ML051100279

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E-mail from Bryan D. Miller Regarding Supplemental Draft Information on Instrument Uncertainty
ML051100279
Person / Time
Site: Waterford Entergy icon.png
Issue date: 04/18/2005
From: Miller D
Entergy Nuclear Operations
To: Alexion T, Kalyanam N
NRC/NRR/DLPM/LPD4
References
Download: ML051100279 (26)
v • d • e


Text

[Tfi-o-m-'as Alexion -------

- RE: ----

SUPPLEMENTAL DRAFT INFORMATION -- ON INSTRUMENT UNCERTAINTY . .......-

1--1---------------------- Pane-- 1 l Pagei IThornas Alexion-RE: SUPPLEMENTAL DRAFT INFORMATION ON INSTRUMENT UNCERTAINTY From: "MILLER, D BRYAN" <dmilIl4@entergy.com>

To: "'Thomas Alexion"' <TWACnrc.gov>, "'KALYANAM, N. KALY'" <nxkCnrc.gov>,

"'aghl @nrc.gov"' <aghl @nrc.gov>

Date: 4/18/05 9:11 PM

Subject:

RE: SUPPLEMENTAL DRAFT INFORMATION ON INSTRUMENT UNCERTAINTY As discussed during last Friday's conference call the supplemental draft information on instrument uncertainty is attached.

'Bryan

Original Message-----

From: Thomas Alexion [1]

Sent: Monday,'April 18, 2005 3:43 PM To: MILLER, D BRYAN

Subject:

SUPPLEMENTAL DRAFT INFORMATION ON INSTRUMENT UNCERTAINTY Bryan, (I haven't seen anything yet?)

Please e-mail the information to me, Kaly, and Allen Howe. Allen's e-mail is aghl @nrc.gov. (I'll be in a training class tomorrow.)

Tom

Ic-itemppGW}Q O1J MP .1 PA 1 Mail Envelope Properties (42645AD6.FE1:10: 20449)

Subject:

RE: SUPPLEMENTAL DRAF INFORMATION ON INSTRUMENT UNCERTAINTY Creation Date: 4/18/05 9:11PM From: "MILLER, D BRYAN" <dmilll4@entergy.com>

Created By: dmnill4@entergy.com Recipients nrc.gov owf4_po.OWFNDO AGHi (Allen Howe)

NXK (N. Kaly Kalyanam)

TWA (Thomas Alexion)

Post Office Route owf4_po.WFNDO nrc.gov Files Size Date & Time MESSAGE 507 04/18/05 09:11PM Draft for NRC 4-18-05.pdf 126305 Mime.822 175076 Options

-Expiration Date: None Priority: Standard Reply Requested: No Return Notification: None Concealed

Subject:

No Security: Standard

[Insert Correspondence Number]

[Insert Date]

U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

License Amendment Request Extended Power Uprate (Aeffl~ent 199) License Condit-i ,egarding Instrument Uncertainty Waterford Steam Electric Stati nit 3 Docket No. 50-382 License No. NPF-38

REFERENCES:

1. NRC letter to M n bnabie d April 15, 2005, "Waterford Steam Electric St Unin mendment Re: Extende d PO aUte (TA o. M 5)~

i we Dear Sir or Madam.

Pursuant P 50. te (Entergy) hereby requests that the

.license n e r~din ostrument uncertainty that was imposed on the Waterford Stear n

--Electr ion, Uni Wate ) license in Reference 1 be deemed complete and rem from the3Wat ord3Iiie Referen pproved the ende ower Uprate (EPU) for Waterford 3 and, as part of thE

-approval, ime the folioWig license condition:

3. As state ,6e ee's letter dated February 5, 2005, the licensee committed as follows: P eeeding 3441 MWt, Entergy will submit, for NRC review and approval, a de'ription of how Entergy accounts for instrument uncertainty for each Technical Specification parameter impacted by the Waterford 3 Extended Power
  • Uprate."Accordingly, subject to completion of this condition, the licensee shall not operate the Waterford 3 facility at a power level exceeding 3441 MWt.

Descriptions of how Entergy accounts for instrument uncertainty for each Technical Specification parameter impacted by the Waterford 3 EPU are provided in Attachment 1.

The information has been evaluated in accordance with 10 CFR 50.91 (a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that the removal of the license condition

[Insert Correspondence Number]

Page 2 of X involves no significant hazards consideration. The bases for these determinations are included in the attached submittal.

Entergy requests approval of the proposed amendment by May 27, 2005 to support power ascension from the Spring 2005 refueling outage. Once approved, the amendment shall be

-implemented prior to exceeding 3441 MWt.

-Waterford 3 can not exceed 3441 MWt and achieve the EPU power level of 3716 MWt following the Spring'2005 refueling outage until the license conditiop osed in Reference 1

-is deemed complete and removed from the license. 'The need f f1icense amendment for this purpose wVas not recognized by Entergy or the 'NRC staff Iii ust prior to the issuance of the EPU license. Therefore, to avoid a derating of Water fo o ng restart from the Spring 2005 refueling outage, Entergy requests that this l coite a n ent request be reviewed and approved on an exigent basis.

If you have any questions or require additional ion, please contact an 'Miller at 504-739-6692.

I declare under penalty of perjury that the foregoing strueancrrect. Executed on [insert date].

Sincerely, J E. able Vice President, Ope 10s Waterford Steam Elect Sa Attachme

1. Analysis f 0oposed T clnical Specification Change

[Insert Correspondence Number]

Page 3 of X 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 4

Winston & Strawn Attn: N.S. Reynolds

'1700 K Street, NW Washington D 7 Louisiana D P ment of n ironme al Quc Office of Irnmental pliance Surveilac liin P.O0.Bo43 Bd igL 2-4312

- QtAtn: Library X

-THwn (Tenter Suit O0S 07-2445

Change DRAFT

Attachment 1 to

[Insert Correspondence Number]

Page 1 of 20

1.0 DESCRIPTION

This letter is a request to amend Operating License(s) NPF-38 for Waterford Steam Electric Station, Unit :3 (Waterford 3), to remove the license condition regarding instrument uncertainty that was imposed on the Waterford 3 with the approval and issuance of the Extended Power Uprate (EPU) amendment. The removal of the license condition will allow Waterford 3 to exceed 3441 MWt and achieve the EPU power level of 3716 MWt.

