ML051160093

E-mail from D. B. Miller on Draft of Waterford 3 Submittal on Instrument Uncertainty License Condition Submittal

ML051160093
Person / Time
Site:	Waterford
Issue date:	04/25/2005
From:	Miller D Entergy Operations
To:	Alexion T NRC/NRR/ADPT
References
Download: ML051160093 (19)

Text

I Thomas Alexion - Draft of Waterford 3 Instrument Uncertainty License Condition Submittal Paae 1 11 Thms A_ o X, Draf _.f Waterford 3__.

Instrument Uncertainty _L i.cense._Cond.i b

_~.... w_

E_

_s s_

to._n__Submittal Paae 1_ s

• From:

'MILLER, D BRYAN" <dmiIl14@entergy.com>

To:

"'Thomas Alexion" <TWA~nrc.gov>

Date:

4/25/05 2:01 PM

Subject:

Draft of Waterford 3 Instrument Uncertainty License Condition Submittal

<<Draft for NRC 4-25-05.pdf>>

Tom, Attached is a draft of the submittal based on Friday's conference call with the staff. Let me know if you have any questions. I am attempting to arrange for our onsite safety review committee to review this submittal on Tuesday (4/26) and for our offsite safety review committee to review this submittal on Wednesday (4/27). If all goes well I may have this submittal to you by the time you come in on Thursday (4/28). (Worse case will be when you come in next Monday.)

I understand that Kaly is back in the office on a part time basis however you will continue to work on the Instrument Uncertainty item.

Bryan CC:

"KALYANAM, N. KALY <nxk nrc.gov>, "LEONARD, THEODORE R"

<TLEONAR@entergy.com>, "MITCHELL, TIMOTHY Gm <TMITCH1 @entergy.com>, BURFORD, FRANCIS G" <FBURFORQentergy.com>

cAtemP\\GW)00001.TMP Pages 1 V c\\Em\\WQO1TMPae1I Mail Envelope Properties (426D307F.D44

2: 56644)

Subject:

Creation Date:

From:

Created By:

Draft of Waterford 3 Instrument Uncertainty License Condition Submittal 4/25/05 2:00PM "MILLER, D BRYAN" <dmill4@entergy.com>

dmi1114@entergy.com Recipients nrc.gov owf4_po.OWFNDO NXK CC (N. Kaly Kalyanam)

TWA (Thomas Alexion) entergy.com FBURFOR CC (FRANCIS G BURFORD)

TMITCH1 CC (TIMOTHY G MITCHELL)

TLEONAR CC (THEODORE R LEONARD)

Post Office owf4_po.OWFNDO Route nrc.gov entergy.com Files MESSAGE TEXT.htm Draft for NRC 4-25-05.pdf Mime.822 Options Expiration Date:

Priority:

Reply Requested:

Return Notification:

Concealed

Subject:

Security:

Size 657 1319 99811 140756 None Standard No None No Standard Date & Time 04/25/05 02:00PM

W3F1 -2005-0032

[Insert Date]

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

SUBJECT:

License Amendment Request Extended Power Uprate (Ari Instrument Uncertainty Waterford Steam Electric Stats Docket No. 50-382 License No. NPF-4,k irding

REFERENCES:

1. NRC lette Steam Eli PowDaSioMaa

Dear Sir or Madam,

Pursuant t staff rev Upon pv lice eondi Electr ti from theN~

Reference 1 approval, im; Or to MrnableX't"O April 15, 2005, "Waterford 3ctric Sta

, Uni suanc m endment Re: Extended te (T A

Sv

c. (Entergy) hereby requests that the NRC ached information submitted in accordance with Reference 1.

rmation, Entergy requests that the NRC staff consider the 3,t uncertainty, that was imposed on the Waterford Steam rd 5ense in Reference 1, to be complete and removed it ided Power Uprate (EPU) for Waterford 3 and, as part of the license condition:

3. As stated in tO icensee's letter dated February 5, 2005, the licensee committed as follows: 'Prior to exceeding 3441 MWt, Entergy will submit, for NRC review and approval, a description 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.

