% tolerance.

The MSRVs are the Target Rock 2-stage design, which has a history of upward setpoint drift significantly greater than +3% in the safety mode (including some found to be effectively stuck closed during testing). Please demonstrate that the 3% tolerance is conservative or provide analysis results for more representative Target Rock SRV setpoint performance.

TVA REPLY 2 The anal y si s used the generic +3'. setpoint tolerance upper limit value which is considered to be a more realistic value. This value is consistent with the ASME/ANSI OM-1 Pressure Relief Valve 3% setpoint limit criteria for in-service surveillance testing of MSRVs and is in accordance with the +3% value accepted by the NRC in its safety evaluation (SE) review letter of March 8, 1993 (TAC NO. M79265), to the Boiling Water Reactor Owners Group (BWROG) licensing topical report NEDC-31753P (BWROG In-Service Pressure Relief Technical Specification Revision). The SE states that analyses of the design basis for the overpressure event use the 3% tolerance limit for MSRV setpoints to confirm that the vessel pressure does not exceed the ASME pressure vessel code upset limits TVA is aware of the Target Rock Two Stage MSRV safety (spring) setpoint drift performance concern. MSRV setpoint drift found outside the 3'. limit will be evaluated and reported in accordance with 10 CFR 50.73 as a Licensing Event Report (LER). For such cases, the specific impact of the "As Found" test results versus the analyzed value is identified and discussed.

E-27

Xn addition, to minimize the effects of setpoint drift, BFN has implemented a modification for Unit 2 that installed pressure switches to automatically open the MSRVs when system pressure exceeds the pressure switch setpoint (See LER 260/97-008, Reference 23). This automatic mode of operation will mitigate high spring safety valve setpoint drift experienced by the Target Rock Pilot Operated MSRVs. For Unit 3 (See LER 296/97-003, Reference 24), this modification is planned for the next refueling outage.

NRC RE VEST 3 You stated that an assessment of flow-induced vibration of the reactor internal component due to power uprate was performed to estimate the vibration levels by extrapolating the recorded vibration data at BFN and by using the operating experience of similar plants (Section 3.3.2.1) . Please provide a detailed sample evaluation and the technical basis for using the operating experience of similar plants in determining that the flow induced vibration at BFN Units 2 and 3 will remain within the acceptance limits.

TVA REPLY 3 BACKGROUND BFN Unit 1 is the NRC designated BWR prototype for TVA plants.

The BFN Unit 1 reactor internals were extensively instrumented during the startup testing of the plant for purposes of vibration monitoring to confirm the structural integrity of major components in the reactor with respect to flow induced vibration.

BFN Units 2 and 3 reactor internals were also instrumented and their reactor internal vibrations measured during startup. At BFN Unit 1, there were a total of 65 sensors installed in the reactor, consisting of 45 strain gages, 3 accelerometers, 3 pressure gages, 13 displacement gages and one thermocouple.

Measurements were made at the following locations: shroud, steam separators, jet pump assemblies, in-core guide tubes, control rod guide tubes, feedwater sparger and in-core instrument tubes.

Extensive vibration measurements were made over a period of two years covering a wide range of operational conditions from pre-operational (without fuel), precritical (with fuel but not critical) and power operational tests. The power operational tests were conducted at 50%, 75% and 100% rod line conditions at various core flows. Using this data, extrapolations were carried E-28

out to project responses at 105% rod line, corresponding to the power uprate condition. The predicted responses were compared to the GE allowables corresponding to 10,000 psi peak stress intensity (acceptance criteria) to determine the acceptability of vibration level. The GE allowable is more conservative than the ASME allowable of 13,600 psi. A value of less than 10,000 psi for the response implies that no fatigue usage is accumulated by the component due to flow induced vibration. The predicted responses for power uprate conditions indicate that all vibrations are within the GE acceptance criteria.

During power uprate the components in the upper zone of the reactor, such as the steam separators and dryer, are most affected by the increased steam flow. Components in the core region and components such as the core spray line are primarily affected by core flow. Components in the annulus region such as the jet pump are primarily affected by the recirculation pump drive flow. Hence, only the steam separator and dryer are significantly affected by power uprate conditions. The flow induced vibrations of all other reactor internals depend primarily on the core flow and the recirculation pump drive flow.

Since power uprate conditions do not require any increase in core flow, and very little increase (approximately 0.8;) in the drive flow, small increases in flow induced vibrations on the components in the annulus and core regions would be expected as a result of power uprate. The extrapolation of measured test data to 105% core thermal power confirmed that all vibrations will be within the allowable criteria.

SAMPLE EVALUATION As stated in the preceding section, the BFN Unit 1 reactor internals were extensively instrumented during the start-up testing of the plant for the purpose of vibration monitoring to confirm the structural integrity of the major components in the reactor with respect to flow induced vibration. The sensor signals were recorded on-line during the test program on magnetic media and on brush chart paper. The tape recorded signals were played back and analyzed using the Dynamic Signal Analyzer to determine the vibration amplitude at different frequencies.

The signals of the components were analyzed at balanced flow test conditions at 75: and 100: rod lines to determine the expected vibration response in the power uprate region. The extrapolated E-29

vibration peak amplitude response in the power uprate region (105% rod line) was compared with the GE allowable design criteria of 10,000 psi peak stress intensity. At this stress level, sustained operation is allowed without incurring any fatigue usage.

In order to apply the vibration criteria, a dynamic structural analysis is performed to relate peak stresses to measured strains or displacements at sensor locations. Mathematical models for each component are developed using finite element methods.

Natural frequencies and modes of vibration are calculated. The location of the peak stress intensity is identified, including the effects of stress concentration factors. The modal strains and displacements at sensor locations are determined relative to peak stress intensity on a normalized basis, such that the highest peak stress intensity is 10,000 psi. The contribution of the various modes are absolute summed for conservatism.

At BFN Unit 1 (the prototype reactor), there were 47 test conditions at various core flows and different rod lines. Of these, the vibrations from test condition H-20 at 75% rod line and H-24 at 100~ rod line, which were measured at about rated core flow, were first extrapolated to 105% core flow by square extrapolation. From the vibration levels at 105. core flow at the two power levels, linear extrapolation was done to the 105%

power level. The analysis is conservative for the following reasons:

1. The GE criteria of 10,000 psi is much more conservative than the ASME criteria of 13,600 psi.

2. The modes are absolute summed.

3. The maximum vibration amplitude in each mode is used in the absolute sum process whereas in reality the vibration amplitude fluctuates.

A sample evaluation is shown in Table 3-1 for the jet pump which has strain gages S1 and S2 mounted on the riser brace in an additive bridge.

BASIS FOR USING OPERATING EXPERIENCE OF SIMILAR PLANTS The BFN units belong to the BWR/4 family of the 251" vessel diameter size ~ BFN 1 is the NRC designated prototype plant for E-30

this family. There are many other plants in the BWR/4-251" size family whose reactor internals are substantially identical to BFN units and which have operated at 105% uprated power conditions (e.g., Peach Bottom 2 6 3, Susquehanna 1 6 2 and Limerick 1 8 2).

Since the rated core flow, steam flow and power rating of these plants are the same as the BFN plants, and since the flow induced vibration evaluation for these plants were themselves based on the BFN 1 plant measurements, their operating experience at uprated conditions provide additional confirmation for the conclusion that the flow induced vibrations at BFN Units 2 and 3 will remain within acceptable limits.

E-31

TABLE 3-1 EXTRAPOLATION FOR JET PUMP VIBRATION DATA, SENSORS S1/S2 (2) (3) (4) (5) (6) (7) (8) (9) 65.9% Power 98.6't Power 105> Power 95 '+o CF 105+o CF 97.56% CF 105>o CF 105+o CF Test Extrapo- 9~ Test Extrapo- ~o Extrapo- ~o Condition lation Criteria Condition lation Criteria lation Criteria H-20 H-24 Frecpxency (microstrains (microstrains (% of (microstrains (micro strains (% of (micro strains (% of Hz I-P) P-P) 10,000 psi) P-I) P-P) 10 000 psi) P-r) 10,000 psi) 15.00 3.64 4.39 1.08 6.76 7.83 1.92 8.50 2.08 23.50 9.00 10.86 2.66 14.10 16.33 4.00 17.40 4.27 33.00 16.14 19.47 8.11 20.60 23. 86 9.94 24.72 10.30 45.50 8.60 10.37 4.32 5.90 6.83 2.85 10.37 4.32 82.00 2.59 3.12 1.30 2.91 3.37 1.40 3.42 1.42 115.00 25.15 30.34 12.64 18.53 21.46 8.94 30.34 12.64 136.00 2.97 3.58 1.06 2.99 3 ~ 46 1.02 3.58 1.06 180.00 1.27 1.53 0.45 1.65 1.91 0.57 1.99 0.59 186.00 11.71 14.13 4. 18 13.14 15.22 4.50 15.43 4.57 190.00 3.87 4.67 1.38 4.45 5.15 1.53 5.25 1.55 203.00 5.41 6.53 1.93 17.43 20.19 5.97 22.86 6.76 207.00 6.54 7.89 2.33 7.68 8.90 2 '3 9.09 2.69 Absolute 41.45< Absolute 45.28+a Absolute 52.26%

Sum All (4145 psi)

Sum All (4528 psi)

Sum All (5226 psi)

Modes= Modes= Modes=

See next a e for notes.

E-32

Notes for Table 3-1 Column Description Lists the various measured frequencies of vibration of the jet pump.

(2) Lists the measured magnitudes of vibration at the corresponding frequencies during 65.9. power and 95.6'. core flow (CF).

(3) Lists the Column (2) values extrapolated to 105. core flow using the square law.

(4) Lists the resulting maximum stress due to each mode expressed as a percent of 10,000 psi. The stresses are absolute summed and shown at the bottom.

(5) (6) (7) Identical in nature to columns (2), (3) and (4), except that the values correspond to 98.6% power.

(5) Lists the measured magnitudes of vibration at the corresponding frequencies during 98.6 ~ power and 97.56 'ore flow.

(6) Lists the Column (5) values extrapolated to 105% core flow using the square law.

(7) Lists the resulting maximum stress due to each mode expressed as a percent of 10,000 psi. The stresses are absolute summed and shown at the bottom.

(8) Lists the Column (3) values at 65.9; power and Column (6) values at 98.6'.

power linearly extrapolated to 105; power. Note that at some frequencies, vibrations decrease with power but no credit is taken for this and the highest measured value is used.

(9) Lists the resulting maximum stresses due to each mode expected at 105% power expressed as a percent of 10,000 psi. The stresses are absolute summed and shown at the bottom. Expected vibratory stress at 105. power = 5226 psi peak (less than GE acceptance criteria of 10,000 psi peak).

E-33

r NRC RE VEST 4 You stated (Section 3.5) that the effects of power uprate have been evaluated for the recirculation loop piping using the present code of record, B31.1 Power Piping Code, 1967 Edition.

You also stated that the piping was evaluated for compliance with the American Society of Mechanical Engineers (ASME) code stress criteria. Provide an explanation of why both B31.1-1967 Power Piping Code and ASME Code were used in the evaluation for the reactor coolant piping. Identify the piping analysis aspects for which both design codes were employed.

TVA REPLY 4 The original code of record for the recirculation piping analysis is USAS B31.1-1967 Power Piping Code for design and analysis.

The B31.1 Power Piping Code does not require fatigue analysis calculations either to piping system or pipe components.

However, calculations were originally performed for the 28" X 24" nominal diameter tees attached to reactor recirculation and residual heat removal (RHR) to calculate CUF using ASME BEPV Code. The power uprate parameters affecting fatigue, are bounded by the parameters used in this pre-uprate evaluation and the CUF calculations for these tees are not affected by power uprate conditions. This is the basis for the reference to the ASME B&PV Code in Section 3.5 of enclosure 5 to Reference 1.

You stated (Section 3.5) that, based upon evaluations performed by GE for similar plants, the BFN recirculation piping is judged to be acceptable for flow induced vibration, due to 105. power uprate conditions. Provide the technical basis for determining that the flow induced vibration levels in the recirculation piping system at BFN are bounded by evaluations performed for other plants.

TVA REPLY 5 For BFN, the reactor recirculation piping system flow rate increases approximately 0.8. with the power uprate condition.

Hence, there is no measurable change to the BFN reactor recirculation piping system flow induced vibration values due to power uprate. Previous 5'. power uprate evaluations for BWR units E-34

with similar recirculation flow rate characteristics to BFN, such as Peach Bottom, have also concluded that power uprate has insignificant impact on the recirculation piping vibration response.

NRC RE UEST 6 Please discuss the functional performanc'e of safety-related mechanical components (i.e., valves and pumps) affected by the power uprate to ensure that the performance specification and technical specification requirements (e.g., flow rates, close and open times) will be met for the proposed power uprate. Confirm that safety-related power-operated valves will be capable of performing their intended function(s) following the power uprate, including such affected parameters as fluid flow, temperature, pressure and differential pressure, and ambient temperature conditions'VA REPLY 6 The system evaluations performed to support the BFN Units 2 and 3 power uprate request considered the effects of increased pressure, flow, process temperature and ambient temperature on the functional performance of safety related components.

The proposed increase in licensed reactor power from 3293 megawatt-thermal (MWt) to 3458 MWt results in a nominal reactor pressure increase of 30 psi and. an increase in main steam flow of 6% The increase in nominal reactor pressure results in a 30 psi increase in the lift setpoints for the MSRVs. Ambient equipment room temperature increases due to power uprate conditions are calculated to be on the order of 3 to 5 degrees F. The maximum post-LOCA containment pressure increase is calculated to be 1.0 psi. Additional conservatism was included in the system evaluations in that the nominal reactor pressure increase was assumed as 35 psi and the MSRV setpoint increase included an additional three percent margin for setpoint drift.

From the changes in plant parameters for the new power uprate conditions, evaluations of plant systems were conducted to ensure that pump and valve functional requirements continued to be met.

Motor operated valves included. in the GL 89-10 program were evaluated separately to ensure that valve operation is acceptable at power uprate conditions. The results of GL 89-10 analysis identified four valves for each BFN unit that require torque E-35

switch adjustments due to power uprate conditions. These modifications will be accomplished prior to power uprate implementation on the respective BFN unit. The results of the system evaluations verified that the functional performance of safety related mechanical components (e.g., flow rates, close and open times) will be met for the proposed power uprate.

Table 6-1 shows a comparison of key pre- and post-power uprate parameters.

The results of the system evaluations performed for the proposed power uprate have identified some GL 89-10 valves as requiring minor modifications for power uprate conditions. These modifications will be accomplished prior to power uprate implementation. Once these modifications are implemented, power operated valves will be capable of performing their intended function following power uprate, taking into account the changes in fluid flow, pressure, differential pressure, process temperature, and ambient temperature.

E-36

~ '

TABLE 6-1 COMPARISON OF 'KEY PARAMETERS Post Uprate Power Uprate Parameter Pre Uprate (value used Analysis in analysis) Results Reactor Slight Pressure used decrease in in Main Steam 1015 psia 1070 psia MSIV closing Isolation time for Valve (MSIV) power uprate analysis Main steam 3,510,000 3,622,500 flowrate per lbs/hr lbs/hr MSIV MSIV closing 3-5 seconds 3-5 seconds time MSRV simmer 100 psi 100 psi No increase margin 110 psi 110 psi in MSRV seat 120 psi 120 psi leakage for power uprate Main steam No increase line flow for 13. 38E6 14. 15E6 in MSRV MSRV ibm/hr ibm/hr leakage vibration RHR pump flow Net positive

[Low Pressure 11, 000 gpm 11,000 gpm suction head Coolant (NPSH)

Injection acceptable (LPCI) mode] for power uprate RHR injection System valve valve maximum stroke times pressure 1341 psig 1386 psig

• acceptable for power uprate RHR pump flow NPSH for 6500 gpm 6500 gpm acceptable suppression for power pool cooling uprate E-37

TABLE 6-1 COMPARISON OF KEY PARAMETERS Post Uprate Power Uprate Parameter Pre Uprate (value used Analysis in analysis) Results Core Spray NPSH pump flow 3125 gpm 3125 gpm acceptable for power uprate Core Spray System valve injection stroke times valve maximum 1020 psig 1055 psig

• acceptable pressure for power uprate Standby SLC system Liquid components Control (SLC) 1225 psig 1279 psig acceptable pump maximum for changes discharge in system pressure parameters SLC pump flow 53 gpm 53 gpm Reactor dome Scram valve pressure for performance CRD scram 1015 psia 1075 psia and CRD scram analysis performance acceptable CRD drive CRD water 1290 to 1300 1325 to 1335 positioning pressure psig psig function acceptable CRD cooling CRD pump water flow capacity 35.2 to 62.9 35.2 to 62.9 adequate for gpm gpm power uprate flow and pressure conditions E-38

~,

TABLE 6-1 COMPARISON OF KEY PARAMETERS Post Uprate Power Uprate Parameter Pre Uprate (value used Analysis in analysis) Results Reactor RBCCW system Building components Closed 3370 gpm 3370 gpm acceptable Cooling Water for changes (RBCCW) in system system flow parameters RBCCW temperature 18.5 F 19.5 F rise Reactor Core RCIC system Isolation components Cooling 600 gpm, 600 gpm acceptable (RCIC) system for changes pump flow in system RCIC parameters pump/turbine 4500 rpm 4600 rpm speed RCIC system 150 to 1120 150 to 1174 test pressure Ps 3.g Ps3.g range High Pressure HPCI system Coolant components Injection 5000 gpm 5000 gpm acceptable (HPCI) pump for changes flow in system HPCI parameters pump/turbine 4000 rpm 4100 rpm speed 4 HPCI system 150 to 1120 150 to 1174 test pressure Ps3.g Ps 3.g range E-39

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TABLE 6-1 COMPARISON OF KEY PARAMETERS Post Uprate Power Uprate Parameter Pre Uprate (value used Analysis in analysis) Results Reactor Water RWCU system Cleanup 133,300 133,300 components (RWCU) system ibm/hr ibm/hr acceptable flow for changes RWCU in system isolation 1020 psid 1045 psid parameters valve maximum Qp Reactor Recirculation Recirculation system system 1233.8 psia 1272.8 psia components discharge acceptable pressure for changes Recirculation 45991 gpm 46553 gpm in system pump flow parameters RHR service RHRSW system water (RHRSW) components flow 4500 gpm 4000 gpm ** acceptable for changes in system parameters

• The limiting condition is the valve operability test mode for these valves. The RHR (LPCI) and Core Spray required system flows and pressures do not change for power uprate.

• • The RHRSW flow rate reduction for power uprate is discussed in section 4.1.1.1 of TS-384 (Reference 1).

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NRC RE VEST 7 Regarding Sections 3.5 and 4.1.2, please provide your evaluation of the piping systems attached to the torus shell, vent penetrations, pumps, and valves, that may be affected by the loss-of-coolant accident dynamic loads (pool swell, condensation oscillation, and chugging) considered in the evaluation for the power uprate.

TVA REPLY 7 The Design Basis Accident (DBA)-LOCA dynamic loads including the pool swell loads, vent thrust loads, condensation oscillation (CO) loads and chugging loads are described in detail in Reference 17. These loads were originally defined and evaluated for BFN Units in Reference 18. The structures attached to the torus shell, such as piping system, vent penetrations, pumps and valves are based on these design basis DBA-LOCA dynamic loads.

For the power uprate condition, the DBA-LOCA containment responses as shown in Reference 18 are re-evaluated as described below.

VENT THRUST LOADS The vent thrust loads were originally provided in Reference

19. These were calculated using the methods described in Reference 17. The postulated DBA causes the most rapid pressurization of the containment system, the largest vent system mass flow rate and, therefore, the most severe vent system thrust loads. The pressurization of the containment for the IBA (intermediate break accident) and SBA (small break accident) is less rapid than for the DBA; thus, the resulting vent system thrust loads for the SBA and IBA are less severe.

Consequently, vent system thrust loads for only the DBA are presented in Reference 19.

The vent thrust load evaluation for power uprate uses the results of the DBA-LOCA containment response at the uprated power condition, with the methods described in Reference 17 to calculate new vent thrust loads. Loads are calculated for the main vents, vent headers, and the downcomers and are then compared to the Reference 19 load values. All vent thrust loads calculated at the power uprate conditions are less than the Reference 19 base case vent thrust loads. Therefore, the vent thrust loads at the power uprate condition are acceptable

and there are no resulting impact on the structures attached to the torus shell.

