ML20217J665

Forwards RAI Re Request to Amend License & TSs to Allow Increase in Licensed Thermal Power from 2652 Mwt to 2775 Mwt

ML20217J665
Person / Time
Site:	Farley
Issue date:	10/14/1997
From:	Jacob Zimmerman NRC (Affiliation Not Assigned)
To:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
References
TAC-M98120, TAC-M98121, NUDOCS 9710210208
Download: ML20217J665 (9)

Text

Mr. D. N. Morey

• Octobcr 14, 1997 Vice President Fcrl:y Pr:J:ct Southem Nuclear Operating Company, Inc.

Post Office Box 1295 Birmingham, Alabama 35201 1295

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION RELATED TO POWER UPRATE SUBMITTAL JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 (TAC NOS. M98120 AND M98121)

Dear Mr. Morey:

By letter dated February 14,1997, you submitted a request to amend the Facility Operating Licenses and Technical Specifications (TS) for the Farley Nuclear Plant (Farley), Units 1 and 2, to allow for an increase in the licensed thermal power from 2652 MWt to 2775 MWt. By letter dated August 21,1997, the NRC staff requested additionalinformation which you responded to by letter dated September 22,1997.

The staff has reviewed your submittals and determined that additionalinformation is required.

The enclosure identifies the requested additionalinformation needed.

in order to maintain a timely review schedule and meet your requested target date for I

completion, it is requested that the information be provided within 30 days of receipt of this letter, if you require any clarification regarding this request, please call me at (301) 415-2426.

Sincerely, ORIGINAL SIGNED BY:

Jacob 1. Zirnmerman, Project Manager l Project Directorate 112 l

Division of LJactor Projects - 1/11 Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364

Enclosure:

Request for Additional Information cc w/ encl: See next page T'-)Fol)/

Distribution:

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b. bbh h@Y 9710210200 971014 PDR ADOCK 05000348 p PDH To receive a copy of this document, indicate in the box: "C" = Copy without attachment / enclosure "E" o Copy with attachment / enclosure "N" = No copy *See previous concurrence OFFICE PM:PJ%[-2 16 LA:PDil-2 i S SRXB* l D:P,DC A j NAME J2fffAMBRMAN:cn L.BEF RY ifP EWEISS DATE. /o 'l M//97 ft) / U/97 ' \ 10/14/97 H. bel 80W

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' NUCLEAR REGULATORY COMMISSION

$ WAsHINs PH, D.C. 30MH001

%,***** October 14, 1997 Mr. D. N. Morey -

Vice President - Farley Project Southem Nuclear Operating Company, Inc.

Post Office Box 1295 Birmingham, Alabama 35201 1295

SUBJECT:

REQUES ( FOR ADDITIONAL INFORMATION RELATED TO POWER UPRATE SUBMITTAL JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 (TAC NOS. M98120 AND M98121)

Dear Mr. Morey:

By letter dated February 14,1997, you submitted a request to amend the Facility Operating Licenses and Technical Specifications (TS) for the Farley Nuclear Plant (Farley), Units 1 and 2, to allcw for an increase in the licensed thermal power from 2652 MWt to 2775 MWt. By letter dated August 21,1997, the NRC staff requested additionalinformation which you responded to by letter dated September 22,1997.

The staff has reviewed your subrnittals and determined that additionalinformation is required.

The enclosure identifies the requested additionalinformation needed.

In order to maintain a timely review schedula and meet your requested target date for completion, it is requested that the information be provided within 30 days of receipt of this letter, if you require any clarification regarding this request, please call me at (301) 415-2426.

Sincerely, cob 1, Zimmerman, Project Manager Project Directorate 112 Division of Reactor Projects - 1/ll Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364

Enclosure:

Request for AdditionalInformation ec w/ encl: See next page a -

i 1 . .

, Joseph M. Parley Nucle:r PI:nt cc:

l Mr. R. D. Hill, Jr.

j General Manager .

