ML11256A026
| ML11256A026 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 09/06/2011 |
| From: | Holbrook K Progress Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| HNP-11-081, TAC ME6169 | |
| Download: ML11256A026 (11) | |
Text
Enclosure to SERIAL: HNP-11-081 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE LICENSE AMENDMENT REQUEST Summary By letter dated April 28, 2011, (ADAMS Accession No. ML11124A180), Carolina Power &
Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc., submitted a proposed amendment for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The proposed amendment will increase the rated thermal power (RTP) level from 2900 megawatts thermal (MWt) to 2948 MWt, and make Technical Specification (TS) changes as necessary to support operation at the uprated power level. The proposed change is an increase in RTP of approximately 1.66 percent. The proposed uprate is characterized as a measurement uncertainty recapture (MUR) using the Cameron Leading Edge Flow Meter (LEFM) CheckPlus System to improve plant calorimetric heat balance measurement accuracy. The proposed change will revise Renewed Operating License NPF-63 Maximum Power Level; Appendix A, TS definition of RTP; Reactor Core Safety Limits; Reactor Trip System Instrumentation; Minimum Allowable Power Range Neutron Flux high setpoint with Inoperable Steam Line Safety Valves; and TS Bases Section 3/4.7.1 to reflect the uprated reactor core power level.
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the information submitted by the licensee, and based on this review determined the following information is required to complete the evaluation of the subject amendment request:
Request 1:
Section IV.1.A.i of Enclosure 2 to Reference 1 states that the reactor pressure vessel (RPV) components at Shearon Harris Nuclear Power Plant (HNP) were originally analyzed for a vessel inlet temperature (Tcold) of 536.6 degrees Fahrenheit (oF). Based on the nuclear steam supply system (NSSS) parameters presented in Table 4 of Enclosure1 to Reference 1, it was stated that a decrease in the vessel inlet temperature of 0.6 oF from the originally analyzed value to the revised value, resulting from the measurement uncertainty recapture (MUR) power uprate, would result in negligible changes in the transient thermal stresses for the RPV at MUR conditions.
Clarify the relationship between the originally analyzed Tcold temperature, stated to be 536.6 oF, and the value presented in Table 4 of Enclosure 1, indicating that the current design conditions for the vessel inlet temperature is 554.4 oF.
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Enclosure to SERIAL: HNP-11-081 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE LICENSE AMENDMENT REQUEST
Response
Current design conditions include a range of Tavg from 572.0 °F to 588.8 °F. Tcold of 536.6 °F corresponds to a Tavg of 572.0 °F at current design conditions. Tcold of 554.4 °F corresponds to a Tavg of 588.8 °F, also at current design conditions. Table 4 of Enclosure 1 only reflects current design conditions associated with a Tavg of 588.8 °F.
HNP has a Tavg range of 572 °F to 588.8 °F. Table 4 shows parameter values at 102 percent power for four Cases. Cases 1 and 2 are at a Tavg of 572 °F; Cases 3 and 4 are at Tavg of 588.8 °F.
The Table 4 column listing current design conditions is only for a Tavg of 588.8 °F, which has an associated Tcold value of 554.4 °F.Section IV.1.A.i of Enclosure 2 discusses the reactor vessel components and indicates that these components were originally analyzed for a Tcold of 536.6 °F (which corresponds to the lowest Tcold value at a Tavg of 572 °F) and a Thot of 623.2 °F (which corresponds to the highest Thot value at a Tavg of 588.8 °F).
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Enclosure to SERIAL: HNP-11-081 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE LICENSE AMENDMENT REQUEST Request 2:
Section IV.1.D of Regulatory Issue Summary (RIS) 2002-03 stipulates that the content of MUR license amendment request (LAR) applications must include the codes of record used in the qualification of structures, systems and components (SSCs) to determine their structural adequacy at MUR conditions.Section IV.1.A.iv of Enclosure 2 to Reference 1 states that primary equipment supports were evaluated and found to be acceptable at MUR conditions.
State the design code(s) of record for the primary equipment supports and confirm that the evaluations performed in support of MUR power uprate implementation at HNP were performed consistent with the provisions in the original design code(s) of record.
Response
LAR Section IV.1.A.iv in the last paragraph states that the evaluations performed concluded that the current analyses of record remain valid for reactor coolant loop piping and supports. The last sentence states that the codes of record are listed in Section IV.1.D and remain unchanged.