A description of how Entergy accounts for instrument uncertainty for Technical Specification parameters impacted by the Waterford 3 EPU is provided in Attachment 1 for NRC staff review and approval in accordance with the license condition. Following NRC staff review 4/25/05 DRAFT

W3F1 -2005-0032 Page 2 of 3 and approval of the information contained in Attachment 1 the condition set forth in the EPU amendment will be complete.

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

The proposed change does not include any new commitments.

Entergy requests approval of the proposed amendment by M 2005, to support power ascension from the Spring 2005 refueling outage. Once a ro e amendment shall be implemented within 60 days after exceeding 3441 MWL Waterford 3 can not exceed 3441 MWt and achiev PU power lev 716 MWt following the Spring 2005 refueling outage until xense condition impos Reference 1 is considered to be complete and removed fro th license.

e need for a e

amendment for this purpose was not recognize nterg r

e NRC staff uni ust prior to the issuance of the EPU license. Therefore, to av a

of Waterford 3 following restart from the Spring 2005 refueling tage, Enterg sts that this license amendment request be reviewed and approved o14 ent basis.

If you have any questions or require ad naclf on, p mcontact D. Bryan Miller at 504-739-6692.

I declare under penal correct. Executed on

1. Analysis Specification Change 4/25/05 DRAFT

W3Fl-2005-0032 Page 3 of 3 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 0-7D1 Washington, DC 20555-0001 Wise, Carter, Child & Caraway Attn: J. Smith P.O. Box 651 Jackson, MS 39205 Winston & Strawn Attn: N.S. Reynolds 1700 K Street, NW Washington, D 17 Louisiana ment of ironme aQuE Office of Ej~nmental pliance Surveillanc ion P. O. Box 431 107-2445 4/25/05 DRAFT

Specification Change to W3F11-2005-0032 Page 1 of 13

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

Review and approve the infom license condition regarding ins' approval and issuance of the E

3.0 BACKGROUND

The amendment approving the

3. As stated in the license follows: "Prior to exceei approval, a description Technical Specification Uprate. "Accordingl operate the W a y

iation below regarding instru _'b ainty and remove the trument uncertainty that wIf'ose aterford 3 with the

PU amendment.

EPU for Waterford e following license condition:

e's leebruary the licensee committed as ding 344 rgy willt, for NRC review and of how E y a for in nt uncertainty for each ameter acte eate 3 Extended Power

,t to conithis

?n, the licensee shall not a

at a p e

vel excee ng 3441 MWt.

)per; s, Inc. (lgy) will account for instrument uncertainty ar e cted e Waterford 3 EPU is provided below for cei e license condition. Following NRC staff ation below h condition set forth in the EPU amendment will

• ondition, Entergy is documenting the treatment of instrument rameters which were revised in association with EPU or fall within the following criteria:

A description of how for Technical Specifics NRC staff review and review anpat be com f

MM' 9

4.0 In accordancej measurement uIS pertinent to EPU

• The parameter is a value which is measured using plant equipment. That is, the parameter is directly indicated to operators using installed plant instrumentation.

and The parameter is a value which is specified by a Limiting Condition for Operation (LCO) of the Waterford 3 Technical Specifications. Parameters listed in Technical Specification Tables which are called out by LCO's are considered within the scope of this effort.

4/25/05 DRAFT to W3F1 -2005-0032 Page 2 of 13 When 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.

The criteria consider 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 least 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. The criteria would capture parameters for which margins to acceptance criteria for analyses discussed in the Bases of applicable Technical Specifications have been impact EPU.

The parameter selection was discussed with the NRC staff du

,onference call on 14 April 2005. The NRC staff concurred with the list of par vided by Waterford 3, with the proviso (agreed to by Entergy) that Containment IRi el (TS 4.6.2.1) also be included.

