L-MT-09-017, Extended Power Uprate: Response to NRC Reactor Systems Branch & Nuclear Performance & Code Review Branch Request for Additional Information (RAI) Dated January 16, 2009
| ML090790388 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 03/19/2009 |
| From: | O'Connor T Northern States Power Co, Xcel Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-MT-09-017, TAC MD9990 | |
| Download: ML090790388 (35) | |
Text
March 19,2009 WITHHOLD ENCLOSURE 1 FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 and 9.17 L-MT-09-0 1 7 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket 50-263 Renewed Facility Operating License License No. DPR-22 Monticello Extended Power Uprate: Response to NRC Reactor Systems Branch and Nuclear Performance & Code Review Branch Request for Additional Information (RAI) dated Januaw 16,2009 (TAC No. MD9990)
References:
I. NSPM letter to NRC, License Amendment Request: Extended Power Uprate (L-MT-08-052) dated November 5, 2008 (TAC No. MD9990)
Accession No. ML0832301 I 1
- 2.
Email P. Tam (NRC) to G. Salamon, K. Pointer, and R. Loeffler (NSPM) dated January 16, 2009, iiMonticello - First portion of draft RAI re.
reactor systems for EPU amendment " (TAC No. MD9990) Accession No. ML090160401 Pursuant to 10 CFR 50.90, the Northern States Power Company, a Minnesota corporation (NSPM), requested in Reference 1 an amendment to the Monticello Nuclear Generating Plant (MNGP) Renewed Operating License (OL) and Technical Specifications (TS) to increase the maximum authorized power level from 1775 megawatts thermal (MWt) to 2004 MWt.
The U. S. Nuclear Regulatory Commission (NRC) Reactor Systems Branch and Nuclear Performance & Code Review Branch provided nineteen (1 9) RAls as described in Reference
- 2. Enclosure I provides the NSPM response. This enclosure contains information which is proprietary to GE Hitachi (GEH). GEH requests this proprietary information be withheld from public disclosure in accordance with 10 CFR 2.390(a)4 and 9.17(a)4. An affidavit supporting this request is provided in Enclosure 2. A non-proprietary version of the NSPM response is provided as Enclosure 3.
Document Control Desk Page 2 of 2 In accordance with 10 CFR 50.91, a copy is being provided to the designated Minnesota Official.
Summaw of Commitments The steady state bypass void fraction for the EPU core will be calculated using the method described by the NSPM response to NRC RAI SNPB-7 of L-MT-09-017.
I declare under penalty of perjury that the foregoing is true and correct.
President, Monticello Nuclear Generating Plant States Power Company - Minnesota Enclosures (3)
I.
Enclosure I
- Response to Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16, 2009 (Proprietary)
- 2. Enclosure 2:
- 3. Enclosure 3:
Response to Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 (Non-Proprietary) cc:
Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC Minnesota Department of Commerce
ENCLOSURE 2 GE HITACHI Affidavit
GE-Hitachi Nuclear Energy Americas LLC AFIFrnAVIT I, Tirn E. Abney, state as follows:
(1) I am Vice President, Services Licensing, Regulatory Affairs, GE-Hitachi Nuclear Energy Americas LLC ("GEH"), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.
(2) The information sought to be withheld is contained in GEH letter, GE-MNGP-MP-1142, GEH Response to NRC MIS 2.8.4.2-1, 2.8.4.2-2, 2.8.4.2-3, 2.8.4.3-4, and SNPB-1 thrz~ 9, dated February 24, 2009. The proprietary information in Enclosure 1 entitled, GEH Response to NRC RAIs 2.8.4.2-1, 2.8.4.2-2, 2.8.4.2-3, 2.8.4.3-4, and SNPB-1 thrzl 9, is identified by a dotted underline inside double square brackets, [ ~ ~ h j ~ s a : g ~ ~ g ~ ~ m m i : s s s ~ n gxam8!&~~.!~]. In each case, the superscript notation {3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.
(3) In malcing this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Proiect v. Nuclear Regulatory Commission, 975F2d871 @C Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 @C Cir. 1983).
(4) Some examples of categories of information which fit into the definition of proprietary information are:
- a.
Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEHts competitors without license fYom GEH constitutes a competitive economic advantage over other companies;
- b.
Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;
- c.
Information which reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;
- d.
Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.
Affidavit for MNGP-AEP-I 142 Affidavit Page 1 of 3
The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. above.