POOL SWELL LOADS In the event of a postulated DBA-LOCA, the drywell and vent system are pressurized and the water initially in the downcomers is accelerated downward into the suppression pool.

After the downcomers are cleared of water, air is discharged into the wetwell below the pool surface and the pool swell transient starts.

The controlling parameter for the pool swell loads which is affected by changes to the vessel conditions due to power uprate is the drywell pressurization rate. Changes to other parameters such as vent submergence, drywell-to-wetwell operating pressure difference and containment volumes also impact the pool swell response. Impact of these parameters was also considered in the analyses.

The design basis pool swell loads for BFN are defined in Reference 19. These loads are based on the plant unique Mark I 1/4 scale tests (Reference 20). The power uprate evaluation included a review of the Mark I 1/12 scale tests (Reference

21) to determine the sensitivity of pool swell loads to changes in submergence and pressurization rate. It was determined that the net effect of reduced pressurization rate due to the use of more realistic breakflows (from the LAMB computer code) and increased submergence is to produce pool swell loads which are bound by the loads defined in Reference

19. Therefore, the pool swell loads are acceptable at the power uprate condition. As such, the integrity of the structures attached to the torus shell subject to the pool swell loads, such as pipings and attached pumps/valves are also acceptable at the power uprate condition.

CONDENSATION OSCILLATION CO LOADS Following the pool swell transient of a postulated LOCA, there is a period during which steam condensation at the vent exit is characterized by steady periodic pressure fluctuations at the vent exit. CO loads are caused by the transmission of these periodic pressure oscillations on the torus shell, submerged structures, and the vent system.

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The loads specified for CO are based on the Full Scale Test Facility (FSTF) tests (Reference 22). Xn these tests, a prototypical segment (one bay) of a Mark l torus and vent system were subjected to steam and liquid blowdown simulating a range of LOCAs. CO loads with the largest amplitude occur during the DBA.

Evaluation of the CO load for the power uprate condition is based on a comparison of predicted CO root mean square (RMS) pressure for FSTF test conditions used to define the CO loads (Reference 17) versus predicted CO loads for the analyzed containment response at the power uprate conditions. The results showed that the predicted power uprate RMS pressure would not exceed the design basis value specified in Reference

17. Therefore, CO loads at the power uprate condition are acceptable, and the integrity of the torus shell, submerged structures, and the vent system are also confirmed at the power uprate condition.

CHUGG1NG LOADS Chugging occurs during a postulated LOCA when the steam flow through the containment vent system falls below the rate necessary to maintain steady condensation at the downcomer exits. The corresponding flow rates for chugging are less than those of the CO phenomenon. During chugging steam bubbles form at the downcomer exits, oscillate as they grow to a critical size (approximately downcomer diameter), and ultimately collapse when the steam energy to the bubble is exceeded by the heat transfer to the surrounding pool water.

The resulting load on the torus shell consists of a low frequency component which corresponds to the oscillating bubbles at the downcomer exits as they grow, and higher frequency components corresponding to the collapsing bubbles.

Chugging occurs during a LOCA when the steam mass flux through the vent becomes so small that a steady steam/water interface can no longer be maintained. The chugging load is thus not sensitive to the initial reactor operating conditions. The design chugging load for BFN is based on the FSTF full scale tests (Reference 22) . These test were run for a range of blowdown and containment conditions to bound all Mark I plants. The containment conditions studied were initial wetwell free-space pressure (nominal to nominal + 5 psig),

downcomer submergence (3.33 to 4.5 feet), and vacuum breaker performance (present to not-present). These conditions continue to be bounding for chugging with power uprate conditions. Therefore, the current BFN design chugging load is not impacted by power uprate. As such, the integrity of the torus shell, submerged structures, and the vent system are also confirmed at the power uprate condition.

NRC RE UEST 8 Regarding Section 3.7, provide a detailed discussion of the effects of the steam flow increase, identified in Table 1-2, on the design basis analysis of the main steam piping due to main steam isolation valve (MSIV) closure and turbine stop valve closure loads. Also, provide an evaluation of MSIV due to the increase in the hydraulic pressure for the higher flow rate following the power uprate, as discussed in Section 4.7 of GE's Generic Evaluation (GE, Licensing Topical Report NEDC-32523P, "Generic Evaluation of General Electric Boiling Water Reactor Power Uprate - Supplement 1, Volume 1," dated June 1996).

The hydrodynamic load induced by closing the MSIV is normally bounded by the corresponding load induced by closing the turbine stop valve (TSV) . The MSIV closing time is between 3 to 5 seconds. The TSV closing time is around 0.1 second; therefore, the hydrodynamic load induced by closing the TSV is bounding for the MSIV closure. This conclusion is also shown in NEDC-31984P, Section 4.7.

An analysis was previously performed. by Engineering Data Systems (EDS) for the BFN units, titled "Shock Loading of Main Steam and Turbine Bypass Piping." This EDS analysis concluded that the load on the BFN main steam line due to shock loading resulting from sudden closure of the TSV is negligible and within the main steam line design basis and, therefore, no additional restraint of the main steam lines is necessary. Any displacements and stresses resulting from this loading are expected to be small and well within allowable. The Power Uprate Generic Evaluations (NEDC-31984P, Section 4.7)- also concluded that at the power uprate conditions, the sudden closure of the TSV will not have any significant effect on main steam piping system stress and displacement including supports load. Therefore, it is concluded for power uprate implementation at BFN, the loads on the main E-44

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steam lines following the TSV closure are within the current allowable design basis.

Regarding the impact on MSIV due to the increase in the hydraulic pressure at higher flow/pressure, the GE report quoted in the Request (NEDC-32523P, Supplement 1, Revision 1) is the Generic Evaluations for Extended Power Uprate. For power uprate application, the corresponding report is NEDC-31984P, Volume 1.

The impact of increase in steam flow rate on the safety function of the MSIV was addressed on a generic basis in Section 4.7.2 of NEDC-31894P. Structurally, the MSIVs are designed to withstand the closure impact from 200% of the pre-uprate steam flow, or 190.5. of the uprated steam flow. This upper flow value is based on the size of the steam flow limiting venturi which is not being changed. Higher operating pressure at power uprate conditions tends to result in higher initial seating pressure across the MSIV after isolation, thus reducing leakage. Therefore, MSIV leakage will be slightly lower for power uprate conditions.

Furthermore, MSIVs are under close scrutiny for leakage and closure time from various surveillance requirements in the plant technical specifications and their safety performance is routinely monitored. Therefore, it is concluded that the increase in the hydraulic pressure on the MSIV at power uprate conditions has no adverse impact on their performance.

NRC RE VEST 9 Please provide the evaluation of the feedwater heater for the power uprate with regard to vibration, stress and fatigue usage (Section 7.4.1).

TVA REPLY 9 The condensate and feedwater systems are non safety related Class II systems with associated piping designed in accordance with USAS B31.1.0-1967. The boundaries of the systems extend from the condenser hotwell to the outboard containment isolation valves downstream of the highest pressure feedwater heaters. Fatigue analysis only applies to Class I piping systems and was not performed for these Class II systems.

For the non safety related feedwater system, evaluations were performed to determine whether this system has adequate capacity at uprate conditions. The BFN feedwater heaters were evaluated at power uprate conditions using the currently available design E-45

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( information to ensure that adequate margin was retained for shell overpressure protection, and that heater tube and shell-side conditions remained within the original design envelope.

The major effect of uprate on the feedwater heaters is increased velocities through the feedwater heater tubes and increased steam flow through the shell-side of the heater. The tube-side velocities were calculated at the power uprate flow condition and determined to be less than the Tubular Exchange Manufacturers Association (TEMA) standards maximum allowable of 10 feet-per-second. On the shell-side, all heaters showed an increase in piping pressures, flows and temperatures at the power uprate condition. Based on these conditions and currently available design information, the feedwater heater piping systems were evaluated and found to be within the original design envelope for these systems. The shell side flows for all feedwater heaters do not exceed the original design specifications. The potential for increased tube vibration at uprate was assessed and found to be insignificant (less than 5 Hz for all heaters) compared to the present normal operating conditions. The shell-side inlet and outlet nozzle mass flow velocity squared terms were also evaluated and determined to meet requirements of the TEMA standards for flow induced loads. The shell side pressure increases retain ample margin below the design pressure and since the design pressures remained the same, no relief valve set point changes are required.

Xn conclusion, based on the current condition of the feedwater heaters and the relatively insignificant increases resulting from power uprate implementation, there is no adverse impacts on this equipment beyond the normal wear and tear associated with heavy usage at rated capacity.

REFERENCES:

TVA letter to NRC dated October 1, 1997, Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 - Technical Specification (TS) Change TS-384 - Request For License Amendment for Power Uprate Operation

2. TVA letter to NRC dated March 16, 1998, Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 Technical Specification (TS) No.

384 Supplement 1 Request for License Amendment for Power Uprate Operation

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TVA letter to NRC dated March 20, 1998, Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 - Technical Specification (TS)

Change TS-384 Request for License Amendment for Power Uprate Operation NRC letter to TVA dated April 22, 1998, Browns Ferry Plant Units 2, and 3 - Request for Additional Information Regarding Technical Specification Change TS-384 Request for License Amendment for Power Uprate Operation (TAC Nos.

M99711 and M99712)

Babcock & Wilcox Company, "Certified Design Document for TVA I and II," December 1970 (GE VPF No. 1805-189-000 revision

1) .

Babcock & Wilcox Company, "Certified Design Document for TVA III," February 1972 (GE VPF No. 1974-094-000 revision 1).

22A5594, Rev. 2, "Feedwater Nozzle Stress Report," San Jose, CA, June 1979.

Nuclear Services Corporation, NSC 1-78-006, "Stress Analysis of Feedwater Nozzle and Safe End for Browns Ferry Unit 2 Power Plant," March 1978 (GE Document No. 299X126-018, Rev.

0) .

23A4303, Revision 0, "Reactor Vessel - Recirculation Inlet Nozzle, Stress Report," San Jose, CA, February 1985.

23A4306, Revision 0, "Reactor Vessel - Recirculation Outlet Nozzle, Stress Report," San Jose, CA, February 1985.

CBI Nuclear Company, Contract 7-CN699, "Stress Report Core Spray Nozzle Replacement Safe End Browns Ferry 1, 2 & 3,"

(GE VPF No. 6033-003-000 revision 1).

American Society of Mechanical Engineers, "Rules for Construction of Nuclear Vessels," ASME Boiler and Pressure Vessel Code,Section III, 1965 Edition with Addenda to and including Summer 1965 (for Unit 2).

American Society of Mechanical Engineers, "Rules for Construction of Nuclear Vessels," ASME Boiler and Pressure Vessel Code,Section III, 1965 Edition with Addenda to and including Summer 1966 (for Unit 3).

14. American Society of Mechanical Engineers, "Rules for Construction of Nuclear Vessels," ASME Boiler and Pressure Vessel Code,Section III, 1974 Edition with Addenda to and including Summer 1976.

15. 24A5890, Rev. 1, "Reactor Vessel -- Power Uprate, Certified Design Specification, Browns Ferry Unit 2," San Jose, CA, June 1997.

16. 24A5895, Rev. 1, Reactor Vessel -- Power Uprate, Certified Design Specification, Browns Ferry Unit 3," San Jose, CA, June 1997.

17. NED0-21888, "Mark I Containment Program Load Definition Report," Rev. 2, November 1981.

18. CEB-83-34 R2, "BFN Nuclear Plant Torus Integrity Long-Term Program Plant-Unique Analysis Report (PUAR)," (CEB 841210 008) .

19. NED0-24580, "Mark I Containment Program Plant Unique Load Definition - BFN Nuclear Plants 1, 2 and 3," Rev. 2, January 1982.

20. NEDE-21944P, "Mark I Containment Program Quarter Scale Plant Unique Tests," Vol. 1, April 1979.

21. NEDE-13456P, "Mark I 1/12 Scale Pressure Suppression Pool Swell Tests", March 1976.

22. NEDE-24539P, "Mark I Containment Program Full-Scale Test Program Final Report", April 1979

23. TVA letter to NRC, dated December 3, 1997, Browns Ferry Nuclear Plant (BFN) - Unit 2 - Docket No. 50-260 Facility Operating License DPR Licensee Event Report 50-260/97008

24. TVA letter to NRC, dated April 9, 1997, Browns Ferry Nuclear Plant (BFN) - Unit 3 Docket No. 50-296 - Facility Operating License DPR Licensee Event Report 50-296/97003 E-48

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ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 BROWNS FERRY NUCLEAR PLANT (BFN) RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) RELATING TO UNITS 2 AND 3 TECHNICAL SPECIFICATION (TS) CHANGE NO. TS-384 POWER UPRATE OPERATION (TAC NOS. M99711 AND M99712)

1. PIPING AND COMPONENT ANALYSIS The following questions are related to TVA's letter dated May 22, 1998 (Reference 8).

NRC Re est 1.a Please provide the Editions of ASME Code for evaluation of the control rod drive mechanism, reactor recirculation and residual heat removal piping.

TVA Re 1 1.a The code used for the original design of the control rod drive (CRD) mechanism is the ASME Boiler and Pressure Vessel Code, 1968 Edition, including Summer 1970 Addenda. For subsequent component replacements, the Code effective date is 1974 Edition, including Winter 1975 Edition.

The design code for the reactor recirculation and residual heat removal (RHR) piping is USAS B31.1.0, 1967 Edition. As part of the Long Term Torus Integrity Program (LTTIP),

applicable portions of the RHR system were evaluated using the ASME Boiler & Pressure Vessel Code, 1977 Edition, including Summer 1977 Addenda.

All power uprate evaluations of the CRD mechanism, reactor recirculation piping and RHR piping were performed consistent with the existing. design basis codes.

NRC Re est 1.b Table 1(c)-2 shows that the stress in the supporting skirt exceeds the Code allowable limits. The skirt support was acceptable by elastic-plastic analysis. Please provide a description of the elastic-plastic analysis, Code-allowable 0 limits and calculation results that demonstrate the reactor pressure vessel support skirt to be acceptable.

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TVA Re 1 1.b The following general procedure is used for calculating the power uprate cumulative fatigue usage factor (CUF) when simplified elastic-plastic analysis is required.

1. Multiply the scaling factor (SCF) for each stress cycle by the original primary plus secondary plus peak (P+Q+F) stress intensity values of the original governing stress report.

2. For each of the limiting stress cycle pairs used in the fatigue analysis of the original governing stress report, determine the absolute value of the difference of 0he power uprate P+Q+F stress intensities (calculated in Step 1).

This value is Spqgqp,new.

3. Determine the power uprate alternating stress intensity, S,l<,,, for each of the original limiting stress cycle pairs as follows:

Salt, new ( 1/2)

• Ko,now * (Ec/Ea)

• Sp+g+p,new where Ko, new = Simplified elastic-plastic factor 1 0I for Sn,new < 3Sm,new

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= 1. 0 + [ (1-n) /n (m-1) j * [ (Sn,n,w/3S,n,) 1], for 3Sm, new < Sn, now < 3mSm, now 1/ng for Sn,new > 3mSm,new (E,/E,) = Elastic modulus correction factor (described in Paragraphs NB-3228.3 and NB-3222.4 of Reference 14).

4. Use S,lt:,, as the value of the ordinate when entering the applicable design fatigue curve in Reference 15 to find the COrreSPOnding allOWable number Of CyCleS (Nl,n,w) fOr eaCh Of the limiting stress cycle pairs.

5. Calculate the power uprate incremental fatigue usage factor (Ul,,~ nl/Nl,n,w) for each of the limiting stress cycle pairs, where nl is the common number of design cycles for each pair.

6. Calculate the power uprate CUF (CUF = ZUl,,). Xf CUF <

1.0, the ASME Code limit is met.

7. Satisfy special stress limits of Paragraph NB-3227 of Reference 14, if applicable.

The following is a summary of the support skirt fatigue and stress calculations.

1. There were two transients considered as significant for this evaluation. These were Heatup/Cooldown (H/C) and Loss of Feedwater Pump (LFWP). The values of S for power uprate conditions of these two transients were 99.2 ksi and 115.9 ksi respectively.

2. The K, Sp,q,~, and S, for these two transients were 1.953, 162.9 ksi, and 159 ksi for H/C and 2.787, 182.7 ksi, and 254.6 ksi for LFWP, respectively. The 3S value was 80.1 ksi and E,/E factor was 1.03. These values resulted in allowable number of cycles of 171 for H/C and 53 for LFWP based on Reference 15 curves. Since the actual cycles are 122 for H/C and 10 for LFWP, the resulting usage factor is 0.713 and 0.189 for H/C and LFWP respectively. Therefore, the CUF value is 0.902 (0.713 + 0.189). Thus, ASME code limits are met.

3. Since the support skirt does not have membrane stress due to pressure, thermal ratcheting requirements are not applicable.

4. For the support skirt, the minimum yield is 50 ksi with a corresponding minimum ultimate strength of 80 ksi. Thus, the calculated ratio of the material specified yield strength to the minimum specified ultimate strength is 0.63, which is within the Code requirement of 0.80.

5. The maximum analysis temperature for the skirt material is 546'F for the transients noted. This is below the maximum temperature in the table in Paragraph NB-3228.3 (i.e., 700'F for carbon steel).

6. The values of S excluding thermal bending stresses were 35 ksi (H/C) and 46 ksi (LFWP). Therefore, the S values were below 3S (80.1 ksi) for both transients.

NRC Re est 1.c Table l(c)-2 also indicates that the fatigue usage factors (CUFs) for the feedwater nozzles are 1.0 and 0.984 for the current rated power and the proposed uprated power conditions, respectively. Please provide description of how these two CUFs were calculated including the location and all transients which were considered in the CUF calculation.

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TVA Re 1 1.c The contributions to fatigue usage in 0he feedwater nozzle comes from two sources: (i) system cycling and (ii) rapid cycling. The methodology for the calculation of the CUFs is described here.

System Cycling: This is the low cycle fatigue usage from the thermal and pressure cycling associated with plant operation (steady-state operation, transients, scram, start-up and cooldown). System cycling typically constitutes the largest component of the CUF. Because there is a small increase in the pressure and temperature during power uprate, a small increase occurs in the system cycling component of the calculated CUF. The system cycling CUF is determined using ASME Section III procedures.

ii. Rapid Cycling: The BFN triple thermal sleeve sparger design has seal surfaces where there could be potential leakage of the feedwater. This potential leakage can mix with the hotter downcomer flow and cause temperature cycling which in turn causes high cycle fatigue usage.

The relationship of temperature ranges to the leakage rate was determined experimentally as part of the triple sleeve feedwater design qualification. The leakage rate was estimated using a conservatively high corrosion rate for the seal surfaces. The corresponding temperature cycles were obtained from the experimental data and the associated CUF can be determined from the fatigue curve (including high cycle fa0igue). Power uprate, coast down at the end of a fuel cycle and final feedwater temperature reduction (FFWTR) impact the calculated CUF due to rapid cycling. The other factor that significantly impacts rapid cycling CUF is the assumed seal refurbishment interval. As the seal refurbishment interval increases, the analytical model used in the GE calculation methodology predicts an increasing rapid cycling CUF. However, based on the successful field experience with the triple sleeve sparger design as well as increased confidence in the NUREG-0619 inspections,

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continued operation with the existing triple sleeve spargers without seal refurbishment is acceptable.

The feedwater nozzle values shown in Table 1(c)-2 of Reference 8 for uprated condii ions include all areas except for the Inner Blend Radius. CUF values have been calculated for both pre-uprate and uprate operation for the inner blend radius and other areas adjacent to the sparger seals. Using a

conservative approach, a 30 year seal refurbishment interval was assumed.

The calculated CUF for the Inner Blend radius exceeds 1 for both the pre-uprate and uprated conditions. Although the CUF is greater than 1.0, the feedwater nozzle is still acceptable for power uprate operation due to the following reasons.

The increase in the calculated fatigue usage is driven by very conservative assumptions involving the leakage past the feedwater sparger seals. This conservatism is demonstrated by 15 years of actual field experience with the triple sleeve sparger design in over 30 boiling water reactor (BWR) units, where there have been no reported incidents of nozzle fatigue cracking. Power uprate conditions will not increase the potential for seal leakage due to no change in the pressure differential across the seals. The increased CUF is driven primarily by increased temperatures which could result only if leakage occurs.