Values relating to applicability (e.g., MODES) of t nical Specificatio considered to be generally out of the scope of the license co

o.

For exa le, Technica ifications 3.2.1 through 3.2.4 for power distribution parame LinearRate, Planar ial Peaking, Azimuthal Power Tilt, and Departure from Nucleate V

Fn (DNBR) margin) are designated as applicable above 20% o ated Therm a

r (RTP). The Entergy license condition scope will not include discu o

strument C ainties with respect to that 20%

power criteria.

Entergy recognizes that safety analyses S acco t

r n uncertainty in all cases.

Since the intent of many 3pecifi s

pro e

surance that the plant is within the assumptions of the ent sis, it i priate that e instrument measurement uncertainties be acc d for in s manne xcept for Limiting Safety System Setting (LSSS) setpoint va there is no ulatory g nce describing specific methods that must be employed to addre e in e

rtain associated with Technical Specification parameters.

Parar r

z]

t°oe of four categories regarding treatment of ins rton uncert^

Categorm s:

Description ment Unceanty is explicitly considered in analyses. There is an explicit A

o etween Technical Specification value and the value assumed in the anapetidwt theTechnical Specification.

Inst uirt~ainty is explicitly considered in plant procedures. There is an B

explicit o etween the LCO value in the Technical Specification and the value specified to be maintained by plant procedures.

C The LCO value may also be the value assumed as initial conditions in safety analyses and the value specified to be maintained by plant procedures.

The Technical Specification value and the plant procedure limit are the same and D

the parameter does not have an explicit analytical basis. The limited numbers of parameters in this category are based on engineering judgment.

4/25/05 DRAFT to W3F1 -2005-0032 Page 3 of 13 Waterford 3 has performed a categorization of Technical Specification parameters within the scope of the license condition. This categorization, shown in the table below, also reflects discussions with the NRC staff on April 14, 15, and 22, 2005.

Consistent with the Waterford 3 licensing basis and HICB-12, Entergy is explicitly applying offsets for instrument uncertainty in the analysis and/or procedures for the Technical Specification parameters impacted by the Waterford 3 Extended Power Uprate as listed below.

Because an explicit offset for instrument uncertainty is being applied, none of the parameters fall into Category C.

The listing of pertinent parameters within the scope of this licendntion and their categorization is provided below.

Categry +

TS e

Category Section Description A

1.24 Rated Thermal Power 3716 MW A

2.2 Table 2.2-1: Linear Power 108% RTP Level-High A

2.2 Table 2.2-1: Logarithmic 0.257% RTP Power Level-High A

2.2 Table 2.2-1: Pressurizer 2350 psia Pressure - High A

2.2 Table 2.2-1: Pressurizer 1684 psia Pressure - Low A 2.2 Table 2.2-1: Containment 17.1 psia

.2 Pressure - High A

2.2 Table 2.2-1: Steam Generator 666 psia Pressure - Low A

2.2 Table 2.2-1: Steam Generator 27.4% Wide Range Level -Low A

2.2 Table 2.2-1: Steam Generator 8.%Wd ag Level - High 87.7% Wide Range A

2.2 Table 2.2-1: Reactor Coolant 19.00 psid Flow - Low B

3.1.1.4 Minimum Tc0 Id for Criticality 5200F A

3.1.2.2 Boric Acid Makeup Tank TS Figures 3.1-1 and 3.1-2 (BAMT) Volume A

3.1.2.8.a Minimum BAMT Volume -

TS Figures 3.1-1 and 3.1-2 MODES 1,2,3,4 7" limit for Control Element B

3.1.3.1 Assembly (CEA) position with 7" (indicated position) respect to rest of Group 4/25/05 DRAFT to W3F1 -2005-0032 Page 4 of 13 Category +

Section Description TS Value 3.1.3.1 CEA Misalignment criteria for 19" (indicated position)

