NL-05-014, Supporting Information for License Amendment Request Regarding Indian Point Unit 3 Stretch Power Uprate

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Supporting Information for License Amendment Request Regarding Indian Point Unit 3 Stretch Power Uprate
ML050380498
Person / Time
Site: Indian Point Entergy icon.png
Issue date: 02/03/2005
From: Dacimo F
Entergy Nuclear Northeast, Entergy Nuclear Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-05-014, TAC MC3552
Download: ML050380498 (43)
v • d • e


Text

AEn tr Entergy Nuclear Northeast Indian Point Energy Center

' Entergy 450 Broadway, GSB PO. Box 249 Buchanan, NY 10511-0249 Tel 914 734 6700 Fred Dacimo Site Vce President Administration

,February 03, 2005 Re: Indian Point Unit 3 Docket No. 50-286 NL-05-014 Document Control Desk U.S. Nuclear Regulatory Commission Mail Stop O-Pl-17 Washington, DC 20555-0001

Subject:

Supporting Information for License Amendment Request Regarding Indian Point Unit 3 Stretch Power Uprate (TAC MC 3552)

Reference. 1. Entergy letter NL-04-069 to NRC, 'Proposed Change to Technicai Specifications: Stretch Power Uorate (4.85%) and Adoption o 'TSlT'F-339", dated June 3, 2004.

2. Entergy letter NL-04-145 to NRC, 'Supporting Information for License Amendment Request Regarding Indian Point 3 Stretch Power Uprate (TAC MC 3552)" , dated November 18, 2004

Dear Sir:

Entergy Nuclear Operations, Inc (Entergy) is submitting additional information to support NRC review of the Stretch Power Uprate (SPU) license amendment (Reference 1) for Indian Point 3 (IP3). This information is being provided as discussed with the staff during teleconferences on January 14, 25 and February 1, 2005.

Attachment 1 provides the information regarding Hot Leg Switchover (HLSO) time; Attachment 2 provides the response to NRC RSB Additional RAI on LOC-4; and Attachment 3 provides Entergy's response to NRC Request for Description of IP3 Compliance with 10CFR50.68.

Attachment 4 provides Entergy's partial response to the NRCs request for clarifications and supporting information regarding Reactor Vessel Internals resulting from the teleconference on January 27, 2005.

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NL-05-014 Docket 50-286 Page 2 of 4 Attachments 5 and 6 contain errata pages based on revised fluence values addressed in for the Stretch Power Uprate Licensing Report transmitted in the original IPEC-Unit 3 License Amendment Request dated June 3, 2004. (Reference 1). Attachment 5 pages are for the proprietary version (WCAP-16212-P) and Attachment 6 pages are for the non-proprietary version (WCAP-16212- NP). Since there is no proprietary information on any of these pages, an application for withholding is not required for these replacement pages. contains additional pages for OPDT/OTDT and Tave calculation to complete the calculation file sent in our November 18, 2004 submittal letter (Reference 2). Included in this submittal are IP3-CALC-RPC-00290 Revision 3, pages 40 to 53 of 56; and Attachment pages 11 and 13 of 13. IP3-CALC-ESS-00281, Revision 2 pages 37 to 41 of 52.

Based on the Staff's request during teleconference on February 1, 2005, Entergy is specifically requesting NRC approval of the Accident Analysis assumption pertaining to auxiliary feedwater operation for the Loss-of-Normal-Feedwater (LONF) event and the Loss of All AC Power to Station Auxilairies (LOAC) event. (See WCAP-16212-P, Section 6.3.1, 6.3.7 and 6.3.8. provides Indian Point Piping Vibration (PV) Plan Logic Diagrams with Examples.

The additional supporting information provided in this letter does not alter the conclusions of the no significant hazards evaluation that supports the subject license amendment requests. There are no new commitments being made in this submittal. If you have any questions or require additional information, please contact Mr. Patric Conroy at (914) 734-6668.

I declare under penalty of perjury that the foregoing is true and correct. Executed on 02/03/05.

Fred R. Dacimo Site Vice President Indian Point Energy Center

NL-05-014 Docket 50-286 Page 3 of 4 Attachment 1: Response to NRC Question regarding Hot Leg Switchover (HLSO) time from January 14, 2005 Teleconference Attachment 2: Response to NRC RSB Additional RAI on LOC-4 from January 14, 2005 Teleconference Attachment 3: Response to NRC Request for Description of IP3 Compliance with 10CFR50.68 Attachment 4: Partial Response to NRC request for clarifications and supporting information regarding Reactor Vessel Internals resulting from teleconference on January 27, 2005.

Attachment 5: Errata Pages for WCAP-1 6212-P Indian Point Nuclear Generating Unit 3 Stretch Power Uprate NSSS and BOP Licensing Report, June 3, 2004 Submittal.

(Proprietary version)

Attachment 6: Errata Pages for WCAP-16212-NP Indian Point Nuclear Generating Unit 3 Stretch Power Uprate NSSS and BOP Licensing Report, June 3, 2004 Submittal (Non-proprietary version)

Attachment 7: OPDT/OTDT and TAve Calculation Pages Attachment 8: Indian Point Piping Vibration (PV) Plan Logic Diagrams with Examples

- cc: nextpage

NL-05-014 Docket 50-286 Page 4 of 4 cc: Mr. Patrick D. Milano, Senior Project Manager Project Directorate I Division of Licensing Project Management U.S. Nuclear Regulatory Commission Mr. Samuel J. Collins Regional Administrator, Region I U.S. Nuclear Regulatory Commission Resident Inspector's Office Indian Point Unit 3 U.S. Nuclear Regulatory Commission Mr. Peter R. Smith, President New York State Energy, Research and Development Authority Mr. Paul Eddy New York State Dept. of Public Service

ATTACHMENT 1 TO NL-05-014 Response to NRC Question regarding Hot Leg Switchover (HLSO) time from January 14, 2005 Teleconference (1 Page)

ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

Attachment 1 to NL-05-014 Docket 50-286 Page 1 of 1 Hot Leg Switchover (HLSO) Time NRC Request:

Confirm that ECCS flows are sufficient to match core boil off rates for the earliest time at which Hot Leg Switchover (HLSO) will be initiated.

Entergy Response for IP3:

Nuclear Safety Advisory Letter NSAL-04-1 addressed concerns related to the time at which HLSO should be initiated relative to the calculated HLSO time and also addressed the need to evaluate the adequacy of core injection flow if earlier HLSO times are implemented in EOPs.

The maximum HLSO time in post-LOCA calculations is based on boric acid precipitation potential and is then checked to ensure that adequate flow is available at that HLSO time. The EOPs are written to instruct operators to initiate HLSO at the specified HLSO time, and it is recognized that the HLSO realignment process requires a finite amount of time. The acceptability of this approach is based on the nature of the HLSO calculations and the conservatism in the methodology used to calculate HLSO time. Most significant is the 4%

uncertainty margin applied to the boric acid saturation limit of 27.53 weight percent (at atmospheric pressure). This 4% reduction in the boric acid saturation limit typically translates to a margin of more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> between the recommended HLSO time and the time at which.

boric acid precipitation may potentially occur.

The EOP-designated HLSO times are interpreted as the beginning of the hot leg recirculation

.,realignment process. This is consistent with the definition of ERG footnote V.01 (Time for transferring to hot leg recirculation). For the reasons described above, there is sufficient margin tin the HLSO calculations such that the realignment can be completed before the potential for boric acid precipitation exists.

Entergy has evaluated the earliest time at which preparations for HLSO actions after a postulated LOCA can start. This early time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is based on radiological dose considerations. IP3 procedures will allow activities to prepare for HLSO to start at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, but specifically state that HLSO is to "commence" at 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

Nevertheless Entergy evaluated an early 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> switchover to hot leg recirculation for Indian Point Unit 3. Breaks in both the cold leg and hot leg were considered. For ECCS injection lines on the leg with the assumed break (e. g., hot leg or cold leg), ECCS spillage was conservatively calculated and considered. Decay heat was based on the Appendix K required decay heat standard (1971 ANS, infinite operation, with 20% uncertainty). No SI subcooling was assumed.

Both active and passive failures were considered.

The calculated core boil-off at 4.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> after the postulated LOCA is 285 gpm. For a cold leg break and the limiting single active failure, the available ECCS flow after realignment to hot leg recirculation is 477.6 gpm. For a hot leg break and the limiting single active failure, the available ECCS flow is 345.5 gpm. For the limiting single passive failure, the evaluation credited LHSI consistent with the actions specified in the EOPs. The available low head ECCS flow for the limiting single passive failure is 751 gpm with one line spilling. Thus, for either hot leg or cold leg breaks, the available ECCS flow is well in excess of the calculated core boil-off at 4.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> after the postulated LOCA.

