W3F1-2005-0001, Review of Draft Safety Evaluation for the Extended Power Uprate

Review of Draft Safety Evaluation for the Extended Power Uprate

ML050100225
Person / Time
Site:	Waterford
Issue date:	01/05/2005
From:	Dodds R Entergy Nuclear South, Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC MC1355, W3F1-2005-0001
Download: ML050100225 (30)

Text

- Entergy Entergy Nuclear South Entergy Operations, Inc.

17265 River Road Killona, LA 70057-3093 Tel 504 739 6379 Fax 504 739 6698 rdodds~entergy.com W3F1 -2005-0001 January 5, 2005 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 R.A. (Al) Dodds, III Director, Nuclear Safety Assurance Waterford 3

SUBJECT:

Review of Draft Safety Evaluation for the Extended Power Uprate Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

REFERENCES:

1.

NRC Letter dated December 16, 2004, "Waterford Steam Electric Station, Unit 3 (Waterford 3) -Draft Safety Evaluation For The Proposed Extended Power Uprate (TAC MC1355)"

Dear Sir or Madam:

By letter (Reference 1), the NRC forwarded the draft Safety Evaluation (SE), prepared in support of the Waterford Steam Electric Station, Unit 3 Extended Power Uprate (EPU) license amendment request, to Entergy Operations, Inc. (Entergy). The NRC requested that Entergy review the draft SE to identify if it contains any proprietary information.

Entergy has completed its review of the draft SE, except for Section 2.7 which was not included in Reference 1, and has determined that it contains no proprietary information.

Additionally, Entergy is taking this opportunity to provide comments on the draft SE for the staff's consideration. These comments are provided in the attachment to this letter.

There are no new commitments contained in this letter.

If you have any questions or require additional information, please contact D. Bryan Miller at 504-739-6692.

Sincerely, RAD/DBM/cbh

Attachment:

Technical Comments on Draft Safety Evaluation

.,ro(

W3F1-2005-0001 Page 2 of 2 cc:

Dr. Bruce S. Mallett U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 NRC Senior Resident Inspector Waterford 3 P.O. Box 822 Killona, LA 70066-0751 U.S. Nuclear Regulatory Commission Attn: Mr. Nageswaran Kalyanam MS O-7D1 Washington, DC 20555-0001 Wise, Carter, Child & Caraway Attn: J. Smith P.O. Box 651 Jackson, MS 39205 Winston & Strawn Attn: N.S. Reynolds 1400 L Street, NW Washington, DC 20005-3502 Louisiana Department of Environmental Quality Office of Environmental Compliance Surveillance Division P. 0. Box 4312 Baton Rouge, LA 70821-4312 American Nuclear Insurers Attn: Library Town Center Suite 300S 29th S. Main Street West Hartford, CT 06107-2445

Attachment To W3F1 -2005-0001 Technical Comments on Draft Safety Evaluation

Attachment to W3F1 -2005-0001 Page 1 of 27 Technical Comments on Draft Safety Evaluation Location Existing Text Recommended Text Bases 1

Section 1.2 The makeup function of the delete This is not a CVCS function stated in (Pages 3 and CVCS is required to maintain the FSAR or in the EPU submittal.

4) the plant in an extended hot Reference 11113/03 Submittal, shutdown condition following Section 2.1.11.

a transient.!

2 Section 2.1.4 Reference 35 establishes a Reference 35 establishes a Reference 35 (WCAP-14577) does Tech. Eval.

threshold of 1 X 102' n/cm2 (E 2 threshold of 1 X 102' n/cm2 (E Ž not cite a threshold value for void (Page 13) 0.1 MeV) for the initiation of 0.1 MeV) for the initiation of swelling.

IASCC, loss of fracture IASCC and loss of fracture toughness, and/or void toughness in PWR RV internal sweliing in PWR RV internal components made from components made from stainless stainless steel (including steel (including CASSs) or Alloy CASSs) or Alloy 600/82/182 600/82/182 materials.

materials.

3 Section 2.1.11 The CVCS and boron recovery The CVCS provides a means The BRS is not a defined system at Reg. Eval.

iystdm (BRS) provide means for...

Waterford 3.

(Page 25) for...

4 Section 2.1.11 The NRC staff has reviewed the The NRC staff has reviewed the The BRS is not a defined system at Summary licensee's evaluation of the licensee's evaluation of the Waterford 3.

(Page 26) effects of the proposed EPU on effects of the proposed EPU on the CVCS a-nd BRIS, and the CVCS, and concludes that concludes that the licensee has the licensee has adequately adequately addressed changes addressed changes in the in the temperature of the reactor temperature of the reactor coolant and their effects on the coolant and their effects on the CVCS anid BRS. The NRC staff CVCS. The NRC staff further further concludes that the concludes that the licensee has licensee has demonstrated that demonstrated that the CVCS...

the CVCS aindiBRS 5

Section 2.2.2 The resultant stresses and The resultant stresses for the Table 2.2-5 does not list CUFs.

Tech Eval CUFS for the RCS main coolant RCS main coolant loop piping for Reference 11113103 Submittal.

(Page 30) loop piping for the hot and cold the hot and cold legs are

Attachment to W3Fl -2005-0001 Page 2 of 27 Location Existing Text Recommended Text Bases legs are determined as shown in determined as shown in Table Table 2.2-5 of the request 2.2-5 of the request submittal.

submittal.

6 Section 2.2.2 For the MSLs and FWLs (Code For the MSLs and FWLs (Code Reference 11113103 Submittal, Tech Eval Class 2 an-d3 piping) inside Class 2 piping) inside Section 2.2.2.2: MSL and FWL are (Page 31) containment, the analyses were containment, the analyses were Class 2.

performed in compliance with performed in compliance with Reference May 13, 2004 requirements of Section I11, requirements of Section III, 1971 Supplement, Response to Su-swectio-s NC and NSl 1971 Edition of the Code, up to and Question 11: 'Class 2 and 3 Piping:

Edition of the Code, up to and including the Wifnter1972 ASME B&V Code,Section III, 1971 including the Summer9 Addenda, which is the Code of Edition, including addenda Addenda, which is the Code of record.

through Winter 1972."

record.

7 Section 2.2.2 In Reference 7, the licensee In Reference 7, the licensee Consider clarifying that this is a BOP Piping, indicated that the evaluation of indicated that the evaluation of current plant issue that is being Components, CCW piping and pipe supports CCW piping and pipe supports addressed per 1 OCFR50 Appendix and Supports would be completed by August 3, would be completed by August B.

(Page 33) 2004. However, in Attachment 3 3, 2004. However, in Reference 8110104 Supplement, to Reference 13, the licensee to Reference 13, : "in addition, it was indicated that, due to the the licensee indicated that it determined that the CCW SDC outlet increase in scope, this issue can a Jermined that the CCW header could, and has exceeded, the not be resolved prior to EPU system return piping design temperature of 1750F at the implementation in the spring of downstream of the SDC heat current (pre-EPU) power level. This 2005.

exchanger could and has issue has been entered into

'exceeded the design Entergy's 10 CFR 50 Appendix B temperature at the current corrective action program."