AACTIONS ACTIONS b, c, &d1"(niae oiin b, c, &d A #

3.1.3.1 CEA Insertion criteria for 145" ACTION f ACTION f Au 3.1.3.5 145" Shutdown CEA Insertion 145" Limit Au 3.1.3.6 CEA Regulating and Group P COLR Figure 5 Insertion Limits 3.2.3 Reduced Thermal Power D

ACTIONS requirements and Reduced 50% RTP b.2 and Linear Power Level - High trip 55% RTP (setpoint) b.3 setpoints

. A.C

>95% Rated Thermal Power 95% RTP b.3 for verifying Azimuthal Tilt B

3.2.5 Reactor Coolant System 148 Million Ibm/hr (RCS) Flow Rate A

3.2.6 Tcold

>5360F and <5490F D

3.2.6

• Tcold

<5590F A

3.2.8 Pressurizer Pressure

>2125 psia and <2275 psia Table 3.3-1 Applicability of A

3.3.1 Logarithmic Power Level-High 104% RTP trip (and NOTES)

Table 3.3-1 Note (a)

A 3.3.1 Logarithmic Power Level-High 3*10.5% RTP trip bypass reset A

3.3.2 Table 3.3-4: Containment 17.1 psia Pressure - High A

3.3.2 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 Generator 666 psia Pressure - Low A

3.3.2 Table 3.3-4: Steam Generator 123 psid Idelta P - High__

4/25/05 DRAFT to W3Fl-2005-0032 Page 5 of 13 Category +

S Description TS Value A

3.3.2 Table 3.3-4: Emergency 36.3% Wide Range Feedwater Control Valve Logic A

3.3.3.1 Table 3.3-6: Control Room 5.45x10-6 pCi/cc 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 SIT volume mode applicability:

B 3.5.1

• 4 tanks operable below 1750

>39% and <77.8%

psia.

SIT volume mode applicability:

B 3.5.1

• 3 tanks operable below 1750

>61% and <77.8%

psia A

3.5.1.d SIT pressure

>600 psig and <670 psig A

3.5.4.a Reactor Water Storage Pool 83%

A______

__(RWSP) volume B

3.5.4.c RWSP Maximum Temperature 100`F A

3.5.4.c RWSP Minimum Temperature 55`F A

3.6.1.4 Containment Minimum 14.275 psia Pressure B

3.6.1.4 Containment Maximum 27" w.g.

Pressure B

3.6.1.5 Containment Maximum 1200F 3.6.1.5_

Temperature B

3.6.1.5 Containment Minimum 900F Temperature B

4.6.2.1 a Containment Spray Riser 149.5 ft MSL Level B

3.6.6.2 Annulus negative Pressure 5" w.g.

Table 3.7-2 allowed reactor A

3.7.1.1 power with Main Steam Safety 85.3% RTP Valve's (MSSV's) Out-of-66.7% RTP Service A

3.7.1.3 Condensate Storage Pool 92% indicated level

________(CSP) volume B

3.7.1.3 CSP Maximum Temperature 1000F A

3.7.1.3 CSP Minimum Temperature 550F 4/25/05 DRAFT to W3F1 -2005-0032 Page 6 of 13 Category Secton Description TS Value D

3.7.1.7 Atmospheric Dump Valve 70% RTP (ADV) (automatic control)

Ultimate Heat Sink Wet A

3.7.4.a Cooling Tower (WCT) basin 97%

level B

3.7.4.b Ultimate Heat Sink WCT 890F Average Basin temp Table 3.7-3: # Fans Required Dry Bulb: 91'F & 980 F A

3.7.4.c.

based on Wet Bulb and Dry Wet Bulb: 750F & 700F Bulb temperatures.

