ML18215A177

Responses to the Request for Additional Information Regarding License Amendment Request to Update the Reactor Coolant System Pressure-Temperature Limits

ML18215A177
Person / Time
Site:	Arkansas Nuclear
Issue date:	08/01/2018
From:	Anderson R L Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2CAN081802
Download: ML18215A177 (18)

Text

• ~Entergy Entergy Operations , Inc. 1 448 S.R. 333 Russellville , AR 7 2 80 2 2CAN081802 August 1 , 2018 Tel 4 7 9-85 8-3110 Richard L. Anderson A NO S~e Vice Pres i dent 10 CFR 50.90 A TIN: Document Control Desk U.S. Nuclear Regulatory Commission Washington , DC 20555 SUBJEC T:

REFERENCE:

Responses to the Request for Additional Information Regarding License Amendment Request to Update The Reactor Coolant System Pressure-Temperature Limits Arkansas Nuclear One, Unit 2 Docket No. 50-368 License No. NPF-6 1 Entergy subm i ttal to the NRC , " License Amendment Request Updating the Reactor Coolant System Pressure-Temperature Limits," dated November 20 , 2017 (2CAN 111702) (ML 17326A387) 2 NRC to Entergy , " Request for Additional Information Regarding License Amendment Request to Update the Reactor Coolant System Temperature Limits ," dated May 10, 2018 (2CNA051801) (ML 18129A425)

Dear Sir or Madam:

Entergy Operations , Inc. (Entergy) submitted a license amendment request to rev i se the Arkansas Nuclear One , Un i t 2 Technical Specifications (TSs) via Reference

1. The proposed changes would replace the current Pressure-Temperature limits , applicable to 32 Effective Full Power Years (EFPY), with new Pressure-Temperature limits applicable to 54 EFPY (approximately 60 calendar years). The Nuclear Regulatory Commission has reviewed the subject request and determined additional information is needed to complete review of the application. The request for additional information was issued to AN0-2 in Reference

2. The purpose of this submittal is to provide the information that was requested. Based on a review of the material provided in this submittal and in Reference 1 , Entergy has concluded the no significant hazards determination provided in Reference 1 , does not need to be revised. In addition , this submittal contains no additional regulatory commitments.

2CAN081802 Page 2 of 2 If there are any questions or if additional information is needed , please contact Stephenie Pyle , Manager, Regulatory Assurance at (479) 858-4704. I declare under penalty of perjury that the foregoing is true and correct. Executed on August 1, 2018. RLA/rwc

Enclosure:

Response to Request for Additional Information

-Reactor Coolant System Pressure-Temperature Limits cc: Mr. K riss M. Kennedy Regional Administrator U.S. Nuclear Regulatory Commission RGN-IV 1600 East Lamar Boulevard Arlington , TX 76011-4511 NRC Senior Resident Inspector Arkansas Nuclear One P. 0. Box 310 London , AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Thomas Wengert MS 0-0881 One White Flint North 11555 Rockville Pike Rockville , MD 20852 Mr. Bernard R. Bevill Arkansas Department of Health Radiation Control Section 4815 West Markham Street Slot #30 Little Rock , AR 72205 Enclosure to 2CAN081802 Response to Request for Additional Information Reactor Coolant System Pressure-Temperature Limits Enclosure to 2CAN081802 Page 1 of 14 Response to Request for Additional Information Reactor Coolant System Pressure-Temperature Limits By application dated November 20, 2017, Entergy Operations, Inc. (Entergy), submitted a license amendment request (LAR) to revise the Arkansas Nuclear One, Unit 2 (AN0-2) Technical Specifications (TSs) by replacing the current Reactor Coolant System (RCS) Pressure-Temperature (P-T) Limits, applicable to 32 Effective Full Power Years (EFPY), with new P-T limits applicable to 54 EFPY (approximately 60 calendar years). The U.S. Nuclear Regulatory Commission (NRC) staff has determined that the following additional information is needed in order to complete its review. RAI MVIB-1 Request: Discounting the shift in RT N OT due to irradiation if the predicted shift is less than 25 °F does not meet the NRC regulation in Title of the Code of Federal Regulations (10 CFR) Part 50 , Appendix G , " Fracture Toughness Requirements." Appendix G IV.A to 10 CFR Part 50 states t hat for reactor vessel beltline materials , including welds, plates, and forgings, the values of RT N OT mu s t account for the effects of neutron radiation.

Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials ," Revision 2 (Reference 4 ), details how ~RT NO T is calculated. The NRC staff notes that Regulatory Issue Summary 2014-11, "Information on Licensing App li cations for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components" (Reference 5), clarifies that 10 CFR Part 50 , Appendix G and 1 0 CFR P a rt 50 , Appendix H , "Reactor Vessel Material Surveillance Program Requirements

," define the beltline as including all RPV materials that will receive a neutron fluence greater than or equal t o 1x10 17 n/cm 2 (E> 1 MeV). The NRC s taff also notes that TLR-RES/DE/CIB-2013-01 is not NRC guidance , and the recommendat i on that the shift due to irradiation can be discounted if it is less than 25 °F is not endorsed i n any NRC guidance document or permitted by the regulations.

Rev i se th e P-T limit evaluation to include RT NOT values for all RPV beltline and extended beltline materials c alculated in accordance with 10 CFR Part 50 , Append i x G, so the NRC staff can verify that the calculations were performed in accordance with the regulations. Entergy's r esponse: T he P-T limit evaluation has been revised as documented in WCAP-18169-NP , Revision 1 (attached). Tables 7-2 and 7-3 of the revision calculate the adjusted reference temperature (ART) values fo r each of the reactor pressure vessel (RPV) beltline and extended beltline materials

[i.e., reactor vessel materials with fluence values exceed i ng 1 x 10 17 n/cm 2 (E > 1.0 MeV)], in accordance with 1 O CFR 50, Appendix G , without the use of the TLR-RES/DE/CIB

-2013-01 (Reference

1) methodology.

Thus, calculated values of the change i n the Reference Temperature for Nil Ductility Transition (6RT N oT) less than or equal to 25 °F are not se t equal to zero. It is noted that the limiting ART values presented in Table 7-4 and the P-T limit c u rves presented in Section 8 remain unchanged in this revision , as the TLR-RES/DE/CIB-2013-01 methodology was not utilized for the limiting materials.

Enclosure to 2CAN081802 Page 2 of 14 RAI MVIB-2 Request: In Section 5 of Attachment 3 of the application, the licensee provides the evaluation of chemistry factors. However, this section does not include the evaluation of USE. Provide the evaluation and the results for USE, based on 54 EFPY fluence values that are discussed in the amendment request. Entergy's response:

There are two methods that can be used to predict the decrease in the Upper Shelf Energy (USE) wit h irradiation, depending on the availability of credible surveillance capsule data as defined in Regulatory Guide (RG) 1.99 , Revision 2 (Reference 2). For vessel beltline materials that are not in the surveillance program or have non-credible data, the Charpy USE (Position 1.2) i s assumed to decrease as a function of fluence and copper content, as indicated i n Reference

2. When two or more credible surveillance sets become available from the reactor, they may be used to determine the Charpy USE of the surveillance material.

The surveillance data are then used in conjunction with Reference 2 to predict the change in USE (Position 2.2) of the vessel material due to irradiation.

The 54 EFPY Position 1.2 USE values of the vessel materials can be predicted using the corresponding 1/4T fluence projection, the copper content of the materials, and Figure 2 in Reference

2. The predicted Position 2.2 USE values are determined for the reactor vessel materials that are contained in the surveillance program by using the reduced plant surveillance data along with the corresponding 1/4T fluence projection. The reduced plant surveillance data was obtained from Table 5-10 of the latest AN0-2 surveillance capsule analysis report (Reference 3). The surveillance data was plotted in Reference

2. This data was fitted by drawing a line parallel to the existing lines as the upper bound of all the surveillance data. These reduced lines were used instead of the existing line to determine the Position 2.2 USE values. The projected USE values were calculated to determine if the AN0-2 beltline and extended beltline materials remain above the 50 ft-lb criterion at 54 EFPY. These calculations are summarized in the table below. The limiting USE value at 54 EFPY is 60.2 ft-lb; this corresponds to the Upper Shell Plate C-8008-2.

All the beltline and extended beltline materials in the AN0-2 reactor vessel are projected t o remain above the USE screening criterion value of 50 ft-lb.