Southern Nuclear Operating Company' i Post Office Box 470 Ashford, Alabama 36312

)

l Mr. Mark A}luni, Licensing Manager

! Southern Nuclear Operating Company Post Office Box 1295 j Birmingham, Alabama 35201 1295 Mr. M. Stanford Slanton Balch and Bingham Law Firm-Post Office Box 306 j 1710 Sixth Avenue North -

Birmingham, Alabams' 35201 Mr. J. D. Woodard Executive Vice President -

Southern Nuclear Operating Company .

Post Office Box 1295 Birmingham, Alabama 35201 -

State Health Officer Alabama Departmort of Public Health i 434 Monroe Gtreet Montgomery, Alabama 36130 1701 Chairman Houston County Commission Post Office Box 6406 -

Dothan, Alabama 36302 -i Regional Administrator, Region 11 U.S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth Street, S.W., Suite 23T85 Atlanta, Georgia 30303 Resident inspector ..

U.S. Nuclear Regulatory Commission

.7388 N. State Highway 95 .!

Columbia, Alabama 36319

. . _ . . . _ . . _ , - _ --_______.a..-_

m.___ _ ..

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE SOUTHERN NUCLEAR OPERATING COMPANY. INC.

REQUEST TO AMEND THE OPERATING LICENSES FOR THE JOSEPH M. FARLEY NUCLEAR PLANT. UNITS 1 AND 2 THERMAL POWER UPRATE REQUEST

1. The response to question 1 in the August 5,1997, submittal does not completely answer the question. The question asked for a Standard Review Plan (SRP) 5.2.2 analysis to be performed at the new power level. The SRP 5.2.2 snalysis differs from a Chapter 15 FSAR type analysis because SRP 5.2.2 does n',t allow credit for the first safety grade reactor trip. Additionally, the SRP 5.2.2 analysis that was performed at initiallicensing, and referenced in the response to the question, no longer applies because the reactor power will be increased. Please provide an SRP 5.2.2 analysis to show that the primary and secondary overpressure protection is adequate. Indicate what percentage of the relieving capacity is used to mitigate the event and the peak primary and secondary i pressure.

I 2. The response to question 4 does not completely answer the question. The response indicates that only a review of the FSAR was performed to identify areas for review. To determine the risk to public health and safety, please perform a review of your Individual Plant Examination and determine if any of the compensatory actions or success paths identified are no longer available c.' the uprated conditions (power, reactor coolant system (RCS) flow, emergency core cooling system (ECCS) flow).

3. The response to question 5 indicates that "the Farley Units are currently operating with zircaloy clad fuel and ZlRLO clad fuel (VANTAGE +)." If both Vantage 5 and VANTAGE +

fuels will be used, please update both the appropriate sections of the TS and FSAR with appropriate references to the VANTAGE + fuel. The safety evaluation referenced in the response to question 5 indicated that the NRC approved the use of ZlRLO clad for lead test assemblies. Please verify that all the analyses presented in WCAP 12610-P A,

" VANTAGE + Fuel Assembly Reference Core R6 port" have been performed with acceptable results.

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4. The response to question 7 indicates that the Technical Specifications (TSs) limit T.,to below $80.3; however, the plant is not analyzed for operation at this temperature.

Because the existing TS does not limit T., to within the accident analysis window, with appropriate instrument uncertainties, provide an accident analysis that uses the appropriate TS T., window, with instrument uncertainties. '

5. The response to question 10 does not provide sufficient detail regarding the analysis of the 50% load rejection. Please describe in greater detail the calculated results including the magnitude (requested in question 10 and not provided) of the power and temperature oscillations. Evaluate the ability of the analysis techniques to adequately evaluate the core transient. Additionally, is there any other operating window where greater oscillations can occur?