2.0 PROPOSED CHANGE

Remove license condition regarding instrument uncertaint a i posed on Waterford 3 with the approval and issuance of the EPU amendment. v

3.0 BACKGROUND

The amendment approving the EPU for Waterfollowing lic ondition:

3. As stated in the licensee's lettr dated Febru 5, the licensee committed as follows: "Prior to exceeding 3 tAX,,t Wt Entergyil.submit, for NRC review and approval, a description of how f tument uncertainty for each Technical Specification parame rli patdbythe Wae ord Extended Power Upra te. "Accordingly,_subject to lto~ot~odt the licensee shall not operate the Wat cility 41 MWt 4.0 TECHNIC ALYSIS In accordance with th t ty perations, Inc. (Entergy) is documenting the treatrE,,,f trume t sure u ainty for parameters which were revised in

- it erht to EPU analyses that fall within the following criteria:

parameter is lue cis measured using plant equipment; That is, the meteris direct dicated o operators using installed plant instrumentation.

  • The pw~ameter is a e which is specified by a Limiting Condition for Operation (LCO) of h Watefl 3 Technical Specifications. Parameters listed in Technical Specificato Swhich are called out by LCO's are considered within the scope of this effort. an LCO refers to values specified in the Core Operating Limits Report (COLR), such values would also be considered within the scope of this effort.

This criteria considers parameters which are pertinent to power uprate analyses, even if the value of the parameter is unchanged for EPU. That is, the parameter is considered if of at lleast moderate importance for analyses pertinent to the parameter (e.g., analyses discussed in Bases of Technical Specifications(TS)) which had to be reperformed to support EPU. This criteria would capture parameters for which margins to acceptance criteria for analyses discussed in the Bases of applicable Technical Specifications have been impacted for EPU.

DRAFT

Attachment 1 to

[insert Correspondence Number]

'Page 2 of 20 The parameter selection was discussed with the NRC staff during a conference call on

-14 April 2005. The NRC staff concurred with the list of parameters provided by Waterford 3, with'the proviso (agreed to by Entergy) that Containment Spray Riser Level (TS 4.6.2.1) also be included.

Values relating to applicability (e.g., MODES) of the Technical Specifications are considered to be generally out of the scope of the license condition. For example, Technical Specifications 3.2.1 through 3.2.4 for power distribution parameters (Azimuthal Power Tilt, Planar Radial Peaking, Linear Heat Rate, Departure from Nucleat ing Ration (DNBR) margin) are' designated as applicable above 20% of Rated Ther a, Power. The Entergy license condition scope will not include discussion of instrum u certainties with respect to that 20% power criteria.

Entergy' recognizes that safety analyses must accoun trument un ity in all cases.

Since the intent of many Technical Specifications ido rovi e assuranc h t the plant is within the assumptions of the accident analysis, itpropriate that the in tnent measurement uncertainties be accounted for i se manner owever, the of rigor applied to documenting the instrument uncertainz nd the abated accountinf the

7applicable analyses and procedures may vary ba h ty significance of the instrument function. Unlike limiting sfely system se iqg(SSS) values, there is no clear regulatory guidance describing spec 'thods at m e employed to address the instrument uncertainties associated wi 'Iaces of T Ical Specification parameters.

Waterford 3 has performed a categorizati n of Te i a pecifitlon parameters within the scope of the license con io. his categho e table below, also reflects discussions with the 2005 Parameters are classified as falling ionto one of fo t des regardi g reatmed instrumentation uncertainty:

XCategory Am A df. Oescription m nidered e n--ef8itF in analyses. There is an explicit

A M ie teeRhe chnical Sp i ication value and the value assumed in the

'A rtine 't 6he Technical Specification.

Instrument certaT isexplicitly considered in plant surveillance requirements or alarm respo

. eMproce i. There'is an explicit offset between the LCO value in Mshe Technical 'pecification and the value specified to be maintained in plant Xa rveillance prdures.

Ti'a' LCO valu iy also be the value assumed as initial conditions in safety C - anas an hg value specified to be maintained in plant surveillance procedures.

The~ value and the plant surveillance limit are the same and

'~ir1lSpecification D the para'er does not have an explicit analytical basis. The limited number of parameters in this category are based on engineering judgment.

More detail is provided below for the parameters of interest based on 14 and 15 April 2005

  • discussions with the NRC staff. 'None'of the Technical Specification parameters impacted by the Waterford 3 Extended Power Uprate are classified in Category C. Although HICB-12 may allow Technical Specification parameters to be the same value as assumed in the safety analyses and specified in the plant surveillance procedures, due to the timing of the review process, Entergy has explicitly applied offsets for instrument uncertainty in the analysis or in DRAFT

Attachment 1 to

[Insert Correspondence Number]

Page 3 of 20 surveillance procedures for the Technical Specification parameters impacted by the Waterford 3 Extended Power Uprate.

Consistent with the approach to instrument uncertainty endorsed in HICB-12, Waterford 3 is

-applying a less rigorous (e.g., 1-sigma) measurement uncertainty to certain of the parameters listed asCategory B items. Regulatory Guide 1.105 provides a methodology for achieving a 95/95 confidence factor for assuring that instrument setpoints are not adversely affected by uncertainty effects. There is no regulatory requirement that specifies the application or the amount of measurement uncertainty for TS LCO values. The LCO values of interest are initial condition values and do not serve as setpoints to actuate efuipe tb mitigate the impact of an accident. Thus, these LCO values are of much lower safety gnificance than instrument actuation setpoints, which serve, for example, to actuate the umps in response to an event. Thus, from a risk and safety significance standpoin tifiation exists consistent with HICB-12 to apply a smaller uncertainty. This has a safet efitihjs of providing an increased operating range to plant operators and thuerently lessei the burden on operations of a decreased operating range.

A discussion of Darameters of interest follows t istina of pe6ent Darametersnd their caegorizati - .-

Icategorization

- --o-n _ _ __

DRAFT to

[Insert Correspondence Number]

Page 4 of 20 TS Category TI Description Tech. Spec. Value A 1.24 Rated Thermal Power 3716 MW A2 Table 2.2-1: Linear Power 108% Rated Thermal

. Level-High Power A 2.2 Table 2.2-1: Logarithmic 0.257% Rated Thermal Power Level-High Power A 2.2 Table 2.2-1: Pressurizer 2350 psia A 2.2 Pressure - High 235 psia A 2.2 ~~Table 2.2-1: Pressurizer164pa A 2.2 ~Pressure -Low164pi Table 2.2-1: Containment Pressure - High 17.1 psia A 2.2 2.2 A Table 2.2-1:Pressure Generator Steam - Low 666 psia Table 2.2-1:Leve Steam 27.4% Wide Range A 2.2 Gator Table 2.2-1: Steam A 2.2 Generator Level - High 87.7% Wide Range