Enclosure to 2CAN081802 Page 3 of 14 Table 1 Predicted USE Values at 54 EFPY Reactor Vessel Material and Wt% 1 / 4T Fluence(bl Initial Identification Number cu(a) (n/cm2, E >1 MeV) USE (ft-lb) Position 1.2(c) Intermediate Shell Plate C-8009-1 0.098 3.06 X 10 19 95 Intermediate Shell Plate C-8009-2 0.085 3.06 X 10 1 9 93 Intermediate Shell Plate C-8009-3 0.096 3.06 X 10 19 134 Lower Shell Plate C-8010-1 0.085 3.10 X 10 19 89 Lower Shell Plate C-8010-2 0.083 3.10 X 10 1 9 94 Lower Shell Plate C-8010-3 0.080 3.10 X 10 19 97 Intermed iate Shell Heat# Longitudinal Welds Multiple 0.05 2.89 X 10 19 110 2-203-A, B , & C Lower Shell Heat# Longitudinal Welds 0.046 2.94 X 10 1 9 125 3-203-A, B, & C 10120 Intermediate to Heat# Lower Shell Girth 0.045 3.05 X 10 19 136 Weld 9-203 83650 Upper Shell Plate C-8008-1 0.13 0.0367 X 10 19 93 Upper Shell Plate C-8008-2 0.13 0.0367 X 10 19 68 Upper Shell Plate C-8008-3 0.08 0.0367 X 10 19 71 Upper Shell Heat# Longitudina l Welds BOLA 0.02 0.0367 X 10 19 92 1-203A, B , & C Heat# 0.22 0.0367 X 10 19 105 10137 Upper to Heat# Intermediate Shell 0.21 0.0367 X 10 19 107 Girth Weld 8-203 6329637 Heat# 0.03 0.0367 X 10 19 118 FAGA Position 2.2(dl Intermediate Shell Plate C-8009-3 0.096 3.06 X 10 19 134 Upper S h ell Plate C-8008-1 0.13 0.0367 X 10 1 9 93 Intermediate to Heat# Lower She ll Girth 0.045 3.05 X 10 19 136 Weld 9-203 83650 Projected Projected USE USE Decrease (%) (ft-lb) 25.0 71.3 25.0 69.8 25.0 100.5 25.0 66.8 25.0 70.5 25.0 72.8 24.5 83.1 24.5 94.4 25.0 102.0 11.5 82.3 11.5 60.2 8.8 64.8 8.8 83.9 18.5 85.6 18.5 87.2 8.8 107.6 34.0 88.4 12.0 81.8 18.5 110.8 Enclosure to 2CAN081802 Page 4 of 14 Notes: (a) Data taken from Table 3-1 of WCAP-18169-NP, Revision 1 (attached). (b) The 1/4T fluence was calculated using the Reference 2 correlation, and the AN0-2 vessel beltl i ne thickness of 7.875 inches. (c) Percentage USE decrease values are based on Position 1.2 of Reference 2 and were calculated by plotting the 1/4T fluence values on Figure 2 of Reference 2 and using the material-specific copper weight percentage values. In calculating Position 1.2 percent USE decreases, the base metal and weld copper weight percentages were conservatively rounded up to the nearest line in Figure 2 of the Reference

2. (d) Percentage USE decrease is based on Position 2.2 of Reference 2 using data from Table 5-1 O of Reference

3. Reference 2 , Position 2.2 indicates that an upper-bound line drawn parallel to the existing lines in Figure 2 of Reference 2 through the surveillance data poin t s should be used in preference to the existing graph lines for determining the decrease in USE. RAI SRXB-1 Request: Provide the following information, considering the new system backpressure and HPSI flow rates, used to support the proposed P-T limits in TS 3.4.9 and P-T limits related TS 3.4.12: a) Provide the values of the system backpressure and HPSI rates used in the L TOP reanalysis and the analysis of record (AOR), and describe the causes (design change or changes in assumptions and methods used in the analysis) that result in a higher system backpressure and higher HPSI flow rates. b) Discuss and justify any changes to the methodologies used in the L TOP reanalysis. c) Discuss the results of the L TOP reanalysis for the most limiting energy addition event and mass addition event, and provide additional information to demonstrate that the RCS pressures are within the proposed P-T limits, and the required relief valve setpoint in TS LCO 3.4.12 remains valid. Provide justification for cases where the reanalyzed limiting events are different from those identified in the AOR. d) Discuss the reevaluation of the L TOP enable temperature and justify that the required enable temperature i n TS LCO 3.4.12 remains acceptable for the proposed P-T limits in in TS 3.4.9. Entergy's response -Part a): Relief Valve Backpressure versus Relief Valve Flow The relief valve flow versus relief valve system backpressure values used in the low temperature overpressure protection (L TOP) AOR and L TOP reanalysis are provided in Table 2 and Table 3. The values in Table 2 and Table 3 were obtained from RELAP models of the piping con fi gurations assuming a starting pressurizer condition with a saturated liquid volume of Enclosure to 2CAN081802 Page 5 of 14 91 O ft 3 and a saturated vapor volume of 323 ft 3 consistent with the TS 3.4.12 requirement and for two different cases of init i al pressur i zer saturated liquid starting conditions (i.e., 417.4 °F @ 300 psia and 444.6 °F@ 400 psia as shown in Tab l e 2 and Table 3 , respectively).