(pre-EPU) power level during normal shutdowns and therefore has entered thils issue into its IOCFR5O Appendix B corrective action programs The licensee also indicated that, due to the increase in scope, this issue can

Attachment to W3F1 -2005-0001 Page 3 of 27 Location Existing Text Recommended Text Bases not be resolved prior to EPU implementation in the spring of 2005.

8 Section 2.2.2 The CCW piping, pipe supports, In Reference 18 the Licensee The evaluations performed were BOP Piping, and components in the line were described evaluations that were uOperability" evaluations not "Code" Components, evaluated at 225 OF and found performed for the CCW SDC compliance evaluations.

and Supports acceptable in compliance with' heat exchanger outlet piping, Reference 101812004 Supplement, (Page 33) tie Co6dieof record for this pipe supports, and components Response to Question 4: "These piping system, namely, ASME to support continued operation.

evaluations determined that the Code,Section III, 1971 Editionr These evaluations determined system will remain operable with a including addenda through the that the system will remain CCW outlet temperature up to winter 1972.;

operable with a CCW SDC heat 225F."

exchanger outlet temperature up Reference 8110/04 Supplement, to 225F. : "Currently a Compensatory action is in place requiring Operations to monitor and maintain CCW return temperature from the SDC outlet header to less than or equal to 2250F. The 2250F limit insures acceptable system and plant protection while the final resolution is identified and implemented. This compensatory action is currently incorporated into plant operating procedures."

9 Section 2.2.2 As a result of the above As a result of the above The CCW SDC heat exchanger BOP Piping, evaluation, and upon_

evaluation the staff concludes outlet evaluations performed were Components, isfacutory implem entatiofiof that BOP piping, pipe supports, "Operability" evaluations not uCode" and Supports the compensatory measures and equipment nozzles, will compliance evaluations.

(Page 33) the staff concludes that BOP remain acceptable and continue Reference 10/8/2004 Supplement, piping, pipe supports, and to satisfy the design-basis Response to Question 4: "These equipment nozzles will remain requirements for the power evaluations determined that the acceptable and continue to uprate (excluding th-CCW system will remain operable with a satisfy the design-basis SOC heat exchanger outlet CCW outlet temperature up to

Attachment to W3Fl -2005-0001 Page 4 of 27 Location Existing Text Recommended Text Bases requirements for the power piping and pipe supports).

225-F."

uprate.~

The CCW SDC heat exchanger Reference 8/1012004 Supplement, outlet piping and piping : "Maintain adequate supports will remain operable compensatory actions in place until for the power uprate, upon the final resolution (e.g., system satisfactory implementation of rerate, etc.) is identified and the compensatory measures',

implemented."

until th e final resolution is implemented per the Licensee's 10CFR50'_Apjpendix B'corrective action program 10 Section 2.2.2 In Reference 1, the licensee In Reference 1, the licensee 1965 Edition of Code is not provided Rx Vessel and indicated that the evaluation of indicated that the evaluation of in Reference 1 Supports the RV was performed in the RV was performed in Reference November 13, 2003 (i.e.

(Page 33) accordance with the Section IRIn accordance with tlieSection'i'il' Reference 1) Submittal, Section 1 965Edition'ofthe Codes The 1971 Editfflo'nof'the Code with' 2.2.2.1.4.1: 'Loads for RV supports results indicate that the Addenda through the Summer and support bearings for power maximum primary plus of 197V The results indicate uprate are bounded by those secondary stresses for all the RV that the maximum primary plus evaluated in the AOR. Therefore, the critical locations other than the secondary stresses for all the stress evaluations in the AOR remain CEDM housing are within the RV critical locations other than valid for EPU."

Code-allowable limits.

the CEDM housing are within the Reference May 13, 2004 Code-allowable limits.

Supplement,

Attachment:

uSection III, 1971 Edition up to and including Summer of 1971 Addendum" 11 Sec 2.2.2 SG The licensee indicated that the The licensee indicated that the Reference May 13, 2004 and Supports Code and Code edition used in Code and Code edition used in Supplement, Response to (Page 34) the evaluation of SGs for the the evaluation of SGs for the Question 7: Each component power uprate is the 1971 Edition power uprate is the 1971 Edition evaluation performed as part of the of the ASME Code,Section III, of the ASME Code,Section III, analysis of the secondary internals "Nuclear Vessel," up to and "Nuclear Vessel," up to and used the 1971 Edition of the ASME including the 197WtIinter including the 197 Su mm ez r

Code with Addenda through the Kddeinda, which is the Code of Addenda, which is the Code of Summer of 1971 as the Code of

Attachment to W3Fl -2005-0001 Page 5 of 27 Location Existing Text Recommended Text Bases record.

record.

Record for both the original analysis of record and for the EPU conditions.

12 Section 2.2.3 The licensee's structural analysis The licensee's structural The design basis seismic analysis is Tech. Eval.,

included the current fuel analysis included the current fuel not affected by EPU.

2nd Para (Page assembly design and mass, and assembly design and mass, and Reference the November 13, 2003

38) modeling refinements to the modeling refinements to the pipe Submittal, Section 2.2.3.1:

seismic analysis, as well as break analyses due to the use of uSeismic loading of the RVI pipe break analyses due to the LBB methodology.

components, as evaluated in the use of LBB methodology.

AOR, is not affected by the extended power uprate (EPU)."

13 Section 2.3.2 Since the MS and FW systems Since the MS and FW systems The MS and FW lines run across the Tech Eval, are located outside of the are located outside of the roof of the Reactor Auxiliary Building HELB (Page containment structure, the Reactor Auxiliaiy Building once they exit containment.

44) environmental conditions structure, the environmental Reference the November 13, 2003 produced by the pipe breaks in conditions produced by the pipe Submittal, Section 2.3.1:

these systems will not affect any breaks in these systems will not "Per FSAR Section 3.6A.2.3, safety-related equipment or affect any safety-related Environmental Effects: "The high components.

equipment or components.

energy piping systems located in the Reactor Auxiliary Building are given in Subsection 3.6A.2. Since the main steam and feedwater systems are located outside of the structure, the environmental conditions produced by the pipe breaks in these systems will not affect any safety-related equipment or components."

14 Section 2.3.2 The licensee evaluated the The licensee evaluated the Reference the May 7, 2004 Main equipment for EPU conditions.

equipment for EPU conditions.