In summary, a HLSO "window" of 4.0 - 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is confirmed acceptable for Indian Point Unit 3.

ATTACHMENT 2 TO NL-05-014 Response to NRC RSB Additional RAI on LOC-4 from January 14, 2005 Teleconference (1 Page)

ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

Attachment 2 to NL-05-014 Docket 50-286 Page 1 of 1 Response to NRC RSB Additional RAI on LOC-4 NRC Request:

For the IP3 Response to LOC-4, the equivalent analysis for IP2 goes out to 1600 seconds. The plots for IP3 only go to 600 seconds. Please provide explanation /justification to conclude that IP3 performance beyond 600 seconds would be consistent with IP2.

Entergy Response:

It is reasonable to conclude the IP3 performance beyond 600 seconds will be consistent with that of IP2 because at 600 seconds both plants exhibit the following trends:

  • Rod cladding surface temperature is stable, decreasing and approaching the saturation level.
  • Downcomer collapsed level is stable and the effects of downcomer boiling have clearly been mitigated.
  • Core collapsed liquid levels are steady.
  • Liquid pool is established and maintained in the upper plenum above the core plate and below the hot leg bottom.

. Loss of inventory through the break is replenished by a steady safety injection flow as evidenced by the increasing reactor vessel mass.

For IP2 these trends are sustained beyond 600 seconds, as demonstrated by the reported extended transient.

With respect to long-term core quench behavior, IP2 and IP3 are plants with very similar features. They are both 4-loop Westinghouse designed PWRs with the same number of fuel assemblies, same power level, and similar peaking factors. Their ECCS systems and containments are sufficiently similar that it is reasonable to conclude that the long term performance will be consistent for the two plants. It is further noted that there are no safety system related differences between the two units that would significantly affect their expected long term performance.

Therefore, the post-600 seconds stable quench behavior predicted for IP2 will also apply to this IP3 analysis.

ATTACHMENT 3 TO NL-05-014 Response to NRC Request for Description of IP3 Compliance with IOCFR50.68 (3 Pages)

ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

Attachment 3 to NL-05-014 Docket 50-286 Page 1 of 3 NRC Request:

Need description of how IP3 meets 10CFR50.68 Entergy Response:

Compliance with each of the requirements of 10CFR50.68 is discussed below.

10CFR50.68(b)(1) Requirement:

Plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.

Compliance:

All fresh fuel assemblies must meet the Fresh Fuel Storage Rack criticality requirement that all fresh fuel above 4.5 w/0 must contain a minimum number of Integral Fuel Burnable Absorber's (IFBAs). Standard Operating Procedure SOP-RP-6, New Fuel Removal from Shipping Container and Inspection, currently permits only one new fuel assembly to be in transit between the associated shipping cask and dry storage rack. This is stated in the Precautions and Limitations section of the procedure.

10CFR5O.68(b)(2) Requirement:

The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.

Compliance:

The new fuel storage facility is used to receive and store new fuel in a dry condition upon arrival on site and prior to loading in the reactor. The new fuel storage racks are designed to store new fuel in a geometric array that precludes criticality. A criticality analysis was done to demonstrate that k-effective is maintained less than or equal to 0.95 when the new fuel racks are fully loaded and dry or flooded with moderator in the event of a design basis fuel handling accident. This analysis was reviewed and approved by the NRC (NRC letter to NYPA, issuance of Amendment for Indian Point Nuclear Generating Unit No. 3 (TAC NO. M96474)", dated April 15, 1997).

I OCFR50.68(b)(3) Requirement:

If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.

Compliance:

Technical Specification 4.3.1.2.b assures compliance with the requirement of 10CFR50.68(b)(3): "The new fuel storage racks are designed and shall be maintained with: k-effective </= 0.95 under all possible moderation conditions (Credit may be taken for burnable integral neutron absorbers)."

Attachment 3 to NL-05-014 Docket 50-286 Page 2 of 3 IOCCFR50.68(b)(4) Requirement:

If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.

Compliance:

Technical Specification 4.3.1.1.b assures compliance with the requirement of 10CFR50.68(b)(4): "The spent fuel storage racks are designed and shall be maintained with: k-effective </= 0.95 if assemblies are inserted in accordance with Technical Specification 3.7.16, Spent Fuel Assembly Storage."

10CFR50.68(b)(5) Requirement:

The quantity of SNM, other than nuclear fuel stored onsite, is less than the quantity necessary for a critical mass.

Compliance:

The total amount of non-fuel SNM on site is such that it meets the 'forms not sufficient to form a critical mass" guidance in Section 1.1 of Regulatory Guide (RG) 10.3 and the total amount of non-fuel SNM is significantly less than the quantities delineated in 10 CFR 70.24(a).

10CFR50.68(b)(6) Requirement:

Radiation monitors are provided in storage and associated handling areas when fuel is present to detect excessive radiation levels and to initiate appropriate safety actions.

Compliance:

Area radiation monitor channel R-5 monitors radiation levels in the Fuel Storage Building. This provides warning thatwater in the spent fuel pool is highly contaminated, that fuel is being improperly handled, or that the pool level is dangerously low for prevailing conditions. High radiation alarms are displayed on the main annunciator, the radiation monitoring cabinets, and at the detector location. In the alarm condition, the supply air tempering units trip if running, the exhaust fan operates to maintain negative pressure in Fuel Storage Building, the face dampers to the charcoal filter will open if closed, the Fuel Storage Building rolling door closes and the air is applied to the door seals. The bypass dampers around the charcoal filters must be manually closed if open.

I0CFR50.68(b)(7) Requirement:

The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to five (5.0) percent by weight.

Compliance:

Technical Specification 4.2.1 restricts the enrichment of reload fuel to no more than 5.0 weight percent U-235.

Attachment 3 to NL-05-014 Docket 50-286 Page 3 of 3 IOCFR50.68(b)(8) Requirement:

The FSAR is amended no later than the next update which §50.71 (e) of this part requires, indicating that the licensee has chosen to comply with §50.68(b).

Compliance:

The UFSAR Section 9.5 will be updated following the next refueling outage to state that IP3 has chosen to comply with 10CFR50.68(b).

ATTACHMENT 4 TO NL-05-014 Partial Response to NRC RAI Regarding Reactor Vessel Internals from January 27, 2005 Teleconference (1 Pages)

ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

Attachment 4 to NL-05-014 Docket 50-286 Page 1 of 1 Partial Response to NRC RAI Regarding Reactor Vessel Internals from January 27, 2005 Teleconference NRC RAI Request #1 (Ref; NL-04-156 dated 12/15/04)

Validate fluence values provided in Table 5.1-3, specifically fluence value should be 0.922 x 10'9 n/cm2) Additionally confirm if other data relies on the fluence data in table 5.1-3 Response to Request #1:

Westinghouse provided, during the teleconference on 1/27/05, assurance that the fluence value was not used in other tables. The revised Table 5.1-3 and page 5.1-5 are provided in Attachments 5 and 6 to this response.

ATTACHMENT 6 TO NL-05-014 ERRATA PAGES FOR WCAP-1 6212-NP INDIAN POINT NUCLEAR GENERATING UNIT 3 STRETCH POWER UPRATE NSSS AND BOP LICENSING REPORT (Non-proprietary version)

Revised Table 5.1-3 and page 5.1-5 (2 Pages)

ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

Table 5.1-3 RT Ts5 Calculations for IP3 Beltline Region Materials at 27.1 EFPY with (3216 MWt) SPU Fluences Fluence (10enlcMI CF ARTpTs() Margin RTNrDT(u)() RTPTs( 3 ) I 0 0 0 Material E>1.0 MeV) FF ( F) ( F) ( F) (OF) (°F)

Intermediate Shell Plate 0.922 0.977 137 133.8 34 5 173 B2802-1 Intermediate Shell Plate 0.922 0.977 152 148.5 34 -4 179 B2802-2 Intermediate Shell Plate 0.922 0.977 136 132.9 34 17 184 B2802-3 Lower Shell Plate B2803-1 0.922 0.977 128 125.1 34 49 208 Lower Shell Plate B2803-2 0.922 0.977 150 146.6 34 -5 176 Lower Shell Plate B2803-3 0.922 0.977 160 156.3 34 74 264 e Using S/C Data 0.922 0.977 170.9 167.0 17(4) 74 258 Intermediate and Lower Shell Weld Longitudinal 0.922 0.977 224 218.8 65.5 -56 228 Weld Seams (heat 34B009) I Intermediate to Lower Shell Circumferential weld Seams 0.922. 0.977 189 184.7 56 -54 187 (heat 13253) 11 Notes:

1. ARpTrs = CF
2. Initial RTNDT values are measured values except for the intermediate and lower longitudinal welds.
3. RTPrTs = RTNDT(u) + ARTPTs + Margin (OF)
4. Using credible surveillance data.