B 3.8.1.1 Diesel Fuel Oil Storage Tank 39,300 gal Level 37,000 gal for 5 days B

3.8.1.1 Diesel Fuel Oil Feed Tank 339 gallons Level B

3.8.1.2 Diesel Fuel Oil Storage Tank 39,300 gal Level 37,000 gal for 5 days B

3.8.1.2 Diesel Fuel Oil Feed Tank 339 gallons Level B

3.9.10.1, 3.9.10.2, 3.9.11 23 feet water over irradiated fuel (over vessel flange when moving fuel) 23 ft

+

tin the rec zation of these Technical Specification cQj~ty is applied to CEA worth which is directly related to Adaiwal discussion is provided below regarding these lof Technical Specification.

4/25/05 DRAFT to W3F1 -2005-0032 Page 7 of 13 4.1 CEA Misalignment Criteria (19")

Technical Specification 3.1.3.1 ACTIONs b, c and d:

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 CEA position. The Waterford 3 treatment of this parameter is consistent with that of other Combustion Engineering nuclear steam supply systems plants.

Technical Specification 3.1.3.1 ACTION b address more than one rippable but misaligned from any other CEA in its group by more than 19 inches (indicate tion); ACTION c addresses the condition of one CEA trippable but misaligned fr y other CEA in its group by more than 19 inches; ACTION d addresses the condition of e

CEA's trippable but misaligned from any other CEAs in its group by between t v1.

alue of 7 inches (indicated position) and 19 inches. While these value not beingc d by EPU, this is considered a pertinent parameter for EPU due to po changes in re tj and rod worths for EPU core designs.

19 inches defines the difference between a large mall isalignment. TS Bases, for small misalignments (less than 19 inches) of the e is (1) a small effect on the time dependent long-term power distribution lative to thos e in generating LCO and LSSS setpoints, (2) a small effect on the av utdown i

and (3) a small effect on the ejected CEA worth used in the safety a As discussed in FSAR Section 7.5.1.6 an r.4.1.

ver dependent CEA position indication systems provid ition in i

he or. The Pulse Counting CEA Position Indication Sys s

CEA si by maint ning a record of the "raise" and "lower" control pulse e to each neticj ontrol element drive mechanism (CEDM).

The Reed Switch 1in Indicati stem us series of magnetically actuated reed switches to provide si arepr H

A po

t. Two independent reed switch position transmitters are rovide o

ce i

indication system uses a series of magneti ee es spaced inch intervals along the assembly and arrange prec sist a voltage divider network. CEA position information based on t e

switch infoai n, i ing CEA deviation information, is provided to the Core Prote o Calculators anthe I Element Assembly Calculators.

The impac EA misalign e ts on power distribution is explicitly accounted for by the Core Protection Ca a or and s p rting analysis. As discussed in TS Bases, the Core Protection Calculator Syste ovid ptection to the core in the event of a large misalignment of a CEA by applying approp ty factors to the calculation to account for the misaligned CEA.

With one or both Con ement Assembly Calculators operable, this increased penalty factor is applied whenever the CEA has an outward deviation of approximately 9.5 inches or greater; supporting analysis has explicitly considered uncertainties in determining this value. Inward CEA position deviations are bounded by the CEA Misoperation (CEA Drop) analysis of FSAR Section 15.4.1.4 which conservatively assumes that the CEA is dropped from an initial full out position to a final full in position; the analysis of this event for 3716 MWt EPU conditions was presented in Section 2.13.4.1.4 of the EPU report, letterW3Fl-2003-0074, Figure 3 of the COLR, which does not require revision for EPU, provides the required power reduction after a CEA drop event. This 19 inch value was also the value specified in NUREG-0212, Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors, and in 4/25/05 DRAFT to W3Fl-2005-0032 Page 8 of 13 NUREG-1432, improved Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors.

The impact of CEA misalignments on shutdown margin and ejected CEA worth is accounted for in the safety analysis through the conservative application of CEA worth uncertainties. As discussed in Section 4.2, instrument uncertainty associated with this parameter is included in the core physics inputs to safety analysis; 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 dditional explicit allowance for CEA position instrument uncertainty to this parame plant procedures since rod worth uncertainties are explicitly applied in the analysis.