Supplement, Attachment 2, Transformers At this time, the licenses At thistimiiethe licens plans Response to Question 14:

(Page 46) expects that additional c6biing to add additional cooling to "Additionally, Main transformer A will for both transformers and main transformer 'B' and be replaced and main transformer B replacement or rewind of thi

'replace main transformeeA cooling will be upgraded to support main transformer"'A' may be in-order to Support EPUM The the higher planned power required in order to support main transformers are being generation. The replacement of main

Attachment to W3F1-2005-0001 Page 6 of 27 Location Existing Text Recommended Text Bases EPU! The main transformers are modified to support the full transformer A is required to restore it being modified to support the full 1333.2 MVA generator post-to the original nameplate capability.

1333.2 MVA generator post-uprate conditions.

The additional cooling for main uprate conditions.

transformer B is required to preclude long term degradation mechanisms which could shorten the useful life of the transformer. These modifications do not change the design function of the equipment, nor will any new system interactions be created."

15 Section 2.4.1 The licensee has indicated that The licensee has indicated that Reference the May 7, 2004, Tech Eval, they have developed the they have developed the Supplement Attachment 2 Inst. Stpt proposed TS Allowable Values proposed Allowable Values Response to Question I and the Method (Page (AVs) in accordance with (AVs) in accordance with fhi November 4, 2004, Supplement

52)

Method 3 of Instrument Society mehodology dscribed In tnhe.

of America (ISA) Standard 67.04, Waterord 3 TS Bases which is aterford 3 does not officially utilize Part 2, issued in 1994. Part 2 of essentially equivalent to "ISA Standard 67.04, Part 2, Method ISA 67.04 has not been Method 3 of Instrument Society 3." Officially Waterford 3 utilizes the endorsed by the staff.

of America (ISA) Standard methodology that was added to the 67.04, Part 2, issued in 1994.

TS Bases via Amendment 113 Part 2 of ISA 67.04 has not been issued on September 5, 1995.

endorsed by the staff.

Which, as illustrated in the November 4, 2004, Supplement, is slightly more conservative then Method 3 (but not enough so to address the staffs generic concern.)

16 Section 2.4.1 AVs are used in the TS as LSSS AVs are used in the TS to The AVs in the Waterford 3 TS are Tech Eval, to provide acceptance criteria for provide acceptance criteria for not the LSSS. The trip setpoints are Generic determination of instrument determination of instrument the LSSS at Waterford 3.

Concern channel operability during channel operability during Reference Waterford 3 TS 2.2.1 Regarding periodic surveillance testing.

periodic surveillance testing.

and the November 4, 2004, Method 3 Supplement.

(Page 53) 17 Section 2.4.1 The staff considers that the plant The staff considers that the plant Entergy understands that not just

Attachment to W3F1-2005-0001 Page 7 of 27 Location Existing Text Recommended Text Bases Tech Eval, instrument readings from instrument readings from any indicator can be used to verify Acceptability indicators can be used for indicators can be used for TS compliance and has a program to of Proposed channel check to detect a gross channel check to detect a gross specify what specific installed Changes failure of an instrument channel.

failure of an instrument channel.

instrument loops and indicators can (Page 54)

ThWe indic ator rea'din'gs acre not The indicator readingsyare be used for verifying specific TS acceptable as the sole basis only acceptable for TS compliance. These specific for TS compliancel There is no compliance provided tie instrument loops and indicators are assurance that the "indicated" in'dicator used (i.e., instrument evaluated and maintained (e.g.,

reading is reliable and loop) is shown to be reliable periodic calibrations) to insure that conservative.

and conservative.

they are reliable and conservative for the purpose of verifying TS compliance. Therefore, specific indicators shown to be reliable and conservative are currently being used and will continue to be used for TS compliance in accordance with industry practice.

18 Section 2.4.1 The numbers specified in the The numbers specified in the The word "indicated" is currently Tech Eval, TSs should be based on safety TSs should be based on safety used in the Waterford 3 TSs (e.g.,

Acceptability analysis and should meet the analysis and should meet the TS 3.1.2.7, 3.1.2.8, 3.4.3.1, 3.5.4.a, of Proposed requirements of 10 CFR requirements of 10 CFR and 3.7.1.3). In these cases the Changes 50.36©)(2) to establish the 50.36©)(2) to establish the Vindicated" value is selected based (Page 54) lowest functional capability or lowest functional capability or on the safety analysis (i.e., protects performance level of equipment performance level of equipment the safety analysis assumption) in required for safe operation of the required for safe operation of the accordance with I OCFR50.36.

facility. Therefore, the words facility. Following two These indicated values are "indicated" and "an indicated" conference calls on March 10 applicable to specific instrument cannot be used in theTSs.

and March 31, 2004, Entergy loops/indicators that have been Following two conference calls concurred with the NRC staff evaluated and determined to be on March 10 and March 31, that the TS bases are an reliable and conservative as 2004, Entergy concurred with the appropriate place to convey this discussed above. The word NRC staff that the TS bases are type of information. By indicated will remain in these TS an appropriate place to convey Reference 10, Entergy post-EPU. Waterford 3 withdrew its this type of information. By informed NRC of its decision request to add the "indicated"

Attachment to W3F1 -2005-0001 Page 8 of 27 Location Existing Text Recommended Text Bases Reference 10, Entergy to withdraw its request to add terminology to additional TS but did Informd NRC of its decislcion this terminology to various not request that this terminology be to withdraw its request to use TSs and will instead retain the removed from the TS where it this terminology in the TSs information in the technical already exists.

and will instead retain the bases. The staff finds this information in the technical decision acceptable.

bases. The staff finds this decision acceptable.

19 Section 2.5.2 The pressurizer relief tank (PRT)

The pressurizer relief tank (PRT)

Waterford 3 does not have power-Reg Eval is a pressure vessel provided to is a pressure vessel provided to operated pressurizer relief valves.

(Page 61) condense and cool the discharge condense and cool the from the pressurizer safety aid discharge from the pressurizer iefIR valves.

safety valves.

In general, the steam In general, the steam condensing capacity of the tank condensing capacity of the tank and the tank rupture disk relief and the tank rupture disk relief capacity should be adequate, capacity should be adequate, taking into consideration the taking into consideration the capacity of the izier capacity of the SVs; the piping to power-operated relief and SVs; the tank should be adequately the piping to the tank should be sized; and systems inside adequately sized; and systems containment should be inside containment should be adequately protected from the adequately protected from the effects of high energy line effects of high energy line breaks breaks and moderate energy line and moderate energy line cracks cracks in the pressurizer relief in the pressurizer relief system.

system.

20 Section The SFP, with a capacity of The SFP, analyzed for a Per TS 5.6.4 the capacity of the SFP 2.5.4.1 Tech 2485 fuel assemblies, can bounding capacity of 2485 fuel is 1849+255+294=2398 Eval (Page accommodate a full-core of 217 assemblies, can accommodate

65) along with 2268 previously a full-core of 217 along with I

discharged fuel assemblies from 2268 previously discharged fuel

Attachment to W3Fl -2005-0001 Page 9 of 27 Location Existing Text Recommended Text Bases 22 previous refueling batches or assemblies from 22 previous partial core of 116 fuel refueling batches or partial core assemblies along with 2369 of 116 fuel assemblies along previously discharged fuel with 2369 previously discharged assemblies from 23 previous fuel assemblies from 23 previous refueling batches refueling batches 21 Section The licensee also confirmed The licensee stated that the Reference November 13, 2003 2.5.5.4 Tech that fluid flow instabilities will iyte-m flow rates considered Submittal, Section 2.5.6.4.