. A 2.2 Table 2.2-1: Reactor 19.00 psid Coolant Flow - Low B 3.1.1.4 Minimum Tcold for Criticality 5200 F 3.1.2.2 Boric Acid Makeup Tank TS Figures 3.1-1 and

. . (BAMT) Volume 3.1-2 Minimum BAMT Volume -- TS Figures 3.1-1 and 3.1.2.8.a MODES 1,2,3,4 3.1-2 7" limit for Control Element B 3.1.3.1 Assembly (CEA) position 7" (indicated position) with respect to rest of Group 3.1.3.1 C A

  • ACTION A Misalignment criteria 19" (indicated position) for ACTIONS

____ ____ b, c, d 3.1.3.1 CEA Insertion criteria for 145" ACTION f ACTION f A* 3.1.3.5 145" Shutdown CEA 145" Insertion Limit A

  • 3.1.3.6 CEA Regulating and Group COLR Figure 5 P Insertion Limits DRAFT to

[Insert Correspondence Number]

Page 5 of 20 TS Category SECTION Description Tech. Spec. Value 3.2.3 Reduced Thermal Power D ACTIONS requirements and Reduced 50% RTP; 55% setpoint b.2 and Linear Power Level - High b.3 trip setpoints 3.2.3 >95% Rated Thermal D ACTION Power for verifying 95%

b.3 Azimuthal Tilt B 3.2.5 Reactor Coolant System 148 Million Ibm/hr

________(RCS) Flow Rate A 3.2.6 Tcold <549 deg F A 3.2.6 Tcold >536 deg F D 3.2.6* Tcold <559 deg F A 3.2.8 Pressurizer Pressure >2125 psia and <2275 psia Table 3.3-1 Applicability of A 3.3.1 Logarithmic Power Level- 104% power

_High trip (and NOTES)

Table 3.3-1 Note (a)

I A 3.3.1 Logarithmic Power Level- 3*10-5% power

_ High trip bypass reset A 3.3.2 Table 3.3-4: Containment 17.1 psia Pressure - High 3.3.2 A Table 3.3-4: Pressurizer 1684 psia Pressure - Low A 3.3.2 Table 3.3-4: Containment 17.7 psia Pressure - High-High A 3.3.2 Table 3.3-4: Steam 666 psia Generator Pressure - Low A 3.3.2 Table 3.3-4: Steam 123 sid Generator delta P - High p Table 3.3-4: Emergency A 3.3.2 Feedwater Control Valve 36.3% Wide Range Logic A 3.3.3 ~Table 3.3-6: Control Room 54106pic Intake Monitor setpoint ___ __ __ ___ __ __ __

A 3.4.3.1.a Pressurizer indicated level >26% and <62.5%

B 3.5.1.b Safety Injection Tank (SIT) >40% and <77.8%

volume DRAFT to

[Insert Correspondence Number]

Page 6 of 20 TS Description Tech. Spec. Value Category SECTION SIT volume mode B 3.5.1.b applicability: 4 tanks >39% and <77.8%

operable below 1750 psia.

SIT volume mode B 3.5.1

  • applicability: 3 tanks >61% and <77.8%

operable below 1750 psia A 3.5.1.d SIT pressure >600 psig and <670 psig A 3.5.4. a Reactor Water Storage>=3 Pool (RWSP) volume B 3.5.4.c RWSP Maximum <=100 deg F

._ Temperature A 3.5.4.c RWSP Minimum >=55 deg F I Temperature A 3.6.1.4 Containment Minimum 14.275 psia Pressure B 3.6.1.4 Containment Maximum 27" w.g. (0.974 psig)

Pressure B 3.6.1.5 Containment Maximum 120F Temperature B 3.6.1.5 Containment Minimum 90F Temperature B* 4.6.2.1.a Containment Spray Riser 149.5 ft MSL Level B

  • 3.6.6.2 Annulus negative Pressure > 5" WG Table 3.7-2 allowed A 3.7.1.1 Reactor power with Main 85.3% and 66.7%

Steam Safety Valve's (MSSV's) Out-of-Service A 3.7.1.3 Condensate Storage Pool > 92%

3.7.1.3 0(CSP) volume A 3.7.1.3 CSP minimum temp >55°F B 3.7.1.3 CSP maximum < 1000 F temperature D 3.7.1.7 Atmospheric Dump Valve > 70% RTP

.__.__.__(ADV) (automatic control) 7 Ultimate heat sink Wet A 3.7.4.A Cooling Tower (WCT) >= 97%

basin level DRAFT

Attachment 1 to

[Insert Correspondence Number]

Page 7 of 20 Category S Description Tech. Spec. Value goySECTION B* 3.7.4.8 Ultimate heat sink WCT <890 F Average Basin temp Table 3.7-3: # Fans A 3.7.4 C Required based on Wet Dry Bulb: 91'F & 980 F

. . Bulb and Dry Bulb Wet Bulb: 750 F & 70'F temperatures.

Diesel Fuel Oil Storage >39,300 gal; >37,000 gal Tank Level for 5 days B 3.8.1.1 Diesel Fuel Oil Feed Tank > 339 gallons Level B B 3.8.1.2 3.8.1.2 Diesel T Fuel aOil Storage>330ga n k ,300 gal Tank Level

~B 3.8.1.2 Diesel Fuel Oil Feed Tank > 339 gallons Level 3.9.10.1, 23 feet water over B 3.9.10.2, irradiated fuel (over vessel 23 ft

. 3.9.11 flange when moving fuel)

Nil

. .A DRAFT

Attachment 1 to

[Insert Correspondence Number]

Page 8 of 20

-4.1 CEA Misalignment Criteria Technical Specification 3.1.3.1 ACTIONs b, c and d:

These parameters are considered Category A, but merit discussion because the treatment of nstrument uncertainty is explicitly built into the rod worth reactivity uncertainties which are then applied to indicated CEA position. Note also that the Waterford 3 treatment of this value is consistent with that of other Combustion Engineering (CE) NSSS plants.

Technical Specification 3.1.3.1 ACTION c addresses the condition ,ne CEA trippable but misaligned from any other CEA in its group by more than 19 inc ACTION d addresses the condition of one or more CEA's trippable but misaligned from nother CEAs in its group by between the 3.1.3.1 LCO value of 7 inches (indicated posi inches. While these values are not being changed by EPU, this is considered ameter for EPU due rtine to potential changes in reactivity and rod worths forE ore design Note the 19 inches is defined as an Indicated , in ACTION b.