6389\sec5_1.doc(060204) 5.1-9 WCAPL16212-NP NSSS and BOP Licensing Report Rev. 0-errata

power level of 3216 MW through 27.1 EFPYs (EOL) for IP3 as shown in Table 5.1-3. The change in RTpTs due to the SPU, as compared to the MUR Program to 3068 MWt, is 1"F. This evaluation also determined that the limiting material is relatively close to the PTS screening criteria of 2700F and is expected to exceed this screening criteria at -36 EFPY.

5.1.2.5 Upper Shelf Energy All beltline materials have a USE greater than 50 ft-lb through 27.1 EFPY (EOL) as required by the Code of Federal Regulations (CFR) 10CFR50, Appendix G (Reference 6). The 27.1 EFPY (EOL) USE was predicted using the EOL 1/4 thickness (1/4t) SPU fluence projections that correspond to a SPU power level of 3216 MWt. Despite the fact that the vessel fluence projections have increase due to the SPU, as compared to the MUR Program to 3068 MWt, the change in USE decrease is zero. The USE values are presented in Table 5.1-5.

5.1.2.6 Inlet Temperature RG 1.99, Revision 2 (Reference 7), which is also the basis for 10CFR50.61 (Reference 5),

states that "The procedures are valid for a nominal irradiation temperature of 5500 F. Irradiation below 5250F should be considered to produce greater embrittlement, and irradiation above 5900 F may be considered to produce less embrittlement." The temperature range of 5250F to 5900 F serves as the basis of the equations and tables that are used in all the RV internal analyses described herein. Therefore, the inlet temperature, which is the temperature to which the reactor vessel is subjected, must be maintained within this range to uphold all existing analyses.

5.1.2.7 Conclusions The fluence projections used for the SPU, while considering actual power distributions incorporated to date, have increased versus the fluence projections developed for the MUR Program (to 3068 MWt). However, this increase has had minimal affect on the analyses of record for reactor vessel integrity since the PTS and USE remain within the acceptance criteria, the PTS curves had less than I EFPY decrease, the ERG category remains unchanged, and there were only minor withdrawal time changes to the withdrawal schedule. The regulatory criteria continue to be met for the SPU conditions. Therefore, there is no significant effect on RV integrity related to the SPU.

6389\sec5_1.doc(060204) 5.1-5 WCAP.16212-NP NSSS and BOP Licensing Report Rev. 0-errata I

ATTACHMENT 7 TO NL-05-014 OPDT/OTDT and Tave Calculation Pages (21 Pages)

ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

ENN NUCLEAR QUALITY RELATED ENN-DC-126 Revision 3 MADMINISTRATIVE PROCEDURE

-Entergy MANUAL INFORMATONALUSE Calculation Page 40 of 56 Calculation No. IP3-CALC-RPC-00290 Revision 3 Project: ER No. 04-3-027

Title:

Instrument Loon Accuracy / Setrpoint Calculation. Overpower Delta-T (OPDT) and Overtemperature Delta-T (OTDT), Reactor Trip 10.0 DETERMINE ALLOWABLE VALUE (AV)

The Allowable Value (AV) can be calculated from the following equation; (Refs. 3.1.3 & 3.2.1)

AV = TS +/- CUCAL Where, TS = Trip Setpoint CUCAL = Channel Uncertainty (CU) as seen during calibration. Therefore, uncertainties due to a harsh environment, process measurement, or primary element are not considered. For conservatism, only RA, DR and ALT uncertainties are considered. The following use AFT and en CAL interchangeably. CUcAL is based on; CUCAL+/-= i e, CAL 2+e, CAL2 Where; excAl = 12

+DR ' + AL T The AV will be calculated using the SRSS method consistent with the method used for the determination of the trip setpoint. Therefore, a check calculation is not required. (Ref. 3.1.3) 10.1 Determine e,>,

10.1.1 Determine AFT 1 (Ref. 3.2.17)

As defined above CUcAL only considers the normal uncertainties as seen during calibration, therefore, the module uncertainty equation el reduces to; AFTi= +/-VRAi+ DRM+ ALT,2 + SH12 The el effects for RA, DR and ALT are substituted in the above equation.

AFT, = +/- o.o32 + 0.0452 + O.072 + o.0o052 AFT = +/-0.089 % of Span The calibrated span for el Is 30-7000 F, therefore; AFT = (+/-0.089%0)* (6707F)

AFT, = 0 592 OF Converting "OF* to "% of AT Span", given AT = 750 F; AFT,= 592 *(100) 0.790% of A T Span 75 Similarly for modules e2 , e3, e4 , es and e8 , the uncertainty associated with the module calibration is; 10.1.2 Determine AFT2 ,

AFT 2 = 1052 + 0412 + 0.52 (Ref. 3.2.17)

AFT2 =+/- 0.817 % of Span e2 Calibration Span = 1200F

ENN NUCLEAR QUALITY RELATED ENN.DC-126 Revision 3 is MANAGEMENT ADMINISTRATIVE PROCEDURE

-Entergy MANUAL INFORMATIONAL USE Calculation Page 41 of 56 Calculation No. IP3-CALC-RPC-00290 Revision 3 Project: ER No. 04-3-027

Title:

Instrument Loop Accuracy / Setnoint Calculation, Overpower Delta-T (OPDT) and Overtemnerature Delta-T (OTDT). Reactor TriD AFT 2 =(+/-O.817 %)*(120OF)

AFT 2 = +/-0.9800 F Converting "OF" to "% of AT Span", given AT = 75°F; AFTz= +/-098*(100)=+1.307%of AT Span 75 10.1.3 Determine AFT 3 (Refs. 3.2.17 & 3.5.13)

The only microprocessor test requirements are for a software check. Therefore there is no As-Found Tolerance.

Therefore, AFT 3 = 0 10.1.4 Determine AFT 4 ,

AFT4 = +/- Jo. 2 + 0252 + 0.52 (Ref. 3.2.17)

AFT = +/- 0.75 % of Span e4 Calibration Span = 120°F AFT4 = (+/-0.75 o) *(120 OF)

AFT4 =+/-0.90°F Converting "°F to "% of AT Span", given AT = 75°F; AFT4=+/- +/- *(100)= 1.20%of AT Span 75 10.1.5 Determine AFTs AFT, = +/- 0.22+0.652+0.52(8/5) (Sect. 7.7)

AFT,=1.350% of Span e5 Calibration Span = 75°F AFT,=(1.350%)*(75OF)

AFT-,= +/- 1.013 OF Converting "'F'to "% of AT Span" given AT = 75°F, AFTs,=+/-i *(100)=+/-1.350% of AT Span 75 10.1.6 Determine AFT6 AFT 6 = F+/-1220.2?+0.6S2 +0.632 (8/5) (Sect. 7.7)

AFT6 +/-1.483 %

e6 Calibration Span = 75°F

ENN NUCLEAR QUALITY RELATED m MANAGEMENT ADMINISTRATIVE PROCEDURE ENN-DC-126 Revision 3

~Enterby MANUAL INFORMATIONAL USE Calculation Page 42 of 56 Calculation No. IP3-CALC-RPC-00290 Revision 3 Project: ER No. 04-3-027

Title:

Instrument Loop Accuracy / Setpoint Calculation. Overpower Delta-T (OPDT) and Overtemperature Delta-T (OTDT), Reactor Trio AFT6 = (+/- 1.483 %) * (75 OF)

AFT 6 = +/- 1.112 OF Converting "OF" to '% of AT Span", given AT = 75°F; AFT6 I 1 *O(100)+ 1.483% of AT Span 75 1i0.1.7 Determine AFTp lAFTP=+/-J4162+0.S2 +0.3 (Drift Bias) (Sect. 7.8)

AFTp=+/-1.676%,+0.3ofSpan ep Process Span = 800 PSI AFTp = +/-(1.676 %o)* (800) + (0.3%) * (800)

AFTp = +/-13.408 P.S.L,+2.4P.S.1.