Eval (Page not occur as a result of the in the FW waterhammer _

'The system flow rates considered in

74) proposed power uprate based analysis bound the feedwater the FW waterhammer analysis on transient computer code system flow rates under EPU bounds the feedwater system flow modeling and analyses.

conditions. Thus the staff rates under EPU conditions.

concludes that fluid flow Therefore, the FW System and instabilities will not occur as a related components are qualified to result of the proposed power handle the waterhammer loads uprate.

under EPU conditions."

22 Section As far as CFS design limitations, As far as CFS design limitations, Reference the July 14, 2004, 2.5.5.4 Tech the licensee found that the the licensee found that the drain Supplement Cover Letter. "Ongoing Eval (Page m

oisture separatorh6t valves for the MSR drain tank analysis indicates that MSR safety

74)

(MSR) safety relief valves, the and for feedwater heater #6, and valve replacement is not necessary."

drain valves for the MSR drain the feedwater heater level tank and for feedwater heater #6, control valves may be and the feedwater heater level undersized for post-EPU control valves may be operation and may have to be undersized for post-EPU modified or replaced in order to operation and may have to be satisfy flow requirements; modified or replaced in order to satisfy flow requirements; 23 Section Table lists peak EPU pressure Table should list peak EPU Reference the July 28, 2004, 2.6.3.2 Tech and temperature as 41.83 psig pressure and temperature as Supplement Cover Letter Eval (Page and 394.4jF respectively.

4.991ig and 194.7F,

'The revised volume increases the

86) respectively.

MSLB peak pressure... from 41.83 psig to 41.88 psig... The MSLB peak containment temperature

Attachment to W3F1-2005-0001 Page 10 of 27 Location Existing Text Recommended Text Bases increased slightly from 394.4-F to 394.7-F."

24 Section 2.6.6 Table entry, "Maximum Table entry should be, Reference the May 12, 2004 Tech. Eval.

RWSP/Spray Temperature (OF)"

uMinimIiuii RWSP/Spray Supplement, Attachment 1, (Page 89)

Temperature (OF)"

Response to Question 1:

'Minimum RWSP / spray temperature" 1

25 Section The NRC staff further concludes The NRC staff further concludes States that SAFDL is not exceeded.

2.8.5.1.1, that the licensee has that the licensee has However, the increase in MS flow Summary demonstrated that the reactor demonstrated that the reactor with a LOOP does violate SAFDL (Page 115) protection and safety systems protection and safety systems and result in fuel failure as stated in will continue to ensure that the will continue to ensure that for Tech Eval section just prior to the SAFDLs and the RCPB pressure thAeAOO the SAFDlLs and the summary.

limits willi bnot be Fxcded as a RCPB pressure limits will not be Reference October 18, 2004 result of these events.

exceeded as a result of these Supplement, Attachment 1.

events. For'thi event with al LOOP, SAFDL violation is of sufficiently limited extent thaI radiological consequences are acceptable!

26 Section However, the analysis results for However, the analysis results for Fuel melt was initially predicted for 2.8.5.1.2, the Steam System Piping Failure the Steam System Piping Failure SLB-Post Trip analysis not the SLB-Tech Eval.

PweTrip Excursion

- Post-Trip analysis indicated Pre Trip Power Excursion.

(Page 116) indicated that there would be that there would be some fuel Acceptance criteria referenced in some fuel damage due to damage due to centerline Section 2.13.1.3.3.2 is for the SLB-centerline melting. This was not melting. This was not pre Trip Excursion only.

acceptable to the staff, since the acceptable to the staff, since the Reference November 13, 2003 licensee's fuel response licensee's fuel response Submittal, Sections 2.13.1.3.1 and methodology is not approved for methodology is not approved for 2.13.1.3.3:

modeling centerline melt, such modeling centerline melt, such that a coolable geometry can be that a coolable geometry can be shown to be maintained in the shown to be maintained in the core, as the fuel rod centers melt core, as the fuel rod centers melt and expand. This wasalso a and expand.

Attachment to W3F1 -2005-0001 Page 11 of 27 Location Existing Text Recommended Text Bases violation of the licensee's acceptance criteria, defined ii Section 2.13.1.3.3.2 (fuel temperature must be

• fuel centerline melt temperature, as demonstrated by PLHR

• 21.0 kWlft)'

27 Section As a result, the licensee As a result, the licensee agreed This was not documented as a 2.8.5.1.2, committed to design the core to design the core loading (e.g.,

regulatory commitment but instead Tech Eval.

loading (e.g., to build in sufficient to build in sufficient scram will become the licensing basis with (Page 116) scram worth), in each reload worth), in each reload core the approval of EPU thus to design a core design to assure that none design to assure that none of the future core with fuel melt for a SLB of the analyzed SLB cases would analyzed SLB cases would lead would require prior NRC approval.

lead to any fuel centerline to any fuel centerline melting Reference October 18, 2004 melting.

(e;.g.,FLHR S24kW/ft)

Supplement, Cover Letter and :

On August 25, 2004, Entergy and members of your staff held a call to discuss the results of the analysis submitted in Reference 1 and the assumptions used for these analyses. As a result of the call, Entergy agreed to reanalyze the main steam line break return to power event with a revised assumption regarding fuel failures 28 Sect. 2.8.5.2.1 Results:

Results:

1177 should be 1186 psia for max Tech Eval, max RCS press = 2732 psia max RCS press = 2732 psia SG pressure without LOOP..

Table, "2.1.3 max SG press = 1177isa-max SG press = fi86psiia Reference November 13, 2003 LOCV" (Page Submittal, Section 2.13.2.1.3.6:

118)

"The peak RCS and SG pressures remained below their respective acceptance criteria of 2750 psia and

Attachment to W3F1 -2005-0001 Page 12 of 27 Location Existing Text Recommended Text Bases 1210 psia. The peak RCS pressure case yielded a peak RCS pressure of 2732 psia. The peak SG pressure case showed a peak SG pressure of 1186 Psia, including accounting for cases with 1 or 2 inoperable MSSVs.

The above cases were also run with a LOOP at the time of the HPPT.

The peak RCS pressure case with a LOOP yielded a peak RCS pressure of 2731 psia. The peak SG pressure case with a LOOP had a peak SG pressure of 1177 psia. Thus, the LOCV with LOOP case is bounded by the LOCV case without LOOP."