19 inches defines the difference between a larget'n misalignment. erTS

'Bases, for small misalignments (less than 19 inche of>, CEA's, there is (1) a small effect on the time dependent long-term pow distribution retive(o those used in generating LCO and LSSS setpoints, (2) a small effe 6 h available Elu'cown Margin, and (3) a small effect on the ejected CEA worth usedt ayse

'As discussed in TS Bases, the Core Protion Cal Ssteprovides protection to the core in the event of a lare ignmentt ofzC.E ai ppropriate penalty factors to the calculation to acco nr , ialigned With one1r both Control Element Assembly Calculator IEAC's) o0sable, thiicreased penalty factor is applied whenever the CEA has an odlrd deviatio f approxim ey 9.5 inches or greater. Inward CEA position de v 3-Tounde t eB E Mis ' sration (CEA Drop) analysis of FSAR Section 15.4.1.4; the an 'MWt EPU conditions was presented in Section Qh eUhAport, letterWSF-2003-0074, Figure 3 of the COLR, which does d~

ire e r cuire revsicjonfo Uoprovides the required power reduction after a CEAA drop dro eve his 19 inch v& ewas he value specified in NUREG-0212, Standard Technical Sped ctions for Comb ;tibn En~ering Pressurized Water Reactors, and in NUREG-I roved Standard hnicaSecifications for Combustion Engineering Pressurized Water Reli-grs. .

The '7 inch val resp o the alarm setpoint for CEA position deviation, with an explicit treatment of instr entertainty.

Because rod worth uncertainties are'determined as a function of indicated rod position, instrument uncertainty is accommodated within the analytical basis for the 19 inch parameter.

.'Thus, it is not necessary to apply any explicit allowance for CEA position instrument uncertainty to this parameter in plant surveillance procedures since rod worth uncertainties are applied in the analysis.

DRAFT

Attachment 1 to

[Insert Correspondence Number]

Page 9 of 20

4.2 CEA Insertion Limits Technical Specification 3.1.3.1 ACTION f Technical Specification 3.1.3.5 Technical Specification 3.1.3.6 Several Technical Specifications provide limits on group CEA positions or involve ACTIONS which are dependent on CEA positions. These parameters are considered Category A, but merit discussion because the treatment of instrument uncertainty is explicitly built into the rod worth reactivity uncertainties which are then applied to indicated C osition. Note'also that the Waterford 3 treatment of this value is consistent with that of CE NSSS plants.

Technical Specification 3.1.3.1.f for a trippable but inopera thin its alignment limits allows operation to continue if the rod is greater than or e o 1 ches withdrawn or if it is within the Long Term Steady State Insertion Limit A groupgroup P. The LCO for Technical Specification 3.1.3.5 requires that all own CEA's be i drawn to greater than or equal to 145 inches. Figures 4 and 5 of I1iej3OLR provide the ins limits required Tby the LCO of Technical Specification 3.1.3.6, enting limits-on reactor po s a function

-of CEA group position in inches.

While none of these values, including QLR Figure ng changed for EPU, these

-parameters are considered pertinent ue to theteial changes in reactivity characteristics associated with EPU. -

As-;'discussed in Technical Specification Byes, th n e 10 lim W f TS 3.1.3.5 and 3.1.3.6 ensure that (1)the mini Sh t own M (2) the potential effects of a CEA ejection acciden 6ltmits. imietaccept

-l, SinaIcEA misalignments would only

-have small effects time de nent Ion power distributions, on shutdown margin, and on CEA worthsassumed for he EA Ejec analyses.

(on Westinghouse prceds inputs to safety analyses do not

'expicitl i ainty in rn group average CEA position. It is not consiod exp tly account for such a factor since the Physics bias and uncQajM' factors thI app iglto the calculated worth of CEA's positioned at nominal inse imits inheret ounte effect of CEA position uncertainty. These bias and uncert ijj'actors were bsed on th statistical analysis of differences between the c Cma worth where the CEA worth measurement was obtained using the CEA Exc ge technicd With this technique, the measured CEA worth is determined by relating the bIge in Noted position of the "reference" bank required to compensate the reactivity inse ' "test" bank. Since no adjustments are made to account for uncertainties in the algal ndicated CEA position, the tolerance limits obtained from the analysis of the raw measured and predicted worth will provide a conservative prediction of the actual worth at the indicated CEA position.

iNote also that if the effects of CEA position uncertainty were explicitly included in the

-.uncertainty analysis, the impact of the overall CEA scram worth uncertainty would be negligible. For an assumed lead bank position 3.7 inches beyond the assumed insertion limit, the associated reduction in CEA scram worth would be less than 0.5%. If this uncertainty component were statistically combined with the remainder of scram worth uncertainty of about 6.5%, the net uncertainty would increase by a negligible 0.02%.

DRAFT

Attachment 1 to

[insert Correspondence Number]

Page 10 of 20 4.3 Thermal Power limits associated with Azimuthal Tilt Actions Technical Specification 3.2.3 ACTIONs b.2 and b.3 Technical Specification 3.2.3 ACTION b.2 requires that Thermal Power be reduced to less than 50% of Rated Thermal Power within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if Azimuthal Tilt exceeds the value specified in the COLR. The Linear Power Level - High trip setpoints are to be reduced to 55% of Rated Thermal Power in the next four hours. ACTION b.3 specifies that the power operation at greater than 50% may proceed provided the Azimuthal Power Tilt is verified acceptable at 95% or greater of Rated Thermal Power. These parameters are cDri ered pertinent to EPU since Rated Thermal Power is being revised, although these pe e tage limits are unchanged.

These parameters are considered Category D.

The ACTION b.2 values for reduced power levels w sen to be arai small. No explicit calculations are performed to support th ues. These value sen to be small enough so that the plant would not be ch l~eging any p wer operating i 1 ts if azimuthal tilt exceeded the Technical Specificat imits. i7ussed in the gSES of

,NUREG-1432, this provides an acceptable level o rote t fm increased power peaking due to potential xenon redistribution hile maintainin er level sufficiently high enough to dampen any resulting azimuthal x 'scillationswFiraintaining sufficient margin to design limits. Similarly, the reduced Li e r Level - WtN setpoint is considered sufficient to ensure the assumptions of ca reg ing power peaking are maintained.

The ACTION b.3 valu Mf49 sen t oe to full 1 % Rated Thermal Power. This provision to allows ring the high f quenc ery hour) surveillances of azimuthal tilt

  • provides an acc bl exit once s da al tilt has returned to an acceptable These v raed gnto be cn nservative as to accommodate Taesloev rhte u ncertainties. Given this nature, it is not necessary to apply anyicit on these values for power levels specified in the ACTr*estatements of Thc ical yfication 3.2.3.