(5) To address 10 CFR 2.390(b)(4), the infolmation sought to be withheld is being submitted to NRC in confidence. The information is of a sost customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my lcnowledge and belief, consistently been held in confidence by GEH, no public disclosure has been made, and it is not available in pt~blic sources. All disclosures to third pasties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary infosmation, and the subsequent steps talten to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.
(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acqtlainted with the value and sensitivity of the infoimation in relation to industiy lcnowledge, or subject to the terms under which it was licensed to GEH. Access to such documelits within GEH is limited on a "need to ~UIOW'~ basis.
(7) The procedure for approval of exteinal release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, st~ppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.
(8) The infolmation identified in paragraph (2) above is classified as proprietary because it contains results of an analysis performed by GEH to support Monticello's Extended Power Uprate license application. This analysis is part of the GEH Extended Power Uprate methodology. Development of the extended power uprate methodology and the supposting analysis techniques and information, and their application to the design, modification, and processes were achieved at a significant cost to GEH.
The develop~nent of the evaluation methodology along with the intelyretation and application of the analytical results is derived fiom the extensive experience database that constitutes a major GEH asset.
(9) Public disclosure of the information sought to be withheld is likely to cause substantial ham to GEH's competitive position and foreclose or reduce the availability of profit-malting opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.
The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply Affidavit for MNGP-AEP-1142 Affidavit Page 2 of 3
the appropriate evaluation pr~cess, In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.
The reseasch, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH.
The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.
GEH's competitive advantage will be lost if its competitors ase able to use the results of the GEH experience to normalize or verify theis own process or if they are able to claim an equivalent understanding by demonstrating that they can asrive at the same or similar conclusions.
The value of this information to GEH would be lost if the information were disclosed to the public. Making such infoimation available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the oppoi-tmity to exercise its competitive advantage to seek an adequate retusn on its large investment in developing and obtaining these very valuable analytical tools.
I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are tn~e and cossect to the best of my knowledge, information, and belief.
Executed on this 23rd day of Febi-uary 2009.
Affidavit for MNGP-AEP-1142 Tim E. Abney Vice President, Services Licensing GE-Hitachi Nuclear Energy Americas LLC Affidavit Page 3 of 3
ENCLOSURE 3 MONTICELLO NUCLEAR GENERATING PLANT Response to Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 (NON-PROPRIETARY)
-Non Proprietary Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 NRC RAI 2.8.4.2-1 Figure 2.8-21 provides a plot of the main steam isolation valve closure with flux scram transient. The Vessel Pressure Rise is plotted in psi, but scaled in percent. Please provide a revised figure with the applicable scale for Vessel Pressure Rise.
NSPM Response The curve on the plot is already in units of psi. It is not scaled to %, but shown in pressure rise from initial. Legend of the plot provides the clarification.
Page 2 of 29
-Non Proprietaw Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 NRC RAI 2.8.4.2-2 Explain why the vessel steam flow is oscillating while all other parameters appear to hold comparatively constant throughout the main steam isolation valve closure with flux scram transient.
NSPM Response The ODYN computer code is capable of modeling steam line dynamic behavior. The model captures the effect of the pressure wave traversing between the MSlV and the reactor dome plenum region. Although the steam flow oscillates, the trend in total steam flow vs. time decreases towards zero, which causes the vessel dome pressure to increase. There are small inflections in the reactor pressure that correspond to the timing of the steam flow oscillations. The perturbation occurs again when SRVs lift where flow finally settles to the SRV flow rate.
Page 3 of 29
-Non Proprietaw Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 NRC RAI 2.8.4.2-3 The main steam isolation valve closure with flux scram transient assume's that three safety relief valves are out of service. Does this assumption reflect a change from the current Monticello licensing basis for the overpressurization transient?
MSPM Response There is no change from the current analysis basis for Monticello as the current licensing basis considers three safety relief valves out of service. Please see USAR Section 14.5.1, Vessel Pressure ASME Code Compliance Model - MSlV Closure.
Page 4 of 29
-Non Proprietary Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 NRC RAI 2.8.4.3-1 Address how the reactor core isolation cooling (RCIC) net positive suction head (NPSH) evaluation aligns with or differs from the current RClC design basis NPSH evaluation, or from the RClC functional capability at current licensing basis conditions.