29 2.8.5.2.4, The licensee must have The licensee currently has The need to secure charging is Tech Eval 4t procedures by which the procedures by which the

'conditioned' based and keys on par before operators can recognize the operators can recognize the pressurizer level therefore the equation condition and act to turn off the condition and act to turn off the procedure will not specify that the (Page 124) charging pumps within the 12 charging pumps priorFto(filgii operator secure charging within 12 minutes allowed by these th pressurizer (e.gi within the minutes. Instead it directs that the analysis results.

12 minutes) as allowed by these operator secure charging prior to analysis results.

filling the pressurizer which, according to the analysis, could be necessary within 12 minutes.

Reference the October 18, 2004 Supplement, Attachment 3:

"Standard post reactor trip actions instruct the Operator to maintain pressurizer level within 33% and 60% by, among other things, operating charging pumps as necessary. This is a continuous action performed throughout the

Attachment to W3Fl-2005-0001 Page 13 of 27 Location Existing Text Recommended Text Bases EOPs whenever pressurizer level is challenged and regardless of which individual EOP is being used. Since Operators continuously monitor pressurizer level during the accident and a high pressurizer level alarm exists, there is high confidence that charging pumps can be secured within the required time to prevent filling the pressurizer."

30 Section The licensee's methodology The licensee's methodology Reference August 10, 2004 2.8.5.3.2 points to calculations that predict points to calculations that predict Supplement Cover Letter and Tech. Eval.

less than 8 percent of the fuel less than 15 perdent of the fuel :

(Page 127) rods will fail. This is the extent of rods will fail. This is the extent "Extended Power Uprate (EPU).

fuel failure that can be incurred of fuel failure that can be Section 2.13.3.3.1, "Single Reactor without violating the offsite dose incurred without violating the Coolant Pump (RCP) Shaft limits.

offsite dose limits.

Seizure/Sheared Shaft," reported a fuel failure limit of 8%.

Reexamination of the radiological consequences analysis for this event indicates that the fuel failure limit can be increased to 15% to better reflect the EPU approach of reporting in the licensing documentation the fuel failure amount which corresponds to the regulatory dose limits for the event."

31 2.8.5.6, None This section of the Draft SE does Analysis for "Small Primary Line

'Decrease in not include any discussion Break Outside of Containment' was Reactor regarding the "Small Primary provided in the EPU submittal.

Coolant Line Break Outside of Reference November 13, 2003 Inventory,"

Containment." Was this Submittal, Section 2.13.6.3.1.

(Pages 139 -

intentional or an oversight?

S

Attachment to W3F1-2005-0001 Page 14 of 27 Location Existing Text Recommended Text Bases 149)9 32 2.8.5.6.3 Tech The staff reviewed the ECCS The staff reviewed the ECCS A LOCA is not 'normal reactor Eval 1st para performance to confirm that the performance to confirm that the operation" nor is it an uAOO."

(Page 142) system design provides an system design provides an acceptable margin of safety from acceptable margin of safety 6ndi-tio'ns-whichwould lead_

follbwFn-gia IC0 and has to fuel damage during normal been accomplished using reactor operation or AOOs and acceptable analytical methods.

has been accomplished using

'acceptable analytical methods.

33 Section The licensee performed these The licensee performed these Previous LBLOCA analyses did not 2.8.5.6.3 and previousi:O'CA analyses EC-db analyses using use Reference 61.

Tech. Eval.

using Westinghouse/CE approved Reference the November 13, 2003 (Page 142)

Westinghouse/CE approved LBLOCA and SBLOCA Submittal, Section 2.12.13:

LBLOCA and SBLOCA methodologies described in uThe LBLOCA ECCS performance methodologies described in Refere nce 6'1 -ahd analysis used the 1999 Evaluation e

a8and Reference 60, respectively.

Model (EM) version of the Reference 59, respectively.

Westinghouse LBLOCA evaluation model for Combustion Engineering (CE) PWRs. The current Waterford 3 LBLOCA ECCS performance analysis, described in Sections 6.3.3.2 and 15.6.3.3.3.1 of the Waterford 3 FSAR (Reference 2.12-3), employs the June 1985 version of the Westinghouse LBLOCA EM for CE PWRs (Reference 2.12-4), which is the version of the evaluation model upon which the 1999 EM is built."

Reference the July 14, 2004 Supplement, Attachment 5:

"The Small Break Loss-of-Coolant Accident (SBLOCA) Emergency

Attachment to W3F1 -2005-0001 Page 15 of 27 I Location I Existing Text I Recommended Text I Bases Core Cooling System (ECCS) performance analysis used the Supplement 2 version (referred to as the S2M or - Supplement 2 Model) of the Westinghouse SBLOCA ECCS Evaluation Model for Combustion Engineering (CE) Pressurized Water Reactors (Reference 2.12-23). This is the same methodology used in the current licensing basis Waterford 3 SBLOCA ECCS performance analysis (Reference 2.12-30)."

34 2.8.5.6.3 Tech Eval, 4 h para (Page 142)

The licensee's commitment to implement processes to assure appropriate input to the analyses (in the following paragraph) further substantiates the applicability of the S2M methodology to Waterford 3.

The licensee's ongoing process to assure appropriate input to the analyses (in the following paragraph) further substantiates the applicability of the S2M methodology to Waterford 3.

This was documented in the supplement as an existing uongoing process" not as a regulatory commitment.

Reference the July 28, 2004,, Response to Question 5:

"2) Waterford 3 and its vendor, Westinghouse Electric Company LLC, continue to have ongoing processes, which assure that LOCA analysis input values bound the as-operated plant values for those parameters."

Reference the July 28, 2004,, "List of Regulatory Commitments" does not list this as a regulatory commitment.

35 2.8.5.6.3 Tech At the staff s request, the At the staffs request, the Reference the July 28, 2004, Eval, 1st para

[IiiCensee, in Reference 12, l

iIcensee, in Reference 1, Response to

Attachment to W3F1 -2005-0001 Page 16 of 27 Location Existing Text Recommended Text Bases after SBLOCA stated that the results of its provided Information enabling Question 6:

results Table assessment indicate that the the staff to conclude that the The Licensee did not provide a (Page 143) value of the calculated pre-value of the calculated pre-summation of the pre and post transient oxidation for the transient oxidation for the oxidation nor did the Licensee Zircaloy fuel through the final Zircaloy fuel through the final compare such a summation to the cycle is sufficiently small that the cycle is sufficiently small that the 17% acceptance criterion.

sum of the pre-transient and sum of the pre-transient and post-LOCA oxidation for the__

post-LOCA oxidation for the zircaloy fuel is less than the 17 zircaloy fuel is less than the 17

'percent' local oxidation percent local oxidation criterion criterion ofl10CFR 50.46(b')(2).

of 10 CFR 50.46(b)(2).