10.1.8 Determine AFT 7 The uncertainty for e7 Is a "lumped" term given as 1.5% Power.

For purposes of determining an AFT value, we will consider only 1.0% as sensible during calibration. Therefore, AFT will be taken as the following:

AFT = [1.5*54/75]*4.0*1.2 AFT,=5.184%of A T Span (Sect. 7.9) 10.1.9 Determine AFT 8 AFFa +/- 10.52 +0.52 (Sect. 7.10)

AFT 8 = i 0.707 % Power AFT 8=[0.707

  • 54/ 75]
  • 4.0
  • 1.2 AFT8 =+/-2.443%ofATSpan 10.1.10 Determine AFT 9 AFT,=+/- N0.a8+ .0 (Sect. 7.11)

AFT 9 =+/-1.131% Power AFTs = [1.131

  • 54/75]
  • 4.0
  • 1.2 AFT 9 = + 3.910% of AT Span 10.1.11 Determine AFT10 AFTo =+/- i0.8662+0o.52 (Sect. 7.12)

AFTo = +/- 1.0 % of A T Span

ENN NUCLEAR OUAUITY RELATEDENDC16Rvso3 MANAGEMENT

?Entergy MANUAL ADMINISTRATIVE PROCEDURE INFORMAnONAL USE Calculation Page 43 of 56 Calculation No. IP3-CALC-RPC-00290 Revision 3 Project: ER No. 04-3-027

Title:

Instrument Loon Accuracy / Setpoint Calculation. OverDower Delta-T (OPDT) and Overtemperature Delta-T (OTDT), Reactor Trip 10.1.12 Determine AFT, AFTi,=+/-1O.8662+0.5s (Sect. 7.13)

AFTX=+/-1.0%of A T Span 10.1.13 Determine AFT12 2

AFT2 =+/-O S2 +02 +O0 2 (Sect. 7.14)

AFT12 = +/-0.735 % of A T Span 10.1.14 Determine AFT13 2

AFT3 +/-i 1 + .2 2 +O5 2 (Sect. 7.14)

AFT, 3 =+/-O.735%ofA T Span 10.2 Determine CUcAL for OPAT Given the above CUCAL definition, the channel uncertainty equation for OPAT from Section 7.17 reduces to; CUCAL = iAFT' + AFT' + Ani + AFT.+ AFT, + AFT2 +AFTJ20+ AFT2 (OPAT) 10.2.1 Since CUcAL is a function of the THOT and TcOLD parameters of AT, the T, CUTHOT, CUTCOLD and CUTavg equations must be calculated for the calibration portion of the loop. Therefore, solving for TcAL; 1o/3( AFI2 + AFT2)

TcAL- 3AT 3 3

TcAj= +/-413(0.7902+1.3072) 3 8

Tcu= +/-0. 8l%Of A T Span 10.2.2 Solving for CUTHOT calibration; CUTHrOT(CAL)=+/-VTCAL + AFT. + AFT!

CUTJOT (CAL) +/- ,O.s 88j + 0.0 + 1.202 CUTIIOT(CAL) = +/- 1.488% of A T Span 10.2.3 Solving for CUTCOLD calibration; CUTcow(CAL)= fT AL + AFT2+ AFT' CU7CoLD(CAL)=+/- 10.88f2 + ).3072 1.202 CUTCoLD(CAL)=i1.981%Of A T Span

EMANAGEMENT ANAGMENT QUALITY RELATED ADMINISTRATIVE PROCEDURE ENN-DC-126 Revision 3

, Entergy MANUAL INFORMATIONAL USE Calculation Page 44 of 56 Calculation No. IP3-CALC-RPC-00290 Revision 3 Project: ER No. 04-3-027

Title:

Instrument Loon Accuracy / Setnoint Calculation, Overpower Delta-T (OPDT) and Overtemnerature Delta-T (OTDT), Reactor Trip 10.2.4 Solving for CUTaVg calibration; CUr4vi;(CAL) = +/- VCUTII JT (CAL) + CU2COW (CAL) 2 CUr,lv(CAL)=+/- A/1-488 + 1.98 i2 CUTAvG (CAL) = 1.238 % of A T Span 10.2.5 Solving for CU AT calibration (CUaTcAL);

CU AT (CAL) iCUTIOT(CAL)+ CU2CO (CAL)

CUATtCAL)=+/-11-4882+1-981 CUATCAL)= +/- 2.477 % of A T Span 10.2.6The term above [AFI +AFT22+AFTJ +AFT,2] represents the total Calibration Uncertainties of Ta, and AT circuits. Therefore, this term can be replaced with;

[CUTAVG(CAL)+ CU1T(CAL)J.

CUcAcop~aT = +/- {[CU,avg,(Cal)*K 612 + CU 2 AT(CAL) + AFT 2 5 + AFT 210 +AFT2 12)"2 CU cAL (OPAT) = +/-([1.238*0.001 5]2 + 2.4772 + 1.3502 + 1.4832 + 1.02 + 0.7352)1/2 CU CAL(OPAT) = +/- V(1.238

  • 0.0015) + 2.4772 +1.350 +1.4832 +1.002 +0.7352 CU cal (OPAT) = +/- 3.348% of AT Span 10.3 Determine CUCAL for OTAT Given the above CUcAL definition, the Channel Uncertainty equation for OTAT from Section 7.18 reduces to; CUCAL+/-+IAFT2+AEI2+AFTj+AFl4+AFTr+AFT62+(AFTp K3 +AFT+AF 8+AFTr+AFTr,+AFrl3(OT1T)

Since IAFT 21 +AFT2 2 + AFT 23 +AFT 241 = ICU2tavg (CAL) +CU2DELTAT(CAL)l 2

CU CAL(OTAT) = +/-[CU.-,(CAL)K.l]- +CUhr(CAL)+A FT. +AF2i+( AFrFT,*K, +AFT + AFT'+ AFT2 + AFTf +AFr2.,

CU CAL (OTDT) = +/- [(1.238 0.022) +2.4772 + 1.3502 + 1.4832+ (13.408*0.0007)2 + 5.1842

+ 2.4432 + 3.9102 +1.02 + 0.7352]112 + 2.4*0.0007 CU cAL(oT&T =+/- 7.736 % of AT Span 10.4 OPAT Allowable Value (AV) Calculation Calculating for OPAT (K4) Allowable Value (AV)

Given, TS = 1.0807 (Sect. 9.1)

CUCAL(OPAT) = +/-3.348% of AT Span Using the conversion 1.3888, from Section 7.9.1

ENN NUCLEAR QUALITY RELATED ENN-DC-126 Revision 3 ADMINISTRATIVE PROCEDURE Enterfegy MANUAL INFORMATIONAL USE Calculation Page 45 of 56 Calculation No. IP3-CALC-RPC-00290 Revision 3 Project: ER No. 04-3-027

Title:

Instrument Loop Accuracy / Setpoint Calculation, Overpower Delta-T (OPDT) and Overtemperature Delta-T (OTDT). Reactor Trip CUCAL(OPAT) = (+/-3.348%)* (1.3888)

CUCAL(OPAT) = +/-4.649% Power The magnitude of CUCAL is combined with the Trip Setpoint in an appropriate direction to determine AV for an increasing signal.

Therefore, AV = TS - CUCAL AV = 1.0807 + 0.04649 AV = 1.127 (OPAT)

Therefore the allowance for uncertainties between AL and AV Is:

1.164 - 1.127 = 0.037 (OPAT) 10.5 OTAT Allowable Value (AV) Calculation Calculating for OTAT (K1) Allowable Value (AV)

Given, TS = 1.42- 0.1795 TS = 1.241 (Sect. 9.2)

CU CAL (OTAT) = +/-7.736% of AT Span Converting "% of AT Span" to a value based on 138.88% full power CU CAL (OTAT) = (+/-7.736%) * (1.3888)

CU CAL (OTAT) = +/-10.74% Power The magnitude of CUCAL is combined with the Trip Setpoint in an appropriate direction to determine AV for an increasing signal. Therefore, AV = 1.241 + 0.1074 AV = 1.348 (OTAT)

Therefore the allowance for uncertainties between AL and AV is:

1.42 - 1.35 = 0.07 (OTAT)

Note: The NRC Staff position regarding AV Is currently uncertain. The Staff has generalized that the application of ISA Method 3 may not be a conservative approach for determining AV. Therefore, the following evaluation is Included to document an ultra conservative ISA Method 2 AV.

This evaluation will combine all uncertainties except RA, DR and ALT ( shown bolded) to determine a channel uncertainty value Utot,, which will be added or subtracted, as appropriate, from the AL to determine an AV which will be closer to the Trip Setpolnt as shown below.