The BFN NUREG-0619 inspection and monitoring program is geared to detect seal leakage and crack indications which might result from the high cycle fatigue generated by potential seal leakage. Previous analysis demonstrates that the driving forces from high cycle fatigue greatly attenuates once an indication reaches a depth of 0.25 inches. Once an indication reaches this depth, structural analysis demonstrates that the time for the flaw to grow to a critical size is on the order of 30 years. TVA's current NUREG-0619 inspection program requires ultrasonic (UT) inspections with the capability to detect 0.25 inch flaws every other refueling cycle. In addition to performing these inspections, TVA has a leakage monitoring system which will provide an indication of increased seal leakage. Monitoring for leakage and conducting inspections at repeated intervals provides assurance that the initiation of conditions for high cycle fatigue will be monitored and that should indications be inii iated they will be detected long before they reach a size to be a concern.

In summary, the CUF has been calculated using very conservative assumptions relative to leakage and seal refurbishment. These conservative assumptions result in high CUF values for the Inner Blend Radius for both pre-uprate and uprated conditions. Experience at BFN and within the industry demonstrates that unacceptable seal leakage has not occurred.

The power uprate conditions have been evaluated and determined that there is no increase in the potential for increased seal leakage. The existing NUREG-0619 program is applicable for power uprate and addresses the CUF levels and, therefore, address concerns of crack initiation at the Inner Blend Radius

location. This program, which involves both t: he monitoring of leakage and inspection of the BFN feedwater nozzles, will continue to provide assurance that the integrity of the nozzles will be maintained for power uprate conditions.

NRC Re est 1.d On page E-21, you indicate that the power uprate temperature and pressure are bounded by the pre-uprate conditions used in the existing piping analysis. Therefore, the existing stress reports are bounding for the power uprate. This is inconsistent with Section 3.12.1, General Electric (GE),

Licensing Topical Report NEDC-32751P, "Power Uprate Safety Analysis for The Browns Ferry Nuclear Plant [BFNP], Units 2 And 3," dated September 1997 (Proprietary), which states that operation at the uprated conditions, would increase the piping and piping component stresses due to slightly higher operating temperature, pressure, and flow rates internal to the pipes.

Please provide the margin between the existing calculated stresses and the Code allowable limits for each line in Figure 3-4 and compare the margin to the stress increases in figures 3-4 and 3-5 based on specific increases in temperature, pressure and flow rate. Also please provide an evaluation of the piping systems attached to the torus shell with regard to the increase in the pool temperature at the power uprate condition.

TVA Re 1 1.d The correct reference is Tables 3-4 and 3-5 in the question rather than Figures 3-4 and 3-5 (Reference 1).

The parameters affecting the piping stress analyses are pressure, temperature and flow. At the power uprate condition, these parameters will increase compared to the pre-uprate conditions, and, thus, the resulting stresses on the piping system will also increase. This is the basis for the statement in Section 3.12.1 of NEDC-32751P (Enclosure 5 of Reference 1). For the uprate analyses, existing BFN design basis documents (e.g., design specifications and piping stress reports) were reviewed to determine t: he design and analytical basis for safety related piping systems. The uprated pressure, temperature and flows were compared with t:he existing analytical basis to det:ermine the impact due t:o power uprate conditions.

As shown in Tables 3-4 and 3-5 of NEDC-32751P, the power uprate operating condition pressure and temperature remain bounded by the pressure and temperature assumed in BFN piping stress analyses. Therefore, even though the'xisting

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the stresses increase due to uprate, the existing piping stress analyses are bounding for the power uprate application and no new stress analyses were performed for the power uprate condition. Since power uprate does not affect the existing piping stress analyses, there is no reduction in the stress margin due to power uprate. One exception is the feedwater piping power uprate operati~ng temperature is approximately 0.8% greater than the temperature used in the feedwater piping stress analyses. This small increase is considered negligible and based upon our evaluation does not have any adverse impact on the feedwater piping structural integrity.

Tables 3-4 and 3-5 of NEDC-32751P also show that the main steam flow will increase by about 8% at the power uprate condition. This is the combination of 5.8% steam flow increase associated with 5% core thermal power uprate and the analytical requirement to assume 102% power for initial condition. An analysis was previously performed for BFN titled "Shock Loading of Main Steam and Turbine Bypass Piping, Browns Ferry Nuclear Power Plant." This analysis concluded that the load on the BFN main steam line due to shock loading resulting from sudden closure of the turbine stop valve (TSV) is negligible and within the main steam line design basis and, therefore, no additional restraint of the main steam lines is necessary. The Power Uprate Generic Evaluations (NEDC-31984P, Section 4.7) also concluded that at the power uprate conditions, the loads on the main steam lines following the TSV closure remain within the current allowable design basis.

The piping systems attached to the torus shell are currently designed for a peak suppression pool temperature limit of 177'F. For the power uprate condition, an analysis and Technical Specification change were proposed as part of the October 1, 1997, power uprate Technical Specification Change TS-384 to maintain the same suppression pool temperature limit of 177'F. Specifically, a Technical Specification change (ITS 3.7.1) was proposed to lower the RHR service water (RHRSW) temperature limit from 95'F at pre-uprate condition to 92'F at power uprate condition (See Reference 1, Section 4.1.1.1(a)).

With no change to the peak suppression pool temperature for the power uprate condition, there is no adverse impact to the integrity of the piping systems attached to the torus.

NRC Re est 1.e In your response 6, you did not address issues relating to pressure locking and thermal binding of valves (Generic Letter (GL) 95-07 issues). Please provide an evaluation of the power uprate effects on the potential pressure locking and thermal binding of safety-related power-operated valves. Also discuss

the potential for over-pressurization of isolated water-filled piping sections per GL 96-06.

TVA Re 1 l.e The BFN evaluations previously performed in response to GL 95-07 (Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves) were reviewed for impact by power uprate. The review determined that valves evaluated in response to GL 95-07 are unaffected by the power uprate conditions.

The evaluations previously performed identified four valves (2 RHR flow control valves and 2 core spray flow control valves) potentially susceptible to pressure locking. These valves were previously modified by providing a hole in the downstream face of each valve disc. These previous evaluations also identified the potential thermal binding of the high pressure coolant injection (HPCI) system steam isolation flow control valve (FCV) 73-16. This situation was applicable to both units.

A modification was accomplished to the Unit 2 FCV-73-16 valve last outage by replacing the existing solid wedge gate valve with a double disc type gate valve. The valve's closed position is based on limit switch position and the valve is mounted with the operator vertically up. The operating conditions for this valve do not induce pressure locking (pressure upstream of valve is always present whenever valve is required to function). In addition, the only condition which causes an external temperature increase is a HPCI steam line break, under which HPCI system would not be required to function. Therefore, the bonnet pressure will always be equal to the upstream line pressure and this valve will not be subject to pressure locking.

The potential for thermal binding on the existing solid wedge Unit 3 HPCI valve has been addressed by procedural controls which require cycling of the HPCI valve to alleviate this potential situation. This valve is currently planned to be replaced with a double disc gate valve during the refueling outage in the fall of 1998.

The evaluation determined that there are no resulting changes in high/low pressure interfaces and no physical relocation is required for these valves. No modifications are required to the safety function of these valves. In addition, power uprate evaluations did not identify any new locations where elevated temperatures would adversely impact the performance of safety-related power-operated valves. Therefore, the

pre-uprate evaluations which identified valve locations potentially susceptible to pressure locking or thermal binding remain valid for power uprate operation.

An evaluation of 0he primary containment penetrations for thermal pressurization per GL 96-06 (Assurance of Equipment Operability and Containment Integrity During Design-Basis Conditions) was performed for uprated conditions. The evaluation assessed the mechanical piping systems that penetrate the primary containment to confirm that postulated thermal expansion of trapped water such that over-pressurization of pressure boundary piping would not occur.

In hot locations, the penetrations not susceptible to thermal pressurization at pre-uprate conditions were reevaluated for power uprate and found acceptable in their current configurations. For those penetrations that were determined to be susceptible to thermal pressurization at pre-uprated conditions, the proposed and previously implemented actions were reviewed for impact by power uprate. The review determined that these actions were acceptable for power uprate operation. Additionally, the review determined that no hot locations were created whereby new penetrations were identified as being potentially susceptible to over-pressurization due to power uprate conditions.

NRC Re est 1.f In its GL 89-10 inspection at Browns Ferry, April 27 to May 1, 1998, the U.S. Nuclear Regulatory Commission staff determined that you had not updated the motor operated valve (MOV) calculations for Unit 2 to reflect the power uprate conditions. Please provide a schedule for revising these calculations (e.g., MOV, and other valve and pump calculations) to reflect the power uprate conditions for Browns Ferry Unit 2 and identify any expected adjustment or modifications. Please confirm that all required modifications will be accomplished prior to the implementation of the proposed power uprate for Unit 2 or 3.

TVA Re 1 1.f The BFN Unit 2 GL 89-10 and other pump/valve calculation revisions are scheduled to be completed by August 31, 1998, which is prior to Unit 2 power uprate implementation planned for the Spring of 1999. The Unit 2 and 3 calculations that support required modifications due to power uprate have been completed. Evaluations have determined that no additional modifications in addition to those discussed in Item 1.g below

will be required on Unit 2 to support power uprate. The calculation revisions performed for power uprate included new design parameters to resolve GL 89-10 issues. The Unit 2 MOV upgrades resulting from implementation of power uprate involve resetting of 4 torque switches. Required MOV modifications will be accomplished prior to the implementation of the proposed power uprate for Units 2 and 3.

NRC Re est 1.

Please clarify whether any MOV modifications (in addition to the torque switch adjustments for the four GL 89-10 MOVs) are planned for the Unit 3 power uprate. Also, indicate if any other power-operated valves (such as air-operated valves or hydraulic-operated valves) were adjusted or modified based on the power uprate conditions.

TVA Re 1 1.

Unit 3 MOV modifications required by power uprate implementation are provided below.

3-FCV-73-02 reset torque switch 3-FCV-73-35 reset torque switch 3-FCV-74-30 replace spring pack and reset torque switch The number of Unit 3 valves requiring modification have been decreased from four as stated in the earlier response in Reference 7, TVA Reply B.6. Further evaluation of valve 3-FCV-73-81 has indicated that the current torque switch setting is adequate and, does not require resetting of the torque switch.

Unit 2 MOV modifications required by power uprate implementation are provided below.

2-FCV-71-34 reset torque switch 2-FCV-73-30 reset torque switch 2-FCV-73-35 reset torque switch 2-FCV-74-30 reset torque switch Power-operated valves (such as air operated valves or hydraulic-operated valves) were reviewed for impact by power uprate. It was concluded that no adjustments or modifications are required for power uprate implementation.

NRC Re est 1.h Please discuss the post-accident containment temperature increase (from 322'F to 336'F) as a result of the "GOTHIC"

analysis (your letter of March 16, 1998) and its effects on MOV output (GL 89-10 issue), pressure locking and thermal binding (GL 95-07 issue) and potential over-pressurization of isolated water-filled piping sections (GL 96-06 issue).

TVA Re 1 1.h This 336'F post accident drywell temperature is calculated using the SHEX code rather than GOTHIC. The maximum temperature has been increased 14'F from 322'F to 336'F for the power uprate condition.

The GL 89-10 MOV calculations utilized the environmental temperatures associated with power uprate conditions. Valves that are located inside of primary containment and required to operate to mitigate a high energy line break (HELB) (i.e.,

loss of coolant accident (LOCA) or main steam line break (MSLB)) inside primary containment were evaluated using the peak power uprate temperature of 336'F. Other valves located inside primary containment (i.e., valves required to mitigate HELBs outside containment) were evaluated at the temperature associated with their design basis operating condition for power uprate.

The 14'F temperature rise in the primary containment environment could affect the motor operator capability of the MOVs located inside the primary containment for those required to operate during a design basis event inside containment.

The capability of each of these motor operators were evaluated at the appropriate temperature associated with the particular design basis event which required the valve to function and appropriate calculation revisions were performed. The evaluation found the motor operator capability to be acceptable.

Power uprate has no adverse impact on the previous BFN evaluations in response to GL 96-06 and GL 95-07. Please refer to the response to TVA Reply 1.e.

ADDITIONAL REQUEST Response to the additional related request noted during the July 9, 1998, meeting is provided below.

NRC Re est 1.i MSRV setpoints increase by 30 psi. How does this compare to the original design values?

TVA Re 1 1.i The main steam relief valve (MSRV) setpoints will be increased by 30 psi corresponding to the reactor dome pressure increase.

Additionally, the tolerance of the valves has been increased to +/- 3% via Technical Specification change TS-386 approved by the NRC on May 18, 1998. The MSRV loads for the proposed power uprate were evaluated at a pressure of 1189.7 psig which is the highest MSRV setpoint pressure for the proposed power uprate (1155 psig) plus a 3% setpoint drift tolerance. Both single and multiple MSRV actuations were evaluated. This compares to the existing design basis for MSRV loads which is at a pressure of 1158.8 psig which is the highest MSRV setpoint pressure (1125 psig) plus a 3% setpoint drift tolerance.

2. SPENT FUEL POOLS NRC Re est 2

a. Please provide the heat load and corresponding peak calculated spent fuel pool (SFP) temperature for both planned and unplanned full core offloads at the current power level and the proposed power uprate level and confirm whether these heat loads and corresponding SFP temperatures include a single failure of SFP cooling (e.g., one of two trains of SFP cooling).

b. Your May 20, 1998 letter states that, no specific calculations were made for the peak SFP normal operation and unplanned full core offloads and that the design basis for the SFP cooling system remains the same for the pre- and post-power uprate conditions. Zf no calculations were performed for the proposed power uprate level, please discuss your basis for assuring that both the heat load and the peak SFP temperature would not increase for the proposed power uprate conditions.

TVARe 1 2 The following fuel pool cooling (FPC) system information is consistent with the 10 CFR 50.59 approved part of the licensing bases. The information is being processed as part of the next scheduled Updated Final Safety Analysis Report (UFSAR) Amendment 17 and will be transmitted to the NRC as part, of that amendment.

Refueling operations at BFN are conducted such that the heat load that will be placed in the fuel pool will not exceed the available cooling capacity. The margins for maintenance of

pool heat load will not be reduced by power uprate parameters.

The capacity of the fuel pool cooling and supplemental decay heat removal systems, considering seasonal cooling water temperatures and current heat exchanger conditions, is utilized to maintain the fuel pool temperature at or below 125'F during normal refueling outages (average spent fuel batch discharged from the equilibrium fuel cycle). The RHR system can be operated in parallel with the fuel pool cooling system (supplemental fuel pool cooling) to maintain the fuel pool temperature less than 150'F if a full core offload is performed. The determination of heat load from the fuel is specific to each outage since the number of fuel assemblies placed in the pool and the time after shutdown for placement in the pool is controlled and determined on a specific outage by outage basis. This determination does not assume a specific single failure and, therefore, a single bounding value of heat load cannot be specifically provided. Amendment 17 to the UFSAR clarifies the basis for fuel pool heat load and provides heat removal capabilities of the systems that support fuel pool cooling.

Although not safety related, TVA has installed an Alternate Decay Heat Removal (ADHR) system whose cooling capacity is capable of cooling the entire pool and core approximately 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown. This provides added assurance of BFN's ability to maintain fuel pool and core cooling during refueling operations.

It is understood that power uprate conditions refueling would slightly compared to increase fuel pool heat load during pre-uprate assuming other parameters remain constant.

However, the method described above ensures that peak allowable pool temperatures will not be exceeded by factoring actual conditions (including uprate) into the calculated expected decay heat load. If uprate heat load becomes limiting, other parameters, such as time after shutdown to core offload, would be adjusted to decrease heat load to acceptable values.

The following table provides parameters associated with the fuel pool cooling and RHR supplemental fuel pool cooling systems.

El-13

Table 2-1 FUEL POOL COOLING SYSTEM PAEMMETERS SYSTEM FUNCTION SYSTEM PARAMETER TOTAL POOL I WELL/ AND PIT VOLUME 106, 881 cu ft FUEL STORAGE POOL VOLUME 51,340 cu ft SYSTEM DESIGN FLOW (1 PUMP) 600 gpm MAXIMUM FLOW (2 PUMPS) 1,200 gpm PUMP CHARACTERISTICS (1 PUMP) 600 gpm, 330 ft TDH FPC HEAT EXCHANGER (1) DESIGN CAPACITY Fuel Pool Flow: 600gpm; Temp. 125'F / RBCCÃ Flow: 4.4 x 10 750 gpm; Temp. 100'F Btu/hr/Hx.

RHR HEAT EXCHANGER (1) DESIGN CAPACITY Fuel Pool Flow: 800gpm; Temp. 150'F / RHRSN Flow: 19.3 x 10'tu/hr/Hx.

4500 gpm; Temp. 95'F Fuel Pool Flow: 2000gpm; Temp. 150'F / RHRSV Flow: 34.0 x 10 Btu/hr/Hx.

4500 gpm; Temp. 95'F

3. REACTOR SYSTEMS NRC Re est 3.a NEDC-32751P, Section 4.3 ECCS performance evaluation: please clarify the following statement: "The SAFER/GESTR code is the pre-uprate analysis for BFN, and therefore an update from a previous analysis ECCS [emergency core cooling system] Code is not a part of the uprate license amendment." Our understanding is that Tennessee Valley Authority (TVA) completed the loss-of-coolant-accident (LOCA) analysis for uprate conditions in 1996, as given in NEDC-32484P. Please confirm that power uprate will not change the limiting break, single failure, or the break spectrum as compared to the existing analysis.

E1-14

TVA Re 1 3.a The SAFER/GESTR-LOCA analysis for BFN as documented in NEDC-32484P Revision 2 is based on core thermal power of 3458 MWt, or 105% of the pre-uprate licensing basis condition of 3293 MWt. The analysis was performed in this manner in anticipation of power uprate conditions. Prior to power uprate, the NRC was notified of this change in methodology via a TVA letter dated February 23, 1996 (Reference 16).

Therefore, the limiting break, single failure and break spectrum assumed for the SAFER/GESTR-LOCA analysis as shown in NEDC-32484P are applicable for the power uprate condition and will not change for the power uprate implementation.

The NEDC-32484P Revision 1 document was attached to the February 23, 1996 let:ter. The February 23, 1996 Enclosure 2 cover sheet erroneously stated that t: he document was at Revision 2 when it was actually at Revision 1. Xn December 1997, the NEDC-32484P document was revised to Revision 2 to include a 10'F adder for the bottom head drain flow path.

Revision 2 is provided in Enclosure 5 for your information.

The NEDC-32484P Revision 2 analysis is the basis for t: he power uprate condition.

NRC Re est 3.b Please provide a baseline comparison run using SAFER/GESTR of pre- and post-power uprate conditions.

TVA Re 1 3.b The SAFER/GESTR-LOCA baseline analysis for the pre-uprate condition was performed for the BP/PSxSR and GE11 fuel for the limiting large break using the Nominal and Appendix K assumptions. As shown by NEDC-32484P, Revision 2 Table 5-1, the GESxSNB fuel has a much lower peak clad temperature (PCT) than the other fuel types and was not included in the pre-uprate baseline analysis. The pre-uprate analysis assumed the same analytical methods and input data as for the power uprate condition, except for the input data that depends on initial power and flow assumptions.

The results of the analysis for design basis accident (DBA)

LOCA performance at the pre-uprate power condition are shown in Table 3.b-l. The existing licensing basis PCT for the power uprate condition is based on the BP/PSxSR fuel at 1590'F. At the pre-uprate condition, the comparable licensing basis PCT is 1580'F for the BP/PSx8R fuel type. The licensing basis PCT is based on the calculated Appendix K PCT with an adder to account for uncertainties. For the GE11 fuel, the

licensing basis PCT at the pre-uprate and power uprate condition are 1550'F and 1580'F, respectively. Table 3.b-1 compares the Nominal, Appendix K and Licensing Basis PCT at pre-uprate and uprated power conditions.

The plots of the system response are shown as follows:

~ Figures 3.b-la through g: BP/PSxSR and GE11, Pre-Uprate, Appendix K

~ Figures 3.b-2a through g: BP/PSxSR and GE11, Pre-Uprate, Nominal

~ Figures 3.b-3a through g: BP/PSxSR and GE11, Uprate, Appendix K

~ Figures 3.b-4a through g: BP/PSxSR and GE11, Uprate, Nominal The performance of GE13 fuel design which will be part of the BFN power uprate core design was also included in the power uprate ECCS-LOCA evaluation. The Appendix K PCT for the limiting large break for GE13 fuel is 2'F higher than that for GE11 fuel (1554'F versus 1552'F), but is less than the Appendix K PCT for the BP/PSxSR fuel and would not increase the Licensing Basis PCT (which is based on BP/PSxSR fuel).