Section IV.1.D lists the code of record for RCS supports as ASME III, Code Class 1, 1974 Edition through Winter 1974.
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Enclosure to SERIAL: HNP-11-081 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE LICENSE AMENDMENT REQUEST Request 3:
Section IV.1.A.ii.5 of Enclosure 2 to Reference 1 summarizes the evaluations performed to demonstrate the continued structural qualification of the reactor vessel internals (RVIs) at HNP following the proposed MUR power uprate implementation. This section of Reference 1 states that the effects of higher heat generation, resulting from the power uprate, were considered in evaluating the structural integrity of the RVIs.
Confirm that all other loads used in the current analyses of record (AOR) for the RVIs remain unaffected by the proposed MUR power uprate implementation (i.e., seismic, LOCA, reactor internal pressure differences, etc.). Additionally, with respect to the evaluation of the RVIs, state the design code of record used to qualify the RVIs for MUR conditions and confirm that the original design code of record was utilized in the evaluations performed to support MUR power uprate implementation.
Response
The HNP reactor internals are not ASME Code internals; therefore, there is no code of record.
The HNP reactor internals were designed and built prior to the introduction of Subsection NG of the ASME Boiler and Pressure Vessel (B&PV) Code,Section III. Therefore, a plant-specific stress report on the reactor internals was not required. The structural integrity of the HNP reactor internals design has been assured by generic and plant-specific analyses using different design codes. These generic and plant-specific analyses makeup the current AOR and were used to evaluate the affected HNP reactor internals components. However, for the HNP MUR power uprate, the reactor internals were analyzed in accordance with Subsection NG of the ASME B&PV Code,Section III, 2004 Edition. Only the reactor internal components affected by the MUR power uprate were analyzed.
The various design parameters and design loads used in the analysis of reactor internals components are shown in Table 1 and were evaluated to determine the components affected by the MUR power uprate. They are consistent with the design parameters and design loads used in the HNP AOR. The MUR power uprate impact on these design parameters/loads is provided in Table 1.
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Enclosure to SERIAL: HNP-11-081 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE LICENSE AMENDMENT REQUEST Table 1. RVI Design Parameters and Loads Parameter / Load MUR Power Uprate Affected Unaffected Geometry Material Thermal Design Transients Heat Generation Rates (Gamma Heating)
Weight Reactor Internals Hold-Down Spring Force Flow Lift Forces and Lateral Forces Reactor Internals Pressure Differences Vibration Loads LOCA System Loads LOCA Acoustic and Hydraulic Loads Seismic System Loads Fuel Assembly Interface Loads Driveline and Control Rod Interface Loads Page 5 of 9
Enclosure to SERIAL: HNP-11-081 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE LICENSE AMENDMENT REQUEST Request 4:
Section IV.1.A.ii.5.b of Enclosure 2 to Reference 1 describes the evaluations performed to structurally qualify the baffle-former bolts for operation at the proposed MUR power level. The basis of this evaluation is stated to be a comparison between a facility similar to HNP (Almaraz Unit 2) showing that the baffle-former bolts at Almaraz Unit 2 are structurally adequate under similar operating parameters. Additionally, it is stated that the MUR power uprate has insignificant impacts on the thermal analyses for these components.
Provide a tabulated comparison of the Almaraz Unit 2 and HNP parameters used to qualify the baffle-former bolts. This comparison should include information which demonstrates that the design basis requirements related to the structural integrity of these components will continue to be satisfied following MUR implementation.
Response
The HNP baffle-former bolts structural qualification was accomplished by comparing the baffle-former bolt data with that from Almaraz Unit 2, which has the same baffle-barrel and baffle-former bolt designs. The following design inputs were used to show the similarity between HNP and Almaraz Unit 2:
Drawings HNP and Almaraz Unit 2 use the same former drawings and baffle plate drawings.
Therefore, the baffle-former bolt locations and baffle plate pressure relief hole locations are the same for both plants.
Operating parameters The operating parameters for the two plants are very similar; the small differences would have an insignificant impact on the thermal analysis used to generate the baffle plate changes in temperature (Ts).
Geometric parameters The geometry used in the thermal analysis of the baffle-barrel region of Almaraz Unit 2 and HNP are identical.