NSPM Response For both CLTP and EPU conditions, the design basis function of the RClC system is to provide coolant to the reactor vessel so that the core is not uncovered as a result of loss of off-site AC power (LOOP) or for a Loss of Feedwater (LOFW) event. In addition, although the HPCl System is the only injection source credited in the design basis Station Blackout (SBO) event, the RClC system may be used during an SBO if available to reduce the HPCl System demand and preserve battery capacity. The suction sources for the RClC System are the Condensate Storage Tanks (normal alignment) or the torus. The torus is limiting with respect to NPSH as the Condensate Storage Tanks are an elevated suction source which is not subject to significant post-event heat addition.
NPSH margin is demonstrated by calculation. The existing calculations for the RClC System are based on conservative design assumptions for the parametric values (e.g.
process flow rates, torus level, torus temperature) that determine the NPSH available.
These parametric values bound those that arise during RClC operation for all the designbasis events of interest. The calculations also develop the NPSH curves used to support Emergency Operating Procedure (EOP) RClC NPSH limits. The EOP RClC System NPSH curve is shown below.
-- I I
I I
I I
0 100 200 300 400 600 RCIC Rmrp Fbw From Torus @pm)
'Ouupsm@rig)=Oryml Ressm +0.4x(T~Lml(fl)+4.1)
Page 5 of 29
-Non Proprietary Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 The EPU changes to the corresponding RClC System operating parameters that affect the NPSH available were compared to the existing design assumptions in the calculations (e.g. torus temperature), and the assumptions remain bounding. Therefore there is-no change to the current RClC design basis NPSH evaluation or functional capability.
Page 6 of 29
-Non Proprietarv Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 NRC RAI 2.8.4.3-2 The PUSAR states that the station blackout (SBO) torus temperature response well bounds the loss of feedwater (LOFW) torus temperature response. State how the SBO torus temperature response was determined.
NSPM Response The SBO torus temperature response was determined for the four hour coping period using the SHEX containment response program. The initial conditions for the model are consistent with the design basis for the SBO event. The initial power and the associated decay heat profile have been changed for EPU. Additional detail is provided in Section 2.3.5 of NEDC-33322P, Rev. 3 which is provided as Enclosure 5 of the License Amendment Request: Extended Power Uprate (TAC MD9990) dated November 5, 2008.
The station blackout (SBO) torus temperature response well bounds the expected loss of feedwater (LOFW) torus temperature response. This is because no torus cooling water would be available during the SBO. For the LOOP or LOFW transients, both trains of torus cooling and shutdown cooling are available.
Page 7 of 29
-Non Proprietaw Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 NRC RAI 2.8.4.3-3 SBO coping analyses typically consider coping periods greater than two hours. What is the significance of using a reference torus temperature of 141 OF at two hours?
NSPM Response The RClC System is not credited in the design mitigation sequence for SBO. Part of the RClC System design basis is to show that the system will operate for two hours following a LOOP. This is a requirement of NUREG-0737 Item ll.K.3.15. The two hour SBO torus temperature response is used to show compliance with this requirement.
Because the reactor water level is recovered quickly, the long term torus temperature response is not determined as part of the transient analysis that credits RClC operation.
The containment response is an essential result of the SBO analysis. Using the SBO containment response as a substitute for the LOOP containment response is a conservative and reasonable method to determine reference torus temperature for the two hour RClC System mission time since both trains of torus cooling will be available during a LOOP (or a LOFW). No torus cooling is available during the SBO. Torus water temperature is a key parameter for the RClC room heatup analysis (torus piping is a significant heat source) and for the NPSH calculation.
The SBO torus response can then be applied to the RClC System design limits, The design basis RClC room temperature is 140°F1 and the RClC System piping operational limit is also 140°F. Given that the two hour peak torus temperature for a design basis SBO at EPU is 141°F and given that torus cooling is available for a LOOP (or LOFW), it is reasonable to conclude that the two hour torus temperature for a LOOP is less than 140°F. It can then be concluded from the RClC NPSH curve and the equipment design temperatures above that adequate NPSH is available and that the RClC System will operate for the required two hour mission time.
Depending on the available decay heat and containment pressure at EPU, torus temperatures may not preclude adequate NPSH over the four hour SBO coping period.
It cannot be posited, however, that the RClC System would be available for the entire four hours in the design basis EPU SBO event as torus temperatures exceed design piping operating temperatures. No credit is assumed for the RClC System in the design basis mitigation sequence for an SBO. This approach demonstrates that the secondary torus suction source for RClC will support RClC operation under EPU conditions.