Table 2 -Relief Valve Flow versus Relief Valve System Backpressure when Pressurizer is Initialized at 417.4 °F @300 psia LTOPAOR L TOP Reanalysis Values Flow (gpm) Backpressure Flow (gpm) Backpressure (psia) (psia) 898.58 154 ----899.2 143.6 --999.1 155.4 1 , 197.88 180 ----1,198.90 177 --1 , 398.80 196.2 1,497.52 204 ----1,598.60 213.3 --1 , 698.30 221.2 --1 , 798.00 229.1 1959.4 6 236 ----1 , 997.90 243.2 2296.27 256 --Table 3 -Relief Valve Flow versus Relief Valve System Backpressure when Pressurizer is Initialized at 444.6°F @ 400 psia LTOPAOR L TOP Reanalysis Values Flow (gpm) Backpressure Flow (gpm) Backpressure (psia) (psia) --768.7 139.6 --899.6 158.6 901.7 3 169 ----999.6 172.3 --1 , 199.50 197.9 1 , 202.1 1 20 1 ----1 , 399.40 221.3 --1,499.30 232.3 --1 , 599.00 242.7 1 , 602.78 24 1 ----1,798.20 262.2 Enc lo s u r e to 2 CA N 0 8 1 802 Page 6 o f 14 Review of Table 2 and T a ble 3 values indicate that the system backpressure versus flow rate for liquid dis c harge in the L TOP reanalysis and the AOR are similar despite the fact that the piping drawings are not the same as a result of the piping modifications made during the replacement steam generator (RSG) project. Per relief valve vendor recommendation , in order to ensure stable valve operation, the relief valve ba c kpressure shall be limited to 50% of the relief valve set pressure.

Per TS 3.4.12 , the set press u re is 430 psig; thus a relief valve system backpressure limit of 229.7 psia is imposed. For the AOR, a maximum relief valve flow limit of 1869 gpm and a maximum relief valve flow limit of 1490 gpm were determined for a pressurizer saturated liquid starting condition of 417.4 °F a t 300 psia and 444.6 °F at 400 psia, respectively.

For the reanalysis, a maximum relief valve flow limit of 1723.5 gpm and a maximum relief valve flow limit of 1416.5 gpm were determin e d for a pressurizer saturated liquid starting condition of 417.4 °Fat 300 psia and 444.6 °F a t 400 psia , respectively. Therefore, the maximum flow permitted to assure stable valve operation is lower in the current reanalysis.

The back p ressure versus relief flow values presented in Table 2 and Table 3 are plotted in Figure 1. Figure 1 shows that the system backpressure for liquid discharge is not all that different between the AOR and reanalysis. Use of system backpressure in the AOR and the reanalysi s is further discussed in the response to Item b ). Figure 1 Backpressure vs. Relief Flow for AOR and Reanalysis 280 2 60 24 0 220 160 140 1 2 0 100 600.00 80 0.00 1,00 0.00 1,200.00 1,400.00 1 , 600.00 1 , 800.00 2 , 000.00 2 , 200.00 2,4 00.00 Relief Valve Flow (gpm) ---AOR B ack p ressure@ 417.4'F -*-AOR B ack pr essu r e@ 444.6'F -+-R eanaly s is Ba ck pr ess ur e@ 417.4'F R ea n a ly sis Bac k p ress ur e@ 444.6'F Enclosure to 2CAN081802 Page 7 of 14 High Pressure Safety Injection Pump Flow The High Pressure Safety Injection (HPSI) Pump flow rates used in the L TOP AOR and the L TOP reanalysis are presented in Table 4 and plotted in Figure 2. HPSI flow is increased for the reanalysis due to modifications made to the HPSI pump impellers.