For the limiting large break, the Nominal PCT for GE13 fuel is 12'F lower than the GE11 fuel (1045'F versus 1057'F) .

E1-16

Table 3.b-1 ECCS Performance Results for BP/PSxSR and GE11 Fuel Types at Pre-Uprate and Uprated Power (DBA Suction Line Break with Battery Failure)

Parameter Pre-Uprate Power Uprate Licensing Basis PCT, 'F BP/PSxSR 1580 1590 GE11 1550 1580 Appendix K PCT, 'F BP/PSxSR 1565 1572 GE11 1525 1552 Nominal PCT, 'F BP/PSxSR 1053 1068 GE11 1052 1057 E1-17

BR FERRY LOT ~L AVERAGE CHANEL TOP OF ACTIVE F 60.

00.

20.

UJ Ld I

0.

0. 60. l20. ISO. 2'I 0.

RAY Ol282 070lel i%12.2 TINE (SECOND)

Figure 3.b-1 a BP/PaxsR Fuel at Pre-uprate Power - Appendix K (DBA Suction - Battery Failure CS + 2LPCI + ADS Available)-

Water Level in Hot and Average Channels

BR FERRY VESSEL PRESQNE 1.2

• 10'.8

0. 'I uJ CL CA CA uJ Q

CL 0.

0. 60. 120. 180. 2'lO.

Ol 282 RAY 070197 11%.2 T IHE (SECOND)

Figure 3.b-1 b BP/P8x8R Fuel at Pre-uprate Power - Appendix K (DBA Suction - Battery Failure CS+ 2LPCI+ ADS Available)-

Reactor Vessel Pressure

0 FERRY IPCI

• 10'R IPCS Nl

~ LPCS 02

~ I LPCI IN LOOP II 5 o

~ 2.LPQUKlQ~

2.

LLI CA CQ I.

UJ I

CL C3 LI 0.

0. 60.' 120 !80. 210.

RAY OI282 070 I Sl I 1%.2 iEECQNDS 'INE Figure 3.b-l c BP/PaxSR Fuel at Pre-uprate Power - Appendix K (DBA Suction Battery Failure CS + 2LPCI+ ADS Available)-

ECCS Flow

BR FERRY PEAK CLAO TE PER TNE 2.

LLI l.

CL LLI a

LLI I

Q.

0. 60. 120. 180. 2'i 0.

RAV OI282 070 IST I%2.2 TIME (SECOND)

Figure 3.b-1d BP/PsxSR Fuel at Pre-uprate Power - Appendix K (DBA Suction - Battery Failure - CS+ 2LPCI+ ADS Available)-

Peak Cladding Temperature

BR FERRY CONY HTC HOT NS C(NV H?C HOI f6 RAD HTC HOT NS RAD HIC HOI P6 QIH B S. ~ IOihL HIC HOI 06 3 0 Lj CO I

CU b

I Cr.'Q S ~ ] g 5 I

~ I I.

C9 D

I

-I

0. 60. 120 ~ ISO. 260.

RAY OI 282 070l91 !ll2.2 T IHE (BECONO)

Figure 3.b-1e BP/P8xSR Fuel at Pre-uprate Power - Appendix K (DBA Suction - Battery Failure - CS+ 2LPCI+ ADS Available)-

Heat Transfer Coefficient

8R FERRY PEAK CLAO TEMPER llHE 5.

II 0'.

0.

0. 00. 80. 120. 360.

AAV 01295 OQNl lOR.1 TINE (SECOND)

- Failure - CS + 2LPCI+ ADS Available)-

Figure 3.b-1 f GE11 Fuel at Pre-uprate Power - Appendix K (DBA Suction Battery Peak Cladding Temperature

0 BR FERRY COW HTC HOT NS C0NV H1C HOT N6 RhD HTC HOT u5 RAD H1C HO1 t6 S.

~

BI LB%.tQI 4 TOTN, HIC HO1 N6

~ ~

5 5

CD ~

C)

I

-l

0. IO. 80. 120. 160.

INV OI 295 062IIIl I052A TINE (SECOND)

Figure 3.b-1 g GH11 Fuel at Pre-uprate Power - Appendix K (DBA Suction- Battery Failure - CS+ 2LPCI+ ADS Available)-

Heat Transfer Coefficient

'g BR FERRY HOT CHAPEL AVERAGE CNNKL TOP OF ACTIVE F .L 60.

t0.

)

u.l 20.

CL LLJ I

0.

0. 60. I20. I80. 210.

RhY 008!2 010lQl 14E2.9 TINE (SECOND)

Figure 3.b-2a BP/P8x8R Fuel at Pre-uprate Power - Nominal (DBA Suction - Battery Failure - CS+ 2LPCI+ ADS Available)-

Water Level in Hot and Average Channels

BR FERRY VESSEL PRESRHE 1.2

• 10 0.8 U)

Q

0. 'I I.> J rY.

CA cn uJ Q

RAY 070L91 0.

009 l2 0.

1~.9 80.

T IHE (SECDND) 120. 180.i 0.

Figure 3.b-2b BP/PSxSR Fuel at Pre-uprate Power - Nominal (DBA Suction - Battery Failure - CS+ 2LPCI+ ADS Available)-

Reactor Vessel Pressure

0 BR FERRY O'CI LPCS Nl LPCS N2 I LPCI IH LSP 0 S

II p 2.

CA CQ 1.

LLJ I

CL C)

Q.

0~ 60. I20. 180. 210.

008 I 2 RAY 070l91 l1%.9 -TINE (SECONCS)

Figure 3.b-2c BP/P8x8R Fuel at Pre-uprate Power - Nominal (DBA Suction Battery Failure CS + 2LPCI+ ADS Available)-

ECCS Flow.~

BR FERRY PEW CLAD TEtPER TLNE 2 ~

l.

CL Z:

QJ 0

0. 60. 120. 180. 210.

RAY 00812 070197 1M.9 TIME (SECONL)

F/gure 3.b-2d BP/P8x8R Fuel at Pre-uprate Power - Nominal (DBA Suction - Battery Failure - CS + 2LPCI + ADS Available)-

Peak Cladding Temperature

BR FERRY COW HTC HOT AS CNN HTC HOT N6 RhD HTC HOT PS

~ RhD HTC HOT N6 s T TOThL HTC HOT t6 5.

b C)

I Ol I

b I

CL I

CQ 1.

C9 C)

I

0. 60. 120. !80. 260.

RAY 008!2 010197 11%.9 TINE (SECOND)

Figure 3.b-2e BPIP8xSR Fuel at Pre-uprate Power - Nominal (DBA Suction - Battery Failure - CS+ 2LPCI+ ADS Available)-

Heat Transfer Coef5cient

BR FERRY PEhK CLhD TE PER TNE 5.

• lo'.

LLI l.

I CK LLI 0

LLI I

0.

0. 80. 80. 120. 160.

RhV 00%5 08691 IOR.9 TINE (SECOND)

- Failure- CS+ 2LPCI+ ADS Available)-

Figure 3.b 2t GE11 Fuel at Pre-uprate Power Nominal (DBA Suction- Battery Peak Cladding Temperature

BR FERRY CONY HTC HOT NS CONV HTC HOT N6 RAD IITC HOT AS RAD H1C HOl t6 S.

~s s IOThL HIC HOT II6 I

s C)

I

0. !Q. 80. t20. 160.

RAY 00%5 ma> Isa.II TIME (SECOND 3 Figure 3.b-2g. GE11 Fuel at Pre-uprate Power - Nominal (DBA Suction - Battery Failure - CS+ 2LPCI+ ADS Available)-

Heat Transfer CoeKcient

BR FERRY AVERAGE ~L IGT CHAOKL TOP CF ACTIVE F 60.

'l0.

20.

y 0.

0. 80. 160. 260. 520.

1QJ 00501 07l995 n51.0 TIME (SECONO)

Figure 3.b-3a Water Level in Hot and Average Channels - DBA Suction - Battery Failure-LPCS+2LPCI+ADS Available - Appendix K

BR FERRY VESSEL PRESQSE l,2 s)0 0 8 CA 0

0. 'I cn CJ) lsl CL 0.

0. 80. l60. 520.

CU 00501 Oll99S I I&.II Trina iSFCOVO)

Figure 3.b-3b Reactor Vessel Pressure - DBA Suction - Battery Failure-LPCS+2LPCI+ADS Available - Appendix K

BR FERRY I Pf 1 lPCS Nl l.PCS N2 l LPCI 1N LOOP n

5. ~~lZCLltLLKPM Il 0 2.

C3 Ld U)

K Q3 I

t~l l.

I IW~LI L I. II

0. 80. 160 ~ 2'IO. 520.

YOJ 00501 01N5 I 151.8 T IHE (SECONDS)

Figure 3.b-3c ECCS Flow- DBA Suction - Battery Failure-LPCS+2LPCI+ADS Available - Appendix K

BR FERRY PEAK CLAD TE PER 5.

IIO 2.

0. 80. t60. 210. 520.

IXI 05N On 995 Ill.5 T IHE (SECONO)

Figure 3.b-3d Peak Cladding Temperature (BP/P8x8R) - DBA Suction - Battery Failure-LPCS+2LPCI+ADS Available - Appendix K

CONY flIC l))f <<5 CNV fllC IGI <<6

~ Nt) NC IIOI <<S

~ IIAO IIIC IIOI <<6 QIhLHIG UGL5

~ 10IAL fllC IKII <<6 3.

I.s I

(U ll I

5I 51 ~ 5 51 5l ~ 5 I

CQ I.

~ 5

~l C)

I I

0. 80. I60. 2 IO. 320.

TINE (SECOND)

Figure 3.b-3e Heat Transfer Coefficient (BP/P8x8R) - DBA Suction - Battery Failure-LPCS+2LPCl+ADS Available - Appendix K

SR FERRY PEN CLAO TBPER tlSE 5.

• t 0',

(9 o

LLI lY uf 0

ul l-0.

0. 80. t60. 2'10. 520.

AO 00241 OI2196 le 51.1 TIME (SECOND)

Figure 3.b-3f Peak Cladding Temperature (GEI I) - DBA Suction - Battery Failure-LPCS t-2I.PCI+ADS Available - Appendix K

BR FERRY C0NV HIC fOT fI5 CONY llIC ffof N6 Re llTC if0T NS ffAD flfC ff0I l6

5. QM Hlf NI

~ IOlhL flIC lof 06 3.

cD I

Al I

4 I

CL

~s I

CQ l,

~ s

( I)

C3 s I

I

0. 80 ~ f60. 2'l0.

ao ooe)

OIR196 f151.1 TINE (SECOND)

Figure 3.b-3g Heat Transfer Coefficient (GE11) - DBA Suction - Battery Failure-LPCS+2LPCI+ADS Available - Appendix K

BR I ERRY lloT OIAML AVERAGE OlANNEL TOP OF ACTlYE Fl L 60.

10 ~

20.

LLj 0.

0. 80. 160. 210. 520.

val 0XE9 01 I 995 I 104.0 . T IHE (SECOND)

Figure 3.bda %'ater Level in Hot and Average Channels - DBA Suction - Battery Failure-LPCS + 2LPCI + ADS Available - Nominal

0 BR FERRY VESSEL PRESQNE l.2 II 0 0.8

0. I 0.

0. 80. l60. 2'IO. 520.

TINE (SECOND)

Figure 3.b4b Reactor Vessel Pressure - DBA Suction - Battery Failure-LPCS + 2LPCI + ADS Available - Nominal

BR FERRY upcl LPCS nl LPCS t2

~

l LPCI IN LOOP A'

~~LI~gP II 2.

CJ LLI CO CQ I .'a Q.

C3 lx

0. 80. 160. 2'IO. 320.

tQI OOCE9 1'INE 07 l99$ n03.0 (SECONDS)

Figure 3.b<c ECCS Flow - DBA Suction - Battery Failure-LPCS+ 2LPCI+ ADS Available - Nominal

BR FERRY PEhK O.N IEGER 1NE 5.

i1 0'.

C9 I.d C3 0.

0. 80. 160. 2'IO. 520.

TINE (SECONO)

Figure 3.bdd Peak Cladding Temperature (BP/PsxSR) - DBA Suction - Battery Failure-LPCS+ 2LPCI+ ADS Available - Nominal

BR FERRY CONV HIC IIOT NS CONV HIC IIOI II6 RAD HTC lIOT IIS

~

RAD lllC HOI II6 5

5. TOTAL HTC HOT t6 t ~ t t ~

3.

4 CO I

CU I

LL.

I CK t l5 ~

I CQ C)

I I

0. 80. 160. 2't 0. 320.

TINE (SECOND)

Figure 3.bde Heat Transfer Coefficient (BP/PsxSR) - DBA Suction - Battery Failure-LPCS + 2LPCI + ADS Available - Nominal I

BR FERRY,; PEAK CLAD TBPER TNE 5.

at 0 2,

LLj l.,

I Q

t.U 0

Ul I-0.

0. 80. 160. 210. 520.

$0 OOXE TIME (SECOND)

OI2%6 lsll. I Figure 3.be Peak Cladding Temperature (GEl1) - DBA Suction - Battery Failure-LPCS + 2LPCI + ADS Available - Nominal

BR FERRY CONV HTC HOT A CONV HIC 161 t6 RAD ITIC,.IGT NS RAD HIC l61 N6

5. 'GLY III'IQIM TOTAL HTC 161 IL'6 I 0 S 3 ~

4 CO I

CU tL.

I I ~

CL

~ 5 I

CQ 1.

CO C)

I I

0. 80. 160. 2'lO. 320.

RO OOXE 4I2%6 I 9 I I. I T IHE (SECOND)

Figure 3.bing Heat Transfer Coefficient (GE11) - DBA'Suction - Battery Failure-LPCS + 2LPCI + ADS Available - Nominal

NRC Re est 3.c NEDC-32751P, Section 6.5 SLCS: The pump discharge pressure is increased from 1275 psig to 1325 psig. Please discuss why the pump discharge relief valve setpoint is not changed and what the setpoint of the relief valve is now. If the new pump discharge pressure is close to 0he present setpoint, there may be inadvertent lifting of the relief valve.

TVA Re 1 3.c As stated in NEDC-32751P, Section 6.5, the change from 1275 psig to 1325 psig involves the surveillance test pressure rather than the expected pump discharge pressure at normal system operation. As discussed below, the pump discharge pressure under normal system operation will be 1279 psig.

At the pre-uprate condition, the standby liquid control system (SLCS) pump discharge relief valve minimum opening pressure is 1350 psig, corresponding to the nominal setpoint of 1425 psig minus 75 psi tolerance. At the power uprate condition, the SLCS pump discharge pressure during injection is the highest analytical limit pressure for the group of MSRVs with the lowest setpoint (i.e., lowest setpoint plus 3% tolerance),

plus 5 psi design specification margin (as discussed in Section 6.5 of Reference 1), plus the reactor water head, plus the system flow loss, and plus the system head loss.

~ MSRV opening setpoint pressure is calculated with an increase of 30 psi and an allowable setpoint tolerance of 3%. The result is an analytical limit for the lowest MSRV opening setpoint of 1169 psig (power uprate).

~ Design specification margin = 5 psig

~ Reactor water head = 15 psig

~ System flow loss = 115 psig

~ System elevation head = -25 psig Thus, at the power uprate condition, the SLCS pump discharge pressure during injection is 1279 psig. With a minimum SLCS pump discharge relief valve opening pressure of 1350 psig, the operational pressure margin between the minimum setpoint for the MSRV and the SLCS pump discharge pressure is 71 psi. Per the anticipated transient without scram (ATWS) rule requirement ("Anticipated Transient Without Scram, Response to NRC ATWS Rule, 10CFR50.62", NEDE-31096-P-A, February 1987), a minimum" pressure margin of 30 to 40 psi is considered necessary for the system to reliably provide full pump discharge flow to the reactor. The power uprate condition E1-46

still meetsdischarge SLCS pump this requirement and relief valve therefore, no change to the setpoint is necessary.

NRC Re est 3.d NEDC-32571P, Section 9.3.1 ATWS: Please confirm that the model ODYN was used for the plant-specific analyses.

TVA Re 1 3.d The ATWS analysis as documented in Section 9.3.1 of NEDC-32751P was performed using the REDY computer code which is the NRC approved code for ATWS event simulation (reference NEDC-31984P, Supplement 1 "Generic Evaluations of GE BWR Power Uprate Supplement 1", October 1991).

Note that the report page headers for Section 9 of NEDC-32751P, as submitted in Reference 1, incorrectly identify the document as NEDC-32571P (as stated in your question) instead of NEDC-32751P. See Reply S.b for corrections to NEDC-32751P.

NRC Re est 3.e Your May 20, 1998 letter, Item D.4, references licensed power.

Please confirm that the power conditions are the uprated conditions.

TVA Re l 3.e The term "licensed power" used in the response to Item D.4 in the May 20, 1998, letter (Reference 7) refers to the power uprate conditions. The rated power level assumed in the BFN power uprate fuel thermal limits transients analyses is 3458 MWt.

est 3.f V

NRC Re In your May 20, 1998 letter, you addressed the issues relating to Maine Yankee lessons learned and described different codes used in your power uprate evaluation. Your response does not indicate whether any third party or independent review of GE calculations was performed. Please discuss the process that was used to verify GE calculations are based on approved methodologies and consistent with all constraints.

TVA Re 1 3.f Many activities have been accomplished by TVA to ensure that the GE products (including computer codes and calculations)

E1-47

are based on approved methodologies and consistent with all appropriate constraints. These activities include steps to ensure that the GE products meet TVA technical and administrative expectations and procedural requirements.

The work format was defined by TVA and agreed to by GE.

Specifically, the work format includes the preparation and review of products by GE per the TVA issued procedures. The GE work was also subject to the TVA procedures that control the handling of issues identified as Conditions Adverse To Quality. Personnel training was also an integral part of the work format defined by TVA.

Each of the GE products is reviewed and comments resolved through the TVA owner of that product prior to issue into the TVA document process. The TVA review ensured that the products were compatible with the licensing bases of BFN and included the latest operating experience. The review scope included a verification of the proper application and approval of computer codes. Many products are being reviewed in detail by TVA prior to issue. These products are selected for this review based on the technical complexity of the associated activity.

A wide variety of additional reviews/assessments of the GE products have also been performed. Detailed technical reviews have been performed by TVA personnel throughout the power uprate design and implementation process activities. Early this year, a group of TVA technical reviewers and managers accomplished an independent assessment of the GE procedural control process and its associated implementation of power uprate products. This assessment also included products other than power uprate developed by GE for BFN. A recent assessme~t included a review by TVA Corporate Engineering to assess the products and processes for GE and the TVA portion of the project.

These reviews provide the needed and required assurance that the products and activities associated with the power uprate project meet the TVA expectations and requirements and are based on approved methodologies and consistent with all constraints.

NRC Re est 3.

Please identify new codes that were used in the power uprate analysis and confirm that they were used in accordance with any conditions associated with the use of these codes. These codes should also identified in the technical specifications bases.

As part of the BFN power uprate project, the computer codes used for the first time for BFN are the GE computer code SHEX, TRACG and the non-GE computer codes COSMO/M and GOTHIC.

The SHEX code is used for BWR Mark I long-term containment heat up analysis for drywell response, peak suppression pool temperature considering the limits due to containment design temperature, net positive suction head, and torus attached piping systems. The application of the SHEX code for BWR power uprate has been reviewed and approved by the NRC in Appendix G of NEDC-31897P-A "Generic Guidelines for General Electric Boiling Water Reactor Power Uprate", May 1992.

Consistent with NRC's request for first time application of computer code in a power uprate project, a baseline case using the SHEX computer code at the pre-uprate condition was performed and the results were documented in Section 4.1.1.1(a) of the October 1, 1997, submittal.

The TRACG (Transient Reactor Analysis Code-GE Version) computer code was used to derive the BFN power uprate acoustic and flow-induced loads on the reactor internals components following a postulated recirculation line break. This improved methodology to calculate the acoustic and flow-induced loads has been used in other recent GE BWR power uprate projects, such as Monticello Extended Power Rerate project and Hatch Extended Power Uprate project.

COSMO/M is a finite element computer program developed and maintained by Structural Research and Analysis Corporation.

