L-99-026, Forwards Response to NRC 990702 RAI Re SG Replacement Related TS Change Request Submitted 981201.Ltr Contains No New Commitments

Forwards Response to NRC 990702 RAI Re SG Replacement Related TS Change Request Submitted 981201.Ltr Contains No New Commitments

ML20210C520
Person / Time
Site:	Farley
Issue date:	07/19/1999
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NEL-99-0269, NUDOCS 9907260080
Download: ML20210C520 (15)

Text

' I j Dave M: rey So1thern Nuclear Vice President Op:r: ting Company. Inc.

Farley Project Post Office Box 1295 g i  ! Birmingham. Alabama 35201  !

Tel 205 992.5131 July 19, 1999 SOUTHERN COMPANY Energy ro Serve nur World" Docket Nos.: 50-348 NEL-99-0269 '

50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant Response to Request for Additional Information Steam Generator Reolacement Related Technical Soecifications Channe Request Ladies and Gentlemen:

By letter dated December 1,1998, Southem Nuclear Operating Company (SNC) submitted a Technical Specifications change request related to the replacement of steam generators (SG) at Farley Nuclear Plant.

A revision to this change package was also submitted on April 21,1999. Your July 2,1999 letter requested additional information in order to complete your review of our submittal. In Attachment 1, SNC provides the additional information requested. ,

There are no new commitments in this response. If you have any questions, please advise.

Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY Dave Morey Sworn to andsubscri before me this 8 day of 1999 0

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' Notary Public / (/ ~

bb My Commission Erpires: 0?' k l>S00l

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CHM / Letter SGR RAI 1. doc Attachments 1. SNC Response to RAI Dated July 2,1999

2. Computer Disk (3.5") with RETRAN Input Deck

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990-[250050 990719 PDR ADOCK 050003 8 P -

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'-0 Page 2 U. S. Nuclear Regulatory Commission cc: Southern Nuclear Operatina Company Mr. L. M. Stinson, General Manager - Farley U. S. Nucle ir Regulatory Commission. Washincton. D. C.

Mr. L. M. Padovan, Licensing Project Manager - Farley U. S. Nuclear Regulatory Commission. Region 11 Mr. L. A. Reyes, Regional Administrator Mr. T. P. Johnson, Senior Resident Inspector - Farley f

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e Attachment 1 Joseph M. Farley Nuclear Plant Response to Request for Additional Information Steam Generator Replacement Related Technical Specifications Change Request (NRC Letter Dated July 2,1999)

Attachment 1 Page1of i1 Response to RAI Dated July 2,1999 i

Joseph M. Farley Nuclear Plant Response to Request for AdditionalInformation Steam Generator Replacement Related Technical Specifications Change Request (NRC Letter Dated July 2,1999)

NRC RAl-1 By letter dated February 11,1999, WCAP-14852-P, "RETRAN-02 Modeling and Qualification Westinghouse Pressurized Water Reactor Non-LOCA [ loss-of-coolant-r accident] Safety Analysis," was accepted for referencing in licensing applications to the exter.t specified and under the limitations delineated in the report and in the associated NRC safety evaluation. Please address each of the conditions delineated in the report and in the conclusion section of the NRC's Safety Evaluation for WCAP-14882-P.

SNC Response to RAl-1 The NRC staff concludes in the safety evaluation report (Reference 1) that the "use of RETRAN as described in WCAP-14882 is acceptable for licensing calculations and RETRAN may be used to replace the L OFTRAN computer code in Westinghouse reload methodology provided that the following conditions are met:

1. The transients and accidents that Westinghouse proposes to analyze with RETRAN are listed in this SER (Table 1) and the NRC staff review of RETRAN usage by Westinghouse was limited to this set. Use of this code for other analytical purposes will require additionaljustification.

2. WCAP-14882 describes modeling of Westinghouse designed 4,3, and 2-loop plants of the type that are currently operating. Use of the code to analyze other designs, including the Westinghouse AP600, will require additionaljustification.

3. Conservative safety analyses using RETRAN are dependent on the selection of conservative input. Acceptable methodology for developing plant-specific input is I discussed in WCAP-14882 and in WCAP-9272-P-A. Licensing applications using RETRAN should include the source of and je ification for the input data used in the j

analysis." j Each of these conditions are addressed below, as they cdate to the Farley Model 54F Replacem'ont Steam Generator Program.

1. The non-LOCA transients explicitly analyzed with RETRAN for this program include the following: steam system piping failures, loss of offsite power, loss of normal feedwater flow, and feedwater system pipe break. All of these events are listed in Table 1 of the SER; therefore, no additional justification is required.