Table 4 -Reanalysis One HPSI Pump Flow Rate AOR Reanalysis RCS Flow Rate RCS Pressure Flow Rate Pressure (gpm) (psig) (gpm) (psig) --0 924.0 0 856 ----100 892.5 --200 858.9 200 789 ----300 825.3 --400 789.6 400 716 ----500 751.8 --600 711.9 600 637 ----700 671.0 --800 626.9 --900 578.6 --1000 527.1 --1100 470.4 --1200 405.3 --1300 328.7 --1400 228.9 --1502 0.3 Enclosure to 2 CAN081 8 0 2 Page 8 o f 1 4 Figure 2 RCS Pressure Versus HPSI Flow (One Pump) 1600 1400 1200 -;;;; ! 1000 QI 800 0. 600 V) u a: 400 200 0 0 100 200 300 400 500 600 700 800 HPIS Pump Flow (gpm) ~AOR -4t-Reana l y sis Entergy's response -Part b): The L TO P Transient AOR 900 1000 The L TOP transient AOR was performed as two parallel efforts. The first effort analyzed the limiting mass addition transient and limiting energy addition transients using bounding design basis input (aside from the L TOP relief valve backpressure for liquid discharge as further discussed below). One assumption was that a maximum of two HPSI pumps would be aligned during L TOP operation.

A second assumption was that the RCS would be water solid at the time of th e limiting transient.

For these transient analysis cases , a nominal L TOP relief valve backpres s ure for liquid discharge of 100 psig was used to determine valve capacity.

The peak pressure at the pressurizer location for the limiting mass addition event was 522.2 psia and for the limiting energy addition event was 539.0 psia. In a paral l el effort , a RELAP5 analysis of the piping network downstream of the L TOP relief valve wa s performed to determine L TOP relief valve backpressure for liquid discharge as a function o f relief valve flow. The focus of this work was to determine the relief valve flow at which the relief valve backpressure for liquid discharge would reach 50% of the set pressure. The potential for unstable valve cycling behavior will be averted by maintaining L TOP re l ief valve backpressure less than 50% of the set pressure.

Comparison of the peak L TOP relief flows for the limiting mass addition transient and limiting energy addition transient to the RELAP5 analysis of relief valve backpressure for liquid discharge as a function of relief valve flow determined that to assure stable L TOP relief valve behavior, the following operating restrictions were imposed via incorporation into the plant TSs: 1. A m aximum of one HPSI pump should be aligned during L TOP operation, and 2. A m aximum nominal pressurizer level of 910 ft 3 should be maintained prior to starting of the first Reactor Coolant Pump (RCP) during L TOP operation.

Enclosure to 2CAN081802 Page 9 of 14 Given these two restrictions on L TOP operation, the peak pressure results of the preliminary mass addition transient and limiting energy addition transients were reported to bound the peak pressure results with L TOP operating restrictions of only one HPSI pump aligned and a maximum nominal pressurizer level of 910 ft3. A peak pressure at the pressurizer location for the limiting mass addition event of 522.2 psia and for the limiting energy addition event of 539.0 psia were reported i n the AN0-2 Safety Analysis Report (SAR) Section 5.2.2.4. The Current L TOP Transient Reanalysis The current mass and energy addition events used explicit L TOP relief valve upstream and downstream pressure values as a function of relief valve flow based on RELAP5. The reanalysis incorporates all of the operating restrictions and all of the latest design inputs of the previous analyses. Reanalys i s of upstream and downstream pressure values was necessary due to changes in the i sometric drawings for the downstream piping. The current RELAP5 analysis modelled the pressurizer and compressed the steam space producing

'wet' steam at a pressure above the saturation pressure.

Pressurizer insurge rates were varied to determine the upstream and downstream pressure versus flow. As in the AOR , pressurizer steam and water temperatures corresponding to saturated conditions at 300 psia and 400 psia were considered.

Subsequent system transient analyses determined that system backpressure would be lower for steam releases t han for liquid releases.