4.4 Tco fllow Reactor ower Cutback Techn o 3caSpecifica 3.2.6 Footnote

  • to Tec cification 3.2.6 allows the upper limit on Tcold to increase to 5590 F for up to 30 s following a reactor power cutback in which (1) regulating groups 5 and/or 6 are dropped or (2) regulating groups 5 and/or 6 are dropped and the remaining regulating groups are sequentially inserted.

This variable is considered Category D.

This value is being revised from 568 0 F to 5590 F for EPU, in conjunction with the change to the Tcold LCO; the LCO is being revised from a range of 5410 F to 558 0 F to a new range of 5360 F to 5490 F. It is noted, as documented in TS Bases, that a 30 F allowance for instrument uncertainty is applied to Tcold in FSAR Chapter 15 accident analyses, resulting in an analysis range of 5330 F to 5520 F. The 5680 F value in Technical Specifications was arbitrarily chosen DRAFT to

[Insert Correspondence Number]

Page 11 of 20 to be 10'F above the upper limit of the LCO, on the basis that it is reasonable to allow some deviation for a short period of time (30 minutes) to allow recovery and subsequent plant stabilization after the reactor power cutback.

This value has been judged to be sufficiently conservative as to accommodate allowances related to instrument uncertainties. Thus, it is not necessary or possible to apply any explicit allowance for instrument uncertainty to this value.

4.5 Containment Spray Riser Level Technical Specification 4.6.2.1.a The NRC staff requested that this parameter be added to tso f this license condition during a conference call on 14 April 2005. This parametek nsd Category B and merits further discussion because a less rigorous inst measure tuncertainty will be applied in surveillance procedures for this paramep is is considere ro nate and consistent with the guidance of HICB-12 due to gebw safety significance his parameter.

Technical Specification surveillance 4.6.2.1 .a c o main a 149 r ser level in the containment spray riser piping. The post EPU3 t.1wrequirement for this instrument is 186 ft indicated which corre ds 54.5 ft MSL urpose of this requirement is to minimize the time before containme santers content to mitigate the impact of containment pressurization transients Wc oss of getAccident (LOCA) or a Main

'Steam Line Break. The acceptance limi chtinetpret i is 44 psig.

Note the lowest centerli ea io of the eaer dr is 158' MSL, only 8.5' MSL above the Technraj$ aon onr ril The oest header is a 6 inch Schedule 40S pipe ,O a 6.065 Ili inside dlaeter. Thus, there is little operational margin above the Technicpecification e uirement ccommodate instrument uncertainty, as a level of less than 8. 5etwabov e chcal S e6ification requirement would result in spray flow into containment t nozi of the risers. Also, significant operator burden is re mai in l t he required band; this burden is reduced, with associi ucl fety, with a more flexible approach to instrument uncertainty forthf2,'aameter.

The cad~ted uncertaint this pameter is less than 5 feet. An allowance of 5 feet for instrumerrhnycertainty wou Result in less than a one second delay in the delivery of spray flow to the c t ment. A second delay has no impact on the peak containment pressure due to LOCAN oci obr to spray flow into the containment air volume. Due to the large total energ a 1 second delay in the start of spray would have negligible impact on the long t rglIculation of containment pressure, where the response over the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is considered to demonstrate that containment pressure has been lowered to no more than half the peak pressure by 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. -A I second delay has an impact of only 0.09 psi on the worst case MSLB peak pressure of 41.88 psig. Given the conservatisms in the analysis and the margin to the 44 psig acceptance limit, the spray riser level instrument uncertainty is considered of very small safety significance. Thus, it is considered consistent with HICB-12 to apply a less rigorous instrument measurement uncertainty in the plant surveillance procedures to demonstrate compliance with Technical Specifications.

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Page 12of20 4.6 Annulus Negative Pressure I Shield Building integrity Technical Specification 3.6.6.2 Technical Specification 3.6.6.2, 'Shield Building Integrity," requires the annulus region to be maintained at a negative pressure of at least 5 inch water gauge (w.g.) during normal operation (i.e., Modes 1, 2, 3, and 4). The Technical Specification limit equals the initial annulus pressure assumed in the post-LOCA annulus pressurization calculation. This parameter is not being changed as a result of EPU but is deemed to be pertinent to EPU since EPU radiological dose calculations are constructed on the b A f assumptions intended to represent or bound this value.

This parameter is considered Category B and merits furthen because a a less rigorous instrument measurement uncertainty will be ap in su iace procedures for this parameter. This is considered appropriate and c7si tent with th idance of HICB-12 due to the low safety significance of this parameter The containment systems consist of the steel t inment ye sel surrounde e Shield Building. The ShieldBuilding provides biologicaielding c ntrolled relea the annulus (region between the containment vessel a wll) atmosphere under accident conditions, and environment issile prote ctfor the containment vessel and the Nuclear Steam Supply System.

The bases for this TS, as stated in TS sT .2, is to nure that the release of radioactive materials from the primary co t nme tgat op ere ii be restricted to those leakage paths and asso e~dil a ses. This restriction, in aratss conjunction with operat o o th ield Buil nggentilation S~'stem (SBVS), will limit the site boundary and contrr radiati Moses to hthin the limits of 10CFR50.67 during accident conditions.

The non-safe -ses n t sure System maintains a vacuum of at

-least 5 in w aue normal op to comply with the TS 3.6.6.2 Limiting Condi dorOpe j Following a LOCA and receipt of a Safety Injection Actuation SigC. KS), the Anns Ne ati e Pressure System is deactivated and a transient condition exis the shield buildnnu1 til the SBVS is in full stable operation. (TS 3.6.6.1 require h~SBVS to be o able.)

In the post-LA~ annulus surization calculation, the initial annulus pressure is assumed to be -5 inch w pro by TS 3.6.6.2. This value is not algebraically adjusted in the calculation by the of the channel instrument uncertainty. The uncertainty for the annulus negative pr r instrument is 0.5 inch w.g. as documented in Waterford 3 uncertainty calculations. Indicated annulus negative pressure is typically maintained more negative than -7.7 inch w.g. by the automatic operation of the exhaust fans to ensure -5 inch w.g. is maintained in all parts of the annulus.

Calculations have shown that under the worst outside atmospheric conditions (e.g.,

temperature, humidity, etc.), and including instrument uncertainty in the pressure measurement, the initial annulus pressure may actually be less negative than the -5 inch w.g specified in the technical specification value. This could result in the annulus pressure becoming slightly positive for a short time early in the accident before the SBVS fans reduce annulus pressure.