Enclosure

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6. Please provide a summary of calevtations that show that the atmospheric dump valves are sized adequately to maintain a 50 F/hr cooldown rate over the entire range that they are necessary using the higher decay heat associated with the power uprate. The response to question 9, the submittal description, and the FSAR description do not

, provide suMelent detall.

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7. Are all codes used in the power uprate and the current licensing basis up to-date with  ;

respect to relevant changes in knowledge, regulations, guidance, and changes in the ,

plant? Discuss this in detail for all codes. Are the codes used in conformance with their I limitations and restictions?

8. Identify all codes used for this power uprate, and of these, which were previously reviewed or approved for use at your plant. To be able to determine the effect on the safety margin from using the new code, if a new code is used, what would a baseline run

, at the old power level reveal in terms of calculated results (i.e., compare the peak l cladding temperature (PCT) at the new power relative to the PCT at the old power)?

Besides this baseline run, identify all cumulative decreases in safety margin from successive plant and procedure modifications. Provide a justification for the new code's applicability to your plant.

9. Document what technical areas should be rereviewed because current analysis is not bounding and identify to what extent the power uprete depends on current analysis. For each analyais relied upon, identify the maximum power level for which it is valid. For example, as discussed in question 1, you referenced the "initiallicensing" SRP 5.2.2 analysis when discussing overpressure protection; however, you did not indicate if this analysis was done at the uprated conditions. Please perform a review to determine other technical areas that shuuld be rereviewed because the current analysis is not bounding.

10 List all uprate assumptions, analyses, and analytical codes. Confirm that these are identified and incorporated in license conditions, technical specifications, or the FSAR, as appropriate. This includes ensuring that the FSAR documents all analytical codes used in the uprate. For example, as discussed in question 4, the TS limits on T,,, are not bounded by the values chosen for the accident analysis. The other uprate assumptions, analyses, and analytical codes should be reviewed by you to verify that they are appropriately incorporated into licensing basis documents.

11. Describe how the steam generator water level bla 1 of -5.0% is modeled in the accident analysis and where the bias comes from. Desc".e the effect on the results of modeling the steam generator levelin this manner. (WCA ?-14723, Section 6 2)

12. The plant-specific modeling and analysis for the large break loss of coolant accident (LOCA) is not provided. Please provide additionat information regarding the analysis assumpions for the best estimate large break LOCA calculations. Information needed is the asomed initial conditions, the limiting transient progression with discussion of why it is the limiting transient, single failure assumptions, loss of-offsite power assumptions,

1 3- l time step assumptions, and major plant parameters with uncertainties. Show that the ,

calculations were performed with the approved version of WCOBRA/ TRAC MOD 7A, I revision 1, and provide information that shows compliance with the code limitations and restrictions. (WCAP 14723, Section 6 3) l l

13. The submittelindicates that the operating ranges for major plant parameter assumptions will be documented in Chapter 15 of the FSAR. r'! ease provide a summary of the parameters and the assumed operating ranges where the analyses is valid. (WCAP.

14723, Gection 6-4) i

14. For the small break LOCA analysis, the submittal ls not clear with regard to single failure assumptions and limiting conditions. The submittal originally states that the limiting single isilure is "that of an emergency power train." A few sentences down the submittal states 1

that the " assumption of LOOP [ loss of offsite power) as the limiting single failure for small l

break LOCA is part of the NRC approved methodology." Please clearty state the j assumed single failure, its affect on ECCS performance and the bases for it being the

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limiting single failure for al! cases. Additionally, discuss the availability of offsite power in the analysis. State the basis for assuming the 1.OOP occurs coincident with the reactor trip. Why is this case more limiting inan assuming offsite power is lost at or prior to the LOCA and why is this more limiting than assuming offsite power is available throughout the event? Describe the modeling of the reactor conlant pumps and explain why not modeling the pump heat is acceptable. (WCAP 14723, Section 6 g)

15. For the small break LOCA analysis, the submittal indicates that the ECCS flow is  ;

delivered to both the intact and broken loops "at RCS backpressure." .Following a LOCA,  ;

the pressure in the broken loop will be lower than in the intact loops. As a result, the injected flow will p,*aferentially go to the broken loop because it is at a lower pressure.