Page 8 of 29
-Non Proprietarv Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16, 2009 NRC RAI 2.8.4.3-4 Summarize the results of the plant-specific LOFW analysis. Address the modeling tool used, describe the sequence of events, discuss any additional failures assumed -
beyond the HPCl - in the analysis and provide plots of the significant parameters to demonstrate performance relative to the applicable acceptance criteria. Compare analytic assumptions to those used in the current licensing basis. If applicable, provide equipment out-of-service assumptions used in this analysis and discuss whether these assumptions change in the EPU analysis as compared to the current licensing basis (CLB) analysis.
NSPM Response As described in the PUSAR, Table 1-1, for Transient Analysis, the modeling tool used is the SAFER04 model, which is the same model used in the ECCS LOCA analysis. The analysis is done consistent with NRC-approved GEH LTR, NEDC-33004P-A.
The general sequence of events in the analysis is as follows. The reactor is assumed to be at 102% of the EPU power level when the LOFW occurs. The initial level in the model is conservatively set at the low-level scram setpoint and reactor feedwater is instantaneously isolated at event initiation. Scram is initiated at the start of the event.
When the level decreases to the low-low level setpoint, the RCIC system and MSlV closure are initiated. The RCIC flow to the vessel begins at 48 seconds into the event, minimum level is reached at 72 seconds and level is recovered after that point. Only RCIC flow is credited to recover the reactor water level. There are no additional failures assumed beyond the failure of the HPCl system. The only other key analysis assumption for the LOFW analysis, discussed in Section 9.1.3 of NEDC-33004P-A, was the assumed decay heat level of ANS 5.1 -1 979 with a two-sigma uncertainty. The assumed decay heat level for the EPU analysis was ANS 5.1-1 979 decay heat + I 0%,
which bounds ANS 5.1-1 979 + two sigma. Thus, the key analytical assumptions are the same or conservative relative to the current licensing basis.
This LOFW analysis is performed to demonstrate acceptable RCIC system performance. The design basis criterion for the RCIC system is confirmed by demonstrating that it is capable of maintaining the water level inside the shroud above the top of active fuel during the LOFW transient. The minimum level (see Figure 2.8.4.23-4-1) is maintained at least 77 inches above the top of active fuel, thereby demonstrating acceptable RCIC system performance. There are no applicable equipment out of service assumptions for this transient.
Page 9 of 29
-Non Proprietaw Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 Figure 2.8.4.23-4-1 Page 10 of 29
-Non Proprietary Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16, 2009 NRC RAI 2.8.4.3-5 Both the PUSAR and the Constant Pressure Power Uprate Licensing Topical Report (CLTR, NEDC-33004P-A, Revision 4) refer to a "transient startup speed peak." Please define this phrase.
NSPM Response The RClC turbine speed is controlled by an electro-hydraulic control system responding to the flow controller demand input. In standby (with no actual pump flow) the flow controller output to the governor is calling for maximum turbine speed. When the system is initiated on low reactor level the steam supply valve to the turbine opens and the turbine rapidly accelerates. Due to system dynamics, during the turbine ramp up (startup transient) the turbine speed overshoots (speed peak) the governor speed setpoint. The turbine acceleration rate is only dependent on the flow controller setpoint, the governor setpoint, and the initial steam supply pressure. Since none of these parameters are changed in the CPPU the transient startup speed peak does not change.
Page 11 of 29
-Non Proprietaw Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16, 2009 NRC RAI 2.8.4.4-1 On Page 2-260, the PUSAR states, "Steam Condensing Mode of RHR is not installed at Monticello," and yet, the USAR for Monticello, Revision 24, Section 6.2.3.1.I, states, "The RHR System, also, provides cooling for the suppression pool so that condensation of the steam resulting from the blowdown due to the design basis LOCA is ensured."
When considered together, these two statements appear to contradict one another.
Please clarify.
NSPM Response The steam condensing mode of RHR was a hot standby operating mode which used the RHR heat exchanger to condense reactor steam and supply the condensate to the RClC pump suction. This mode of RHR was never installed at Monticello. USAR Section 6.2.3.1.I refers to the suppression pool cooling mode of RHR. This mode of operation cools the suppression pool following design basis events. This cooling results in the condensation of steam from the LOCA blowdown in the suppression pool.
Page 12 of 29
-Non Proprietarv Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 NRC RAI 2.8.4.4-2 Regulatory Guide I
.139, "Guidance for Residual Heat Removal," has been withdrawn because it is, according to Federal Register Volume 73, No. 1 12, Tuesday, June 10, 2008, "the guidance provided in this regulatory guide is no longer necessary."