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Page 13 of 20 To account for this conditioh, it is conservatively assumed that after the first 30 seconds post-accident, the annulus reaches a positive pressure and remains positive for 30 more seconds before the 'SBVS is able to maintain negative pressure in the annulus for the remainder of the event. This is consistent with BTP CSB 6-3, which requires that the total allowed containment leakage be assumed to be an unfiltered direct release to the environment when the shield building annulus pressure may be greater than -0.25 in. w.g. (i.e., TS 3.6.6.1 SBVS requirement) Thus, the dose contribution due to the assumed positive pressure period

'between 30 and 60 seconds after a LOCA is included in the total L dose results. Note that only 40% of the total allowed containment leakage is assu eak into the shield building annulus region.

An informal calculation using the GOTHIC computer cod aso t operation of one

SBVS fan restores negative pressure in the annulus iapproely conds This result demonstrates that the assumption of 100% unfilter tainment leaka the environment for 30 seconds is a longer release time than it w d~take for the SBVS fan gestore a negative pressure.

'Furthermore, under the Alternate Source Term me 3Ido,§3ypued for the Waterford 3 EPU radiological dose analyses (W3F1-20 0053), the seqrm available for release is time dependent. jPer RG 1.183, For the 3 ds after the cident, only activity in the 'RCS water is available for release. The fue lcdia gas ajty is released starting at 30 seconds for a duration of one half hourmfi t slue t RG 1.183 timing assumptions, the relative activity in the c~ainme i I at Mistime under AST tom assumptions compared ainmen th the early in-vessel release phase, at 1.8 hours in tteve hus, th a of instru ent uncertainty in the annulus negative pressure as rement n ot e cal teoffsite and control room dose is very small.

Thus, there would bey~na ne it on LOCA offsite and control room dose calculation if those cc er assuming no initial vacuum in the shield building.

The the assum of a econd unfiltered release results in a conservatively high offsi ontrol room F iogi se for this small contributor.

Becausee extremely I safety significance of this parameter, it is acceptable and consistent w CB-12 to p ly a less rigorous instrument measurement uncertainty to plant surveillan ure Technical Specification compliance.

4.7 Power Level for OPERABILITY of ADV Automatic Actuation Technical Specification 3.7.1.7 This parameter is Category D.

New Technical Specification 3.7.1.7 is being added due to EPU to specify OPERABILITY required for the'Atmospheric Dump Valves. This TS is being added since the'EPU Small Break LOCA Emergency Core Cooling System (ECCS) analysis; the ADV's were previously credited only for cooldown to shutdown cooling entry conditions and for their containment isolation function. Thus, ADV operability, which had previously been addressed in the DRAFT

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Page 14 of 20 licensee' controlled in Technical Specification 3.6.3 and the Technical Requirements Manual (TRM), will now be addressed in the new Technical Specification.

The small break LOCA analyses assume a maximum ADV setpoint of 1040 psia. This value is specified in the footnote to TS 3.7.1.7 and explicitly accounts for the instrument uncertainty offset from the nominal setpoint of 1007 psia.

The footnote to the LCO also documents that the ADV automatic actuation channels are not required to be operable when the reactor has-been at less than ordepl to 70% Rated Thermal Power for greater than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (following long-term op0 tio at EPU Rated Thermal Power of 3716 MWt). The 70% is considered an arbitrary valu hich uncertainty need not

'be applied. In support of this arbitrary value, analyses wer d8ted to demonstrate that the decay heat load associated with operation for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> QR i Thermal Power is such that the ADV's need not be credited to demonst cceptable performance.

The value of 70% is specified based on reasonabl e enng judgme l a power level' below which automatic actuation of the ADV's s quired. Note that V's are not credited in the Waterford 3 Cycle 13 Small Bred OCA ECCS nalyses, wh ads to the conclusion that long-term operation at power le 34414 (92.6%of EP Oated Thermal Power) is acceptable without crediting A BLOCA analysis. The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> time frame is consistent with ACTION of new TS 3 M;7hich calls for reducing power to less than or equal to 70% of Rated e Power withil6 urs if the automatic actuation channel for one ADV is inoperable an n esorable status.

It is noted that there is generall no anal I basf4 CTiQ times in Technical

'Specifications. The 6 ho lati ime her as coesistency with Technical Specifications for siml fn ut that s rbitrary timbased upon shared engineering judge _hich con rs oper' g experience.

'Margin exists in the y heat lysis4betwee t at where ADV's are not required (e.g.,

long term operation at y at corresponding to operation at 70%

Rated Tfor, hous or less. Aht analytical approach would result in a curve of incr Sig React erm Pwer as a function of time, that is, the reactor power could be

'slow §jdrea sed up oxi tel, 92.6% in order for this decay heat logic to be main aied. In conside of argin and the fact that the decay heat load associated with 7 wer operation decre e with longer times, it is not considered necessary to apply any plicit offset to cc unt for power measurement uncertainty to the 70% value specified in ical Spe fictions.

4.8 Wet Coolin To e asin Temperature Technical Spe a ion 3i.7.4.B Technical Specification 3.7.4.b requires the wet cooling tower (WCT) basin water to be less than or equal to 890 F as a limiting condition of operation (LCO) for the ultimate heat sink (UHS). This limiting condition-of operation ensures that the UHS can dissipate the peak accident heat load assuming the worst case meteorological conditions as required by Regulatory Guide 1.27. This parameter is considered pertinent to EPU due to higher decay heat for EPU conditions.

This parameter is considered Category B and merits further discussion because a a less rigorous instrument measurement uncertainty will be applied in surveillance procedures for DRAFT

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Page 15 of 20 this parameter. This is considered appropriate and consistent with the guidance of HICB-12 due to the low safety significance of this parameter.

Following the implementation of EPU, calculations demonstrate that the WCT basin temperature is required to be 89.30F or less for the UHS to dissipate the peak accident heat load at the worst case meteorological conditions. The current LCO does not need to be changed since the LCO bounds the analysis value, ensuring the safety function of the UHS following the implementation of EPU will be met.

Performance testing of the component cooling water heat exch e ensures margin in WCT basin temperature will exist. Analysis uncertainty is also app ensure actual UHS heat loads will not exceed the design basis limits. Thus testing fd sis provide conservatisms to the WCT basin water temperature.