Modeling both the intact and broken loops at RCS pressure will overpre dict the flow going to the intact loops and ur' der: edict the flow in the broken loops. Explain why your ,

modeling assumptions acceptably bound plant conditions. Additionally, describe why  ;

your modeling approach eliminates the need fer the 150*F peak clad penalty. l (WCAP 14723, Section 6 g/10) '

16. The submittal indicates that a break of an injection line is less limiting than a break at the bottom of the cold leg. Please provide additional bases for this conclusion. Discuss the competing effects of degraded injection flow (decreasing ECCS performance) and the more rapid core depressurization (increasing ECCS performance) as a result of the break

- at the top of the cold leg. A quantitative assessment of the relative magnitude of these effects would be helpful. (WCAP 14723, Section 6 g/10)

17. The submittal indicates that the reactor trip is modeled to occur at 1840 psia with a setpoint of 1865 psig; however, the uncertainty presented on pg 6-2 indicates that a t50 psi would be used, Explain why the *50 uncertainty was not used for small break LOCA. (WCAP-14723, Section 610)

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18. The submittal indicates tnat units "may be subject to SI [ safety injection) interruption or reduction while switching over to cold leg recirculation." Please explain how the calculated core temperature is maintained at an acceptably low value if flow can be interrupted for both small and large breaks. How is long-term cooling acceptability calculated and verified? (WCAP 14723, Section 611) 1

19. The sensitivity study for your analysis !ndicates that the lower Tm yields more limiting

results than a higher T.. Is there any physical reason you would expect that this would i be the. case for your plant design?

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20. Verify that the acciderit analyses assume a stuck rod following a reactor trip, the most limiting single failures, and a LOOP for each transient. Describe how uncertainties in the

! Fo and Fm are reflected in the analysis.

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21. For a number of transients, the analysis performed to support the overtemperature delta e

temperature and overpressure delta temperature setpoint license amendment change, i- the Vant6ge5 fuel license amendment change, or the existing analyses are credited, j Please describe the modeling for these analyses and verify that all the changes associated with this license amendment request are correctly modeled. Modeling assumptions include the effects of the increased power, reduced RCS flow, reduced o ECCS flow, reduced T , control rod control optimization and ZlRLO If the analyses that l are credited have not been submitted under a different approved license amendment

{ request, please provide additional details regarding assumptions and approved codes. If j the credited analyses do not mod 31 all the changes associated with this submittal, provide a justification for not needing to explicitly model them. -

22. The verbiage in chapter 6 of the topical report appears inconsistent with the introduction in section 6.2.0 For example, in section 6.2.0 the uncontrolled rod cluster control assembly withdrawal at power transient is listed as a transient that " Reanalysis performed in support of the FNP [Farley Nuclear Plant) uprate." However, in section 6,2.2 this transient is dispositioned by stating that NRC review and approval for this event was ,

al eady received and no details are included. To clarify, please provide a list of all the Chapter 15 transients; for each, indicate if the ana;ysis was performed to support the uprate, if analysis was performed earlier at the uprated conditions (including all other

' changes), or if analysis is not necessary with an explanation justifying the conclusion.

23. The submittal indicates that the optimized rod control system parameters (WCAP-14723, Section 6-62) are modeled in the accident analysis; Please describe these changes to the rod control system and how it affects the accident analysis. '

24. On page 6-62 of WCAP-14723, the plant changes modeled in the accident analysis are l described, The reduction in the required RCS flow is not included in this discussion. l Please verify that the reduction in RCS flow is modeled in all the accident analyses '

credited in the submittal. l l

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25. It is appropriate to use the Revised Thermal Design Procedure when evaluating the critical heat flux o' departure from nucleate boiling (DNB) and then use nominal values for .

power, pressure, and temperature. However, the uncertainties should be included in evaluating the other limits such as peak pressure and linear heat rate. For the transients j that challenge the limits other than DNB and the revised thermal design procedure was l used, please evaluate the need to perform an analysis to verify that the other limits are j not exceeded.