Consequently, its availability has been significantly restricted. Please describe the RHR alternate shutdown cooling performance evaluation without reference to this Regulatory Guide. Provide the time required to cool the reactor below 212°F.
NSPM Response This PUSAR section requires some clarification. The reference to Regulatory Guide I.
139 really doesn't belong here as it refers to the Appendix R alternate shutdown analysis.
In addition to the normal shutdown cooling evaluation described in PUSAR Section 2.8.4.4, an alternate shutdown analysis was performed to evaluate the ability to attain cold shutdown following a non-LOCA event coincident with a single failure. MNGP is designed with redundant safety grade systems that are capable of attaining and maintaining cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> of an event using either on-site or off-site power. As discussed in USAR Section 6.2.3.3.4, the specific assumptions for this analysis are:
- 6. Off-site power is not available
- 7. The main condenser is not available
- 8. Condensate and Feedwater systems are not available
- 9. No credit is taken for RClC system operation since it is not a safety grade system.
- 10. HPCl is assumed to be inoperable.
The analysis is performed based on the Alternate Shutdown Cooling Method (ASCM) which depressurizes the reactor using the Automatic Depressurization System (ADS).
Either a Core Spray pump or RHR pump is used to flood the reactor vessel. Safetylrelief valves are then used to discharge heated reactor water to the suppression pool which is cooled with the suppression pool cooling mode of the RHR system. The water is then re-injected to the reactor vessel by aligning the injecting pump to take suction from the suppression pool.
The design basis Appendix R event utilizes the ASCM to attain cold shutdown.
This Appendix R alternate shutdown analysis was performed using this criteria at EPU conditions which demonstrates that the reactor can be cooled to cold shutdown (5200°F) in 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. This analysis is discussed in more detail in PUSAR Section 2.5.1.4.
Page 13 of 29
-Non Proprietaw Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 NRC RAI SNPB-I The disposition of Condition 1 from the NRC SER (safety evaluation report) for the IMLTR (interim methods licensing topical report or NEDC-33173P) specifies that TGBLAOG/PANACl 1 methods were used in the safety analysis. Please confirm that the version of TGBLAOG used is consistent with the modification made during the staff's review of the IMLTR (specifically TGBLAOGAE5 or later).
NSPM Response The Monticello Extended Power Uprate (EPU) analysis is based on a representative equilibrium core design performed with the TGBLAOGE4 and PANACI I Engineering Computer Programs. The TGBLAO6E5 modification is associated with the evaluation of the 1.058 eV resonance peak under high void conditions (>70% void fractions). The impact of this change for the Monticello representative core design is not significant based on the previously submitted RAI response of Reference SNBP-1-1.
The actual EPU core design will be performed prior to actual cycle operation. The most up-to-date TGBLAOG version will be used in the safety analyses to create the nuclear libraries for new fuel bundles in the EPU core. Existing bundles will be based on the earlier version of TGBLA06.
Reference RAI SNPB-1
- 1.
GEH letter, MFN 06-297 Supplement I, "Supplemental Response to Portion of NRC Request for Additional Information Letter No. 53 Related '
to ESBWR Design Certification Application - DCD Chapter 4 and GNF Topical Reports - RAI Number 4.3-3", November 8, 2006.
Page 14 of 29
-Non Proprietary Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 NRC RAl SNPB-2 The disposition of Condition 3 from the NRC SER for the IMLTR is not consistent with the guidance in GEH Letter (MFN 08-693) dated September 18, 2008. The power to flow ratio point considered is the lowest flow point at the highest thermal power, in other words the intersection of the high flow control line (HFCL) and the licensed thermal power line (LTPL). This occurs for the proposed Monticello Nuclear Generating Plant (MNGP) extended power uprate (EPU) operating domain at 100 percent EPU thermal power and 99 percent rated core flow (RCF). Please confirm that this condition in future cycles will be met, consistent with MFN 08-693.
NSPM Response PUSAR Table 2.8-2 provides the core thermal power to total core flow for two statepoints: 120% power (EPU) at 100% flow and 120% power (EPU) at 80% flow. In both cases the ratio of the thermal power to core flow does not exceed 50 MWtlMlbmIhr. The 80% flow statepoint bounds the 99% flow statepoint; therefore, the ratio at the low flow point at rated power (120% powerl99% flow) for Monticello would not exceed 50 MWtlMlbmIhr.