10.6 AV Evaluation using method 2 variation for Un The Allowable value will be calculated by the following;

ENN NUCLEAR QUAUTY RELATED ENN-DC-126 Revision 3 MANAGEMENT ADMINISTRATIVE PROCEDURE Entergy MANUAL INFORMATIONAL USE Calculation Page 46 of 56 Calculation No. IP3-CALC-RPC-00290 Revision 3 Project: ER No. 04-3-027

Title:

Instrument Loon Accuracy / Setpoint Calculation, Overpower Delta-T (OPDT) and Overtemperature Delta-T (OTDT), Reactor Trip 10.6.1 Determine U1 , for module e1 :

Where: Bias = 0 el = +/-(O.032 +0.0452 +0.0 +0.0 +0.0 +0.032 +0.072 +0.00152)11 U. = +/-0.03% for the RTD (Cold Leg and Hot Leg) calibrated span is 30 - 7000 F U1 = +/-[(0.03%)(6700 F) = 0.200 F for the RTD (Cold Leg and Hot Leg) calibrated span Converting T"°F to "% of AT Span", given AT = 750 F; U. = +/- (0.200 F / 750 F) * (100)

U1 =t+/-0.268% of AT Span 10.6.2 Determine U2 for module e2 :

Where: Bias = 0 e2 = +/-(O.52 +0.412 +0.272 +0.252 +0.112 +0.5V2)'tO U2 = +/-0.38% of span Given the R/E calibrated span is bounded by 5200F to 6400 F (or 1200 F) U2 effect in terms of IFMs, U2 = +/-(0.38%)(1201F) = 0.1 770 F Converting "'F" to "% of AT Span", given AT = 750 F; U2 = +/-(0.1770 F / 750 F) '(100)

U2 = +/- 0.236% of AT Span 10.6.3 Determine U3, for module e3 :

Where: Bias = 0 e3 = +/-(0.102 +0.502)"t0 U3 = +/-0.50% of span Given that the R/E calibrated span is 5200 F to 6400 F (or 120 0F), U3 effect in terms of "F" is, U3 = (1 20'F) (0.50%) = +/-0.6000 F Converting 0 1F"to "% of AT Span", given AT = 750F; U3 = +/-(0.6000 F / 75°F) *(1 00)

U3 = +/- 0.800% of AT Span

ENN NUCLEAR QUALITY RELATED ENN-DC-126 Revision 3

- MNG ADMINISTRATIVE PROCEDURE

- Entergy MANUAL INFORMATIONAL USE Calculation Page 47 of 56 Calculation No. IP3-CALC-RPC-00290 Revision 3 Project: ER No. 04-3-027

Title:

Instrument Loop Accuracy / Setpoint Calculation. Overpower Delta-T (OPDT) and Overtemperature Delta-T (OTDT), Reactor Trip 10.6.4 Determine U4 for the module e4:

Where: Bias = 0 e4= +/-(o.52 +0.25 +0.243 +0.502 +0.11' +0.50)'+/-O U4 = +/-0.57% of span Given that the E/i calibrated span is 520'F to 640'F (or 120 0F), U4 effect in terms of "OF" is, U4 = +/-(0.57%)(120'F)

U4 = +/-0.68OF Converting "IF" to "% of AT Span", given AT = 750 F; U4 = +/-(0.680°F / 75°F) *(1 00)

U4 = +/- 0.907% of AT Span 10.6.5 Determine Us, for the module eS:

Where: Bias = 0 e5 = +/-(0.202 +0.652 +0.302 +.5025+0.122 +0.502)M+/-0 U5 = +/-0.595%

Because of the input to output relationship discussed above, we will also multiply this uncertainty by 8/5 to account for the gain effect. Therefore:

Us = +/-.95% of span Given the dynamic compensator (Cold Leg and Hot Leg) calibrated span is 540 to 615°F or 75 0F, Us effect in terms of 0F" is, (ref. 3.5.4)

U5 = +/-(0.95%)(750F)

U5 = +/-.7140 F 10.6.6 Determine U6 , for the module e6 :

Where: Bias = 0 e6 = +/-(o.202 +0.652 +0.3022 +0.502 +0.122220.632) +/-0 Us = +/-.597% of span Because of the input to output relationship discussed above, we will also multiply this uncertainty by 8/5 to account for the gain effect. Therefore:

U6 = +/-95% of span Given the bistable (Cold Leg and Hot Leg) calibrated span is 540°F to 615°F or 75°F, U6 effect in terms of UOFF is, (ref.3.5.4)

U6 = +/-(0.95%)(75 0F)

U6 = +/-.716'F

ENN NUCLEAR QUALITY RELATEDEN-C16Rvso3 MANAGEMENT ADMINISTRATIVE PROCEDURE ENN-DC-126 Revision 3

-Enter rgy MANUAL INFORMATIONAL USE Calculation Page 48 of 56 Calculation No. IP3-CALC-RPC-00290 Revision 3 Project: ER No. 04-3-027

Title:

Instrument Loop Accuracy / Setpoint Calculation. Overpower Delta-T (OPDT) and Overtemperature Delta-T (OTDT). Reactor Trip 10.6.7 Determine U7, for the module e7 :

Where: Bias = 0 e-,= +/-5.184% of AT span For this module it is considered conservative to use a value of 100% of e7 to determine U7.

U7 = +/-5.184% of AT span 10.6.8 Determine U8, for the module e8 :

Where: Bias = 0 e8 = +/-(0.562 +0.042 +0.502) 0o U8 = +/-0.04% of full span power Using the full power range span, f (Al) penalty and conversion for full power AT U8 = +/- (0.04^ 54/75)* 4.00*1.2 U8 = t 0.138% of AT Span 10.6.9 Determine U9, for the module eg:

Where: Bias = 0 eq = +/-(o.802 +0.042 +0.802)"t0 U9 = +/-0.04% of full span power Using the full power range span, and conversion for full power AT U9 = t (0.04^ 54/75)* 1.2 Ug = +/- 0.035% of AT Span 10.6.10 Determine U10, for the module el0 :

Where: Bias = 0 e10 = *(O.802 +0.802)1 to U20 is considered to be negligible for this module, therefore, U10=ta0.0 of AT Span 10.6.11 Determine Usi, for the module ell:

Where: Bias = 0 ell = +/-(0.866 +0.502)%+/-0 U., is considered to be negligible for this module, therefore, U., = +/- 0.0 of AT Span

ENN NUCLEAR QUALITY RELATED ENN-DC-126 Revision 3 MANAGEMENT ADMINISTRATIVE PROCEDURE

-- Entergy MANUAL INFORMATIONAL USE Calculation Page 49 of 56 Calculation No. IP3-CALC-RPC-00290 Revision 3 Project: ER No. 04-3-027

Title:

Instrument Loop Accuracy / Setnoint Calculation, Overpower Delta-T (OPDT) and Overtemperature Delta-T (OTDT), Reactor Trio 10.6.12 Determine U12 for the module e12:

Where: Bias = 0 e,2 = +/-(0.52 +0.20 +0.2432 +0.502 +0.1912 +0.50e)f+/-0 U12 = +/-0.588% of AT Span (OP AT) 10.6.13 Determine U13 for the module e13 :

Where: Bias = 0 e13 = +/-(0.52 +0.20 +0.2432 +0.502 +0.2412 +0.502) +/-0 U13 = +/-0.606% of AT Span (OT AT) 10.6.14 Determine Up for the module ep:

Where: Bias = +0.30 ep = +/-(1.602 +1.28 +0.312 +0.102 +0.952 +0.502)1 +0.3 UP = +/-1.627%, +0.30%

Converting this uncertainty to process units (2500 - 1700 = 800psi) the following equation is used; UP = (+/-1.627%, +0.30%) *800 psi Up==13.02 psi, +2.4 psi 10.6.15 Determine UPMI The PM, total uncertainty is identified as +/-3.744% of AT Span. It is considered conservative to assume that 20% of the uncertainty is comprised of random accuracy and drift. However, since the incore instrumentation system components are not directly evaluated for uncertainties, we will conservatively include the full PM, value. Therefore:.

UPMI = +/-3.744%

10.6.16 Determine UPMvg The PMw, total uncertainty is identified as +/-8.645% of AT Span. It is considered conservative to assume that 10% of the uncertainty is comprised of random accuracy and drift in excore system components evaluated for uncertainties in this calculation. Therefore, a value of

+/-7.7805% of AT Span (+/-8.645 *90%) will be as UPML'e UPMV, = +/-7.7805% rounded to 7.781% of AT Span 10.7 Determine UTOTAL for OPAT and OTAT In order to solve for total U for both OPAT and OTAT; Tu must be determined. TcOLD will include the cold leg streaming effect of -1.0% of AT span. A THOT streaming PM random effect of +/-1.0 0 F or 1.33% of AT span will be included. UTOTAL is a function of the THOT and TCOLD parameters of AT, the T, UTHOT. UTCOLD and UTAVG equations must be calculated for the U portion of the loop. Therefore, solving for Tu, 10.7.1 Determine Tu Tu = +/- [3(U 12 +U22)]V/3

ENN NUCLEAR QUALITY RELATEDENDC16Rvso3 AM MANAGEMENT ADMINISTRATIVE PROCEDUREENN-DC-126 Revision 3

,,Entergy MANUAL INFORMATIONAL USE Calculation Page 50 of 56 Calculation No. IP3-CALC-RPC-00290 Revision 3 Project: ER No. 04-3-027

Title:

Instrument Loop Accuracy / Setooint Calculation, Overpower Delta-T (OPQT) and Overtemperature Delta-T (OTDT), Reactor TriD Tu= 3(0.268%2 +0. 2 3 6%2)]fl /3 Tu = +/- 0.206%

