2CAN081802, 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: | Richard Anderson Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2CAN081802 | |
| Download: ML18215A177 (18) | |
Text
~ Entergy Entergy Operations, Inc.
1448 S.R. 333 Russellville, AR 72802 Tel 479-858-3110 Richard L. Anderson ANO S~e Vice President 10 CFR 50 .90 2CAN081802 August 1, 2018 ATIN : Document Control Desk U.S. Nuclear Regulatory Commission Washington , DC 20555
SUBJECT:
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
REFERENCE:
1 Entergy submittal to the NRC, "License Amendment Request Updating the Reactor Coolant System Pressure-Temperature Limits," dated November 20, 2017 (2CAN 111702) (ML17326A387) 2 NRC to Entergy, "Request for Additional Information Regarding License Amendment Request to Update the Reactor Coolant System Pressure-Temperature Limits," dated May 10, 2018 (2CNA051801) (ML18129A425)
Dear Sir or Madam:
Entergy Operations, Inc. (Entergy) submitted a license amendment request to revise the Arkansas Nuclear One, Unit 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. Kriss 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 NOT 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 that for reactor vessel beltline materials, including welds, plates, and forgings, the values of RTNOT must account for the effects of neutron radiation. Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials," Revision 2 (Reference 4 ), details how ~ RT NOT is calculated . The NRC staff notes that Regulatory Issue Summary 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components" (Reference 5), clarifies that 10 CFR Part 50, Appendix G and 10 CFR Part 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 to 1x10 17 n/cm 2 (E> 1 MeV).
The NRC staff also notes that TLR-RES/DE/CIB-2013-01 is not NRC guidance, and the recommendation that the shift due to irradiation can be discounted if it is less than 25 °F is not endorsed in any NRC guidance document or permitted by the regulations.
Revise the P-T limit evaluation to include RT NOT values for all RPV beltline and extended beltline materials calculated in accordance with 10 CFR Part 50, Appendix G, so the NRC staff can verify that the calculations were performed in accordance with the regulations.
Entergy's response:
The 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 for each of the reactor pressure vessel (RPV) beltline and extended beltline materials [i.e., reactor vessel materials with fluence values exceeding 1 x 10 17 n/cm 2 (E > 1.0 MeV)], in accordance with 10 CFR 50, Appendix G, without the use of the TLR-RES/DE/CIB-2013-01 (Reference 1) methodology. Thus, calculated values of the change in the Reference Temperature for Nil Ductility Transition (6RTNoT) less than or equal to 25 °F are not set equal to zero. It is noted that the limiting ART values presented in Table 7-4 and the P-T limit curves 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) with 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) is assumed to decrease as a function of fluence and copper content, as indicated in 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 to 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 Initial Projected Projected Reactor Vessel Material and Wt% 1 / 4T Fluence(bl cu(a) USE USE USE Identification Number (n/cm2, E >1 MeV)
(ft-lb) Decrease (%) (ft-lb)
Position 1.2(c) 19 Intermediate Shell Plate C-8009-1 0.098 3.06 X 10 95 25 .0 71.3 19 Intermediate Shell Plate C-8009-2 0.085 3.06 X 10 93 25.0 69.8 19 Intermediate Shell Plate C-8009-3 0.096 3.06 X 10 134 25.0 100.5 19 Lower Shell Plate C-8010-1 0.085 3.10 X 10 89 25.0 66.8 19 Lower Shell Plate C-8010-2 0.083 3.10 X 10 94 25.0 70.5 19 Lower Shell Plate C-8010-3 0.080 3.10 X 10 97 25.0 72 .8 Intermediate Shell Heat#
Longitudinal Welds 0.05 2.89 X 10 19 110 24.5 83.1 Multiple 2-203-A, B, & C Lower Shell Heat# 19 Longitudinal Welds 0.046 2.94 X 10 125 24.5 94.4 10120 3-203-A, B, & C Intermediate to Heat# 19 Lower Shell Girth 0.045 3.05 X 10 136 25.0 102.0 83650 Weld 9-203 Upper Shell Plate C-8008-1 0.13 0.0367 X 1019 93 11 .5 82.3 19 Upper Shell Plate C-8008-2 0.13 0.0367 X 10 68 11 .5 60.2 19 Upper Shell Plate C-8008-3 0.08 0.0367 X 10 71 8.8 64.8 Upper Shell Heat#
Longitudina l Welds 0.02 0.0367 X 1019 92 8.8 83.9 BOLA 1-203A, B, & C Heat#
0.22 0.0367 X 10 19 105 18.5 85.6 10137 Upper to Heat#
Intermediate Shell 0.21 0.0367 X 10 19 107 18.5 87.2 6329637 Girth Weld 8-203 Heat#
0.03 0.0367 X 1019 118 8.8 107.6 FAGA Position 2.2(dl 19 Intermediate Shell Plate C-8009-3 0.096 3.06 X 10 134 34.0 88.4 Upper Shell Plate C-8008-1 0.13 0.0367 X 1019 93 12.0 81.8 Intermediate to Heat#
Lower Shell Girth 0.045 3.05 X 10 19 136 18.5 110.8 83650 Weld 9-203
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 ine 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 points 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 LTOP 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 LTOP reanalysis .