36 2.8.5.6.3 Tech With regard to the 1 percent With regard to the 1 percent Waterford 3 does not use Eval, 2nd para core-wide hydrogen criterion of core-wide hydrogen criterion of Framatome fuel.

after SBLOCA 10 CFR 50.46(b)(3), the staff 10 CFR 50.46(b)(3), the staff results Table notes that the pre-existing notes that the pre-existing (Page 143) oxidation of the Fraimatome fuel oxidation of the Wetinihouise would not contribute to the LOCA fuel would not contribute to the maximum core-wide hydrogen LOCA maximum core-wide generation.

hydrogen generation.

37 2.8.5.6.3 Tech

1. Lists the Waterford 3 time to

1. The Waterford 3 time to Reference November 13, 2003

Eval, saturation as 3.0_oiuiis saturation is U5houri.

Submittal, Section 2.12.5.3:

Comparison of ult shows that without simultaneous H3B03

4. Lists the Waterford 3

4. The Waterford 3 saturation hot and cold leg injection, the boric Accumulation saturation limit at 23.58 3

limit is B76.

acid concentration in the core Characteristics hours:!

exceeds the solubility limit at Table, Items 1 approximately 4.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> post-LOCA.

"Time to When 372 gpm of simultaneous hot Reach and cold leg injection is initiated at 3 Saturation" hours post-LOCA, the maximum and 4 boric acid concentration in the core is "Assumed 23.58 w/o, a margin of 4.02 w/o to H3B03 the solubility limit of 27.6 w/o."

saturation limit (wo)

/

Attachment to W3F1 -2005-0001 Page 17 of 27 Location Existing Text Recommended Text Bases (Page 146) 38 Section 2.9 None This section of the Draft SE does Analysis for these events was not contain discussion of:

provided in the EPU submittal

- Inadvertant ADV Opening (Sections 2.13.1.2.4, 2.13.1.2.3, and

- Excess MS Flow w/ LOOP 2.13.2.3.1 respectively) and

- FW Line Break supplements as well as the AST Was this intentional or an Submittal and Supplements. Note:

oversight?

IADVO and FWLB can be considered to be bounded by the Inside Containment MSLB case, based on AST results (July 15, 2004 AST Submittal). Excess Main Steam, Flow with LOOP can be considered bounded by CEA Ejection.

39 Section 2.9.1 Other inputs and assumptions None Reference e-mails from Bryan Tech Eval 2nd were unchanged from the Miller to Kaly and Paul Sicard to para (Page original Waterford 3 design basis Michelle dated 12115/04 and 154) as specified in UFSAR 12116104 respectively.

Section 11.1. The NRC staff Section 2.7.1, "Source Terms for finds that the licensee has used Radwaste Systems Analysis," of the the appropriate core power November 13, 2003, Submittal assumptions for the EPU. The stated:

NRC staff also finds that the "Fission product escape coefficients,,

EPU would not impact any of the purification flow rate, and ion other inputs and assumptions to exchange removal efficiency are the maximum coolant unchanged from the original concentration calculations, SO Waterford 3 design basis as continued use of the curient specified in Section 11.1."

USAR values is acceptable' The post-EPU product escape coefficients and the ion exchange removal efficiencies did change for some of the radionuclides. The post-EPU results submitted in Table 2.7-1, "Comparison of Maximum Reactor

Attachment to W3F1 -2005-0001 Page 18 of 27 Location Existing Text Recommended Text Bases Coolant Radionuclide Concentrations Based on 1 % Fuel Failure," are correct since they were the result of the analysis performed using the changed product escape coefficients and the ion exchange removal efficiencies. Therefore, no conclusions are impacted by the inaccuracy contained in the submittal.

Information to be docketed with SGTR 3-sec TD RAI response in mid-January.

40 Section 2.9.1 Entergy continued to use the Entergy continued to use the Note that the TID-14844 source term Tech Eval, 6h Waterford 3 current design basis Waterford 3 current design basis is applicable for LOCA. An ORIGEN para (Page Technical Information Document Technical Information Document based source term has been used 155)

TID-14844 core isotopic source TID-14844 core isotopic source for clad activity pre-EPU and, as term, adjusted for the increase in term for LOCA, adjusted for the described in the AST submittal July thermal power to 100.5 percent increase in thermal power to 15, 2004, for the AST submittal in of the current RTP or 3735 MWt 100.5 percent of the support of EPU.

to bound the current Waterford 3 EPU RTP or 3735 MWt to bound power measurement uncertainty.

the current Waterford 3 power measurement uncertainty.

41 Section 2.9.2 The NRC staff further concludes The NRC staff further concludes Fuel failure was not assumed for Summary that the plant site and the that the plant site and the MSLB outside of containment.

(Page 156) dose-mitigating ESFs remain dose-mitigating ESFs remain Reference the October 18, 2004 acceptable with respect to the acceptable with respect to the Supplement, Attachment 1:

radiological consequences of a radiological consequences of a "The impact of the above changes.

postulated MSLB iutside postulated MSLB.1nid0 along with the iterated SCRAM worth containment, which assumes fuel containment, which assumes results in a small number of fuel pins failure, since the calculated fuel failure, since the calculated predicted to experience DNB SAFDL whole-body and thyroid doses at whole-body and thyroid doses at violation for inside containment (IC) the exclusion area boundary the exclusion area boundary break location LOOP cases. No (EAB) and the low population (EAB) and the low population violation of SAFDLs occurs for inside

Attachment to W3F1-2005-0001 Page 19 of 27 Location Existing Text Recommended Text Bases zone (LPZ) outer boundary meet zone (LPZ) outer boundary meet containment no-LOOP cases or the the exposure guideline values the exposure guideline values outside containment (OC) break specified in 10 CFR 100.11.

specified in 10 CFR 100.11.

locations."

42 Section 2.9.3 Entergy states that cycle-specific Entergy states that cycle-specific Reference August 10, 2004 Tech. Eval.

fuel failure evaluations for future fuel failure evaluations for future Supplement Cover Letter and (Page 157) cores will be performed to cores will be performed to :

ensure that the 8 percent ensure that the 15 percent "Extended Power Uprate (EPU).

allowed fuel failure limit will not allowed fuel failure limit will not Section 2.13.3.3.1, "Single Reactor be exceeded.

be exceeded.

Coolant Pump (RCP) Shaft Seizure/Sheared Shaft," reported a fuel failure limit of 8%.

Reexamination of the radiological consequences analysis for this event indicate's that the fuel failure limit can be increased to 15% to better reflect the EPU approach of reporting in the licensing documentation the fuel failure amount which corresponds to the regulatory dose limits for the event."

43 2.9.11 For the offsite dose analysis of For the offsite dose analysis of Conclusion contradicts Draft SE Conclusion DBAs, eicept the SGTR, the DBAs the NRC staff finds that Section 2.9.6, 'Radiological (Page 167)

NRC staff finds that Entergy Entergy used analysis, methods, Consequences for Steam Generator used analysis, methods, and and assumptions consistent with Tube Rupture."

assumptions consistent with the the conservative regulatory conservative regulatory requirements and guidance requirements and guidance identified in Section 2.0 above.

identified in Section 2.0 above.