10.7.2 Solving for UTHOT, UTHOT = +/- (UPM 2 +TU2 + U3 + U42)% +/-Bias UTHOT = +/- (1.33%/62 + 0.206%/62 +0.800%2+ 0.907%2 )* +/- 0.0 of AT span UTHOT = +/- 1.807% of AT span 10.7.3 Solving for UTCOLD, UTCOLD = +/- (Ul2 +U22+ U42)P-Bias UTCOLD = +/- (0.268%2 +0.236%2+ 0 .9 0 7 %2)%-1.0% of AT span UTCOLD = +/-T 1.034% -1.0% of AT span 10.7.4 Solving for UTAVG, (This represents the total UTAVG uncertainty at the input to module e5)

UTAVG = +/- (U2THOT + U2TCOLD)) /2- Bias/2 (Sec. 7.10.3)

UrAVG = +/- [(1.807 + 1 .03 4 2)]" /2- 1.0% /2 UTAVG = a- 1.041 % with -0.5 Bias of TAVG span 10.7.5 Solving for U.T, (This represents the total U,&T uncertainty at the input to module 66)

U'iT = +/- (U2TCOLD + U2THOT) (Sec. 7.10.3)

U,&T = +/- [(1.8072 + 1 .0342)fl] -1.0% of AT span UT = +/- 2.082%, -1.0% of AT span 10.7.6 Calculate total OPAT channel U value (UopjT)

UOPT is determined from the following; UOPAT = +/- (1PM2 +U12 +U122 _U32 +U4 2 + US2 + U62 + PM c +el02 +6122) 1t2 +/- B In the above combination of terms, (UPM 2 +u,2 +U2 2 _u32 +142) represents the total U of both the tAVG and the AT portions of the hot and cold let temperature circuits. Using UTAVG and UAT, UOP,,T becomes; UOPRT = +/- [(UTAVO K) +UKAT +U5 +U6 +PMCal + U,0 + U12 B of (TAVG K6, AT, IRETAvG K6, IREAT)

Where, UTAVG = +/- 1.041 % with -0.5% Bias of TAVG span U&T = +/- 2.082%, -1.0% Bias of AT span U5 = +/- 0.95% AT span U6 = +/- 0.95% AT span PMcaj = +/- 0.907% AT span U10 = +/- 0.0%of AT Span

ENN NUCLEAR QUALITY RELATED Am MANAGEMENT ADMINISTRATIVE PROCEDURE ENN-DC-126 Revision 3 Entergy MANUAL INFORMATIONAL USE Calculation Page 51 of 56 Calculation No. IP3-CALC-RPC-00290 Revision 3 Project: ER No. 04-3-027

Title:

Instrument Loop Accuracy / Setpoint Calculation. Overpower Delta-T (OPDT) and Overtemnerature Delta-T (OTDT), Reactor Trip U12 = +/-0.588% of AT Span (OPAT)

K6 = 0.0015 AT SpanPF TAVG, EASRE = IRETAVG = - 0.477% Bias of AT Span

= IREAT = -0.613% Bias of AT Span UOPT = +/- [(1.041 *0.0015)2 +2.0822 +0.952 +0.952 +o.90712 + 0.02 + 0.5882]1"2 B of (-0.5*0.0015, -1.0, -0.477

  • 0.0015, -0.613)

UOPAT = +/- (0.00156152 +2.0822 +0.952 +0.952 +0.9072 + 0.02 + 0.5882)12 +/- B of (-0.00075, -1.0, 0.0007155, -

0.613)

UOP,,T = +/- 2.703, -1.614% AT Span UOPAT = + 2.703, -4.317% AT Span 10.7.7 Calculate total OTAT channel U value (UOTAT)

UOT,&T is determined from the following; UOTA = +/- (UP% 2 +U12 +U22 +U32 +U42 + U52 + U62 + PMCa12 +UP2 +PM 2 2 2 1 +PM ie +U7 +U8 +U9 2 2

+U11 +U132) 1 t B In the above combination of terms, (UPM2 +U12 +U2 2 _U32 +U42 ) represents the total U of both the TAVG and the AT portions of the hot and cold let temperature circuits. Using UTAVG and UAT, UOPAT becomes; UOTAJ +/- [(UTAVG *K2)2 +UAT 2+U 52 + UU 2 + pPMCal2 +(Up K3) 2 +UPMI 2 +UPMve 2 +U72 +U8 2 +U92 +U11 2

+U13 ] +/- Biases of (TAVG *K2, AT, UP *K3)

Where, UTAVo = +/- 1.041 % with -0.5% Bias of TAVG span UAT = t 2.082%, -1.0% Bias of AT span U5 = +/- 0.95% AT span U6 = +/- 0.95% AT span PM,,, = +/- 0.907% AT span Up = +/- 13.02 psi, +2.4 psi Bias UPMI = +/-3.744% AT span UPMvO = +/-7.781 % AT span U7 = 5.184% AT span U8 =+/-0.138% AT span U9 = +/- 0.035%of AT Span U1, = +/-0.00% of AT Span (OPAT)

U13 = +/- 0.606% AT span K2 = +/- 0.022% AT span/0 F TAVG K3 = +/- 0.0007 AT Span/psi

ENN NUCLEAR QUALITY RELATED A MANAGEMENT ADMINISTRATIVE PROCEDURE ENN-DC-126 Revision 3

- Enfergy MANUAL INFORMATIONAL USE Calculation Page 52 of 56 Calculation No. IP3-CALC-RPC-00290 Revision 3 Project: ER No. 04-3-027

Title:

Instrument Loop Accuracy / Setnoint Calculation. Overpower Delta-T (OPDT) and Overtemperature Delta-T (OT.T). Reactor rip UOT,1T = +/- [(1.041 *0.022)2 +2.0822 + 0.952 + 0.952 + 0.90712 +(13.02 *0.0007)2 +3.7442 +7.7812

+5.1842 +0.1382 +0.035 +0.002 +0.6062112 +/- Biases of (-0.50 *0.022, -1.00, +2.4 *0.0007)

UOTAT = +/-[ 0.0232 +2.0822 + 0.952 + 0.952 + 0.907,2 +.0092 +3.7442 +7.7812 +5.1842 +0.1382

+0.0352 +0.002 +0.6062¶12 +/- Biases of (-0.011, -1.00, +0.0017)

UOTAT = +/- 10.3845, +0.0017, -1.011 % AT Span UOTAT = + 10.3862, -11.396% AT Span 10.8 Determine Nominal AVs The nominal AVs can be calculated from the following equation; AV = AL+/-U 10.8.1 OPAT (K4) Allowable Value For OPAT, the relationship of the Analytical Limit (K4 (MAX)), Allowable Value (K4 (Av)) and Uncertainty (UOPAT) is as follows; K4 (max) ATo - K4(AV) ATO = UOPAT K4 (max) - K4 (AV) = UOPAT AT.

Solving for the Allowable Value (K4sAv))

AV = K4(AV) = K4maX- UoPar AT.

From above, UOPAT=-4.317% of AT Span The above OPAT uncertainty UOPAT is the uncertainty at a condition of measured Full Power AT (equaling 75°F, or 100% of Span). However, lP3's full power AT will be assumed to be 54°F, which is a bounding lowest loop measured AT compared to a AT calibrated Span of 75°F. The following may be determined for; (Ref. 3.5.12)

Uopar (Refs. 3.2.22 & 3.2.23)

AT.

Given, AT. = AT at 100% Full Power AT Span = (75/54) ' 100 AT. = 138.888% of AT, Uophr (4.317%of ATSpan)*(138.888%Full Power)

AT, (100% Full Power) *(100% AT Span)

UoP= - 0.0599 AT, The negative value of E2EaT is used to determine AV since the process is increasing towards the AT, analytical limit. Therefore, calculating the Allowable Value for OPAT; (Ref. 3.1.3)

AV(oPAT) = AL - UoPar AT.

AVcopAn = 1.164 - 0.0599 (Refs. 3.2.13 & 3.2.26)

AV(OPAT) = 1.1041 (OPAT)

ENN NUCLEAR QUALITY RELATED is MANAGEMENT ENN-DC-126 Revision 3 ADMINISTRATIVE PROCEDURE

--- Entergy MANUAL INFORMATIONAL USE Calculation Page 53 of 56 Calculation No. IP3-CALC-RPC-00290 Revision 3 Project: ER No. 04-3-027

Title:

Instrument Loon Accuracy / Setpoint Calculation, Overpower Delta-T (OPDT) and Overtemperature Delta-T (OTDn, Reactor Trip 10.8.2 OTAT (K,) Allowable Value For OTAT, the relationship of Analytical Limit (Kim,), Allowable Value (KlAv) and the uncertainty (UOT&T) are as follows; Kl(.) ATo - KiAv ATo - UOTAT Ki(niax) - KiAV = UOtiar AT.