This computer code has been used extensively in many other nuclear plants. For the TVA power uprate application, this computer code is used for finite element analysis and stress calculations. Prior to its plant-specific application, this computer code is also benchmarked against classical problems solutions to confirm its accuracy.

Temperatures and pressures in the reactor building that result from HELBs were determined using the GOTHIC version 5.0c computer code. The original Browns Ferry Equipment Qualification analyses were performed using MONSTER. The GOTHIC computer code was written by Numerical Applications Incorporated (NAI). For the Browns Ferry power uprate analyses, the GOTHIC options were set up to match the previous analyses that were performed using MONSTER to the extent possible. The only significant difference in the two codes as set up for these analyses is the buoyancy flow model in GOTHIC. The MONSTER reactor building model used for the equipment qualification analysis was converted to be

compatible with GOTHIC. The basic model and input assumptions were not changed during the model conversion. Changes were made only as necessary to be consistent with GOTHIC input requirements. These changes did not impact the results. The GOTHIC results were compared with the previous MONSTER analyses for BFN. The codes indicated excellent agreement where buoyancy effects did not dominate the junction flow between nodes. The results obtained for power uprate as described in this submittal are consistent with physical processes and the results from previous analyses. Similar GOTHIC analyses have been performed by others in support of power uprate programs. It is concluded that GOTHIC is a suitable replacement for MONSTER for these types of analyses.

The GOTHIC software program was transmitted to NRC through the Electric Power Research Institute Electric power Software Center in July of 1996.

ADDITIONAL REQUEST Response to the additional related request noted during the July 9, 1998, meeting is provided below.

NRC Re est 3.h Provide the basis for the reactor vessel overpressurization analysis with regards to initial steam dome pressure. NRC stated that the use of 1055 psig as an initial condition is a change of design/licensing basis. Original basis was the high dome pressure scram function and now is a steam dome pressure LCO with manual actions. May need a 10 CFR 50.92 specific to this issue for legal acceptability.

TVA Re 1 3.h The response to the request related to steam dome pressure is addressed in a supplement to the license amendment for TS-384 (Reference 13).

4. ELECTRICAL POWER AND AUXILIARY SYSTEMS The following questions are related to TVA's May 20, 1998, letter (Reference 7).

NRC Re est 4.a Although no hardware changes or modifications are needed for power uprate, the electrical power requirements for the condensate, condensate booster, and recirculation pumps are expected to increase. You response (Item B.2) did not quantify what increases in the electrical loads are required E1-50

for these pumps, and concludes that the pre-uprate electrical calculations would be valid for power uprate. Please provide the basis for your conclusion including any supporting analysis to show that the onsite electrical distribution system voltage is adequate to handle the increases in the electrical loads required by the power uprate or demonstrate that the electrical load increases for the above pumps are minimal. Also, submit the one line-diagrams (from load flow cases) which illustrate the load and voltage changes before and after power uprate cases under the worst expected grid voltage. As part of its review, the staff will examine the bus loadings and voltage changes for the onsite and the offsite electrical power system.

TVA Re 1 A.a The additional loading of the electrical equipment due to power uprate is minimal and is well within the existing analyzed capability of the plant onsite electrical distribution system. The table below identifies the major load changes due to power uprate. These changes are limited to increased power requirements to the reactor recirculation pump motor-generator sets, the condensate pumps and the condensate booster pumps. There are other minimal load changes but all are within the motor nameplate ratings. The electrical calculations for the auxiliary power system have used nameplate data for the affected loads. The name plate loads are higher than the operating loads at either pre-uprate or uprated conditions. Therefore the pre-uprate electrical calculations and the uprate calculations are the same.

Additionally, the electrical single lines drawings show the motor nameplate ratings and do not require revision. No electrical equipment requires increased voltage to operate at the proposed power uprated conditions. Therefore, the 4kV and the 480V distributions system are adequate for power uprate.

E1-51

Table 4.a-1 Changes In Browns Ferry Units 2 and 3 Onsite AC Distribution System Loads System/Component Pre-Uprate Power Uprate Nameplate Remarks Condition Requirements Rating Reactor 8378 HP 8 8678 HP 8 9000 HP Note 1 Recirculation 100% core 100% core Pump flow flow Motor/Generator Set Motors Condensate pumps 849 HP 867 HP 900 HP Note 2 Condensate 1519 HP 1559 HP 1750 HP Note 2 booster pumps Notes:

1. Power requirement per recirculation system loop in service to obtain 100% core flow.

2. Power requirement per pump with combination of 3 reactor feedwater pumps, 3 condensate booster pumps, and 3 condensate pumps.

There are no load increases in the DC equipment required for the proposed power uprate. Therefore, the 250VDC 1E plant system is adequate for power uprate.

Additionally, the following power systems are not affected by power uprate:

~ 125VDC Diesel Batteries

~ 250VDC Distribution Boards (Shutdown board switchgear controls)

~ 120VAC Instrument and Control bus

~ 48VDC Distribution System

~ 24VDC Distribution System As discussed with the NRC reviewer on July 9, 1998, Figures 4.c-1 and 4.c-2 provide one-line diagrams for grid load information.

NRC Re est 4.b In a previous request for additional information (RAI), the staff requested the list of grid stability cases performed to support the power uprate and summary of the findings for each case. Please provide the list and discuss new stability El-52

limits that would result in an Operations Standing Order (response item B.3).

TVA Re 1 4.b Generator stability studies have been performed with the increased generator capability and minimum gross MVAR limits per unit have been established. The limitations (as shown below) are based on transmission system configurations. The limits will be incorporated into the Operations Standing Orders as part, of the implementation of power uprate. With all transmission lines in service (i.e., none out of service),

the minimum MVAR has changed from -250 MVAR to -200 MVAR.

Table 4.b-1 Minimum MVAR Values

¹ OF BFN 500KV LINES PRE -UPRATE UPRATE BFN OUT OF SERVICE MINIMUM MINIMUM UNITS MVAR/UN TI MVAR/UN TI NONE -250 -200 BFN MADISON +100 BFN-WEST POINT -150 -50 BFN-TRINITY -200 -100 ANY LINE ON BUS 2 -250 -150 BFN-MADISON -250 -200 NRC Re est 4.c In response to staff's question that an increase of 57.5 MW generation to each unit could have an impact on grid voltage profile, you indicated that (response Item B.3) PSB-1 does not require reanalysis since the methodology and software have not changed, and the degraded grid setpoint is not affected by the added generation because there are no load changes. The staff believes that the offsite grid voltage may be impacted as result of power uprate and by increases in the onsite loads which, in turn, could affect previous PSB-1 analysis and the degraded grid setpoints. Please reassess the degraded grid setpoint calculation and PSB-1 based on the new grid voltage El-53

0 to ensure its adequacy, or show that there is no impact on the previous PSB-1 analysis.

TVA Re 1 4.c For offsite power, the pre-uprate switchyard voltage is calculated to be 525.9kV with both Units 2 and 3 operating at 1100MW at 0 MVAR. The switchyard voltage for power uprate is 524.1kV with both units operating at 1180MW at 0 MVAR which bounds the expected increase of 55 MWe. This is a 0.34%

decrease in voltage and has negligible impact to the plant auxiliary power system. Additionally, the station service transformers which supply the safety loads have automatic load tap changers. Additionally, the degraded grid setpoint is not impacted by the change in switchyard voltage. See Figures 4.c-1 and 4.c-2 E1-54

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!I':DI<<II BFN AREA PLOT Fr I, Hal:  !=. 1998 1L(138 KVI 5140 . SL944,&44 r(igure 4.c-2 One-Line Diagram for Grid Load - 1180 MWe

NRC Re est 4.d Besides containment spray and residual heat removal pumps, please identify and discuss acceptability of other electrical equipment for which name plate horsepower values are not used.

TVA Re 1 4.d As stated above in TVA Reply 4.a, electrical calculations use motor nameplate data when performing load flows and voltage drop calculations. Core spray (not containment spray as stated in the NRC Request: ) and RHR pumps are the only loads which do not use motor nameplate data when performing these types of calculations. The flow requirements for these loads have not changed nor did the injection pressure change as a result of power uprate.

NRC Re est 4.e Based on the review of the radiological doses for the safety-related electrical equipment before and after power uprate, please confirm that all safety-related electrical equipment is bounded by the original design basis.

TVA Re 1 4.e The response to this item is provided by the TVA Reply to item C of Reference 11.

ADDITIONAL REQUESTS Responses to additional related requests noted during the July 9, 1998, meeting are provided below.

NRC Re est 4.f The NRC would like to see a composite profile vs. current profiles for temperature .over time. Did total curves bound extended time not just during the peak? The NRC would like to see an example of calculations (Curves) for one component.

Show Arrhenius review with justification that pre and post uprate temperature curves are bounding TVA Re 1 4.f The environmental qualification (EQ) of electrical equipment is based on testing equipment to conditions more severe than expected during accident mitigation. In cases where the equipment test profiles do not envelope the plant accident profile for the entire post accident, time, the Arrhenius E1-57

methodology has been used to extrapolate the test profiles to the full duration of the post accident requirement plus required margin.

Excerpts from the following current calculations are provided in Enclosure 2 and provide examples of the comparison of the test profile to the composite temperature profile.

~ MD-Q0999-980041

~ MD-Q0999-980056

~ MD-Q0999-980068 NRC Re est 4.

Provide a table listing line breaks used for the temperatures analyzed for EQ.

TVA Re 1 4.

Table 4.g-1 High Energy Line Breaks Analyzed Inside Primary Outside Primary Containment Containment Recirculation Line Main Steam Break (DBA LOCA)

Feedwater 0.01 ft Main Steam Reactor Water Cleanup Line Break 0.10 ft Main Steam HPCI Steam Line Line Break Reactor Core Isolation 0.50 ft'ain Steam Cooling (RCIC) Steam Line Line Break 5 . RADIOLOGICAL ISSUES In the February 18, 1998 RAI, the staff regarding parameters for the design basis accidents (DBA) analyses. TVA responded in a letter dated April 1, 1998. TVA has determined the radiological doses due to DBAs for power uprate conditions by

'scaling the pre-uprate doses upward by 5%. Although TVA provided great detail on how the scaling factor was developed and applied, the input data to each analysis was not provided.

In its letter of May 7, 1998, the staff identified a large E1-58

number of apparent discrepancies between the Updated Final Safety Analysis Report (UFSAR) and your October 1, 1997 analyses and requested that TVA provide a tabulation of analysis parameters. TVA responded in a letter dated June 12, 1998. After reviewing the response, the staff has the following additional questions:

a. Please resolve the following discrepancies between the data provided in your June 12, 1998 let:ter and the UFSAR and other regulatory documents.

NRC Re est S.a.1 It:em 1.b.(l) of the table identifies that the. iodine concentration in the containment sump water is 25% of the core inventory. This is inconsistent with regulatory guidance that the activity should be based on 50% core inventory.

TVA Re 1 5.a.1 As stated in TVA's June 12, 1998, response (Reference 9), ECCS leakage is not currently part of BFN's design or licensing basis. It was additionally noted in Reference 9 that some of the values given were subject to change. The proposed iodine concentration in the containment sump water was an estimated value based on the current licensing basis. In Reference 9, TVA proposed a license condition that will require TVA to perform an analysis of the design basis LOCA to confirm compliance with GDC-19 and offsite dose limits considering main steam isolation valve leakage and ECCS leakage.

Subsequent review of Standard Review Plan (SRP) 15.6.5, Appendix B by TVA in preparation for this analysis identified that 50% of the core iodine inventory should be postulated to be mixed in the sump water being circulated through the containment external piping systems. The iodine concentration of containment: sump water used in the analysis will be consistent wit:h regulatory guidance. TVA wants to emphasize that the percentage of core iodine inventory contained in the containment sump water is not material for the current licensing/design basis, since ECCS leakage is not postulated to occur and thus, has no effect on the power uprate submittals.

NRC Re est 5.a.2 Item 3.c specifies that main steam isolation valves (MSIVs) will close in 5.5 seconds. The analysis described in the UFSAR assumes an isolation time of 10.5 seconds.

E1-59

NRC Re est 5.a.3 Item 3.d lists main steam line break steam and water release quantities of 19,874 and 43,740 ibm, respectively. The UFSAR analysis lists 25, 000 and 160, 000 ibm, respectively.

TVA Re 1 5.a.2 6 5.a.3 Inconsistencies between the design basis calculation for the MSLB, the UFSAR, and the Technical Specification Bases discussion of the MSLB has been previously identified as a condition adverse to quality during the implementation of Improved Technical Specifications (ITS). The ITS bases state:

The basis for the equilibrium coolant iodine activity limit is a computed dose to the thyroid of 36 rem at the exclusion distance during the two-hour period following a steam line break. This dose is computed with the conservative assumption of a release of 140,000 lbs of coolant prior to closure of the isolation valves...

Since the design basis calculation of the MSLB event is bounded by the ITS bases for both dose (32.05 rem vs. 36 rem) and mass release (63614 lb vs. 140,000 lb), TVA determined at the time of discovery that no unreviewed safety question existed. The use of a MSIV closing time of 5.5 seconds is consistent with the ITS requirement that the MSIV isolation times are between 3 and 5 seconds. It should also be noted that the 5.5 second closure time used in the MSLB calculation is consistent with the guidance contained in Regulatory Guide 1.5, "Assumptions Used for Evaluation the Potential Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors" which states "The steam line isolation valves close in the maximum time incorporated or to be incorporated in the technical specifications." The different mass releases between the UFSAR and the design basis MSLB calculation are a direct function of the assumed MSIV closure time, thus the noted discrepancies between the UFSAR and the design basis MSLB calculation are expected. A 10 CFR 50.59 evaluation was performed which concluded that the 5.5 second closure time does not involve an unreviewed safety question.

Therefore, the submitted value for closure time of 5.5 seconds is TVA's current approved licensing basis. The change to the UFSAR will be included in the next UFSAR update.

NRC Re est 5.a.4 Item 4.a states that there are 48,132 fuel rods in the BFN core (764 bundles). This appears to assume 63 rods per

bundle. In the generic GE 8 x 8 fuel there are two water rods and 62 rods that contain fuel, for a total of 47, 368 rods in the core. The UFSAR analysis assumes 62 rods per bundle. In this application, a larger number of rods is less conservative.

TVA Re 1 5.a.4 In TVA's June 12, 1998, response (Reference 9), item 4.a stated the value of 48132 as the number of fuel rods in the core for the fuel handling accident (FHA) analysis. The value of 48132 fuel rods in the core was obtained from the design basis calculations for the radiological consequences of a FHA.

The TVA FHA calculation assumed that each bundle typically consisted of 63 fuel rods (for a total of 48132 rod in the core), but noted in the calculation that if all bundles only contained 62 rods t:here would be an insignificant variation in bundle source strength of only about 1.6%. As noted in the TVA calculations which determined the radiological consequences of the FHA, if all bundles contained only 62 rods there would be an insignificant increase (1.6%), which would not significantly affect the results of the calculation (including the power uprate submittal). Based on the above discussion, the design basis calculations for the FHA will be revised to reflect the UFSAR assumptions.

NRC Re est 5.a.5 Your response, Item 4.h. st:ates that 5500 cubic feet of release occurs from the reactor building prior to isolation and transfer to standby gas treatment system ('SGTS). This is apparently based on a 15-second ventilation flow at the rate of 22,000 cfm. This flow rate is inconsistent with the second bulleted item under Item 5 which states that'he air flow from the reactor building prior to isolation is 95,000 cfm.

TVA Re 1 S.a.5 The second bulleted item under item 5 is referring to the normal reactor zone ventilation flow rate (95,000 cfm) per unit. The normal refuel fan flow rate per unit is 45,000 cfm.

This is the tot:al flow exhausted from the refuel floor on a per unit basis and is the flow rate of concern for a FHA.

There are various suction points on the refueling floor and it:

would be overly conservative to assume all 45,000 cfm is instantaneously flowing from the point of the FHA. There are two suction points which are considered applicable in the consideration of a FHA:

~ The reactor cavit:y with a total flow rate of 9,000 cfm.

~ The spent fuel pool with a total flow rate of 18,000 cfm.

Since the spent fuel pool has a significantly higher flow rate, it was chosen as representing the worst case flow rate.

For additional conservatism, 22,000 cfm (as noted in UFSAR, section 14.6.4.5.h) was used in the analysis as the bypass flow rate. This is, coincidentally, similar to the SGTS system flow rat:e.

NRC Re est 5.a.6 The new power after uprate would be 3458 Mwt. Your submittal indicates that safety analyses have been performed at 102% of the uprated power level. This is inconsistent with Section 8.3.2 of your submittal which states that "...fission product inventories are prepared based on irradiation of BFNP fuel for 1400 days at the uprated power of 3458 Mwt...." However, RG 1.49, Power Levels of Nuclear Power Plants, paragraph C.3, provides that analyses of the offsite radiological consequences of postulated design basis accidents should be performed for an assumed core power level equal to 1.02 times the proposed licensed power level.

TVA Re 1 5.a.6 Section 1.2.1 of the BFN Power Uprate Safety Analysis Report has the following statement "the pre-uprate safety analysis basis assumed that the reactor has been operating continuously at a power level of at least 1.02 times the licensed power level; ...".

A review of the pre-uprate BFN licensing analyses bases showed that the extent of this 2% power factor has been applied to the design calculations pertaining to reactor transients, containment, and LOCA analyses. In the May 20, 1998, submit:tal to NRC RAI, TVA has provided the same information in the response t'.o Question D.4. The review also showed that the pre-uprate licensing design bases for BFN do not include the commitment to Regulatory Guide 1.49.

Regarding the methods and assumptions for radiological evaluations, Appendix H of NEDC-31897P-A, "Generic Guidelines for GE BWR Power Uprate", May 1992 stated in Section H.3 that "For those cases where the calculations require reanalyses, the new analysis will be based on the methods, assumptions and Regulatory Guides that are currently used in the existing SARs." In the NRC Safety Evaluation Report (SER) associated with the NEDC-31897P-A, Section 2.7(a), confirmed the continued use of the licensed radiological evaluation methods and assumptions.

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Since BFN is not committed to Regulatory Guide 1.49, it concluded that the off-site radiological calculations for BFN is do not require the application of the 2% power factor. The calculated offsite doses as shown in References 1 and 9 are well within the 10 CFR 100 limits. If the 2% power factor had been utilized, the offsite doses would still remain bounded by 10 CFR 100 limits. The approach used in the power uprate off-site radiological evaluations is consistent with the BFN licensing design bases, the GE BWR generic guidelines for power uprate analyses and the associated NRC SER.

However, a conservative 1.05 multiplier can be used to account for the 2% power factor. A review of the power uprate off-site radiological consequences for LOCA (Table 9-3, Reference 1, Enclosure 5), Main Steamline Break (Table 9-4), Fuel Handling Accident (Table 9-5), and Control Rod Drop Accident (Table 9-6) showed that with this 1.05 multiplier, the resulting power uprate radiological doses still remain below regulatory limits.

NRC Re est S.b Please review the above discrepancies and any other UFSAR discrepancies in the data relating to power uprate issues and determine whether they involve an unreviewed safety questions.

Please inform us of your schedule for resolving these discrepancies. This is necessary for the staff to make a finding on the acceptability of the power uprate values obtained by the scaling approach.

TVA Re 1 S.b TVA has performed a review of the power uprate submittals including the license amendment, supplements, and RAI responses (References 1 through 12) to determine if these submittals were consistent with the information provided in the UFSAR. Inconsistencies that were identified are discussed further below. This review did not identify any technical issues which would invalidate the results or conclusions in the power uprate submittals.

Inconsistencies identified during this review are categorized as follows:

2'ncreased Core Flow (2'CF) The UFSAR review ident:ified two cases in which t: he information presented in the UFSAR had not been updated to reflect the approved operation of the plant in ICF and the Maximum Extended Load Line Limit (MELLL) domain through the ICF region. Operation with ICF E1-63

and MELLL have been previously approved by the NRC and does not involve a unreviewed safety question.

Main Steam line Break radiological analysis As identified in NRC Request 5.a above, the UFSAR did not contain the results of the latest radiological analysis for the MSLB.

Accordingly, TVA has revised the UFSAR in accordance with 10 CFR 50.59 to reflect the current MSLB analysis. No unreviewed safety question was involved in this change.

Control Room Emergency Ventilation System (CREVS) As a result of the ongoing discussions with the NRC involving the CREVS, new analyses and results were provided to the NRC with the power uprate submittals. Since these analyses were recently revised, the results had not yet been incorporated into the UFSAR. These changes will be incorporated into the UFSAR as part of the UFSAR update process in accordance with 10 CFR 50.59.