44 Table 2.9-2 1-131 0.12 All isotopes (except Kr-85) 0.10 Reference Reg. Guide 1.89, through Table Kr-85 0.30 Kr-85 0.30 Section C.2.c(2):

2.9-4, gap Other iodines and noble gases 10% for all isotopes except for Kr-85, fractions, 0.10 for which 30% should be assumed.

(Pages 169-171)

Attachment to W3F1-2005-0001 Page 20 of 27

[ Location I Existing Text I Recommended Text I Bases 45 Table 2.9-2 through Table 2.9-5 (Pages 169 - 172) primary coolant mass, lb 480,000 SG initial water mass, lb 153,700 priary coolant mass, lb 433,778 SG initial water mass, lb 241,450 Reference the October 13, 2004, AST Supplement, Attachment 1, Response to Question 3.

uExcept for SGTR, the dose analyses assuming primary-to-secondary leakage assume the following Reactor Coolant System (RCS) masses:

34,260 Pressurizer Liquid Mass (Ibm) 4,016 Pressurizer Steam (ibm) 395,502 Non-Pressurizer Liquid Mass (Ibm)"

433,778 Total (Ibm)

Reference the October 13, 2004, AST Supplement, Attachment 1, Response to Question 2" The assumed secondary inventory values of 153,700 Ibm at Hot Full Power (HFP) and 241,450 Ibm at Hot Zero Power (HZP) are values of the liquid mass for a single SG. The larger HZP value was considered more representative of the SG mass present in an intact SG that is being cooled to Shutdown Cooling (SDC) entry conditions, and maximizes the initial activity present for the case of releases from a faulted SG.

Therefore, the HZP values were used as the basis for all of the release calculations excent for Steam Generator Tube Rupture (SGTR).

Attachment to W3F1 -2005-0001 Page 21 of 27 Location Existing Text Recommended Text Bases 46 Table 2.9-4, Percent containment leakage Percent containment leakage Reference July 15, 2004 AST Percent filtered, percent 92 filtered, percent 94 Submittal, Section 10.0, CEA Containment Ejection:

Leakage "The containment building release Filtered, (Page pathway model is similar to and 171) derived from the LBOCA release model."

Reference July 15, 2004 Submittal, Table 5-1, Assumptions Used for LOCA Radiological Analysis:

"Containment Leakage Pathway:

Controlled Ventilation Area System (CVAS) 54%

Filtration (Reactor Auxiliary Building)

Shield Building 40%

Unfiltered Direct Bypass 6%"

47 Table 2.9-6 Primary coolant mass, Ibm Primary coolant mass, Ibm Reference the October 13, 2004, Primary R8-0O,0 6467,00 AST Supplement, Attachment 1, Coolant Mass Sec. coolant mass, Ibm Sec. coolant mass, Ibm Response to Question 3.

(Page 173) 307,400

.106,300 "For SGTR, the dose analyses are based on the detailed CENTS analysis performed in support of the information presented in Section 2.13.6.3.2 of the Extended Power Uprate (EPU) report submitted'via W3F1-2003-0074 dated November 13, 2003. The total initial RCS mass of 467,000 lb is assumed for that analysis."

Reference the October 13, 2004, AST Supplement, Attachment 1, Response to Question 2" "For the SGTR event, a low initial SG level (corresponding to 106,300 Ibm

Attachment to W3F1 -2005-0001 Page 22 of 27 Location Existing Text Recommended Text Bases per SG) was conservatively assumed since this results in earlier uncovery of the top of the SG tubes, which results in increased flashing and a lower decontamination factor."

48 Table 2.9-7, Primary Containment Free Volume (Page 174)

Elemental iodine spray removal coefficient, hr' Injection phase 1.82 Recirculation phase 1.35 Organic iodine spray removal coefficient, hr' 0.4 Particulate iodine spray removal coefficient, hr' Initially 3.9 When <4 percent of initial concentration 6.75 Elemental iodine spray removal coefficient, hr' Injection phase 20 Recirculation phase 13.3 73 Organic iodine spray removal coefficient, hr1' Particulate iodine spray removal coefficient, hr-'

Initially 3.5 When <4 percent of initial concentration 0.35 Is this table supposed to be assumptions used for EPU analysis before the AST submittal or assumptions used in the AST analysis?

For EPU (pre-AST) Spray removal coefficients are per NUREG-0800 SRP Section 6.5.2.

Reference the May 7, 2004 Supplement, Attachment 1, Response to Question 15:

"Was spray removal assumed in the containment for the LOCA and for the CEA ejection? If so, what spray removal assumptions were used?

Response

Containment spray is credited for the removal of fission products in containment in the LOCA radiological dose calculations. Spray removal coefficients per NUREG-0800 SRP Section 6.5.2 are used.

Specifically, the formula from Section 6.5.2 is used for particulate and elemental Iodine. The maximum allowed elemental spray removal value of 20/hr is assumed and supported for Waterford 3; a maximum decontamination factor of

Attachment to W3F1 -2005-0001 Page 23 of 27 Location Existing Text Recommended Text Bases 200 is imposed per the SRP. The SRP formula for spray removal of particulate iodine is applied; this value is reduced by a factor of 10 per the SRP guidance once a decontamination factor of 50 has been achieved. No spray removal is assumed for organic Iodine, consistent with the SRP guidance.

No credit is taken for containment spray removal of iodine for the CEA Ejection analysis."

Reference the October 19, 2004, AST Supplement, Table 4-1.

Particulate 3.596Thr (until PF = 50) 0.3596/hr (once PF > 50) 49 Section 2.11

'There would be a change in "ADVs will have a TS-mandated Reference the July 14, 2004 (Page177) the controllers-to the ADV automatic setpoint, and Supplement Cover Letter.

being converted to digital operators will be required to uAs a result of the reanalysis of the

'controllers for, improved perform channel checks once post-EPU SBLOCA, Entergy has accuracy. The new contiollers every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to ensure ADV determined that digital ADV will have a TS-mandated automatic actuation operability controllers will not be needed in automatic setpoint, and when operating above 70 support of the EPU."

operators will be required to percent RTP."

perform channel checks once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to ensure ADV automatic actuation operability when operating above 70 percent RTP."

50 Section 2.11 "Table 2-11.1 of the submittal, uTable 2-11.1 of the submittal, Reference the April 15, 2004, third para titled "impact of EPU on HRA titled "Impact of EPU on HRA Supplement Response to (Page177)

[Human Reliability Analysis]

[Human Reliability Analysis]

Question 1.