2. Farley Nuclear Plant Units 1 and 2 are 3-loop, Westinghouse-designed, pressurized water reactors that are currently in commercial operation. Therefore, no additional justification is required.

Attachment 1 Page 2 of 11 Response to RAI Dated July 2,1999 i I SNC Response to RAI-1 (continued)

3. The non-LOCA RETRAN analyses were performed in accordance with the methodologies discussed in WCAP-14882-P-A (Reference 2) and WCAP-9272-P-A (Reference 3). Using these methodologies insures that the analysis conservatively bounds operations at Farley.

References

1. USNRC Letter, " Acceptance for Referencing of Licensing Topical Report WCAP-14882,

'RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analysis' (TAC NO. M99107)," Akstulewicz, F. (USNRC) to Sepp, H. (LV), February 11,1999.

2. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses," Huegel, D. S., et al., April 1999.

3. WCAP-9272-P-A, " Westinghouse Reload Safety Evaluation Methodology," Bordelon, F.

M., et al., Approved .luly 1985.

NRC RAI-2 Please provide an electronic copy of the input deck used in the RETRAN-02 analyses of non-LOCA transients performed in support of the steam generator (SG) replacement.

SNC Response to RAI-2 The requested input deck is provided on the attached diskette (Attachment 2). The file was also forwarded to the NRC via email.

NRC RAI-3 I We understand that departure from nucleate boiling ratio (DNBR) was evaluated in RETRAN using a partial derivative method as discussed in WCAP-14882-P. Provide values for the partial derivatives used and justify that these values are conservative for DNBR analysis of Farley.

SNC Response to RAI-3 No event analyzed for the Farley RSG Program used the RETRAN model to calculate minimum DNBR values.

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! Attachment i Page 3 of 11 Response to RAI Dated July 2,1999 i

4 NRC RAI-4 in Section 2.1.2.1,"LOCA Forces," provide a description how Leak Before Break was applied to generate the LOCA forces.

SNC Response to RAI-4 The LOCA hydraulic forcing functions (LHFF) and loads that occur as a result of a postulated LOCA are calculated assuming a limiting break location and break area. The NRC's revision to GDC-4 allowed main coolant piping breaks to be " excluded from the design basis when analyses reviewed and approved by the commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping." This exemption is generally referred to as " leak-before-break."

The analysis presented in WCAP-12825 (Reference 1) is technicaljustification for eliminating primary loop pipe mptures from the design basis for Farley Units 1 and 2. The applicability of a LBB design basis eliminating primary loop piping breaks for Farley Units I and 2 was approved by the NRC staff (Reference 2). SNC went on to pursue elimination of the pressurizer surge line from consideration with WCAP-12835 (Reference 3). The NRC reviewed and approved SNC's request to eliminate the pressurizer surge line (Reference 4).

Thus, the primary loop piping and pressurizer surge line breaks did not need to be considered when generating the Farley Units I and 2 LOCA hydraulic forces. The breaks that were considered were the accumulator and RHR line breaks.

References

1. WCAP-12825," Technical Justification For Eliminating Large Primary Loop Pipe Rupture As The Structural Design Basis For Joseph M. Farley Units 1 And 2 Nuclear Power Plants."

2. NRC Letter Dated August 12,1991, " Safety Evaluation Of Elimination Of Dynamic Effects Of Postulated Primary Loop Pipe Ruptures From Design Basis For Joseph M.

Farley Units 1 And 2 (TAC NOS. 79660 And 79661)."

3. WCAP-12835,"rechnical Justification For Eliminating Pressurizer Surge Line Rupture From The Structural Design Basis For Farley Units 1 And 2."

4. NRC Letter Dated January 15,1992," Safety Evaluation Of Elimination Of Dynamic Effects Of Postulated Pipe Ruptures In The Pressurizer Surge Line From Structural Design Basis For Joseph M. Farley Nuclear Plant Units 1 And 2 (TAC NOS. M80367 And M80368)."

Attachment i Page 4 of 11 Response t.; RAI Dated July 2,1999 a

NRC RAl-Sa in Section 2.1.2.2.1, " Method of Analysis" Provide verification that the damping used in the time-history seismic analysis was based on that specified in Regulatory Guide (RG) 1.61.

SNC Response to RAI-Sa Per the Farley FSAR, Section 3, Appendix A, Regulatory Guide 1.61 is intended to apply to nuclear power plants docketed after April 1,1973; consequently, RG 1.61 was not considered applicable to Farley. The damping values used in this analysis are based on the values listed in FSAR Table 3.7-1, which are based on a paper by N. M. Newmark and W. J. Hall, "Scismic Design Criteria for Nuclear Reactor Facilities," and another paper by N. M.