A brief UHS system flow path description during ac mode is prov do understand how the testing and analysis briefly discussed a emonstrate the UH af ty function is met. The UHS consists of two systems, comp et cooling water (CCW) andui iary component cooling water (ACCW) systems. The jr co pehts in the CC stem for heat removal are the CCW pump, dry cooling towe( nd the CCW heat exchanger (CCWHx). The major components inhe ACCW sys heat removal are the ACCW pump, CCW temperature control val e(b and the he CCW is a closed loop

system with heat removal first being p the DCT4eh DCT contains 5 cells of cooling coils with each cell being coole 3 fashe rema accident heat load will then be'dissipated by the CCWHx. CCW flow ers thlCWandecooled by the ACCW system. The CCW TCV h e ACCWIo eir e ~intain CCW at the desired outlet temperature to auxiliarir remove accident heat loads. The heat removed by the AC then dis ated to t tmosphere by the WCT. Each WCT contains a basin am o cooling is, each ce lconsisting of 4 fans. Additional system description details of5 Section 9.2.5.

The AC t plwiaing w toC CWHx directly from the WCT basin.

Theref efthe spifo WCT basin temperature LCO ensures the CCWHx will rrftain desired UItemTQature to cool the plant auxiliaries and remove accident heat loadTe CCWHx is te c w wto ith the requirements of the Generic Letter (GL) 89-13 prograr he purpose of eBL 89-'3 testing is to demonstrate that heat exchangers will perform tr 1esign basis et removal function following a design basis accident. For CCWHx t ,ata is col] d while a more typical heat duty is being dissipated by the CCWHx. All m me ntment uncertainties are applied to the measured data collected during tjg e data collected, after applying uncertainties, is analyzed to determine the overa1Nlth of the CCWHx and ensure the projected accident CCW outlet temperature meets the test acceptance criterion of 1.00F or more less than the analyzed limit.

If the acceptance criterion is met, -a margin of 2.60F exists in WCT basin temperature with respect to the basis of the LCO. In other words, if the WCT basin temperature was 91 .60 F (89°F + 2.60 F), the CCWHx would dissipate its required peak accident heat load and maintain the CCW outlet temperature at the analyzed limit. This margin bounds the instrument uncertainty by most accurate available indication for this parameter.

The latest testing on the CCWHx demonstrated the CCW outlet temperature could be maintained more than 3YF below the analyzed limit. The current CCWHx testing assumes pre-DRAFT

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Page 16 of 20 EPU conditions which are, more severe than the EPU heat removal requirements, as discussed in Section 2.5.5.:4 ofthe Power Uprate Report in W3F1-2003-0074.

The UHS peak heat load analysis assumes a CCW temperature control valve (TCV) uncertainty of -3.00F which would increase the assumed UHS heat load post-accident and therefore is conservative. Cooler CCW temperatures to the plant auxiliaries result in removing more heat from the plant due to improved heat removal efficiency. The increase of CCWHX heat duty is expected to be 7.3MBtu/hr, the WCT basin temperature can be as high as 92.2 0F and maintain the CCW outlet temperature at the analye it.

Thus, conservatisms in the analysis methodology and inhere inL 89-13 heat exchanger testing lead to the conclusion that the 890 F value is sufficie eosative as to accommodate allowances related to instrument uncertain The agins described above provides assurance that the UHS will fulfill its safety f i and doe vide a bases for not explicit applying instrument uncertainty for the Tec t Specification teme ature limit.

However, to insure Technical Specification comr pe, I an explicit, but les grous, uncertainty will be applied, consistent with HIC l2 to plant s eillance proces

5.0 REGULATORY ANALYSIS

5.1 A plicable Regulatou Reauire mn rteria

'Per 10CFR50.36(c)(2), Limitin Conditio for Op nCO are the lowest functional

-3capability or performanc ee quipm peration of the facility. It is

-not necessarily requir inu eplicit o dor instrum nt measurement uncertainty to

.provide this require ttional caaility N r does 10CFR50.36(c)(2) prescribe any 4specific approach fbrie treatme instrume t'measurement uncertainty. Thus, consistent with other industry pr cents secification values on indicated values or to tie analyses to no a cach maintains compliance with I OCFR5 6 The red approach~itru ~nt uncertainty is explicitly endorsed in Regulatory Guide (RG 05, Revision 3, embe 9, "Setpoints for Safety-Related Instrumentation," and Branch nical Position 1B-12, uldance on Establishing and Maintaining Instrument Setpoints Je 1997).

RG 1-1 05 app 1 y t et oints, which are considered of greater risk and safety significance than ti ition values. However, given that RG 1.105 endorses a graded approach to be app o setpoints, this provides a precedent for also using a graded approach in addressing parameters which are initial condition values. Setpoints are of far greater safety significance since a setpoint results in actuation of mitigation equipment; the availability of mitigation equipment is of far greater impact on analyzed results of a transient than slight variations in the initial conditions assumed for the analysis. For example, there would be far greater impact on Chapter15 Nuclear Steam Supply System (NSSS) analyses if the control element assemblies did not insert on a reactor trip signal than if there was a slight variation in the control rod worth from the assumed value. Si milarly, small variations in temperature of safety injection fluid would have a much smaller impact than if the safety injection pumps did not respond to the event.

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Page 17 of 20 The different nature of the significance of an instrument setpoint, compared to an assumed initial condition value, is highlighted by Generic Design Criterion6(GDC) 29: " Protection against anticipated operational occurrences". The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences." GDC29 focuses on the performance of safety functions, that is, on the ability of mitigating systems to respond to events, rather than on the small variability in analysis results associated with slight variations in the value of assumed initial conditions for the safety analyses.

No regulatory requirements exist for the incorporation of instrumsuncertainty in the operating envelope limits used as inputs to the safety analysi roess, with the exception of initial power level. Regulatory Guide 1.49 establishes the q irei et that safety analyses be performed for an initial power level that accounts for powent uncertainty.

However, for plants other than Waterford 3, some ap e analysis hodologies credit other uncertainties to support performing analyses itfout explicit consiation of power measurement uncertainty. Also, licensing basis Tes for low iI nts that are considered "beyond design basis" are perform di the licen se power level,"~ifhut uncertainty (e.g., Station Blackout, Anticipated Trsient Wit O&Scram (ATWS))I The determination of the safety signifil ce of instrum nctions should consider all available information. This would in vew of dete 4,:nistic requirements, the impact on risk, and other available information. on of thergin of safety associated with applicable parameters would be within scop This appr \Xensures reactor safety, complies with;regulatory requements, is se og giepring practices, and avoids unnecessary operating r 6ThF hs upoth tention tae to be focused in a manner to maximize t sa~~ft.