c) Discuss the results of the LTOP 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 LTOP enable temperature and justify that the required enable temperature in 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 (LTOP) AOR and LTOP 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 configurations 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 initial pressurizer saturated liquid starting conditions (i.e., 417.4 °F @
300 psia and 444 .6 °F@ 400 psia as shown in Table 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 LTOP Reanalysis Values Backpressure Backpressure Flow (gpm) Flow (gpm)
(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 1959.46 236 - -
- - 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 LTOP Reanalysis Values Backpressure Backpressure Flow (gpm) Flow (gpm)
(psia) (psia)
- - 768.7 139.6
- - 899.6 158.6 901 .73 169 - -
- - 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
Enclosure to 2CAN081802 Page 6 of 14 Review of Table 2 and Table 3 values indicate that the system backpressure versus flow rate for liquid discharge in the LTOP 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 ckpressure shall be limited to 50% of the relief valve set pressure. Per TS 3.4.12, the set pressure 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 at 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 determined for a pressurizer saturated liquid starting condition of 417.4 °Fat 300 psia and 444.6 °F at 400 psia , respectively. Therefore, the maximum flow permitted to assure stable valve operation is lower in the current reanalysis.
The backpressure 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 280 260 240 220 160 140 120 100 600.00 800 .00 1,000.00 1,200.00 1,400.00 1,600.00 1,800 .00 2,000.00 2, 200.00 2,400.00 Relief Valve Flow (gpm)
- - - AOR Backpressure@ 417 .4' F -+- Reanalysis Ba ck pressure@ 417.4' F
- * - AOR Backpressu re@ 444 .6' F Rea nalysis Backpressure@ 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 LTOP AOR and the LTOP 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 2CAN081 802 Page 8 of 14 Figure 2 RCS Pressure Versus HPSI Flow (One Pump) 1600 1400 1200
! 1000 QI
~ 800
~
0.
V) 600 u
a:
400 200 0
0 100 200 300 400 500 600 700 800 900 1000 HPIS Pump Flow (gpm)
~ AOR -4t- Reana lysis Entergy's response - Part b):
The LTOP Transient AOR The LTOP 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 LTOP relief valve backpressure for liquid discharge as further discussed below) . One assumption was that a maximum of two HPSI pumps would be aligned during LTOP 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 LTOP relief valve backpressure 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 lel effort, a RELAP5 analysis of the piping network downstream of the LTOP relief valve was performed to determine LTOP relief valve backpressure for liquid discharge as a function of 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 LTOP re lief valve backpressure less than 50% of the set pressure.
Comparison of the peak LTOP 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 LTOP relief valve behavior, the following operating restrictions were imposed via incorporation into the plant TSs:
- 2. A maximum nominal pressurizer level of 910 ft 3 should be maintained prior to starting of the first Reactor Coolant Pump (RCP) during LTOP operation.
Enclosure to 2CAN081802 Page 9 of 14 Given these two restrictions on LTOP 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 LTOP 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 in the AN0-2 Safety Analysis Report (SAR) Section 5.2.2.4.
The Current LTOP Transient Reanalysis The current mass and energy addition events used explicit LTOP 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 .
Reanalysis of upstream and downstream pressure values was necessary due to changes in the isometric 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 than 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 LTOP mass addition AOR. The relief valve inlet piping pressure drop was calculated based on this equilibrium flow rate. This 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 LTOP operation . The flow for the single HPSI pump used in the analysis has increased compared to the 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 included explicit LTOP 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 of the transient). The maximum 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 (as in the LTOP 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 and closing of the (single credited) LTOP relief valve, the relief valve steam flow, and the eventual relief valve liquid flow as the primary system became water solid .