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Enclosure to SERIAL: HNP-11-081 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE LICENSE AMENDMENT REQUEST Design transients The HNP MUR power uprate design transients are the same as those used in the Almaraz Unit 2 analysis. Both plants use the transients in the Standard System Design Criteria (SSDC) 1.3, Revision 2.
Heat generation rates The HNP MUR power uprate heat generation rates are the same as those used in the Almaraz Unit 2 analysis. Both plants use the heat generation rates specified in WCAP-9620, Revision 1, Reactor Internals Heat Generation and Neutron Fluences.
In conclusion, the design input comparison shows that the Almaraz Unit 2 results are applicable to HNP. The total cumulative fatigue usage factors for the short and long bolts, which are less than one, are applicable to HNP; therefore, the baffle-former bolts are structurally acceptable.
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Enclosure to SERIAL: HNP-11-081 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE LICENSE AMENDMENT REQUEST Request 5:
Section IV.1.A.vi.3.b of Enclosure 2 to Reference 1 discusses the impact of the proposed MUR power uprate implementation at HNP on the flow-induced vibration (FIV) and tube wear in the HNP steam generators (SG). It is stated that the fluid-elastic stability ratio will increase by as much as 3.4% while the tube vibration amplitude will increase by as much as 6.9%.
Discuss the methodology used to extrapolate the stability ratio and vibration amplitude to the values expected at MUR conditions. Confirm that the methodology is consistent with that used in the current AOR for the SG tubes. If the methodology varies from that in the current AOR, provide a technical justification for the use of the alternate methodology.
Response
The analysis of record (AOR) uses the thermal-hydraulic analysis results and geometry to evaluate the HNP steam generator (SG) tubes for the effects of fluid-elastic instability, turbulence and wear. These evaluations make use of computer programs that consider the SG geometry and the fluid parameters to determine the acceptability of the tubes for these FIV effects. The methodology used to evaluate the effects of the MUR make use of the FIV AOR and considers the changes to key parameters, i.e., fluid velocity and density, based on a revised thermal-hydraulic analysis performed to address the MUR. This analysis is consistent with the methodology used in the current AOR.
A review of the equations used to evaluate the FIV state of the SG tubing shows that since the geometry of the SGs is unchanged as a result of the uprate, changes to tube effects become a function of density and velocity. The FIV AOR results for the uprate are adjusted by a ratio of the change in these key parameters, pre- and post-uprate, based on how they are used to define the FIV results. Justification follows:
Fluid-Elastic Instability Fluid-elastic instability is a function of various parameters as indicated in the design basis analysis. Changes in operating conditions will modify some of these parameters and result in changes to the previously calculated values. The principle values that will change as a result of the uprate will be the secondary side fluid velocity and the density.
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Enclosure to SERIAL: HNP-11-081 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE LICENSE AMENDMENT REQUEST The fluid-elastic stability ratio is related to the fluid velocity and density through the following relationship:
Stability ratio 1/2 x V For the purpose of the MUR evaluation, the ratio used is conservatively taken as (V2).
Turbulence Turbulent displacements are a function of various parameters including density () and velocity (V) of the form (V(3+s))0.5, where S (a turbulence constant) can range from 0.304 to 2.34. This means that displacements can be a function of Va where a can range from 1.65, [(3+0.304)*0.5] to 2.67, [(3+2.34)*0.5] or ~ V2. The nominal relationship of y (V2)2 is conservatively used to evaluate the effect of the uprate on turbulence induced displacements in SG tubes.
Wear Potential Tube wear is a function of both fluid forces and tube displacement. The fluid forces acting on the tube are proportional to V2. As described above, the tube turbulent displacements are also a function of V2. Combining these two terms shows that tube wear, for a given geometry, will be proportional to (V2)2.
Conservatism is added by considering the ratio of the quantity of V2 produced for a representative 100 percent power operating condition considered in the AOR, to each of the 100 percent power operating conditions considered for the uprate. The maximum ratio is then used to evaluate the MUR FIV effects. The percent increase in the parameters above is the maximum (V2) ratio of 1.034, and 1.069 is the square of that ratio for an increase of 3.4 percent and 6.9 percent, respectively.
Since there is no change in the geometry of the tube bundle, evaluating only the change in thermal-hydraulic results calculated for the MUR is consistent with the AOR methodology. The algorithms used to evaluate the MUR FIV effects are the same as those used for the AOR, except that conservative ratios are used to modify the results as justified above. Therefore, the methodology is considered to be appropriate.
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