Number oS f'uel rods used in analyses This discrepancy is discussed in NRC Request 5.a.4 above.

Typographical errors in the power uprate submi ttals These errors have no technical impact on the results or conclusions of the power uprate program. Correction to the associated pages of NEDC-32751P have been included in ~ Since these are minor editorial errors in the submittals (not the UFSAR), they do not involve an unreviewed safety question. The following typographical errors were identified in the power uprate submittals.

~ NEDC-32751P, Tables 1-2 and 9-1, list the pre-uprate vessel steam flow as 13.37 Mlb/hr. The correct value is 13.38 Mlb/hr as previously provided in our RAI response dated May 22, 1998, Table 6-1 (Reference 8).

~ NEDC-32751P, Table 4-1, lists the pre-uprate peak drywell gas temperature as 294'F. The correct value is 295'F as previously stated in our RAI response dated June 26, 1998 (Reference 11).

~ NEDC-32751P, Section 3.6, "Main Steamline Flow Restrictors," references UFSAR Section 5.4. UFSAR Section 5.4 does not exist. The correct reference is UFSAR Section 4.5.

~ NEDC-32751P, Section 9, as discussed in Reply 3.d, incorrectly identifies the document as NEDC-32571P in the report page headers.

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~ The RAI response dated June 12, 1998, (Reference 9),

TVA Reply 8 listed the secondary containment volume as 1, 931, 500 ft . The correct value is 1, 931, 502ft'.

This difference is inconsequential and no replacement page is provided.

The UFSAR changes identified above will be processed in accordance with the site procedures and included in the next amendment to the UFSAR in accordance with 10 CFR 50.71(a).

NRC Re est 5.c The first bulleted item under Item 5 states that the reactor zone volume is 1,335,000 cubic feet prior to secondary containment isolation. TVA has previously stated that the secondary containment is being treated as a single zone.

Please explain (1) the applicability of the phrase "prior to isolation" and (2) why the secondary containment volume of 1, 931, 500 cubic feet used in other analyses is not applicable.

TVA Re 1 5.c On September 27, 1994, Technical Specification change request TS-322 was approved by the NRC which eliminated the main steam line radiation monitor isolation signal for Units 1, 2, and 3.

When the main steam line radiation monitor isolation signal for the recirculation sample line was eliminated, a potential primary containment to secondary containment bypass leak path was identified. The volume identified in the first bullet under item 5 is the volume of a single reactor zone (1,335,000 ft ), whereas the volume of 1,931,502 ft used in other analysis is the secondary 'containment mixing volume. For this particular case, prior to secondary containment isolation, it is appropriate to use the volume of a single reactor zone, since normal ventilation is in service and the SGTS has not been initiated. After SGTS initiation and secondary containment isolation, the secondary containment mixing volume used in other analyses (1,931,502 ft') is used. It should also be noted that the smaller volume used initially (prior to isolation) will give more conservative dose results.

Additionally, the term "prior to isolation" refers to the configuration of the secondary containment prior to'econdary containment isolation and initiation of SGTS.

ADDITIONAL REQUESTS Responses to additional related requests noted during the July 9, 1998, meeting are provided below.

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0' NRC Re est 5.d.1 Are off-site dose affected by uprate?

TVA Re 1 5.d.1 The offsite doses are affected as described in the October 1, 1997, submittal, Section 9.2; the April 1, 1998, submittal, Question A.1.a; and the June 12, 1998, submittal Question 1.

As described in these submittals, the offsite doses were scaled according to the scaling factors in Table 1 of the April 1, 1998, submittal.

NRC Re est 5.d.2 The June 26, 1998, submittal, page E2-6 states that the total accident radiation doses are not increased by power uprate, yet the October 1, 1997, submittal noted a 5% increase.

Provide clarification.

TVA Re 1 5.d.2 The post accident doses outside of primary containment are affected as discussed in the October 1, 1997, submittal in accordance with the scaling factors discussed in the reply to Request A.l.a of the April 1, 1998, submittal. Some components of dose contribution increase while others decrease. The post accident doses outside of primary containment remain within the doses previously assumed in the evaluations of equipment with the exception of the standby gas treatment building as discussed on page E2-7 of the June 26, 1998, submittal. The statement in the June 26, 1998, submittal was referring to the fact that the doses do not increase above the dose assumed previously.

6. HUMAN FACTORS ISSUES NRC Re est G.a In response to the staff's RAI, by letter dated April 28, 1998, you discussed issues relating to operator actions that are particularly sensitive to the power uprate, including operator response times, or performance. You stated that you have reviewed all operator responses used in your probalistic safety analysis (PSA) and confirmed that the effect on operator response times [due to power uprate] at Browns Ferry are consistent with the GE generic findings. For certain

.operator actions which would be sensitive to power uprate, the required operator response time has decreased. You concluded that because these operator actions are controlled by El-66

emergency operating procedures, the slight reduction in response times noted for the power uprate condition "will not significantly affect the operator's ability to safely complete the required actions."

For certain scenarios, operator responses are required to be achieved in less than 5 minutes. In general, minimal reduction in response times should not significantly affect the operator's ability to complete their actions, but the staff is concerned that those actions assumed to be performed in 5 minutes or less by the PSA may not be achievable under realistic conditions. The staff refers to guidance contained in ANSI/ANS Standard 58.8, "Time Response Design Criteria for Safety-Related Operator Actions" (1994), which indicates that safety-related operator actions that must be initiated within 5 minutes or less (for events that occur with an estimated frequency of -10 ') "shall be initiated by automatic protection systems." Therefore, please provide evidence that operators can perform the required tasks under accident conditions in the times assumed by the PSA.

TVA Re 1 6.a For special events, the manual actions required for mitigation are not evaluated utilizing American Nuclear Society/American National Standard-58.8-1994, "Time Response Design Criteria for Safety-Related Operator Actions." ANSI/ANS-58.8 specifies its scope to include "...criteria ...used to determine the minimum response time intervals for safety-related operator actions that are taken to mitigate design basis events (DBEs) which result in an automatic reactor trip." The purpose of this standard is thus, "not to serve as a basis for determining actual operator action times in procedures...,"

but to specify minimum time intervals which will be sufficient for plant design to allow operators to perform a manual, safety-related action. This methodology is most applicable to new designs or to event-oriented scenarios where well-defined, chronologically-sequenced plant response may be expected such as the UFSAR design basis transients and accidents. As such, the standard criteria given by ANSI/ANS-58.8 are used in the "design of safety-related systems" and "to determine whether safety-related systems can be initiated by operator action or require automatic action." Special events include plant conditions beyond design basis events (e.g., multiple equipment failures).

For the scenarios of this issue, the Emergency Operating Instructions (EOI),govern operator response. The EOIs do not require the operator to perform detailed diagnosis of the event in order to be able to correctly respond to the event.

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Instead, the operator responds to predetermined symptoms and specified event entry conditions and controls plant shutdown and cooldown by the established procedural guidance. The NRC SER of the BWR Owner's Group (BWROG) Emergency Procedure Guidelines (EPGs) Revision 4, dated September 12, 1988, is included as part of the licensing bases of Browns Ferry. This NRC SER stated, "we find the actions specified in the Emergency Procedure Guidelines to be generally correct and appropriate and within the operator's capability...The continued use of symptoms, rather than events as bases for actions, should serve to minimize errors resulting from incorrect diagnosis of events and addresses the possibility of multiple failures and operator errors. We therefore find the guidelines acceptable for implementation." As such, the operator actions of concern are procedurally performed as the symptoms are recognized, using controls immediately available to the reactor operators in the main control room.

The actions specified to occur within 5 minutes are all in response to an ATWS event. The ATWS scenarios are addressed by operator response through EOI-1. While performing these actions, the operator is focused on completing the EOI steps under the command and control of the Unit Supervisor, ensuring timely completion of the mitigating actions. For an ATWS event, there are no other required actions which would distract the operator from completing the necessary mitigating actions and the accomplishment of the assigned duties.

Initiation of SLCS pump start and inhibit of automatic depressurization system (ADS) during an ATWS are specific actions found only in BFN's EOIs. The NRC's SER specifically addresses ATWS evaluation with respect to the EPGs, stating, "ATWS guidelines are symptom oriented...(and) give improved guidance...to prevent and mitigate severe accidents." At BFN, SLCS pump start and ADS inhibit are performed from direct procedural guidance and both can be completed using switches immediately available to reactor operators in the main control room's control area. SLCS pump start is accomplished by turning the SLCS keylock switch" (key located next to switch on panel) to either the 'START A'r 'START B'osition. ADS inhibit is accomplished by turning two keylock switches (keys located next to switches on panel) to the "INHIBIT" position.

These are straightforward operator actions and, thus, can be performed rapidly.

Emergency depressurization of the reactor pressure vessel (RPV) is not initiated until RPV water level has lowered to the top of the active fuel (TAF). Since RPV water level is one of the key parameters of control, specific monitoring of RPV level continues during the time RPV level is lowering and

the control room staff is attempting to reestablish RPV injection. RPV water level conditions requiring are expected to be noticed immediately, using emergency'epressurization control room instrumentation. MSRVs are then manually opened using control switches immediately available to the reactor operator in the main control room's control area. Individual MSRVs are opened by placing the control switch to the "OPEN" position. This is a straightforward operator action and, thus, can be performed rapidly.

From a PSA perspective, the BFN PSA includes a detailed methodology and corresponding results of the human actions evaluations performed to support the generation of the PSA.

The types of actions evaluated included human solutions and human errors that are a vital part of nuclear plant operation and accident response.

The following types of human actions were evaluated:

~ Routine actions before the initiating event

~ Actions that can cause initiating events

~ Dynamic operator actions accomplished during the plant response to an initiator

~ Recovery actions Detailed human action evaluation work was accomplished to assign an unique numerical value for selected human actions including the actions associated with dynamic actions accomplished during the plant response to an initiator. These human action evaluations were accomplished by including an interview with operating personnel. These interviews were guided by selected factors. These factors included: a) plant man-machines interfaces and indications of conditions, b) task complexity, c) adequacy of time to accomplish action, d) significant preceding and concurrent actions, e) stress, f) training and experience, and g) procedural guidance. The interviewee's were encouraged to utilize simulator experience and observations as input to the question responses. Post interview activities also included informal review of simulator experience.

Each of these factors had a systematic approach to provide the input to support the interview process. Note the fact that the human actions evaluations includes the assessment of the time to accomplish the associated action. To determine the response time available for the human actions within the scope of the PSA, a plant unique analysis was accomplished. BFN used the MAAP code to do this. The MAAP code provides the necessary input to assess and determine the unique numerical

0 values for each of the human actions that are a part of the PSA.

The times determined by the MAAP code must not be construed to be the absolute time responses of the plant for the input to operating procedures. These values are determined to support the PSA and are not intended for general use in other applications.

Operator actions for mitigation of these special events are not provided as an alternative to an automatic action. The actions are taken in immediate response to monitored actual plant conditions using main control room indications. The actions are completed by using control switches readily available to the reactor operator in the main control room' control area.

In conclusion, the operator action times specified in t: he April 1, 1998, submitt:al necessary to mitigate the postulated special events can be accomplished wit:h a high degree of reliability within the specified time frames following power uprate.

NRC Re est 6.b Your April 28, 1998 letter also indicated that one manual action, i.e., termination of the High Pressure Coolant Inject:ion [HPCI] system injection following an Appendix R fire event has a reduction in response time from 10 minutes to 7 minutes. Your letter explained t:hat the reduction in response time is a result of using different models to predict pre- and post-uprate operator action times (i.e., the GE SAFE model for pre-uprate predictions and the SAFER model for post-uprate).

Your letter further indicated that this action which involves closing a valve from the 250V DC reactor MOV board located just outside the main control room, has been performed in a shorter time than allowed by the SAFER model, i.e., 7 minutes.

NRC Re est 6.b.1 Your letter, page EI-4 stat:es: "TVA has previously demonstrated that this action, close one valve from the 250V DC reactor MOV board, located on the same elevation just outside the main control room, can be performed within the shorter time predicted by the SAFER model." Please provide a reference for this demonstration including the following factors considered: Environmental conditions expected; procedural guidance for the required actions; support personnel and/or equipment required to carry out the required E1-70

actions; information requirements including qualified instrumentation.

TVA Re 1 6.b.1 The Appendix R fire event which requires manual isolation of HPCI from outside of the control room is a fire in fire zone 16, which is a control building fire in plant elevation 593 through plant elevation 617. The procedural guidance to take the manual operator action is found in TVA procedure 2/3 SSI-16. TVA Operations personnel performed in plant drills on October 25, 1995 to validate all manual operator action times.

Drill performance was observed by the NRC and documented by NRC as part of an NRC inspection (NRC Inspection Report 95-60) .

For Unit 2, the 10 minute HPCI isolation steps were steps 33 and 34 of 36 total action steps for the associated Unit 2 procedure. The entire associated procedure's performance was completed in 8 minutes, 20 seconds. Since all actions of this procedure are completed at control boards in the same Shutdown Board Room, except travel time from the main control room, each step of the procedure can be assumed to take approximately the same amount of time. For power uprate, the procedure is being revised to move the HPCI isolation steps to the first steps of the procedure. As a result, this action will be performed in considerably less than 7 minutes.

For Unit 3, the 10-minute HPCI isolation steps were steps 10 and 11 of 62 total action steps for the associated Unit 3 procedure (containing 10 and 20 minute actions). The entire associated procedure's performance was completed in 12 minutes, 15 seconds. All actions of this procedure are not completed in the same Shutdown Board Room. In fact, travel from the shutdown board room to locations ending in the reactor building add. considerable time not used in performing control switch manipulations for this procedure. As described above for Unit 2, the Unit 3 procedure is being revised to move the HPCI isolation steps to the first steps of the procedure. Therefore, this action will be performed in considerably less than 7 minutes.

In conclusion, two steps used to perform the HPCI isolation for power uprate remain the same, except they have been relocated to the first steps in the respective procedures and further ensure the design time requirement of 7 minutes will be met.

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The required manual action to isolate HPCI is performed inside of a Shutdown Board Room. This room has a safety-related ventilation system, independent from fire zone 16. Therefore, a fire in fire zone 16 does not create smoke in the areas in which the required manual operator actions must: take place.

For manual isolation of HPCI, a unit operator is dispatched to the 250V Reactor MOV board. To isolate HPCI, the operator will use existing controls on the HPCI Turbine Steam Supply Valve breaker compartment. No special equipment or support personnel are required to take this action.

NRC Re est 6.b.2 The SAFER model is used to analyze LOCA conditions and fuel heatup activities. Your October 1, 1997 letter, (enclosure 5) indicates that GE applied the model to analyze an Appendix R fire event and stated that, "Sufficient time is available for the operator to perform the necessary actions" (p.6-9). How does the SAFER model evaluate human actions? What is the basis for GE's conclusion that operators have sufficient time to perform necessary actions?

TVA Re 1 6.b.2 The SAFER model does not evaluate human actions. The SAFER model was used to determine the reactor vessel response for Appendix R fire events. The demonstration of operator response times is discussed in TVA Reply 6.b.l.

NRC Re est 6.b.3 It is our understanding that the SAFER model was for the used for the pre-uprate power uprate analysis and the SAFE model analysis. According to your April 28, 1998 letter, the SAFE model predicted a 10-minute required operator action time to shut down HPCI. The SAFER model predicted 7 minutes to shut down HPCI. Which action time is TVA taking credit for?

TVA Re 1 6.b.3 BFN will implement the 7 minute action time concurrent with the implementation of power uprate and the associated adopt:ion of the SAFER code for Appendix R events.

NRC Re est 6.b.4 In addition, what are credible errors that operators could make in taking this action? What are the consequences of the operator failing to accomplish the action and how will recovery from the failure(s) be accomplished? How does TVA El-72

know that operators can successfully recover from credible errors, i.e., provide evidence that operators can recover from credible errors.

TVA Re 1 6.b.4 No changes are being made in the design approach of the Appendix R program due to the power uprate. The procedural step changes being made do not introduce any new human factor engineering issues. The steps used to perform the HPCI isolation remain the same except they have been relocated to the first steps in the procedure to further ensure the time requirement will be met. In addition, performance-shaping factors of Appendix R procedure usage (inclusion in licensed operator requalification training, control board layout and location, and detailed procedures) provide a combined influence for a low probability of operator error. Should this, or any other error affect reactor control parameters, reactor water level or reactor pressure, resulting conditions would be recognized by the operators using indications at the backup control panel.

Consequences of not making the 7 minute HPCI time action could potentially fill the reactor vessel up to the main steam lines and result in RCIC being disabled by liquid carryover.

Recovery actions by the operators would be to drain RCIC and utilize other alternate injection paths in that time frame.

High confidence exists that this is a very recoverable scenario should the action not be performed within specified time frames.

NRC Re est 6.c Please describe all changes the power uprate will have on the operator training program and the plant simulator. Provide a copy of the post-modification test report (or test abstracts) to document and support the effectiveness of simulator changes as required by ANSI/ANS 3.5-1985, Section 5.4.1 within 60 days of implementing power uprate on each unit.

TVA Re 1 6.c This item was addressed in Reference 5. Please refer to the TVA response to Item 5 of that submittal for the details of the requested information and to Enclosure 2 of Reference 5 for the associated licensing commitments.

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7. CONTAINMENT ADDITIONAL REQUESTS Responses to additional requests noted during the July 9, 1998, meeting are provided below.

NRC Re est 7.a How long is temperature is above 336 and 281? State how long is duration that temperature exceeds containment design parameters. How long do we exceed 281? Same question for 336 inside containment.

TVA Re 1 7.a As shown in Section 4.1.1.2 of the BFN Power Uprate Safety Analysis Report NEDC-32751P, the power uprate analysis of the DBA recirculation line. break predicted a peak drywell airspace temperature of 297'F. The total time duration for which the drywell airspace temperature exceeds the BFN containment structural design basis temperature limit of 281'F is approximately 11 seconds.

For the MSLB inside containment analysis used in the EQ evaluations, the calculated peak drywell airspace temperature at the power uprate condition is 336'F. The total time duration for which the drywell airspace temperature exceeds the BFN containment structural design basis temperature limit of 281'F is approximately 12 minutes. However, the calculated pe'ak drywell shell temperature remains at 277'F and did not exceed the 281'F drywell shell design temperature limit.

NRC Re est 7.b Provide containment analysis information (Pressure and Temperature Curves at uprate and pre-uprate conditions vs.

FSAR values) .

TVA Re 1 7.b The BFN primary containment response to a postulated LOCA at the pre-uprate condition is provided in UFSAR Section 14.6.

The current design basis calculated peak drywell pressure and temperature are 49.6 psig and 295'F, respectively. Subsequent to the original FSAR containment analyses, the BFN short-term DBA LOCA analyses were re-analyzed during the Mark I Long-Term (LTP) Containment Program to validate the containment structural integrity with regards to the hydrodynamic loads on the Mark I type of containment. These Mark I LTP analyses El-74

were based on the GE containment model M3CPT using the break flows and enthalpies internally calculated with the M3CPT built-in break flow model.

For the power uprate condition, the computer codes LAMB and M3CPT were used to calculate the peak primary containment responses. While M3CPT has the capability to predict break flow rates, the LAMB computer code with its more detailed reactor pressure vessel model was used to determine more realistic break flow rates for input into the M3CPT code.

This LAMB/M3CPT approach is consistent with the NRC approved generic guidelines for power uprate (Appendix G, NEDC-31897P-A, May 1992) . The results of this analysis, as presented in Table 4-1 of Reference 1, Enclosure 5, are 50.6 psig for peak drywell pressure and 297'F for peak drywell temperature. The peak drywell pressure result is bounded by the BFN containment design limit pressure of 56 psig.- The peak drywell temperature exceeds the containment design temperature for a short duration (approximately 12 seconds) and does not represent any concern to the containment structural integrity.

To provide a meaningful comparison of the power uprate impact on the BFN containment short-term peak containment pressure and temperature, a sensitivity analysis was performed at the pre-uprate and uprate conditions, using the M3CPT model with internally calculated break flows (Mark I LTP methodology).

Except for the initial core thermal power and dome pressure, there is no change in the containment input parameters between the pre-uprate and power uprate sensitivity cases.

Additionally, the inputs are the same used in the power uprate design basis containment analyses with the exception of the use of the built-in break flow model.