Entergy's response - Part c):
Comparison of the Current LTOP 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 LTOP 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 bubble. 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 is 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).
Rated Steam Flow For transient initiated with 417.4 °For with 444.6 °F saturated steam flow is greater than 17,000 gpm Rated 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 LTOP 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 Justification 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 LTOP 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 LTOP 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 limiting energy addition event (RCP start event) is modelled using the CENTS code. It is assumed that the primary system Tcold is initially at the LTOP enable temperature of 220 °F and that the steam generators are filled with water at the limiting temperature of 100 °F above the primary system Tcold* The shutdown cooling (SOC) system is isolated as are charging and letdown.
This historical 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 psia) 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 the primary coolant temperature to a condition 100 °F cooler than the secondary side water inventory. At that point, SOC is isolated.
The LTOP mass addition event (one HPSI pump and three charging pumps) is also assumed to occur with the primary system Tcold initially at the LTOP enable temperature of 220 °F in order to maximize decay heat.
The peak transient pressures are reduced at lower RCS temperatures.
Comparison of the Current LTOP Transient Peak Pressure to the P-T Limits A bounding 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 energy addition and to replace inventory lost through the relief valve discharge) of 2.2 psi is added to the LTOP transient peak pressure of 498 psia prior to comparison of the peak pressure result 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 limiting P-T point is shown on the LAR heatup curve of TS Figure 3.4-2A is 588 psia at 60 °F.
Enterqy's response - Part d}:
The LTOP enable temperature in TS Limiting Condition for Operation (LCO) 3.4.12 was reevaluated per ASME Code Case N-641 (Reference 4) based on the limiting 1/4T ART in WCAP-1 8169-NP, Revision 1, Table 7-4. The P-T limits in TS 3.4.9 are also based on the limiting ART values in WCAP-18169-NP , Revision 1, Table 7-4.
Enclosure to 2CAN08 1802 Page 12 of 14 ASME Code Case N-641 presents alternative procedures for calculating P-T relationships and LTOP 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 LTOP effective temperature calculations.
Per ASME Code Case N-641, the LTOP system shall be effective below the higher temperature determined in accordance with (1) and (2) below. Alternatively, LTOP 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 RTNOT - 85 °F.
(3) A coolant temperature(a) corresponding to a reactor vessel metal temperature(bl, for all vessel beltline materials, where Te is calculated on a plant-specific basis for axial and circumferential 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 ix 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 coolant temperature. Since the enable temperature is based on the coolant I
I L
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 LTOP system minimum enable temperature based on the fastest heatup rate for 54 EFPY.
Determination of the minimum enable temperature per Item (2):
Per Table 7-4 of WCAP-18169-NP, Revision 1 (attached), the conservative limiting 1/4T RTNOT 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 temperature 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 Determination of the minimum enable temperature per Item (3):
Te = RTNOT+ 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) 112 s 3.464 for an inside surface axial flaw
= 2.599 -Yin.
p = 2.485 ksig Ri = 79.719 inches Calculation 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
Enclosure to 2CAN081802 Page 14 of 14 Determination of the Overall Minimum Enable Temperature:
The LTOP system shall be effective below the higher temperature determined in accordance with (1) and (2) above, which was determined to be 200 °F for 54 EFPY. Alternatively, LTOP systems shall be effective below the higher temperature determined in accordance with (1) and (3) above, which was also determined to be 200 °F for 54 EFPY. Therefore, the minimum required enable 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. NRC Technical Letter Report, TLR-RES/DE/CIB-2013-01, "Evaluation of the Beltline Reg ion for Nuclear Reactor Pressure Vessels," Office of Nuclear Regulatory Research
[RES], November 14, 2014 (ML14318A177).
- 2. Reg ulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials,"
dated May 1988 (ML003740284 ).
- 3. Entergy submittal to NRC, "Reactor Vessel Surveillance Capsule Technical Report," dated October 17, 2016 (2CAN101602) (ML16293A583).
- 4. Am erican Society of Mechanical Engineers (ASME) Code Case N-641, "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System RequirementsSection XI, Division 1, ASME International, January 17, 2000 ATTACHMENT - WCAP-18169-NP, Revision 1
Enclosure Attachment to 2CAN081802 WCAP-18169-NP, Revision 1