Newmark, " Design Criteria for Nuclear Reactors Subjected to Earthquake Hazards." These values are consistent with plant specific seismic input. ASME Code Case N-41I was not used. He damping values used in this analysis are more conservative than required by Regulatory Guide 1.61.

NRC RAI-Sb in Section 2.1.2.2.1," Method of Analysis" Indicate if the seismic analysis of the Reactor Coolant Loop model was performed with all 15 steam generator snubbers removed.

SNC Response to RAI-Sb The seismic analysis of the Reactor Coolant Loop with the RSG was performed with all 15 steam generator snubbers removed.

NRC RAI-6a in Section 2.1.2.2.4, "RCL Supports," and Table 2.1-4: Provide the basis and the values for the Faulted Condition allowable load or stress in compression for the SG columns, the Reactor Coolant Pump (RCP) Columns and the RCP tie-rods.

SNC Regonse to RAI-6a For the faulted condition, stress and interaction equations are solved for each member for the combined load. RCP and SG columns take tension and compression loads; however, the RCP tie-rods only take tension loads. The value for the faulted conditions allowable values, P, & Pm are given below.

The interaction and stress equations use.d are similar to the equations used for the normal condition,but they are modified ts te.flect faulted condition limits as described in AISC-69, Part 2, and shown below.

I Attachment 1 Page 5 of 11 Response to RAI Dated July 2.1999 i

SNC Response to RAl-6a (continued) e Buckling interaction equation P

-+ C", M s 1.0 (1 P,) M .

. Yield interaction equation P M

-P,+ 1.18 M, s 1.0 M S M, Notation:

P= applied axialload.

P, = axial force at member yield. This value is 2196 kips for the SG columns, and 2196 kips for the RCP columns.

P, = 2g2 A F, (where F,' is defined in AISC-69 Section 1.6.1)

P, = 1.7 A F, This value is 1964 kips for the SG columns, and 2063 kips for the RCP columns.

F, = allowable compressive stress if member is subjected to compression loads only. F, is a function of the member slenderness ratio, effective buckling length, and material properties.

A= cross sectional area of member.

M = applied moment.

i C,,, = coefficient defined in Section 1.6.1, AISC-69 (conservat vely taken to be 0.85) .

M,= plastic moment of member.

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Attachment i Page 6 of 11 Response to RAI D:ted July 2,1999 I '

SNC Response to RAl-6a (continued)

M, = maximum moment that can be resisted by the member in the absence of axial load:

Af. = Af, For columns braced in the weak direction, i ,

M,= 1.07 '

g Af"sAf#

For columns un-braced in the weak direction, 3160 1= member length.

I r, = minimum radius of gyration.

NRC RAI-6b In Section 2.1.2.2.4, "RCL Supports," and Table 2.1-4: Provide the largest compressive load acting on SG columns, the RCP columns and the RCP tie-rods.

SNC Response to RAl-6b l l

The SG columns have a maximum compressive load of 1,452 kips. The RCP columns have a j maximum compressive load of 826 kips. The RCP tie-rods are tension only members. They do not take any compression load due to gaps at the tie-rod pins.

s NRC RAI-6c In Section 2.1.2.2.4, "RCL Supports," and Table 2.1-4: For the Reactor Vessel Support ,

Structure, provide the limiting load or stress for the support structure under Faulted Condition I compressive loads, in accordance with American Society of Mechanical Engineers (ASME)

Section 111, Subsection NF and Appendix F.  ;

i SNC Response to RAl-6c The design code for the Farley plant is AISC Specification for the Design, Fabrication and Erection of Structural Steel for Buildings - 1969, therefore, ASME Section Ill, Subsection NF and Appendix F do not apply to the Farley Reactor Vessel Supports. Under AISC the maximum permissible load is 3,400 kips per Reactor Vessel Support, in the vertical direction.

He maximum actual load is 1,092 kips.

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Attachment 1 Page 7 of 11 Response to RAI Dated July 2,1999 NRC RAI-7 In Section 2.1.2.2.5,"RCL Equipment Nozzle Load Evaluation," provide the comparison of the RCL primary equipment nozzle loads to the umbrella allowable loads given in the equipment design specification.

SNC Response to RAI-7 The comparison was done by comparing actual versus allowable stress intensities (SI) from nozzle loadings and was done for every case. The comparison was not done between actual load vs. umbrella load. Table I represents a summary of the equipment nozzle evaluation.