The results of the analyses are summarized in Table 7.b-1 below. Using the M3CPT with built-in break flow calculation, an increase in the peak DBA-LOCA drywell pressure of 1.9 psi and an increase in the peak drywell temperature of 0.8'F were obtained at the power uprate condition. Figures 7.b-l through 7.b-4 show the drywell and wetwell pressure and temperature time histories for the 30 second analysis duration.

The above power uprate impact on drywell pressure and temperature can be used to estimate the primary containment response at the pre-uprate condition for a LAMB/M3CPT type of calculation. The peak containment response at the pre-uprate condition for such evaluation would be expected to be approximately 48.7 psig (i.e., 50.6 psig minus 1.9 psi) for drywell pressure and 296 F (297'F minus 0.8'F) for drywell

temperature. These results remain bounded by the BFN containment design pressure and temperature limits.

Table 7.b-l Summary of BFN DBA-LOCA Containment Analyses Results Using M3CPT with Built-in Break Flow Model*

Pre-Uprate Power Uprate Peak Drywell Pressure, psig 44.8 46.7 Peak Drywell Temperature, 'F 289.2 290.0

• The actual peak containment pressure using LAMB/M3CPT is as shown in Table 4-1 of Reference 1.

It is noted that the power uprate design basis peak containment response (50.6 psig) is based upon a conservative representation of the LAMB results to maximize the predicted containment pressure. A less conservative representation of the LAMB results would be expected to give a peak pressure very similar to the value predicted using the M3CPT built-in break flow model.

El-76

TVA IPIP ORAL PRESBHE CONT PRESS WETWELL PRESSlHE RESP RE BRK 102P/81F - FFWTR 60.

't0.

20.

0.

0. to. 20. 50 ~

l11N 00202 070998 1 629.4 TIME (SECONDS)

Drywell and Wetwell Pressures as a Function of Time - 102% Current Rated Power /100% Core Flow

TVA IPIP ORPULL TENPERAT WET4ELL TBPERAT CONT TBP RESP RE BRK l02P/SIF - FFWTR

't 50.

500.

CA

'0.

CL-Z:

0 ~

0 ~ 10 50. , 00.

N!N 00202 070%8 1 629.1 TINE (SECONDS)

Figure 7.b-2 Drywell and Wetwell Airspace Temperatures as a Function of Time- 102%

Current Rated Power/100% Core Flow

TVA IPIP ORWELL PRESSLSE NEThELL PRESSNE CONT PRESS RESP RE BRK 102P/81F - FFNTR 60.

't0.

CA CL 20.

PllM 07!09S 0.

00E9A 0.

l 537.5 10 '0.

TIME (SECONDS)

SO. 'l o.

Drywell and Wetwell Pressures as a Function of Time 102% - Uprated Power /100% Core Flow

TVA IPIP ORWELL TEHPERAT NETNELL TEHPERAT CONT TEMP RESP RE BRK 102P/81F - FFNTR 150.

300.

150.

'0 CL uJ CL uJ 0 ~

0. 10 30. 'to.

HIH 071098 00E9A l557o5 (SECONDS) 'IME Drywell and Wetwell Airspace Temperatures as a Function of Time - 102%

Uprated Power /100% Core Flow

NRC Re est 7.c Table 4.1 Provide explanation of 50.3 vs. 50.6 numbers and analysis. State models used: State pre and post model and pre and post uprate values.

This is an error from the October 1, 1997, submittal and the correct value for the peak drywell pressure response at the power uprate condition is 50.6 psig. NRC was notified of this error by letter issued March 20, 1998 (Reference 3).

Please see response to 7.b for containment pressure/temperature analysis models and results for pre-uprate and power uprate conditions NRC Re est 7.d The June 17, 1998, submittal, Item 4, stated that there would be a reduction in the time required to initiate the containment atmospheric dilution system. The submittal stated

"...that CAD initiation will be required between 1 and 2 days...". Please provide a more exact time.

TVA Re 1 7.d The evaluation of time for containment air dilution (CAD) system initiation is based on the UFSAR Section 5.2.6.3 and Figures 5.2-14 and -15. The UFSAR states that the oxygen concentration will remain below the threshold limit of 5%

oxygen for greater than 2 days in the drywell and greater than 1 day in the wetwell. A more exact time for dilution is difficult to ascertain since the figures in the UFSAR are logarithmic. However, operator action is based on oxygen concentration and the 1 to 2 day time frame is sufficient for appropriate operator actions necessary to mitigate the incident.

REFERENCES:

Power U rate Submittals

.1. TVA letter to NRC dated October 1, 1997, "Browns Ferry Nuclear Plant (BFN) Units 2 and 3 Technical Specification (TS) Change TS-384 Request for License Amendment for Power Uprate Operation"

2. TVA letter to NRC dated March 16, 1998, "Browns Ferry Nuclear Plant (BFN) Units 2 and 3 Technical h

Specification (TS) No. 384 Supplement 1 Request for License Amendment for Power Uprate Operation" TVA letter to NRC dated March 20, 1998, "Browns Ferry Nuclear Plant (BFN) Units 2 and 3 Technical Specification (TS) Change TS-384 Request for License Amendment for Power Uprate Operation" TVA letter to NRC dated April 1, 1998, "Browns Ferry Nuclear Plant (BFN) Response to Request for Additional Information (RAI) Regarding Units 2 and 3 Technical Specification (TS) Change TS-384, Request for License Amendment for Power Uprate Operation, (TAC Nos. M99711, M99712) and Resolution of Control Room Emergency Ventilation System (CREVS) Issues (TAC Nos. M83348, M83349, M83350)"

TVA letter to NRC dated April 28, 1998, "Browns Ferry Nuclear Plant (BFN) Response to Request for Additional Information (RAI) Regarding Units 2 and 3 Technical Specification (TS) Change TS-384, Response to the Request for Additional information Relating to License Amendment for Power Uprate Operation, (TAC Nos. M99711, M99712)"

TVA letter to NRC, dated May 1, 1998, "Browns Ferry Nuclear Plant (BFN) Supplemental Response to Request for Additional Information (RAI) Regarding Units 2 and 3 Technical Specification (TS) Change TS-384, Request for License Amendment for Power Uprate Operation, (TAC Nos.

M99711, M99712) and Resolution of Control Room Emergency Ventilation System (CREVS) Issues (TAC Nos. M83348, M83349, M83350)"

TVA letter to NRC dated May 20, 1998, "Browns Ferry Nuclear Plant (BFN) Response to Request for Additional Information (RAI) Regarding Units 2 and 3 Technical Specification (TS) Change TS-384, Request for License Amendment for Power Uprate Operation, (TAC Nos. M99711, M99712)"

TVA letter to NRC dated May 22, 1998, "Browns Ferry Nuclear Plant (BFN) Response to Request for Additional Information (RAI) Regarding Units 2 and 3 Technical Specification (TS) Change TS-384, Request for License Amendment for Power Uprate Operation (TAC Nos. M99711, M99712)"

TVA letter to NRC dated June 12, 1998, "Browns Ferry Nuclear Plant (BFN) . Response to Request for Additional E1-82

Information (RAI) Relating to Units 2 and 3 Technical Specification (TS) Change No. TS-384 Power Uprate Operation (TAC Nos. M99711, M99712)"

10. TVA letter to NRC dated June 17, 1998, "Browns Ferry Nuclear Plant (BFN) Response to Request for Additional Information (RAI) Relating to Units 2 and 3 Technical Specification (TS) Change No. TS-384 Power Uprate Operation (TAC Nos. M99711, M99712)"

TVA letter to NRC dated June 26, 1998, "Browns Ferry Nuclear Plant (BFN) Supplemental Information and Response to Request for Additional Information (RAI)

Relating to Units 2 and 3 Technical Specification (TS)

Change No. TS-384 Power Uprate Operation (TAC Nos.

M99711, M99712)"

12. TVA letter to NRC dated June 26, 1998, "Browns Ferry Nuclear Plant (BFN) Units 2 and 3 Technical Specifications (TS) Change-384 Request for License Amendment for Power Uprate Supplement 2"

.13. TVA letter to NRC dated July 17, 1998, "Browns Ferry Nuclear Plant (BFN) Units 2 and 3 Technical Specifications (TS) Change-384 Request for License Amendment for Power Uprate Supplement 2, Revision 1"

~

Other References

14. American Society of Mechanical Engineers, "Rules for Construction of Nuclear Vessels," ASME Boiler and Pressure Vessel Code,Section III, 1974 Edition with Addenda to and including Summer 1976.

15. American Society of Mechanical Engineers, "Rules for Construction of Nuclear Vessels," ASME Boiler and Pressure Vessel Code,Section III, 1965 Edition with Addenda to and including Summer 1965 (Unit 2) and Summer 1966 (Unit 3).

16. TVA letter to NRC dated February 23, 1996, "Browns Ferry Nuclear Plant (BFN) Units 1, 2, and 3 Adoption of the General Electric (GE) SAFER/GESTR Loss of Coolant Accident Methodology" E1-83

ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 BROWNS FERRY NUCLEAR PLANT (BFN) RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) RELATING TO UNITS 2 AND 3 TECHNICAL SPECIFICATION (TS) CHANGE NO. TS-384 POWER UPRATE OPERATION (TAC NOS. M99711 AND M99712)

EXCERPTS FROM ENVIRONMENTAL QUALIFICATION CALCULATIONS (See Attached)

Calculation Number MD-Q0999-980041 The purpose of this calculation is to determine ifthe accident simulation testing applicable to Limitorque Valve Actuators is more severe than the plant environmental qualification requirements for operation at 3458 MWt. This calculation documents the results derived by the System 1000 computer software for incorporation into the appropriate EQDP.

2.0 References 2.1 TVA Calculation MD-Q0999-980029, Rev. 1 (RIMs PR14 980520 105).

2.2 Fulcrum Group, Inc. 'System 1000 User's Manual', Program Revision 15.2, July 23, 1996.

3.0 Design Input Data The plant postulated accident requirement time and temperature points (Drywell Bounding Temperature Profile) were taken from Reference 2.1. The accident simulation time and temperature points for Limitorque test reports 600376A and B0212 were taken from Appendix B. The activation energy value (Ea) used in the System 1000 "Calculation Input" was taken from Appendix B. The "Aging Temperature" value shown in Appendix A is actually the reference temperature used for the Accident Degradation Equivalency (ADE) calculations. The System 1000 ADE calculation mode compares the "Accident Test" profile and the "Accident Requirement" profiles based on equivalent operating time at the selected reference temperature. The reference temperature selected has no effect when calculating the test margin as a percentage of the accident requirement. The last temperature point from the "Requirements" profile is used to satisfy the System 1000 program's need for a reference temperature.

4.0 Assumptions None.

5.0 Requirements/Limiting Conditions None.

6.0 Computations and Analysis The "Accident Requirements" data points were taken from Reference 2.1 and the simulation time and temperature points were taken from Appendix B. These real time data points were converted into durations with a starting and ending temperature format for input into the System 1000. The calculations are performed internal to the System 1000 computer software. The accident degradation equivalency calculations performed by the System 1000 are based on the following Arrhenius model:

Rev. Preparer/Date Checker/Date Page7of 9

Calculation Number MD-Q0999-980041 n

tA = Z t,/exp((E,/ka)(1/Ty 1/TA))

y=l where:

tA = Equivalent time at temperature TA ty = Time at temperature T exp = exponent to base e E, = Activation energy ka = Boltzman's constant Ty = Accident test temperature TA = Reference (baseline) temperature Since the System 1000 program calculates an area under a curve, a duration of zero in the accident simulation data is meaningless and is not accepted by the System 1000 program as input. Any zero time points were omitted from the following table to establish the "Accident Test" data shown in Appendix A. The actual "Accident Requirements" data and "Accident Test" data input to the System 1000 are included in Appendix A.

LIMITORQVE(RELIANCE MOTOR) TEST REPORT NUMBER 600376A SIMULATION SYSTEM 1000 FORMAT TIME TEMP DUR START END (SEC) ('F) (SEC) TEMP TEMP 162 23 336 23 162 336 240 340 217 336 340 10800 340 10560 340 340 11700 332 900 340 332 12600 325 900 332 325 13500 322 900 325 322 14400 320 900 322 320 21600 320 7200 320 320 25200 252 3600 320 252 27000 249 1800 252 249 345600 249 318600 249 249 345600 190 249 190 2592000 190 2246400 190 190 Rev. Preparer/Date Checker/Date Pagesof 9

Calculation Number MD-Q0999-980041 6.0 Computations and Analysis (continued)

LIMITORQUE (VITON SEAL) TEST REPORT NUMBER B0212 SIMULATION SYSTEM 1000 FORMAT TIME TEMP DUR START END (SEC) ('F) (SEC) TEMP TEMP 120 415 120 415 10 418 415 418 22 425 12 418 425 107 420 85 425 420 190 425 83 420 425 212 427 22 425 427 256 441 44 427 441 278 435 22 441 435 300 400 22 435 400 390 360 90 400 360 540 342 150 360 342 630 324 90 342 324 780 312 150 324 312 2100 305 1320 312 305 21600 252 19500 305 252 1080000 252 1058400 252 252 7.0 Supporting Graphics See Appendix A for the System 1000 calculation plots of the Requirements versus Tests.

8.0 Summary of Results The accident tests were more severe than the plant requirement.

9.0 Conclusion The, plant requirement is satisfied by the accident test for plant operation at 3458 MWt.

Rev. Preparer/Date Checker/Date Page9of 9

Calculation Number MD-Q0999-980041 PAGE Plant Requirement versus Test Time 0 Temperature Plot (RPT8 600376A).... A-2 Test vs. Requirements Input Data and Calculation Results (RPTP 600376A).... A-3 Plant Requirement versus Test Time 5 Temperature Plot (RPTP B0212)........ A-5 Test vs. Requirements Input Data and Calculation Results (RPTI/ B0212)........ A-6 The Peak Accident Transient was deleted ofF the front of both the Plant Profile(s) and Test Profile(s) in order to compare the severity of the Post Accident Period. The same amount of time or more was removed from the test depending on the recorded time points to insure that the Plant Peak Accident Transient was enveloped. 1450 seconds were removed from the Requirements; 1450 seconds were removed from Test 600376A and 2100 seconds were removed from Test B0212..

Rev. Preparer/Date Checker/Date Page A -1 of 6

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FULCRUM GROUP, INC. SYSTEM 1000, REVISION 15.2 Analysis Report, Page 1 Print Date: 06/20/98 Time: 12:32 TEST VS REQUIREMENTS CALCULATION Item

Description:

600376A VS ROOM 0 CALCULATIONINPUT Activation Energy: 1.0200 Aging Temperature: 106.00 F ACCIDENT REQUIREMENTS STARTING ENDING TIME UNITS TfMPERATURE UNITS TEMPERATURE UNITS 17755.00 S 274.00 F 178.00 F 20931.00 S 178.00 F 167.00 F 46264.00 S 167.00 F 153.00 F 1.00 D 153.00 F 141.00 F 3.00 D 141.00 F 130.00 F 5.00 D 130.00 F 119.00 F 30.00 D 119.00 F 112.00 F D 112.00 F 108.00 F D 108.00 F 106.00 F ACCIDENT TEST STARTING ENDING TIME UNITS TEMPERATURE UNITS TEMPERATURE UNITS 9350.00 S F 340.00 F 900.00 S 340.00 F 332.00 F 900.00 S 332.00 F 325.00 F 900.00 S 325.00 F 322.00 F 900.00 S 322.00 F 320.00 'F 7200.00 S 320.00 F 320.00 F 3600.00 S 320.00 F 252.00 F 1800.00 S 252.00 F 249.00 F 318600.00 S 249.00 F 249.00 F 37440.00 M 190.00 F 190.00 F

FULCRUM GROUP, INC. SYSTEM 1000, REVISION 15.2 Analysis Report, Page 2 Print Date: 06/20/98 Time: 12:32 TEST VS REQUIREMENTS CALCULATION Item

Description:

600376A VS ROOIVI 0 CALCULATIONRESULTS: 4292.08%

Prepared By: Date: ~ >>

Reviewed By: Date C zy

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FULCRUM GROUP, INC. SYSTEM 1000, REVISION 15.2 Analysis Report, Page 1 Print Date: 06/20/98 Time: 12:34 TEST VS REQUIREMENTS CALCULATION Item

Description:

B0212 VS ROOM 0 CALCULATIONINPUT Activation Energy: 1.1600 Aging Temperature: 106.00 F ACCIDENT REQUIREMENTS STARTING ENDING TIME UNITS TEMPERATURE UNITS TEMPERATURE UNITS 17755.00 S 274.00 F 178.00 F 20931.00 S 178.00 F 167.00 F 46264.00 S 167.00 F 153.00 F 1.00 D 153.00 F 141.00 F

~"

3.00 D 141.00 F 130.00 F 5.00 D 130.00 F 119.00 F 30.00 D 119.00 F 112.00 F D 112.00 F 108.00 F D 108.00 F 106.00 F ACCIDENT TEST STARTING ENDING TIME UNITS TEMPERATURE UNITS TEMPERATURE UNITS 19500.00 S 305.00 F 252.00 F 17640.00 M 252.00 F 252.00 F CALCULATIONRESULTS: 7710.08%

Prepared B Date: 4 Z Reviewed By; Date: 4 Z3

Calculation Number MD-Q0999-980041 PAGE Accident Simulation data interpretation from test 600376A B-2 Accident Simulation data interpretation from test B0212 .B-3 0

Rev. Preparer/Date Checker/Date Page B-1 of 3

Limitorque Valve Actuator s CALCULATZON

SUMMARY

844 '890606 816 PEA EGR COH T"

'MP

-AC CIO 0

AT N E IVA V ISM ACT RE S

ES:

U I:'.22 ON:

TIVATION ENERGY (eV): 1.02 ve ope with 100 d s 30d s, NCY'70. 760 025 28 d s SI 0 LE ILE AME: . Sl 10 064A 0 OOR 0 (I OEC AZ AC"1 1 S MULATIOH TES RT NO.: 600376A NOTES: RELIANCE ORS f 463489&X, LIMZTORQUE OPERATOR SMB-0 8 135809A qui erne s Simulation im T perat e Time Temperature (s on ) (c) (seconds) (F) {c) 47 63.8 1 162 72 I 2222 295 3 .5 1

1 illl

.611 23 240 336

.340 168.8889 171.1111 61.1 10800 340 171.1111 322 161. 111 11700 332 166.6667 gsso 280 13 .777 12600 325 162.7778

.7 4, 13500 322 161.1111 0 0 14400 320 160 20 200 93 21600 320 160 500 1 25200 252 -

122.2222 7 92 0 82. 27000 249 120.5556 0008 175 7 .4444 .

345600'45600 249 120.5556 12407 17 .67 .15 190 87.7778 148 0 .33 '8

76. 2592000 190 87.7778 17 06 75.

27 165 .26 220 1 .33 2. 11 2 04 7 .6667 47 15 . .37

69. 22 31 2 54 668 15 .67 64 9. 65. 3 3 00 147 .8889 400 14 .88 8IHOER HO: BFH2EQ-MOV<01 FAC-1144, PLOT 1 LOCA (R3.0)

Page B-2 of 3

LIHITORQUE ACTUATOR QUALIFICATION - VITON SEAL cALcULATI0N OUNNABY 844 8906 0 6 8 1 6 ACTIVATION ENERGY (eY): 1.16

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OST CCI T ERA IO RES:

RE IR:

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ILE FILE E:

AHE:

1000 IH90 Cl r

DEG ADA ON E C SER TIS ACT '3 IVAL CY: 446. 97BS

.299 816 41 0 R day 0 NSIQ ONTAI C"113 SIHULATION TEST REPOR NO.: 80212, F E quir ent Simulation i ime mper ure Time Temperature (se nds) (F) (c) (seconds) (F) (c) 147 63.8 9 1 120 48.8889 1 .1111 9 415 212.7778 3 .5 .611 10 418 214.4444 22 61.11 22 425 21S.3333 322 161. 11 107 420 215.5556 330 28 137 8 190 425 218.3333 3 .7 1 .8333 212 427 219.4444 00 0 .3333 256 441 227.2222 2520 200 93. 278 435 223.8B89 500 4 1 85 300 400 204.4444 7 2 0 82 390 360 182.2222 OOOS 75 7 .4444 540 342 172.2222 12407 172 S.15 630 324 162.2222 148 40 17 .33 76. 780 312 155.5556 17 06 8 75. 6 2100 305 151.6667 27 65. 7 .2611 21600 252 122.2222 220 163. .961 1080000 252 122.2222 2 04 16 1.6 2 470 ,67 70. 22 56.33 .0722 3166 154 .777 3 68 15 66.