4.2 CEA Insertion Limits (145" and COLR Figure 5)

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 o up CE oions or involv TIONS which are dependent on CEA positions. These par trronsidered Category A, but merit discussion because the treatmen instrument cty is explicitly built into the rod worth reactivity uncertainties which ar lied to in ed CEA position.

Technical Specification 3.1.3.1.f for a triplebn rable w

ithin its alignment limits allows operation to continue if the rod is gier th a to inches withdrawn or if it is within the Long Term Ste sertio C

ap 6 or group P. The LCO for Technical Specification s that down CE s be withdrawn to greater than or equal to 145 inches.

es 4 an df the C provide the insertion limits required by the LCO of Technical ecation 3.1

, presenti mits on reactor power as a function of CEA group position in inch While non lue ding COL e 5, are being changed for EPU, these param econ ed pit to EPU due to the potential changes in reactivity char etics associa ith As disc in Technical ifica Bases, the insertion limits of TS 3.1.3.5 and 3.1.3.6 ensure tha he minimum utdown Margin is maintained and (2) the potential effects of a CEA ejection ent are li d to acceptable levels. Small CEA misalignments would only have small effec the ependent long-term power distributions, on shutdown margin, and on CEA worth for the CEA Ejection analyses.

Westinghouse procedures for calculating core physics inputs to safety analyses require the application of uncertainty factors to these inputs. The uncertainty factors are determined from benchmarks of the Physics code (e.g., DIT/ROCS for Waterford 3) to plant measurements. For all parameters except power peaking, the uncertainty is defined to bound the 95/95 tolerance limits of the population of total difference between the calculation and the measurement. Since the uncertainty factor is based on the total difference between the calculation and the measurement, it accounts for the measurement uncertainty as well as the pure calculational uncertainty.

4/25/05 DRAFT to W3F1 -2005-0032 Page 9 of 13 The method for measuring control rod worth used by most Combustion Engineering (CE) plants (including Waterford 3) is the CEA Exchange Technique. In this technique a reference bank is defined to be used such that its worth will be exchanged for the various test bank worths. The worth of the reference bank is first measured by boron dilution. As the other "test banks" are inserted one at a time, their reactivity is compensated by movement of the reference bank. The worth of these test banks are inferred by the indicated position of the reference bank.

The uncertainty in the measured control rod worth using this technique is due to many components: (1) control rod position uncertainty; (2) measured boro oncentrations errors; (3) differences between actual values of the kinetics parameters and lues used in the reactivity computer; (4) changes in the reference bank worth du st bank exchange; and (5) effects of spatial flux redistribution on the excore detector sig s

are used to drive the reactivity computer. Since these effects are difficult to qu iser ely, the uncertainty method used by Westinghouse for the CE plants is to ag o

thes rtainty components to the calculational uncertainty.

The uncertainty factors defined by this method ar ed in the safety ana n a conservative manner. For example, scram wo re reducedy the 95/95 lo olerance limit of the total difference between calculation a asure hereas CEA Siks worths used in the Inadvertent CEA Withdrawal Accident a r

by the 95/95 upper tolerance limit. Since these tolerance limits incluhthe measur ncertainty as well as the pure calculational uncertainty, the impact o el CEA posit certainty is thus accounted for in the safety analysis.

i Waterford 3 procedures call for not chang tests to measure Isotherma erature Coefficient (MTC). Sinc 1

r rods on the ITC or MTC redisrssocia e with dus dJ1 e performance of physics MMaderator Temperature I dgvhe testing, there is no impact measurement uncertainty.