_ Time Available," listed several Time Available," listed several "It was assumed that there was not

Attachment to W3F1 -2005-0001 Page 24 of 27 I Location I Existing Text I Recommended Text I Bases

, operator recovery actions for which the available time to accomplish each task will be reduced as a consequence of t he EPU. Because t fe~ii iis available for successful completion are decreased the licensee stated that those actions are assumed to be failed in their risk assessment, and the failure probabilities for those actions have been set to 1.0.

le'retfo-r; credit is not taken in the licensee's risk assessment for manuial recovery actions with-decreased available' time-s under the EPUV operator recovery actions for which the available time to accomplish each task will be reduced as a consequence of the EPU. Because thimUfies i1ib-lefor successful completion for these ev aint are shorter than the times required to perform the actions,4 the licensee stated that those actions are assumed to be failed in their risk assessment, and the failure probabilities for those actions have been set to 1.0. Althoughtheremaining opertor actions are shown tin the table to have increased times available for EPU, this Is a result'of conservatisms'in the original (pre-EPU) modeF.

These operator actions are also expected to have decreased times available asa result of EPU, but the impact of these time decreases is small compared to the effect

'of decreases in loss of offsite power recovery times, which are explicitly included in the licensee's risk assessment for EPU."

enough time available to successfully accomplish the tasks and thus the failure probabilities for these actions were set to 1.0."

uThe post-power uprate times (14 and 2.83 minutes) are from realistic, plant-specific thermal-hydraulic analyses performed in support of the power uprate risk assessment. The pre-and post-uprate times for these events are not at all comparable.

Table 2.11-2 shows the estimated decrease in time available for three operator actions for which pre-and post-uprate times were estimated using the same plant-specific thermal-hydraulic analysis method; these time changes are more representative of the possible effects of power uprate."

51 Section 2.11 Table 2.9-1 of Reference 1 Table 2.9-1 of Reference 1, as Reference the May 7, 2004 The last provided a summary of the m

odfi hd by Refierenc-s 5 and Supplement Cover Letter.

paragraph on control room controls, displays, 10, provided a summary of the "The "after EPU" QSPDS Cold Leg

Attachment to W3FI-2005-0001 Page 25 of 27 Location Existina Text Recommended Text Bases the page and alarms that would be control room controls, displays, Temperature alarm is changed from (Pagel 77) affected with details on the and alarms that would be 562F to 549F.'

setpoints/ranges before and after affected with details on the the EPU. TiheTable includes setpoints/ranges before and Reference the July 14, 2004 the change from analog to_

after the EPU.

Supplement Cover Letter and digital control for the ADV, as.

discussed previously) uAs a result of the reanalysis of the post-EPU SBLOCA, Entergy has determined that digital ADV controllers will not be needed in support of the EPU."

52 Section 2.12.1 Generator Output Breakers - The Generator Output Breakers -

Reference January 29, 2004 Tech. Eval.

high capacity gas circuit The existing oil circuit Supplement, Attachment 1:

(Page 185) breakers will be replaced to breakers will'be repla With "Replace both existing generator accept the proposed EPU high capacity gas circuit output oil circuit breakers with higher operating conditions. The PMT breakers to accept the proposed capacity gas circuit breakers.

will be executed to verify proper EPU operating conditions. The (generator output breaker 'B' was operation of trip circuits, perform PMT will be executed to verify replaced and tested during RF12 in AC and DC acceptance tests, proper operation of trip circuits, the Fall of 2003)."

perform electrical tests to verify perform AC and DC acceptance breaker operation, and verify tests, perform electrical tests to proper calibration of verify breaker operation, and synchronization check circuit.

verify proper calibration of synchronization check circuit.

53 Section 2.12.1 SIT-TP-724, "Temperature SIT-TP-724, "Temperature The DNB limit is not changing (TS Summary Decalibration Verification" - The Decalibration Verification" - The 2.1.1.1). The "latest values of DNB" (Page 187) algorithms contained in the CPC algorithms contained in the CPC are the estimated" DNB values were unchanged for the EPU.

were unchanged for the EPU.

determined by the "update" algorithm TCPI be updated The' "update" algorithm within for the existing core conditions at reflect the latest values of DNB CPCs accommodates for that time during plant operation.

and will not impact the changes in cold leg Reference the January 29, 2004, minimum and maximum cord_

temperature. The DNBR Supplement, Attachment 4:

leg temperature' limits required algorithm limits on minimum "The "update" algorithm within CPCs by TSsI and maximum cold leg accommodates for changes in cold

Attachment to W3F1-2005-0001 Page 26 of 27

[ Location I Existing Text I Recommended Text l Bases temperature of 4950F and 580?F are' not impacted by thee proposed change in Technical Specification limits on cold leg tempe'rature and thus do not require revision for EPUI leg temperature and applies a conservative temperature shadowing factor. Therefore, no special test is required."

Reference the May 7, 2004, Supplement, Attachment 2, Response to Question 12:

'The CPC algorithms are unchanged for EPU, as stated in letter W3F1-2004-004 dated January 29, 2004.

The "UPDATE" algorithm is a-program within the CPC System that updates the latest values of departure from nucleate boiling ratio (DNBR) and quality margin available from another CPC algorithm (the "STATIC" program) and determines the DNBR and local power density based on current temperature, pressure, core power, core flow, and power distribution. The "UPDATE" program does not calculate an actual DNBR; this is done by the "STATIC" program. However, the "UPDATE" program looks at the rate of change of the input variables, including cold leg temperature, to calculate the rate of change of DNBR over the 100 milliseconds it takes to run "UPDATE." This allows updates for the changes in DNBR to be done 20 times over the two seconds required for the "STATIC" program to run a detailed DNBR calculation. The U __________________________________

L

Attachment to W3FI-2005-0001 Page 27 of 27 Location Existing Text Recommended Text Bases DNBR algorithm limits on minimum and maximum cold leg temperature of 4950F and 5800F, documented in Technical Specification Section 2 Bases, are not impacted by the proposed change in Technical Specification limits on cold leg temperature and thus do not require revision and are not being revised for EPU."

54 Section The EPU will modify the The EPU will rescale plant Reference the July 14, 2004 2.12.1.3 Eval remnoteshutdown panel with instrument meters to reflect the Supplement Cover Letter.

Section on new ADV controllers and new EPU operating parameters.

"As a result of the reanalysis of the Page 188 rescale plant instrument meters post-EPU SBLOCA, Entergy has to reflect the new EPU operating determined that digital ADV parameters.

controllers will not be needed in support of the EPU."

55 Section 2.12.1 Main Condenser - Additional Main Condenser - Additional Reference the January 29, 2004, Tech Eval, main condenser tube support will main condenser tube support will Supplement, Attachment 1:

(Page 185) be added to minimize the effects be added to minimize the effects

" System: Main Condenser of flow-induced vibration for the of flow-induced vibration for the Modification: Additional support proposed EPU operating proposed EPU operating staking of the main condenser tubes conditions. The PMT will consist conditions. The PMT will consist to minimize effects of flow induced of monitoring secondary system of monitoring secondary system vibration.

viIbratio~n parameters.

Elihiiilstry parameters.

Testing: Monitor secondary chemistry."