Waterford 3 setpo lor Engine fety F t re Actuation System (ESFAS) are listed in Technical Specificati ble 3 -4. Section 3.3.2 requires that the ESFAS trip setpoints be consis e in Table 3.3-4. Reactor protective instrume t9 a0 ipints e Ii chnapecification Table 2.2-1. The Limiting Safety tem Se LS r Reactor trip setpoints, Technical Specification 2.2.1, req sthat reactor p ftive rumentation setpoints be set consistent with the values of Tabl? -1. The Bases fAS 3/ 313and 3/4.3.2 describe the basis for the explicit treatment of instr int uncertainty SEASnd Reactor Protection System (RPS) setpoints:

NPSIESFA$ p Setpoint values are determined by means of an explicit setpoin1 culati alysis. A Total Loop Uncertainty (TLU) is calculated for each RPsFtrument channel. The Trip Setpoint is then determined by adding or subtcfing the TLU from the Analytical Limit (add TLU for decreasing process value; subtract TLU for increasing process value) .......

The methodology used by Waterford 3 for RPS/ESFAS setpoints has been previously reviewed and approved by the NRC as documented in Amendment 113 issued September 5, 1995.

It is noted that, aside from RPSIESFAS setpoints, flexibility exists for licensees to determine what methodology to use when instrument uncertainty is to be explicitly accounted for.

Branch Technical Position HICB-12 Revision 4 dated June 1997 states that licensees may apply a less rigorous setpoint determination method for certain functional units and LCO's.

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Page 18 of 20 The accounting of instrument Uncertainty for other than ESFAS or RPS setpoints is discussed in an NRC Task Interface Agreement Evaluation (TAC No. M95177) dated July 22, 1996.

The NRC staff has previously recognized that, for instrumentation other than ESFAS or RPS, instrument uncertainty can be accounted for through plant safety analyses, Technical Specification limiting values, measured values, surveillance testing, or emergency procedures. The use of ISA standard S67.04 is not required and other methodologies can be used to account for instrument uncertainty.

Note some of the parameters in the Table above do not serve asi Sument setpoints, but rather are assumed initial conditions for parameters in safety a lses. Since these parameters are not instrument setpoints, they are beyond o f RG 1.105. Consistent with this philosophy of ISA-S67.04, which is endorsed b IG-12 G 1.105, it is recognized that there is far greater safety and risk sigr iance for par ers which serve as setpoints for accident mitigation equipment than foarmeters which oerve as initial conditions for analyses of postulated events.

Entergy has determined that the proposed chan does no, eire any exem s or relief from regulatory requirements and does not affect c ori ncewith any General Design Criterion (GDC) differently than describ d in the Upd td Fnal Safety Analysis Report (UFSAR.)

5.2 No Siqnificant Hazards Conside 1in This letter is a request te eratin ce se( ior Waterford Steam Electric Station, Unit 3 (Waterf d3 to h ense condition regarding instrument uncertainty that was imposed aerford 3 the ap al and issuance of the Extended Power Uprate (EPU) am mnent (i.e., A dment 1) The license condition required that additional informatioi e ardin t ent hertainty is accounted for in Technical Specification paramete im ci >P itted for NRC staff review and approval.

The requ rpInThf- ation bmitted license amendment request and approval of this re q9e docur te Retion of the NRC staffs review and approval as required by the i66Age condition' hrem f the license condition will allow Waterford 3 to proceed abo 41 MWt and ac Ae the U power level of 3716 MWt as authorized in Amendment 199 to l aterford 3 Op raing Linse.

Entergy Op t s, Inc. h aluated whether or not a significant hazards consideration is involved with t os endment(s) by focusing on the three standards set forth in 10 CFR 50.92, Isn amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed amendment is to remove a license condition imposed on Waterford 3 with the issuance of Amendment 199 approving the EPU. The removal of the license condition will allow Waterford 3 to operate at the power level of 3716 MWt which has previously been evaluated and approved by the NRC staff as documented in Amendment 199 to the Waterford 3 Operating License.

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Page 19 of 20 Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or differenrt kind of accident from any accident previously evaluated?

Response: No.

The proposed amendment is to remove a license condit osed on Waterford 3 with the issuance of Amendment 199 approving the Ejl. he removal of the license condition will allow Waterford 3 to operate at the povie gof 3716 MWt which has previously been evaluated and approved by the sta aocumented in Amendment 199 to the Waterford 3 Operatin ense.

Therefore, the proposed change does nqt Prfeyate the possibility of ageor different kind of accident from any previously ev ted.

3. Does the proposed change involve a signifi i~hion in a margin of safety?

dnt Response: No.

The proposed amendment is t o ise ction imposed on Waterford 3 with the issuance of Amendment99 appynhe EPw~he removal of the license condition will allo aord 3 to level of 3716 MWt which has previously be au an d by the NRC staff as documented in Amendmento96o the Wame ord 3 0perating License.

Toes o olve a significant reduction in a margin of Based ea rg udes that the proposed amendment(s) present no si ican azards co ideratio der the standards set forth in 10 CFR 50.92(c), and, accooly, a finding o . igni c hazards consideration" is justified.

5.3 En idmental Cop drations The propo dme s not involve (i) a significant hazards consideration, (ii) a significant changs or significant increase in the amounts of any effluent that may be released offsite a i iia significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore,' pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

7.1 Entergy letter to the NRC dated November 13, 2003, "License Amendment Request NPF-38-249, Extended Power Uprate" (W3F1 -2003-0074)

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'Page 20 of 20 7.2 Waterford 3 Final Safety Analysis Report 7.3 Waterford 3 Technical Specifications (through Ameridhient 199) 7.4 NUREG-0212, "Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors," Revision 3, December 1981 7.5 NUREG-1432, Improved Standard Technical Specifications Combustion Engineering Plants," Revision 3, June 2004 7.6 Regulatory Guide 1.27, "Ultimate Heat Sink for Nuclear Power Plants" 7.7 Generic Letter (GL) 89-13, "Service Water System Problems Affecting Safety-Related Equipment 7.8 10CFR50.36, "Technical Specifications" 7.9 Regulatory Guide 1.105, "Setpoints for Safety-Rel tdIstrumentation," Revision 3 7.10 Branch Technical Position HICB-12, "Guidance Sbihing and Maintaining Instrument Setpoints," June 1997 7.11 Generic Design Criterion 29, "Protection A, 'st Anticipate Operational Occurrences"A 7.12 NRC Task Interface Agreement Evalua TAC No. M95177) 22,1996 7.13 Regulatory Guide 1.49, "Power Levst Nuclear P wer Plants," 1, December 1973 RIfon DRAFT

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