Thus, CEA position' were also explicitly, accounting frtthatL to be alsx 7I bias activity woul negligible.

inserti l

, the a uncertairiWbpon4 uncertainty au there would be uncertainty twice r I is unted forWtin the Westinghouse methodology. If it

.Qe ial Specification values, this would be de. Ho he effects of CEA position uncertainty were ite uncertainty analysis over and above the inherent inclusion in

, gents, the impact of the overall CEA scram worth uncertainty ssu ead bank position 3.7 inches beyond the assumed edue CEA scram worth would be less than 0.5%. If this tisticay combined with the remainder of scram worth et uncertainty would increase by a negligible 0.02%. Thus, i associated with the additional burden of accounting for this process.

5.0 REGULATORY ANALYSIS

5.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36, Technical Specifications Paragraph (c)(1 )(ii)(A) requires, in part, that, where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded. No I

44/25/05 DRAFT to W3F1 -2005-0032 Page 10 of 13 Technical Specification limiting safety system settings are changed or affected by this license amendment request since this request is administrative in nature in that it provides descriptions of how Entergy Operations, Inc, (Entergy) accounts for instrument uncertainty at Waterford Steam Electric Station, Unit 3 (Waterford 3).

Paragraph (c)(2), Limiting Conditions for Operation (LCO's) are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Paragraph (c)(2) does not prescribe any specific approach for the treatment of instrument measurement uncertainty. Waterford 3 maintains compliance to 10CFR50.36 for arameters listed above by applying an explicit offset for instrument uncertainty in the an sand/or procedures consistent with the Waterford 3 licensing bases and HICB-12.

General Design Criterion (GDC)

GDC 13, "Instrumentation and Control," requires, a ther things, th rumentation be provided to monitor variables and systems and t trols be provided to in these variables and systems within prescribed operati nges. No' trumentation ntrols are being added or deleted by this license amendme uest sic is request is a inistrative in nature in that it provides descriptions of how Enterg

, Inc, (Entergy) accounts for instrument uncertainty at Waterford St Electnc Sta n

nit 3 (Waterford 3).

GDC 20, "Protection System Function mong o K ings, that the protection system be designed to initiate operation sprastem nsure that specified acceptable fuel design limits are not exce

. No r ec i sy functions or protective system initiation setpoints ed or a ethis amendment request since this request is administrativ tu hat it p descriptio s of how Entergy Operations, Inc, (Entergy) accou instrum uncertai at Waterford Steam Electric Station, Unit 3 (Waterford 3).

Miscellaneous No reg o

i ts e r the incorporation of instrument uncertainty in the operating enve mits (i.e., no S

• nts) used as inputs to the safety analysis process, with the exce initial power

. Re ory Guide 1.49 establishes the requirement that safety analyse erformed for itial p er level that accounts for power measurement uncertaint wever, for pi other than Waterford 3, some approved analysis methodologie dit other c rtainties to support performing analyses without explicit consideration o o

r m ement uncertainty in the power value itself. Also, licensing basis analyses for low pr b

vents that are considered "beyond design basis" are performed at the licensed power le ithout uncertainty (e.g., Station Blackout, Anticipated Transient Without Scram (ATWS).

The determination of the safety significance of instrument functions should consider all available information. This would include review of deterministic requirements, the impact on risk, and other available information. Consideration of the margin of safety associated with applicable parameters would be within this scope. This approach ensures reactor safety, complies with regulatory requirements, is based on sound engineering practices, and avoids unnecessary operating restrictions upon the plant. This allows attention to be focused in a manner to maximize the safety benefit.

4/25/05 DRAFT to W3FI-2005-0032 Page 11 of 13 The accounting of instrument uncertainty for setpoints other than Reactor Protection System (RPS) and Engineered Safety Features Actuation Systems (ESFAS) 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. HICB-12, provides additional guidance for accounting for instrument uncertainty.

Entergy has determined that the proposed change does not re y exemptions or relief from regulatory requirements and does not affect conforma y General Design Criterion (GDC) differently than described in the Updated

)Safe alysis Report (UFSAR.)