The current mass addition analysis determined the equilibrium relief valve discharge flow rate which is equal to the charging flow rate and HPSI flow rate with allowances added for fluid expansion due to heat sources including:

decay heat, heat from two RCPs, and pressurizer heaters working at their maximum heat rate. This method is the same as the L TOP mass addition AOR. The relief valve inlet piping pressure drop was calculated based on this equilibrium flow rate. Th i s inlet piping pressure drop was added to a pressure equal to 110% of the relief valve setpoint to obtain the peak transient pressure at the pressurizer location.

The peak pressure at the pressurizer location for the limiting mass addition event was 498.0 psia. The relief valve flow remained below the flow at which the relief valve backpressure would reach 50% of the set pressure. The reduction in peak pressure compared to the analysis of record peak pressure is mostly due to reducing the number of HPSI pumps aligned during L TOP operation. The flow for the single HPSI pump used in the analysis has increased compared to t he analysis of record to account for modifications made to the HPSI pump impellers.

The most limiting energy addition event is a single idle RCP start with a maximum nominal pressurizer level of 910 ft 3 and with a secondary-to-primary temperature differential of 100 °F. The current energy addition analysis used the CENTS code (see next paragraph for additional details regarding the CENTS code) to evaluate the system transient.

The CENTS code was used to model the liquid and vapor phases in the pressurizer in lieu of the OVERP code because the OVERP code only models a water solid RCS/Pressurizer.

Input to the CENTS code incl u ded explicit L TOP relief valve upstream pressure characteristics calculated using RELAP5 with a conservative calculated inlet piping pressure drop. The peak pressurizer pressure occurred during the liquid relief portion of the transient (as opposed to the steam relief portion o f the transient). The maximum i nlet piping pressure drop was added to a pressure equal to 110% of the relief valve setpoint to obtain the peak transient pressure at the pressur i zer location (as in the L TOP energy addition analysis of record). The peak pressure at the pressurizer location for the limiting energy addition event was 497.5 psia.

Enclosure to 2CAN081802 Page 10 of 14 Explicit modelling of the pressurizer steam space was not possible with the analysis method of record (OVERP code). The OVERP code was expressly developed to treat water solid systems. The CENTS code evaluated the expansion of the primary system liquid volume including heat sources such as nuclear fuel decay heat , pressurizer heater heat input, RCP heat, and heat from the secondary system which is transferred to the primary system. CENTS modelled the increase in pressurizer pressure due to pressurizer insurge flow , the dynamic opening a nd closing of the (single credited)

L TOP relief valve, the relief valve steam flow , and the event u al relief valve liquid flow as the primary system became water solid. Entergy's response -Part c): Comparison of the Current L TOP Reanalysis to the AOR The limiting cases for the energy addition event and mass addition event AORs were re-evaluated using the same design inputs except as discussed below. In each case, it was determined that if these limiting cases were initiated at a pressurizer pressure of 400 psia versus 300 psia , the L TOP relief valve would need to open more fully because valve capacity is adversely affected by increased initial saturation temperature in the presence of a pressurizer steam bu b ble. Also , relief valve backpressure is adversely affected by increased initial saturation temperature in the presence of a pressurizer steam bubble as is shown in Figure 1. The margin between the peak flow for the transient and the maximum flow which assures valve stability i s reduced at an initial pressurizer pressure of 400 psia versus 300 psia. From the RELAP5 reanalysis

The analysis value for predicted rated steam and water flows are as follows at 10% above the set pressure (487. 7 psia). Rat e d Steam Flow For transient initiated with 417.4 °For with 444.6 °F saturated steam flow is greater than 17 , 000 gpm Rat e d Water Flows For transient initiated with 417.4 °F saturated water = 1 , 856 gpm For transient initiated with 444.6 °F saturated water = 1,600 gpm The maximum flows for steam are not a concern with regard to valve backpressure for L TOP transients because the imposed backpressure at analyzed steam flow conditions remained below 25 psia which is significantly less than the backpressure imposed by the maximum liquid flow. However , the maximum relief valve water flows which assure valve stability are as follows. For transient initiated with 417.4 °F saturated water= 1,723.5 gpm For transient initiated with 444.6 °F saturated water = 1,416.5 gpm Enclosure to 2CAN081802 Page 11 of 14 Justifica t ion for Cases Where the Reanalyzed Limiting Events Are Different From Those Identified In the AOR The RELAP5 reanalysis determined that higher initial pressurizer steam pressure is a more limiting condition.