26. For the loss of electric load analysis, what was assumed for the accumulation in the j safety relief valves? What are the peak primary and secondary pressures calculated?

i- 27. - For the analysis of the loss of nonemergency ac power to plant auxiliaries analysis please

- describe the sequence of events in greater detail. Describe why the loss of flowis a more limiting DNB event and explain why the assumptions are chosen to assure limiting l results. Additionally, describe the initial conditions relating to the important parameters.

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28. For the inadvertent ECCS actuation, the description in the submittal is not consistent with regard to the credit given the power operated relief valve (PORV) to actuate on demand.

l Page 6-124 states, "PORVs are not assumed as an automat'c pressure control function for the pressurizer filling case." However, later discussion states, "PORV availability must l be assured by manually opening the block valve to allow the associated PORV to actuate on demand." Please clarify the situation by describing the reliance of the PORVs. For the overfilling case, please indicate if the acceptance criteria (pressurizer is permitted to go water solid), the operator actions (7 minutes to open the block valve), and reliance on y the actuation of the PORV have been approved by the NRC for this transient. Describe

the single failure assumptions.

l 29. Please provide more information regarding the analysis of the main steamline break (MSLB). Please provide an explanation why the analyses performed, with the assumed single failures, are bounding (with assumptions made regarding rod motion). Discuss the

LOOP assumptions and justify that they are bounding. The conclusions for the MSLB

only state that the DNB design basis is met. Please state the design basis and provide j' the results, What is the minimum DNB and what percentage of the rods experience DNB l_ (if applicable)? Verify that the primary and secondaiy pressure is maintained below j acceptable design limits, considering the potential for brittle and ductile failures. The response to question 8 references WCAP-g226, Rev.1, " Reactor Core Response to Excessive Secondary Steam Releases" to justify the use of a conservatively low flow i rather than the use of a conservatively high flow. Please provide a copy of this topical (or provide a reference if it has been docketed) and justify that the conclusion is applicable to your plant. If maximum vs. minimum RCS flows are used, would the cooldown be more j severe and cause a violation of the cooldown limits associated with brittle or ductile failures? Does the flow area through the main steamline nozzle assumed in the analysis

] account for or bound the design tolerances and thermal expansion of the metal?

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30. The analysis results are not presented in the submittal or the FSAR for the main feedline rupture. Please identify the acceptance criteria and present the results of the analysis. Is fuel failure predicated, what is the lowest DNBR, are Part 100 limits met? The FSAR analysis is referenced for this event. Please describe how the changes with this amendment are modeled in the FSAR analysis and justify why the results continue to be bounding. Discuss the LOOP assumptions and justify that they are bounding.

31. The locked rotor event assumes a " conservatively large (absolute value) of the Doppler-only power coefficient." Please justify this assumption. A large absolute Doppler power coefficient would appear to be non-conservative in a heat uplover power event.

32. For the rod ejection accident, provide the results of the analysis (percentage of rods experiencing DNB, percentage of rods experiencing fuel melt, peak pressure). The analysis description in the submittalis not sufficient to determine if the analysis techniques are in conformance with the approved methodology. Verify that the analysis

inputs and analysis technique was performed in compliance with Regulatory Guide 1.77, "Assumptior's Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors" and the approved methodology, WCAP 7588-1-A. Any deviations should be justified.

33. For the steam generator tube rupture analysis, please provide a summary of the analysis with the calculationwl results. What are the assumption and limiting single failures?

Provide an analysis for the steam generator overfill calculations with the analysis, assumptions, and conclusions presented.

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