Solving for the Trip Setpoint (Ker(s)),

TS = K1Av Ki(rriax) - UoTar AT, From above, UOTAT= - 11. 3 9 6 % of A T Span The above OTAT uncertainty Is the uncertainty for at a condition of a measured Full Power AT equaling 75 0F. Similarly to OPAT, OTAT is also based on a Full Power Rating of 138.888% for 540 F, which is the lowest loop measured AT compared to a AT calibrated Span of 75 0F.

The following may be determined for; UOTT (Ref. 3.2.22)

AT.

Given, ATO = AT at 100% Full Power AT Span = (75/54)

  • 100 AT0 = 138.888% of AT, UOTAT = (11.396% of AT Span)*(138.888% Full Power)

AT. (100% Full Power)*(100oAT Span)

UaTaT = - 0.1583 AT.

Therefore, calculating the AV for OTAT; AV = 1.42 - 0.1583 (Refs. 3.2.13 & 3.2.26)

AV = 1.2617 (OTAT)

ATTACHMENT 3 AV ALTERNATIVE EVALUATION Attachment Page 11 of 13 Calculation No. IP3-CALC-RPC-00290 Revision 3 Project: ER No. 04-3-027

Title:

Instrument Loop Accuracy I Setpoint Calculation. Overpower Delta-T (OPDT) and OvertemDerature Delta-T (OTDT).

Reactor Trip UOPAT = + 1.610%, - 3.224% AT Span (The process is increasing toward the AL, therefore, the negative uncertainty value will be subtracted from the AL to determine the OPAT AV)

Determine Allowable Value for OPAT Given: OPAT (K4) AL = 1.164 OPAT UOPAT = + 1.610%, - 3.224% AT Span For OPAT, the relationship of the AL (K4(MAx)), AV (K4(AV)), and the AV/AL Channel Uncertainty (UOPAT) are as follows; K4(MAX) ATO - K4(Av) ATo = UOPAT K4(MAM - K4(AV) = UOPAT/ ATO Solving for the Allowable Value (K4(Av)

AV = K4(AV) = K4(MAX) - (UOPAT/ ATo)

The above AVWAL OPAT uncertainty is the instrumentation uncertainty at a condition of measured Full Power AT (equaling 75°F, or 100% of Span). However, IP3's full power AT will be assumed to be 54°F, which is a bounding lowest loop measured AT compared to a AT calibrated Span of 75 0F. The following may be determined for; UOPAT/ ATO = AT at 100% Full Power AT Span = (75 / 54) X 100 ATo = 138.888% of ATo UOPAT/ ATO = -[(3.224% of AT Span X 138.888% Full Power) / (100% Full Power X 100% AT Span)]

UopaT/ ATo = -0.0448 The negative value of UopA&T/ATO is used to determine AV since the process is increasing towards the AL. Therefore, calculation the AV for OPAT; AVOPAT = AL - (UopAT/ ATo)

AVOPAT - 1.164 - 0.0448 AVOPAT - 1.1192 OPAT NOTE: AL =1.164(OPATK4)

METHOD 3 AV = 1.127 METHOD 2 AV = 1.119 (1.104 using the more conservative Method 2 approach in Section 10.8.1)

TechSpecAV =11.100 TS (calculated) = 1.0807 TS (implementing) = 1.074 Therefore, Method 2, being the more conservative AV determination methodology, will be the basis for establishing the SPU AV for the OPDT function.

ATTACHMENT 3 AV ALTERNATIVE EVALUATION Attachment Page 13 of 13 Calculation No. IP3-CALC-RPC-00290 Revision 3 Project: ER No. 04-3-027

Title:

Instrument Loop Accuracy / Setpoint Calculation. Overpower Delta-T (OPDT) and Overtemperature Delta-T (OTDT).

Reactor Tr2i The following may be determined for; UOTAT/ ATO ATO = AT at 100% Full Power AT Span = (75 / 54) X 100 ATo = 138.888% of ATo UOTaT / ATo -[(5.320% of AT Span X 138.888% Full Power) / (100% Full Power X 100% AT Span)]

UOTAT / ATo = -0.0739 The negative value of UOTAT/ ATO is used to determine AV since the process is increasing towards the AL. Therefore, calculation the AV for OTAT; AVOTAT AL - (UoPAT/ ATo)

AVOTAT 1.420-0.0739 AVOTAT = 1.346 OTAT NOTE: AL = 1.420 (OTAT K1)

METHOD 3 AV =1.348 METHOD 2 AV = 1.346 (1.2617 using the more conservative Method 2 approach in Section 10.8.2)

Tech Spec AV = 1.260 TS (calculated) = 1.241 TS (implementing) = 1.22 Therefore, Method 2, being the more conservative AV determination methodology, will be the basis for establishing the SPU AV for the OTDT function.

ENN NUCLEAR QUALITY RELATEDEN-C16Rvso3 MANAGEMENT ADMINISTRATIVE PROCEDURE ENN-DC-126 Revision 3 MANUAL INFORMATIONAL USE Calculation Page 37of 52 Calculation No. IP3-CALC-ESS-00281 Revision 2 Project: ER No: 04-3-027

Subject:

instrument Loot Accuracv/Setooint Calculation Low Temperature Average (Lo T.) Si and SLI Actuation The random and independent errors include:

ACCICENT NORMAL PM = +/- 1.000 F PMm = +/-1.000 F (sec. 7.1)

PMd= -0.75OF Bias PMd,= -0.75OF Bias (sec. 7.1)

PE = 0.0 PE = 0.0 (sec. 7.2) en = +/- 0.64'F e1 = +/- 0.630 F (sec. 7.3.12) e2 = +/- 1.25'F e2 = +/- 1.08OF (sec. 7.4.11) e3 = +/- 0.612'F e3 = +/- 0.612OF (sec. 7.5.11) e4 = +/- 1.26'F e4 = +/- 1.13 0 F (sec. 7.6.11) e5 = +/- 1.42'F es = +/- 1.24OF (sec. 7.7.11) e 6 =+/- 0.78'F e 8 = +/- 0.700 F (sec. 7.8.11)

The channel uncertainty (CU) is determined by calculating the propagation of the individual error components through the Lo Ti, circuit.

Lo Tavg is determined by the following:

THOT = (TI + T2 + T3) /3, TAVw (THOT + TcOLD) / 2 Where, T1 + T2 + T3 = Hot Leg RTD/Transmitter Output THOT = Hot Leg Average Temperature TcOLD = Cold Teg Temperature TAVg = THOT and TcOLD Average Temperature 7.10.1 Calculate Total THOT Channel Uncertainty (CUTHOT)

To calculate the total THOT channel uncertainty (CUTHOT), first, the average THOT uncertainty (T) must be determined with individual random uncertainties of modules el and e2 propagated through the THOT circuit using the Square Root Sum of Squares (SRSS), as follows:

T = +/-[(e21 + e') + (e 1 +e+ (e21 + e22)]' /3 Where, T = Average Uncertainty of the three Hot Leg measurements.

T = +/-[3(e2 l + e22 )J" /13 T (ACCIDENT) = +/-[3(0.642 + 1.252 )]1/ 3 T (ACCIDENT)= +/- 0.81 1-F T (NORMAL) = +/-[3(0.632 + 1.08 2)]4 / 3 T (NORMAL)= _ 0.7220 F Please note that the Hot and Cold Leg IRE effects are included in the overall CU uncertainty determination of Section 7.10.4.