3 4734 .33 65.

8880 47 889 8640 0 147 .8 C

'I BINDER NO: BFN2EQ+OV<01 FAC"1144. PLOT 2 LOCA (R3.0)

Page B-3 of 3

Calculation Number MD-Q0999-980068 1.0 Purpose The purpose of this calculation is to determine ifthe accident simulation testing applicable to Honeywell OP-AR and OPD-AR limit switches is adequate to satisfy the plant environmental qualification requirements for operation at 3458 MWt. This calculation will document the results derived by the System 1000 computer software for incorporation into EQDP BFNOEQ-IZS-003.

2.0 References 2.1 TVA Calculation MD-Q0999-980104 (RIMS PR14 980520 104).

2.2 Fulcrum Group, Inc. 'System 1000 User's Manual', Program Revision 15.2, July 23, 1996.

3.0 Design Input Data The plant postulated accident requirement time and temperature points for Outside Primary Containment Temperature Profiles were taken from Reference 2.1. The accident simulation time and temperature points for Honeywell Limit Switch test report CCL A-686-85 were taken from Appendix B. The activation energy value (Ea) used in the System 1000 "Calculation Input" was taken from Appendix B. The "Aging Temperature" value shown in Appendix A is actually the reference t temperature used for the Accident Degradation Equivalency (ADE) calculations. The System 1000 ADE calculation mode compares the "Accident Test" profile and the "Accident Requirement" profile based on equivalent operating time at the selected reference temperature (Reference 2.2). The reference temperature selected has no effect when calculating the test margin as a percentage of the accident requirement. The last temperature point from the "Requirements" profile is used to satisfy the System 1000 program's need for a reference temperature.

4.0 Assumptions None.

5.0 Requirements/Limiting Conditions None.

6.0 Computations and Analysis The "Accident Requirements" data points were taken from Reference 2.1 and the "Accident Test" time and temperature points were taken from Appendix B. These real time data points were converted into durations with a starting and ending temperature format for input into the System 1000. The calculations are performed internal to the System 1000 computer software. The accident degradation equivalency calculations performed by the System 1000 are based on the following Arrhenius model:

Rev. Preparer/Date Checker/Date Page 7 of 9

Calculation Number MD-Q0999-980068 6.0 Computations and Analysis (continued) n t~= Z tgexp((E,/ka)(1/Ty 1/TA))

y=l where:

tA Equivalent time at temperature TA ty Time at temperature T exp exponent to base e E, Activation energy ka Boltzman's constant Tg Accident test temperature TA Reference (baseline) temperature Time durations of zero in the accident simulation data contribute nothing to the calculated area under a curve and are not accepted by the System 1000 program as input. These zero time points were omitted from the following table to establish the "Accident Test" data shown in Appendix A. The actual "Accident Requirements" data and "Accident Test" data input to the System 1000 are included in Appendix A.

CONVERSION OF ACCIDENT TEST DATATO SYSTEM 1000 FORMAT TEST REPORT CCL A-686-85 SIMULATION SYSTEM 1000 FORMAT TIME TEMP DUR START END (SEC) ('F) TEMP TEMP 1.0 303 6.6 309 6.6 303 309 10.0 330 3.4 309 330 30.0 337 20.0 330 337 136.6 347 106.6 337 347 210.0 316 73.4 347 316 2674.4 225 2464.4 316 225 3600.0 219 925.6 225 219 18000.0 219 14400.0 219 219 86400.0 165 68400.0 219 165 929405.6 165 843005.6 165 165 929405.6 171 0.0 165 171 1304074.8 171 374669.2 171 171 1304074.8 165 0.0 171 165 2592000.0 165 1287925.2 165 165 Rev. Preparer/Date Checker/Date Page 8 of 9

Calculation Number MD-Q0999-980068 7.0 Supporting Graphics See Appendix A for the System 1000 plots of the Requirements versus Test.

8.0 Summary of Results The accident test was more severe than the plant requirements.

9.0 Conclusion The plant requirements for operation at 3458 MWt. are satisfied by the accident test.

Rev. Preparer/Date Checker/Date Page 9 of 9

Calculation Number MD-Q0999-980068 PAGE Plant Requirement versus Test Time & Temperature Plot A-2 Test vs. Requirements Input Data and Calculation Results . A-3 The Peak Accident Transient was deleted offthe front of both the Plant Profile(s) and Test Profile(s) in order to compare the severity of the Post Accident Period. The same amount of time or more was removed from the test depending on the recorded time points to insure that the Plant Peak Accident Transient was enveloped. 1500 seconds was removed from the Requirements and 2674.39 seconds was removed from the Test.

Rev. Preparer/Date Checker/Date Page A-1 Of 3

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Time (Seconds)

Item

Description:

BFNOEQ-IZS-003

FULCRUM GROUP, INC. SYSTEM 1000, REVISION 15.2 Analysis Report, Page 1 Print Date: 06/20/98 Time: 10:32 TEST VS REQUIREMENTS CALCULATION Item

Description:

CCL A-686-85 VS OPC EXC RM 7 CALCULATIONINPUT Activation Energy: 0.7500 Aging Temperature: 135.00 F ACCIDENT REQUIREMENTS STARTING ENDING TIME UNITS TEMPERATURE UNITS TEMPERATURE UNITS 500.00 S 195.00 F F 1600.00 S 190.00 F F 18000.00 S 155.00 F F 2394.00 H 135.00 F F ACCIDENT TEST STARTING ENDING TIME UNITS TEMPERATURE UNITS TEMPERATURE UNITS 925.61 S 225.00 F 219.00 F 14400.00 S 219.00 F 219.00 F 68400.00 S 219.00 F 165.00 F 843005.61 S 165.00 F 165.00 F 374669.19 S 171.00 F 171.00 F 21465.42 M 165.00 F 165.00 F CALCULATIONRESULTS: 20.29%

Prepared By: Date: 6 a Reviewed By: Date:

Calculation Number MD-Q0999-980068

\

PAGE Accident Simulation data interpretation from test CCL A-686-85. .B-2 Rev. Preparer/Date Checker/Date Page 8-1 of 2

Limit SMitch Qualification 844 'Q6 P Qg CALCULATION

SUMMARY

TIVATION ENERGY (eV): 0.75 TU S: Env ope PEAK POS

-A P

ACC ENT I S GUI: ON:

1 days days H FILE FILE N 'GN0800

':SIH 3 GRA ATIO EGUIVA NCY: 41.758 0474 day ON .RVAT M FACT R .41 585004 5 a R D EN 1 LO OOHS EXC T 0 8 7, AC-1138

. '.ELATION REPORT NO.:. A&86-85, Dated 02/14/86 rR.q Ti rement econds (F)

Te erature (C)

Simulation Time (seconds) (F)

Temperature (c) 1 135 57. 1. 303 150.5556 1.0 13 137. 9 58.4 6.6 309 153.8889

1. 216 102 667 10 330 165.5556 2 .66 .5889 30 337 169.4444 6.8 2.6667 i3i.6 347 175 3 16.89 102.7167 210 316 157.7778 4 217.01 102.7 2674.39 225 107.2222 4 494 217.0 9 102.8 7 3600 219 103.8889

.5 219. 104. 67 18000 219 103.8889 5 .65 10 .1389 86400 165 73.8889 5.5 2.94 1.6333 929405.61. 165 73.8889 6 8.2 14.5556 929405.61 171 77.2222 6 242.7 117.08 1304074.8 171 77.2222

~ 246. 119.2 8 1304074.8 165 73.8889 7.5 25 .11 121. 22 2592000 165 73.8889 8'.5 1 .7778

.35 4.0833 9 257.24 25.1333 9 258.76 125.977 259. 126. 9 13.181 261 33 127 241 15 .31 13 .95 20 .22 .5667 287.87 142.15 286.2 141.2444 35 286. 141.53 40 28 .83 143.

45 2 .32 142 6 50 86.61 1 .45 278.79 7.1056 264.7 129.311 62.2767 258 352 125.79 65 25 .42 124. 8 70 .06 1 .Bill gS 75 49.96 1 .0889 8 253.35 22.97 256.1 124.

90 4 125

'57.

95 .28 1 .1556 100 4.41 .5611 BINOER NO: BFN2EQ-IZS<03 FAC-ii77, PLOT i LOCA (R3.0) PAGE 1 OF 5 Page B-2 of 2

Calculation Number MD-Q0999-980056 1.0 Purpose The purpose of this calculation is to determine ifthe accident simulation testing applicable Limitorque DC Operators is adequate to satisfy the plant environmental qualification requirements for operation at 3458 MWt. This calculation will document the results derived by the System 1000 computer software for incorporation into EQDP BFNOEQ-MOV-003.

2.0 References 2.1 TVA Calculation MD-Q0999-980104 (IUMS ¹R14 980520 104).

2.2 Fulcrum Group, Inc. 'System 1000 User's Manual', Program Revision 15.2, July 23, 1996.

3.0 Design Input Data The plant postulated accident requirement time and temperature points for Outside Primary Containment Temperature Profiles were taken from Reference 2.1. The accident simulation time and temperature points for Limitorque test reports B0212 and B-0009 were taken from Appendix B. The activation energy values (Ea) used in the System 1000 "Calculation Input" was taken from Appendix B.

The "Aging Temperature" value shown in Appendix A is actually the reference temperature used for the Accident Degradation Equivalency (ADE) calculations. The System 1000 ADE calculation mode compares the "Accident Test" profile and the "Accident Requirement" profile based on equivalent operating time at the selected reference temperature. The reference temperature selected has no effect when calculating the test margin as a percentage of the accident requirement. The last temperature point from the "Requirements" profile is used to satisfy the System 1000 program's need for a reference temperature.

4.0 Assumptions None.

5.0 Requirements/Limiting Conditions None.

6.0 Computations and Analysis The "Accident Requirements" data points were taken from Reference 2.1 and the "Accident Test" time and temperature points were taken from Appendix B. These real time data points were converted into durations with a starting and ending temperature format for input into the System 1000. The calculations are performed internal to the System 1000 computer software. The accident degradation equivalency calculations performed by the System 1000 are based on the following Arrhenius model:

Rev. Preparer/Date Checker/Date Page 7 of 9

Calculation Number MD-Q0999-980056 6.0 Computations and Analysis (continued) n tA= Z tgexp((E,/ka)(1/T~1/TA))

y=1 where:

4 = Equivalent time at temperature TA ty = Time at temperature T exp = exponent to base e E, = Activation energy kB = Boltzman's constant T)p = Accident test temperature TA = Reference (baseline) temperature Time durations of zero in the accident simulation data contribute nothing to the calculated area under a curve and are not accepted by the System 1000 program as input. These zero time points were omitted from the following table to establish the "Accident Test" data shown in Appendix A. The actual "Accident Requirements" data and "Accident Test" data input to the System 1000 are included in Appendix A.

CONVERSION OF ACCIDENT TEST DATATO SYSTEM 1000 FORMAT TEST REPORT B0212 SIMULATION SYSTEM 1000 FORMAT TIME TEMP DUR START END (SEC) ('F) TEMP TEMP 1 120 9 415 9 120 415 10 418 1 415 418 22 425 12 418 425 107 420 85 425 420 190 425 83 420 425 212 427 22 425 427 256 441 44 427 441 278 435 22 441 435 300 400 22 435 400 390 360 90 400 360 540 342 150 360 342 630 324 90 342 324 780 312 150 324 312 2100 305 1320 312 305 21600 252 19500 305 252 1080000 252 1058400 252 252 Rev. Preparer/Date Checker/Date Page 8 of 9

Calculation Number MD-Q0999-980056 6.0 Computations and Analysis (continued)

TEST REPORT B0009, SMULATION SYSTEM 1000 FORMAT TIME TEMP DUR START END (SEC) ('n TEMP TEMP 1 120 3600 340 3600 120 340 7200 340 3600, 340 340 7200 330 0 340 330 14400 330 7200 330 330 14400 310 0 330 310 25200 310 10800 310 310 212 0 310 212 '5200 90000 212 64800 212 212 7.0 Supporting Graphics t See Appendix 8.0 A for the System 1000 plots Summary of Results of the Requirements versus Tests.

The accident tests were more severe than the plant requirements.

9.0 Conclusion The plant requirements. for operation at 3458 MWt. are satisfied by the accident tests.

Rev. Preparer/Date Checker/Date Page 9 of 9

Calculation Number MD-Q0999-980056 PAGE OPC Bounding Profile versus Test B0212 time Ec temperature plot .............. ...A-2 Test B0212 vs. OPC Bounding Input Data and Calculation Results .A-3 OPC Bounding Profile versus Test B0009 time A temperature plot .............. ...A-5 Test B0009 vs. OPC Bounding Input Data and Calculation Results .............. ...A-6 k

Room 7 (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) Bounding versus Test B0009 time temperature plot ... ...A-7 Test B0009 vs. Room 7 (first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) Input Data and Calculation Results . ...A-8 The Peak Accident Transient was deleted offthe front of both the Plant Profile(s) and Test Profile(s) in order to compare the severity of the Post Accident Period. The same amount of time or more was removed from the test depending on the recorded time points to insure that the Plant Peak Accident Transient was enveloped. 50 seconds were removed from the OPC Requirements and 107 seconds were removed from Test B0212. 50 seconds were removed from the OPC Requirements and 3600 seconds were removed from Test B0009. 50 seconds were removed from Room 7 Requirements and 3550 seconds were removed from Test B0009 entry that was constant at 340'F.

Rev. Preparer/Date Checker/Date Page A-1 of 8

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FULCRUM GROUP, INC. SYSTEM 1000, REVISION 15.2 Analysis Report, Page 1 Print Date: 06/21/98 Time: 07:51 TEST VS REQUIREMENTS CALCULATION Item

Description:

B0212 VS OPC CALCULATIONINPUT Activation Energy: 1.1600 Aging Temperature: 180.00 F ACCIDENT REQUIREMENTS STARTING ENDING TIME UNITS TEMPERATURE UNITS TEMPERATURE UNITS 3550.00 S 290.00 F 210.00 F 18000.00 S 210.00 F 180.00 F 2394.00 H 180.00 F 180.00 F ACCIDENT TEST STARTING ENDING TIME UNITS TEMPERATURE UNITS TEMPERATURE UNITS S 420.00 F 425.00 F S 425.00 F 427.00 F S 427.00 f 441.00 F S 441.00 F 435.00 F S 435.00 F 400.00 F S F 360.00 F 150.00 S 360.00 F 342.00 F S 342.00 F 324.00 F 150.00 S 324.00 F 312.00 F 1320.00 S 312.00 F 305.00 F 19500.00 S 305.00 F 252.00 F 17640.00 M 252.00 F 252.00 F

FULCRUM GROUP, INC. SYSTEM 1000, REVISION 15.2 Analysis Report, Page 2 Print Date: 06/21/98 Time: 07:51 TEST VS REQUIREMENTS CALCULATION Item

Description:

B0212 VS OPC CALCULATIONRESULTS: 596.86%

Prepared By: Date: 5 Reviewed By: Date:

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~~

H- Q Time (Seconds)

Xylem Descri peion: B=NOEQ-HOV-003

FULCRUM GROUP, INC. SYSTEM 1000, REVISION 15.2 Analysis Report, Page 1 Print Date: 06/21/98 Time: 07:52 TEST VS REQUIREMENTS CALCULATION Item

Description:

B0009 VS OPC CALCULATIONINPUT Activation Energy: 1.6300 Aging Temperature: 180.00 F ACCIDENT REQUIREMENTS STARTING ENDING TIME UNITS TEMPERATURE UNITS TEMPERATURE UNITS 3550.00 S 290.00 F F 18000.00 S 210.00 F F 2394.00 H 180.00 F F ACCIDENT TEST STARTING ENDING TIME UNITS TEMPERATURE UNITS TEMPERATURE UNITS 3600.00 S 340.00 F F 7200.00 S 330.00 F F 10800.00 S 310.00 F F 64800.00 S 212.00 F F CALCULATIONRESULTS: 3885.05%

Prepared B Date: 0 By:

'eviewed Date: 0 Z> j~

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FULCRUM GROUP, INC. SYSTEM 1000, REVISION 15.2 Analysis Report, Page 1 Print Date: 06/21/98 Time: 10:24 TEST VS REQUIREMENTS CALCULATION Item

Description:

B0009 VS 24 HOURS IN ROOM 7 CALCULATIONINPUT Activation Energy: 0.8000 Aging Temperature: 180.00 F ACCIDENT REQUIREMENTS STARTING ENDING TIME UNITS TEMPERATURE UNITS TEMPERATURE UNITS 50.00 S F 225.00 F S 225.00 F 210.00 F 2600.00 S 210.00 F 210.00 F t

18000.00 S 210.00 F 180.00 F 18.02 H 180.00 F 180.00 F ACCIDENT TEST Test Name: LIMITORQUE Report Number: B0009 STARTING ENDING TIME UNITS TEMPERATURE UNITS TEMPERATURE UNITS 3600.00 S 120.00 F 340.00 F 3550.00 S 340.00 F 340.00 F 7200.00 S 330.00 F 330.00 F 10800.00 S 310.00 F 310.00 F S 212.00 F 212.00 F II CALCULATIONRESULTS: 2461.44%

Prepared B Date: Co ~

Reviewed By: Date: 4 >3 II

Calculation Number MD-Q0999-980056 PAGE Accident Simulation data interpretation from test B0212 .B-2 Accident Simulation data interpretation from test B0009 .B-3 0

Rev. Preparer/Date Checker/Date Page B-1 of 3

LIMITORGUE ACTUATOR QUALIFICATION VITON SEAL ~ a CALCULATION SUMMARy C .SY K. BY~ ~Q 4

ACTIVATION eV ES: nv

'AT oped ith M GIN N tC 83 1,3, 6.7, 9.10, 1,19,2 T CCI I day at L IHULATION TEST 021 FI6URE 2 SI DE A IN'N OH E Y.

TI 12. days 7.3 1797 day at 180F HAHE: SI}80 IO QUIY MARS 4 .6618 486 ays AT FA OR .84 91 R uir ents Simulation ime T mpera re Time Temper ature (se onds) (F) (C), (seconds) (F) (C) 1 151. 1 120 48.8889 25 149 44 9 415 212.7778

.1 1 .1111 10 418 214.4444 0 .111 22 425 218.3333 125 135 107 420 215.5556 20.8 110. 78 190 425 218.3333 8 212 427 219.4444 40 4, 256 441 227.2222 600 04. 278 435 223'.8889

.1 101. 300 400 204.4444 4400 15 10 .6667 390 360 182.2222 2160 180 .2222 540 . 342 172.2222 180 630 324 162.2222 780 312 155.5556 2100 305 151.6667 21600 252 122.2222 1080000 252 122.2222 IOp 00 BINDER NO: BFNOE(HSY<03 MHN~20125 Ri, FIBSE B.l ADEPLT IO.O)

Page B-2 of 3

l LIHITORQUE OC OPERATORS OUTSIDE CONTAINMENT HEET CALCULATION

SUMMARY

CM. BY K. BY ACTIVATION ENERGY (eV)

P K T HP ATUR 've 1.63 ped wi HARG I AL HO: L i. 5 6 7 9,10 1.19,20

-A IDEHT QU Tr 100 s FILE '

NLATIOH TEST HO.: 8-0009. DATED ril 30. 1976 OE H.

TI A

QUIV.

E . 10 .

.041 GIH: 911.8 23747 days d sat days OF SIH 'S 1118 CO VAT FACT 84 812 7 X H

' OLIC LACK TERH BL Bequ'ments Simulation Ti Te perat e Time Temperature econds) (F) (c (seconds) (F) (c) 1 305 .6667 . 120 48.8889 25 30 49.4444 3600 340 171.1111

.1 146.11 7200 340 171.1111 00 146.1 1 7200 330 165.5556 125 135 14400 330 165.5556 20 .78 .6278 14400 310 154.4444 12.7778 25200 310 154.4444 40 0 104. 25200 212 100 600 220 104. 90000 212 100

.1 215 101 4400 2 1 .6667 2160 0 00 180 82.

N0 ~-OO3 BINDER HO: BFHOEQM>'-"" HtH)09~20125 Ri, FI6URE 8.2 ADEPL T IO.O) Page B-3 of 3