The mass addition and energy addition L TOP transients were both shown to be acceptable even if initiated at the higher initial condition of 400 psia. The valve capacity and the opening setpoint (430 psig) of the L TOP relief valve were demonstrated to be adequate by the current RELAP5 reanalysis. The peak pressure consequences for the most limiting energy addition event and the most limiting mass addition event are less severe than those currently quoted in the SAR. Choice of RCS Temperature for Limiting Transient Pressure Results The limiti n g energy addition event (RCP start event) is modelled using the CENTS code. It is assumed that the primary system T co ld is initially at the L TOP enable temperature of 220 °F and that the steam generators a r e filled with water at the limiting temperature of 100 °F above the primary system T c old* The shutdown cooling (SOC) system is isolated as are charging and letdown. This hist o rical worst case approximates a cooldown accomplished by steaming the nuclear steam supply system down to the SOC window. The goal is to be in the SOC window (300 °F/300 p s ia) in three hours. Initializing the event at this primary temperature maximizes the decay heat contribution to the reactor coolant expansion.

Steaming is secured and SOC reduces t he primary coolant temperature to a condition 100 °F cooler than the secondary side water inv e ntory. At that point, SOC is isolated.

The LTO P mass addition event (one HPSI pump and three charging pumps) is also assumed to occur wit h the primary system T c old initially at the L TOP enable temperature of 220 °F in order to maximiz e decay heat. The pea k transient pressures are reduced at lower RCS temperatures.

Comparison of the Current L TOP Transient Peak Pressure to the P-T Limits A boundi n g value for the pressure drop in the pressurizer surge line (i.e., flow induced pressure drop due to the insurge flow into the pressurizer from the RCS as a result of the mass or ene r gy addition and to replace inventory lost through the relief valve discharge) of 2.2 psi is added to the L TOP transient peak pressure of 498 psia prior to compar i son of the peak pressure resul t to the actual allowable pressurizer pressure versus RCS temperature values. The most limiting P-T point shown on the LAR cooldown curve TS Figure 3.4-2B is 543 psia at 60 °F. The most lim i ting P-T point is shown on the LAR heatup curve of TS Figure 3.4-2A is 588 psia at 60 °F. En t erqy's response -Part d}: The L TO P enable temperature in TS Limiting Condition for Operation (LCO) 3.4.12 was reevalua t ed per ASME Code Case N-641 (Reference

4) based on the limiting 1/4T ART in WCAP-1 8 169-NP, Revision 1, Table 7-4. The P-T limits in TS 3.4.9 are also based on the l i miting ART values in WCAP-18169-NP , Revision 1, Table 7-4.

I I L Enclosure to 2CAN08 1 802 Page 12 of 14 ASME Code Case N-641 presents alternative procedures for calculating P-T relationships and L TOP system effective temperatures, Te, and allowable pressures.

The procedures provided in ASME Code Case N-641 take into account alternative fracture toughness properties, circumferential and axial reference flaws, and plant-specific L TOP effective temperature calculations.

Per ASME Code Case N-641, the L TOP system shall be effective below the higher temperature determined in accordance with (1) and (2) below. Alternatively, L TOP systems shall be effective below the higher temperature determined in accordance with (1) and (3) below. (1) A coolant temperature(aJ of 200 °F (2) A coolant temperature(aJ corresponding to a reactor vessel metal temperature(b), for all vessel beltline materials, where Te is defined for inside axial surface flaws as RT NOT + 40 °F, and Te is defined for inside circumferential surface flaws as RT NOT -85 °F. (3) A coolant temperature(a) corresponding to a reactor vessel metal temperature(bl, for all ve s sel beltline materials, where Te is calculated on a plant-specific basis for axial and cir c umferential reference flaws using the following equation:

Te= RTNoT + 50 In [((F

• Mm (pRi / t))-33.2) / 20.734] Where, F = 1.1, accumulation factor for safety relief valves prescribed by Code Case N-641 Mm = the value of Mm determined in accordance with ASME Section XI, Append i x G, Paragraph G-2214.1, v'in. (See WCAP-18169-NP, Revision 1, Section 6.2) p = vessel design pressure, ksig Ri = vessel inner radius, in. t = vessel wall thickness, in. Notes: (a) The coolant temperature is the reactor coolant inlet temperature. (b) The vessel metal temperature is the temperature at a distance one-fourth of the vessel section thickness from the clad/base metal interface in the vessel beltline region. RT NOT is the highest adjusted reference temperature (for weld or base metal in the beltline region) at a distance one-fourth of the vessel section thickness from the vessel clad/base metal interface as determined by RG 1.99, Revision 2 (Reference 2). During a cooldown, the 1/4T metal location will be at a higher temperature than the coolant temperature.