The total THOT Channel Uncertainty (CUTHOT) is calculated by including PM, e3 and e4 module uncertainties in SRSS as follows:

CUTHOT= +/-(PM 2 +T2 +e3 2 +e4 2) +/-B

ENN NUCLEAR QUALITY RELATED ENN-DC-126 Revision 3 MANAGEMENT ADMINISTRATIVE PROCEDURE MANUAL INFORMATIONAL USE Calculation Page 38of 52 Calculation No. IP3-CALC-ESS-00281 Revision 2 Project: ER No: 04-3-027

Subject:

instrument LooD Accuracv/Setpoint Calculation T Si and SLI Actuation Where bias = 0 ACCIDENT CUTHOT= +/-(1.002 +0.8112 +0.6122 +1.262)15 ACCIDENT CUTHOT= +/-1.90°F NORMAL CUTHOT= +/-(1.002 +0.7222 +0.6122 + 1.132)%

NORMAL CUTHOT= +/-1.780 F NORMAL CUTHOT not including PM = +/-1.474°F 7.10.2 Calculate TcOLD Uncertainty (CUTcOLD)

To calculate CUTCOLD, the random uncertainties of modules el, e2 and e4 are combined using SRSS:

CUTCOLD = +/- (Pm 2 + e21 + e2 2 + e24)A + Bias ACCIDENT CUTCOLD = +/- (0.642 + 1.252 + 1.262)X -0.75 ACCIDENT CUTCOLD = +/- 1.89°F - 0.75°F NORMAL CUTCOLD = +/- (0.632 + 1.082 + 1.132)¶ -0.75 10 NORMAL CUTCOLD = +/- 1.69°F - 0.75°F 7.10.3 Calculate T. Uncertainty (CUTnVg)

To calculate CUTAvG the CUTHOT and CUTCOLD random uncertainties must be propagated through the T. circuit using SRSS as follows:

CUTAVG = +/-(CU 2THOT + CU 2 TCOLD)" / 2 ACCIDENT CUTAvG = [+/-(1.902 + 1.89 2)V-0.75] / 2 ACCIDENT CUTAvY = [+/-2.68°F - 0.75 0F] /2 ACCIDENT CUTAVG = +/-* 1.34°F, - 0.375°F NORMAL CUTAVG = [+/-(1.782 + 1.6929) -0.75] /2 NORMAL CUTAVG = [+/-2.45°F -0.75°F] /2 NORMAL CUTavg = +/- 1.23°F -0.375°F

- ENN _

NUCLEAR QUALITY RELATED ENN-DC-126 Revision 3 MANAGEMENT ADMINISTRATIVE PROCEDURE END- 6Rvso MANUAL INFORMATIONAL USE Calculation Page 39of 52 Calculation No. IP3-CALC-ESS-00281 Revision 2 Project ER No: 04-3-027

Subject:

instrument Loop AccuracW/Setooint Calculation Low Temperature Average (Lo T . Si and SLI Actuation 7.10.4 Calculate the Total Channel Uncertainty (CU)

To calculate Total CU, modules es, e6 and IRE (IRE for ACCIDENT only) are combined with CUT.,g using SRSS.

CUTOTAL = +/- (CU2 TAVG + e25 + e 6) +/-B ACCIDENT CUTOTAL = +/- (1.342 + 1.422 + 0.782)%4+0,1.2,-0.375 ACCIDENT CUTOTAL = +/-2.100F +0, -1.575 ACCIDENT CUTOTAL = +2.10 0F, -3.675°F NORMAL CUTOTAL = +/- (1.232 + 1.242 + 0.702)' -0.375 NORMAL CUTOTAL = +/-1.88°F, -0.375°F NORMAL CUTOTAL = +1.88 0F, -2.25°F

ENN NUCLEAR QUALITY RELATEDENN-DC-126 Revision 3 MANAGEMENT ADMINISTRATIVE PROCEDURE CNlculto Revison 5 MANUAL INFORMATIONAL USE Calculation No. IP3-CALC-ESS00281 Revision 2 Project ER No: 04-3-027

Subject:

instrument Loot AccuracvlSetpoint Calculation Low Temperature Average (Lo T Si and Si Actuation 8.0 OBTAIN ANALYTICAL LIMIT (AL) 8.1 The modeled/credited value used in the Safety Analyses for Steamline Break Lo Tas9 coincidence is5350F.

Therefore, LO Tvg AL = 5350F (see Attachment 4) 8.2 The alarm limit used as an NPL for the Hi T,,9 setpoint for operator convenience is the COLR DNB limit for Tavg of 574.80F (Ref 3.2.24). This is not a modeled or credited function in the Unit 3 safety analyses. However, Westinghouse correspondence supporting critical parameter values for power uprate identifies a value or 572°F Therefore, Hi T. , NPL = 574.80F (see Attachment 6)

(

ENNI

___ NUCLEAR OUALrY RELATED ENN-DC-126 Revision 3 En MANAGEMENT ADMINISTRATIVE PROCEDURE Energy MANUAL INFORMATONAL USE Calculation Page 41of 52 Calculation No. IP3-CALC-ESS-00281 Revision 2 Project ER No: 04-3-027

Subject:

Instrument Loon AccuracvlSetpoint Calculation Low Temperature Averace (Lo T.,<): Si and SLI Actuation 9.0 DETERMINE SETPOINTS(TS)AND RTD CONVERTER CALIBRATIONS The nominal trip setpoint can be calculated from the following equation:

TS = AL + (CU + Margin) (ref. 3.1.2)

If Margin = 0.0 Channel Uncertainty is:

ACCIDENT CU = +2.10 0F, -3.6750 F (Section 7.10.4)

NORMAL CU = +1.88 0 F, -2.250 F 9.1 LO Ta,, TS calculation:

The negative value of CU is not used to determine TS since the process is decreasing towards the analytical limit. Accident conditions are possible during a LO Tavg event. Therefore, TS = 535'F +2.100 F TS= 537.1 00 F (DEC),or when scaled for 400mV and 750 F span, the decreasing mV signal is [400 mV(537.10-540) /75]+100 = 84.5mV(DEC)which is below scale. NOTE: The existing setpoint is conservatively set at 5420 F or 110.67 mV decreasing (Ref. 3.5.7).

9.2 Hi Tag TS calculation:

The positive value of CU is not used to determine HI TagTS since the process increases towards the alarm limit. Also, harsh environment accident conditions are not considered present during a Hi Tavg event. Therefore:

TS = 574.80F - 2.250 F TS = 572.550 F (INC), or when scaled for a 400mV and 750 F span the increasing mV signal is

[400 mV(572.55-540) / 75]+100 = 273.6 mV (INC). NOTE: The existing setpoint is conservatively set at 569.890 F or 259.46 mV Increasing. (Ref. 3.5.7).

The above setpoint at 569.89 'F adequately supports the current operating Full Load Tavg of 5670 Fas well as the Hi Tavg alarm NPL of 574.8 0F. However, it may be necessary to increase Full Load Tavg as much as 3 degs to achieve adequate Main Stem Pressure for acceptable turbine first stage performance. Potential changes will be considered in 1 deg increments, specifically at 568, 569 and 570 "F. We will therefore configure the following setpoint changes if and at what value the full load condition is actually changed to:

FL Tavq (OF) Hi Tava SetDoint (OF) Hi Tavq SetDoint (mV) 568 570.90 264.80 569 571.90 270.13 570 572.55 273.60

ATTACHMENT 8 TO NL-05-014 Indian Point Piping Vibration (PV) Plan Logic Diagrams with Examples (3 Pages)

ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

Indian Point Piping Vibration (PV) Plan Logic fionm Develop l Recommend IDPipingI cenn Insa Wll lFDenet Evaluatel iew for PV I orPV _I

. prknI w DPiping/ Perform ID Pipin or E~xec Supports Analysis _ Support ModL to be It Uprae Required for Uprde /

t ~~~Conditlon Ys/eid R&fl~f GUbiMtUM rObe perrrmsed in coordiuaion with the testing program put in place

.-. r-

KJ Indian Point Unit 3 \J P2 Piping Vibration Collection Data Baseline Power Power Power Prior to Ascension Ascension Uprate +7 Uprate to Uprate to Uprate Days (100%

(100% RX (96.5% RX (100% RX RX Pwr)

Test Attribute Pwr) Pwr) Pwr) r Monitor Point ID Ef E 1 . '

Pipe Location Description P&ID or Piping Drawing F02IO7i(Maii Sam)

Nominal Pipe Diameter (in) 1'1 Photo Number(s) 12 Test Date / Time Percent Reactor Power Velocity (in/sec) .

X-Axis Dispi (mils)

- Freq (Hz)

Test Data (Max Velocity (in/sec)

Value of Velocity s and, Displacement a Y Displ

_Axis (mils)

Given Frequency) Freq (Hz)

Velocity (in/sec)

. In-Line

.Z-Axis Displ (mils)

Freq (Hz)

Data Collected By: = _

Remarks:

IP3 Piping Vibration Collection Data Sheets - Revised.)ds 20tf28 1/31/2005

K> Indian Porint Unit 3 <_J Piping Vibration Collection Data P3 Baseline Power Power I Power Prior to Ascension Ascension Uprate +7 Uprate to Uprate_ to Uprate Days ( 00%

(100% RX (96.5% RX (100% RX RX Pwr)

Test Attribute Pwr) Pwr) - Pwr) P Monitor PointID ID ;2 .

Pipe Location Description P&ID or Piping Drawing 2017 Nominal Pipe Diameter (in) 314" Photo Number(s) 13 Test Date / Time Percent Reactor Power Velocity (in/sec)

X-Axis Dispi (mils)

Freq (Hz)

Test Data (Max Velocity (in/sec)

Value of Velocity . (

and Displacement a Y-Axis Displ (mils)

Given Frequency) Freq (Hz)

.In-Line Velocity (in/sec)

Z-Axis Displ (mils) _

Freq (Hz)

Data Collected By.

Remarks:

IP3 Piping Vibration Collection Data Sheets - Revised.xls 3 of 28 1131/2005

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