Table 1 Farley Equipment Nozzle Load Evaluation for the RSG Snubber Elimination Nozzle Loading Actual Allowable Ratio SI SI RPVIN OBE 2 4.2 0.48 SSE+ MAX (LOCA,MSB,RVB) 23.5 26.8 0.88

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RPVON OBE 4.5 4.9 0.92 SSE+ MAX (LOCA,MSB,RVB) 21.4 30.9 0.69  !

SGIN OBE 4.7 14.4 0.33 SSE 6.3 18.7 0.34 MAX (LOCA,MSB,FWB) 12.8 57.2 0.22 SGON OBE 2.4 10.9 0.22 SSE 3.2 15.2 0.21 7 MAX (LOCA MSB,RVB) 10.9 76.9 0.14 RCPIN OBE 1.8 9.3 0.19 SSE 2.4 13.7 0.18 l MAX (LOCA,MSB,FWB) 9.4 77.9 0.12 RCPON OBE 2.4 13.1 0.18 i SSE 3 4.9 0.61 l MAX (LOCA,MSB,RVB) 15.3 26.4 0.58

! Attachment 1 Page 8 of 11 l

Response to RAI Dated July 2,1999 a

f NRC RAI-8 What values did you use in the dose analyses for the reactor coolant system mass and volume?

SNC Response to RAI-8 The RCS mass is 410,000 lbm and the CVCS mass is 30,900 lbm for a total mass of 440,900 4

lbm. The RCS volume is 1.02 x 10 ft'.

NRC RAI-9 What values did you use in the dose analyses for the mass and/or volume initially in the steam generators?

SNC Response to RAl-9 The initial steam generator water mass is 168,000 lbm. The initial steam generator volume is 2,700 ft'.

NRC RAI-10 Did you address SG tube uncovery in the locked rotor accident?

SNC Response to RAI-10 SNC did not directly address tube uncovery for a locked rotor accident. This issue was discussed during power uprate with the NRC staff. The following response was provided to the NRC during the Power Uprate project under an SNC letter dated April 13,1998.

NRC Ouestion No. 2 (Reference Anril 9 & 13.1998 NRC/SNC Conference Call)

With respect to Farley power uprate analyses and alternate repair criteria (ARC) and the steam generator tube uncovery program analyses and conclusion presented in WCAP-13247 for the RCP locked rotor and control rod 6jection events, is Farley considered to be a representative plant?

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Attachment 1 Page 9 of 11 Response to RAI Dated July 2,1999 I I SNC Response to RAI-10 (continued)

SNC Response to Ouestion No. 2 The issue of tube bundle uncovery was considered in a generic Westinghouse Owners Group (WOG) program as presented in WCAP-13247, " Report on the Methodology for Resolution of the Steam Generator Tube Uncovery Issue," March 1992. The program concluded that the effect of tube uncovery is essentially negligible for the limiting SGTR transient. It also concluded that for non-SGTR events, such as locked rotor and rod ejection, the probability that an event could result in off-site radiological consequences that exceed the acceptance limits was estimated to be sufficie :tly low so as to place this issue on the exclusion category as defined in NUREG-0933. Therefore, the program concluded that the steam generator tube uncovery issue could be closed without any further investigation or generic restrictions. The NRC review of the WOG program concluded Nhe Westinghouse analyses demonstrate that the effects of partial steam generator tube uncovery on the iodine release for SGTR and non-SGTR events is negligible. Therefore, we agree with your position on the matur and consider this issue resolved." (Reference NRC letter from Robert C. Jones to Lawrence A. Walsh,

" Westinghouse Owners Group Steam Generator Tube Uncover Issue," dated March 10, j 1993.)

The conclusions of the WOG program apply to the Farley units. The implementation of  ;

a power uprating and ARC (in accordance with NRC Generic Letter 95-05, " Voltage- {

Based Repair Criteria for Westinghouse Steam Generator Tubes AfTected By Outside I Diameter Stress Corrosion Cracking," dated August 3,1995) have no impact on the i conclusions of the WOG program. The effects of partial steam generator tube uncovery on the radiological consequences of SGTR and non-SGTR events, including locked rotor and rod ejection, are negligible and do not present a safety concern for Farley.

End of Response i

SNC has confirmed with Westinghouse that Model 54F steam generators are bounded by the representative plant as described in WCAP-13247, and therefore, SG tube uncovery is not considered in the RCP locked rotor accident analysis. As noted in our steam generator replacement licensing submittal, the analysis submitted for the Power Uprate project continues to be bounding.