5.2 No Significant Hazards Consideration This letter is a request to amend Operating Licen NPF-Waterford StElectric Station, Unit 3 (Waterford 3) to remove the license darding instrument uncertainty that was imposed on Waterford 3 with X approval an nce of the Extended Power Uprate (EPU) amendment (i.e., Amendment 9

e license c iton required that additional information regarding how instrument u accunj r in Technical Specification parameters impacted by EPU be submitti or a revie d approval. The required information was submitted with this licens end uest approval of this request documents the completio C sta v

Ind oI as required by the license condition. The remova f ic onditi allow Wate ord 3 to proceed above 3441 MWt and achieve th P power Ihe of 371 t as authorized in Amendment 199 to the Waterford 3 Opera Icense.

Entergy Operations, Inc.

an ot a significant hazards consideration is involved (as edent(s) by o sing on the three standards set forth in 10 CF "Is of a ement," as discussed below:

1.

s the propose cng e a significant increase in the probability or c

quences of an ident eviously evaluated?

Resr No.

X The propss administrative in nature and does not result in a change to any structure, sys r component (SSC). The accident mitigation features of the plant for previously evaluated accidents are not affected by the proposed change. The proposed change has no impact on the safety analysis because the application of an explicit offset to the Technical Specification parameters for instrument uncertainty provides additional assurance that the plant will operate within the operating envelop previously analyzed.

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.

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

4/25/05 DRAFT

- to W3FI-2005-0032 Page 12 of 13

2.

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

Response: No.

The proposed change is administrative in nature and does not change the design function or operation of any SSC. The proposed change introduces no new mode of operation. The proposed change does not affect the functio=

pability of safety-related equipment. The removal of the license condition low Waterford 3 to operate at the power level of 3716 MWt which has pren evaluated and approved by the NRC staff as documented in Amen to the Waterford 3 Operating License.

Therefore, the proposed change does not of accident from any previously evaluated

3.

Does the proposed change involve a sign Response: No.

The proposed change is admini structure, system, or component C) previously evaluated accidents are affE change has no im on h safety I

y to the Technica fi narame r

assurance thi plant wil erate wi I Existing Te II Specifi io operabi reduced by th osed a he re Waterford 3 to o SF e possibility od or different kind t

a margin sty?

Iature an os not result in a change to any dent tion features of the plant for ed change. The proposed es ade application of an explicit offset r instrume uncertainty provides additional the operating envelop previously analyzed.

End surveillance requirements are not al of the license condition will allow 716 MWt which has previously been is documented in Amendment 199 to the of Based on the hazards consided finding of 'no sigi ncludes that the proposed amendment(s) present no significant standards set forth in 10 CFR 50.92(c), and, accordingly, a consideration" is justified.

5.3 Environmental Considerations The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a 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.

4/25/05 DRAFT to W3F1 -2005-0032 Page 13 of 13

7.0 REFERENCES

7.1 Entergy letter to the NRC dated November 13, 2003, 'License Amendment Request NPF-38-249, Extended Power Uprate" (W3Fl-2003-0074) 7.2 Waterford 3 Final Safety Analysis Report 7.3 Waterford 3 Technical Specifications (through Amendment 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 Specifica ombustion Engineering Plants," Revision 3, June 2004 7.6 Regulatory Guide 1.27, "Ultimate Heat Sink for Nu ea r Plants" 7.7 Generic Letter (GL) 89-13, "Service Water Syste roble ecting Safety-Related Equipment 7.8 10CFR50.36, "Technical Specifications" 7.9 Regulatory Guide 1.105, "Setpoints for R

elated Instrument Revision 3 7.10 Branch Technical Position HICB-12, nce on Es blishing and i

ining Instrument Setpoints," June 1997 7.11 Generic Design Criterion 29, "Protectio A n

5 ated Operational Occurrences" 7.12 NRC Task Interface Agreeme Evaluation ( G

. M95177) dated July 22, 1996 7.13 Regulatory Guide 1.49, "Po s at Nucle wer Plants," Revision 1, December 1973 4/25/05 DRAFT