At steady-state, the 1/4T metal location will be at the same temperature as the coolant temperature.

During a heatup, the 1/4T metal location will be at a lower temperature than the c oolant temperature.

Since the enable temperature is based on the coolant Enclosure to 2CAN081802 Page 13 of 14 temperature corresponding to the 1/4T metal temperature, the enable temperature determined for the fastest heatup rate will result in the highest enable temperature and will bound the enable temperature for all other heatup, cooldown, and isothermal conditions.

The following calculations show the determination of the AN0-2 L TOP system minimum enable temperature based on the fastest heatup rate for 54 EFPY. Determi n ation of the minimum enable temperature per Item (2): Per Tabl e 7-4 of WCAP-18169-NP, Revision 1 (attached), the conservative limiting 1/4T RT NOT value for 54 EFPY P-T limit curve development is 122 °F. Therefore, Te= RTNoT + 40 °F equals 162 °Fat the 1/4T location for 54 EFPY. Using the vessel temperatures at the 1/4T location and the coolant temperatures that were calculated for the 80 °F/hr heatup rate case (see WCAP-18169-NP, Revision 1, Table A-1), the coolant t e mperature corresponding to the Te per item (2) was calculated to be 178 °F. Minimum Enable Temperature for 54 EFPY per Item (2) = = 175 + (180 -175) * (162 -159.011) / (163.927 -159.011) = 178 °F Determi n ation of the minimum enable temperature per Item (3): T e= RT NOT+ 50 In [((F

• Mm (p

• Ri / t)) -33.2) / 20.734] Where, RTNoT = 122°F for 54 EFPY (see WCAP-18169-NP, Revision 1 Table 7-4) F = 1.1 t = 7.875 inches Mm = 0.926 (t)112 , since (t)112 = 2.806, which is 2 s (t)1 12 s 3.464 for an inside surface axial flaw = 2.599 -Y in. p = 2.485 ksig R i = 79.719 inches Calculati o n of Te for 54 EFPY using the above parameters yields a Te of 153.2 °F. Using the vessel temperatures at the 1/4T location and the coolant temperatures, that were calculated for the 80 °F/hr heatup rate case (see WCAP-18169-NP, Revision 1, Table A-1 ), the coolant temperature corresponding to the Te value per Item (3) was calculated to be 169 °F. Minimum Enable Temperature for 54 EFPY per Item (3) = 165 + (170-165)

• (153.2-149.195) / (154.099-149.195)

= 169 °F Enclosur e to 2CAN08 1 802 Page 14 of 14 Determination of the Overall Minimum Enable Temperature:

The L TOP system shall be effective below the higher temperature determined in accordance with (1) a nd (2) above, which was determined to be 200 °F for 54 EFPY. Alternatively, L TOP systems s hall be effective below the higher temperature determined in accordance with (1) and (3) abov e, which was also determined to be 200 °F for 54 EFPY. Therefore, the minimum required e nable temperature (without uncertainties) for the AN0-2 reactor vessel is 200 °F for 54 EFPY. An instrument uncertainty of 20 °F is added resulting in the minimum enable temperature with uncertainties of 220 °F. REFERENCES

1. NR C Technical Letter Report, TLR-RES/DE/CIB-2013-01, "Evaluation of the Beltline Reg i on for Nuclear Reactor Pressure Vessels," Office of Nuclear Regulatory Research [RE S], November 14, 2014 (ML 14318A177).

2. Reg u latory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," dated May 1988 (ML0037 40284 ). 3. Ent e rgy submittal to NRC, "Reactor Vessel Surveillance Capsule Technical Report," dated Oct o ber 17, 2016 (2CAN101602) (ML 16293A583).

4. Am e rican Society of Mechanical Engineers (ASME) Code Case N-641, " Alternative Pre s sure-Temperature Relationship and Low Temperature Overpressure Protection Syst e m RequirementsSection XI, Division 1, ASME International, January 17, 2000 ATTACHMENT

-WCAP-18169-NP , Revision 1 Enclosure Attachment to 2CAN081802 WCAP-18169-NP , Revision 1