Attachmerf 1 Page 10 of 11 Response to RAI Dated July 2,1999 i

n NRC RAI41 Regarding your SG replacement containment analyses model, please indicate the key input parameters and assumptions that are different from the parameters and assumptions used in your SG uprate containment analyses model.

SNC Response to RAI.Il LOCA - Changes from Uprate Analysis

a. The primary change to the .RSG containment analysis model from the power upree model was the blowdown mass and energy release. No other changes were made to the model which would significantly effect the calculated peaks,

b. The "end of blowdown" times and integrated blowdown energy for LOCA which are used to determine the switchover from Tagami to Uchida HTC was changed slightly due to the revised blowdown data.

c. Revised times for RHR and Containment Spray switchovers were modeled in the LOCA analyses based on recent RWST level setpoint evaluations. In general, the switchover times decreased.

d. A small change to the RHR Heat Exchanger heat transfer surface area was modeled to represent plant design data with a margin for tube plugging. In addition, a small change to the RHR flow rate was made based on the revised blowdown flow rates ,

after 3,600 seconds. )

l MSLB - Changes from Uprate Analysis l l

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a. The prinu y change to the RSG containment analysis model from the power uprate model u . die blowdown mass and energy release.

b. Another key input change was the specification of 8% condensate revaporization, as allowed by NUREG 0588 Appendix B for the durati . in which the atmosphere is superheated.

LOCA & MSLB - Key Inputs Which Were Not Changed

a. Contaimrent Spray flow rates and temperatures

b. Fan Cooler heat removal rates

c. Initial temperature, pressure and relative humidity {'

d. Containment Heat Sinks (walls, structural steel, etc.)

e. ESF response delays (slight changes in time to reach the containment pressure set ,

points occurred due to the revised blowdown data) l l

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3 Attachment i Page 11 of 11 Response to RAI Dated July 2,1999 i

e NRC RAI-12 In your SG replacement containment analyses model, the peak LOCA pressure increased slightly from 43.0 psig to 43.8 psig. However, peak main steam line break pressure decreased from $2.4 psig to 52.0 psig and peak containment temperature decreased from 383 degrees F to 367 degrees F. Please discuss the reasons for the above changes in pressure and-temperature.

SNC Response to RAI-12 LOCA - Increase in Results from Power Uprate ne increase to the LOCA peak results is due to the increased blowdown mass and energy releases associated with the RSGs. The Model 54F steam generator has more tubes than the Model 51. This results in an increase in RCS mass at the start of the LOCA event.

MSLB - Decrease in Results from Power Uprate The decrease in the MSLB peak temperatures and pressures from the power uprate results will be discussed in two parts: 1) the decrease in pressure; and 2) the decrease in temperature. The changes that lead to a decrease in te nperature also have a small effect on pressure, and vice versa; however, the primary cause of the decrease in pressure verses temperature is different.

Pressure: There are two primary causes of the decrease in the MSLB RSG peak pressures as compared to the Power Uprate analysis. The first is a change in the RSG operating water level, based on reactor power. In general, for power levels 30%, 70%,

and 102%, there was a net decrease in the initial SG inventory. For the 0% power case, however, there was a significant increase in the initial SG mass. Consequently, for cases 1, 8, 9, and 12, the peak pressures all decreased when compared to the results for the Power Uprate analysis. For case 13, however, the peak pressure increased from the power uprate case 13. As such, the limiting pressure case has shifted from case 12 for power uprate to case 13 for RSG. (See BOP Licensing Report section 3.2 for a description of each Case.) The second significant change that contributed to the decrease in MSLB peak pressures for RSG is a change in the method of analysis from LOFTRAN to RETRAN for the mass and energy releases.

Temperature: De secondary side pressure increased from 798 psia for the uprate analysis to 817 psia for the RSG analysis which resulted in an increase in the enthalpy of the break flow. His increased enthalpy would result in an increase in peak CTMT Temperature. As noted in the response to RAI-l 1, above, for the MSLB analyses, 8%

revaporization was credited as allowed by NUREG 0588, Appendix B. Crediting 8%

revaporization reduced the containment temperature response resulting in the slightly lower peaks when compared to the power uprate results. During preparation of the analysis, informal sensitivity studies were performed without 8% revaporization in order to observe the increased temperature response and verify that the expected changes due to RSGs did in fact occur. He final calculation, however, only documents the 8%

revaporization cases.

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Attachment 2 Faseph M. Farley Nuclear Plant Response to Request for Additional Information Steam Generator Replacement Related Technical Specifications Change Reauest (NRC Letter Dated July 2,1999)

Computer Disk for RETRAN Input Deck 1

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