The power-flow map is independent of fuel design and does not change cycle to cycle.
Therefore, the power to flow ratio for Monticello's future EPU cycles will also remain below 50 MWtlMlbmIhr.
Page 15 of 29
-Non Proprietaw Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 NRC RAI SNPB-3 The disposition of Condition 4 from the NRC SER for the IMLTR does not appear to be consistent with the condition as described in the NRC SER. The condition requires that a 0.02 adder be applied to the safety limit minimum critical power ratio (SLMCPR) as well as the single loop operation (SLO) SLMCPR. This 0.02 adder is intended to account for potentially increased uncertainties in the nuclear methods for modern fuel designs and operating strategies. Please confirm that the SLMCPR for dual loop operation (DLO) as well as SLO incorporates a 0.02 adder to account for nuclear methods uncertainties. Also confirm that sufficient margin is applied consistent with ELTRI (NEDC-32424P-A) and ELTR2 (NEDC-32423P-A) to account for SLO.
NSPM Response Although not clearly stated in the PUSAR, the application of Limitation 4 is not limited to single loop (SLO). GEH will apply the SLMCPR adder to the SLMCPR values calculated for both SLO and dual loop operation.
The SLMCPR adder is related to the IMLTR and is not part of ELTRI or ELTR2.
Attached are revised pages of the Monticello EPU PUSAR, NEDC-33322Pl which clarifies the intended implementation of Limitation 4. Revision bars indicate the changes.
Page 16 of 29
-Non Pro~rietarv NEDC-33322P, Revision?, Revised RAI SNPB-3 The Safety Limit MCPR (SLMCPR) can be affected slightly by EPU due to the flatter power distribution inherent in the increased power level. Experience has shown that the power uprate flatter power distribution results in an increase in the SLMCPR of ( 0.01. This effect is not changed by the EPU approach (Reference 1). The SLMCPR analysis reflects the actual plant core-loading pattern and is performed for each plant reload core (see Reference 4).
((
I1 I]
The calculated values will be renorted in the Sun~lemental Reload Licensing Report (SRLR) for the EPU core....- --...-...........-.---. - -.-- ----..-. --- ae,
As required by Reference 2526, -value shall be added to the cycle-specific SLMCPR value. In addition, the 0.02 adder for EPU will also be included in the single recirculation loop operation (SLO) SLMCPR methodology.
MCPR Operating Limit Per the CLTR, the EPU operating conditions have only a small effect on the MCPR Operating Limit. The MCPR Operating Limit is calculated by adding the change in MCPR due to the limiting A 0 0 event to the SLMCPR and is determined on a cycle specific basis. EPU does not change the method used to determine this limit. The effect of EPU on A 0 0 events is addressed in Section 2.8.5. Based on experience with previous plant specific power uprate submittals, the effect on the MCPR Operating Limit due to EPU is small and does not significantly affect plant operation. ((
11-I1 MAPLHGR and LHGR Operating Limits Page 17 of 29
-Non Proprietary NEDC-33322P, Revision 2, Revised RAI SNPB-3 Page 18 of 29 Number 4
5 6
7 Title SLMCPRI SLMCPR2 R-Factor ECCS-LOCA 1 Limitation Description For EPU operation, a 0.02 value shall be added to the cycle-specific SLMCPR value. This adder is applicable to SLO, which is derived from the dual loop SLMCPR value.
For operation at MELLLA+, including operation at the EPU power levels at the achievable core flow statepoint, a 0.03 value shall be added to the cycle-specific SLMCPR value.
The plant specific R-factor calculation at a bundle level will be consistent with lattice axial void conditions expected for the hot channel operating state. The plant-specific EPU/MELLLA+ application will confirm that the R-factor calculation is consistent with the hot channel axial void conditions.
For applications requesting implementation of EPU or expanded operating domains, including MELLLA+,
the small and large break ECCS-LOCA analyses will include top-peaked and mid-peaked power shape in establishing the MAPLHGR and determining the PCT. This limitation is applicable to both the licensing bases PCT and the upper bound PCT. The plant-specific applications will report the limiting small and large break licensing basis and upper bound PCTs.
Disposition As stated in Section 2.8 slJvlwk --.. 1-&*
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-- 3 0.02value shall be added to the cvcle-specific SLMCPR value. In addition, the 0.02 adder for EPU will also be included in the sinale recirculation loop operation (SLI-SLMCPR methodoloal Not applicable to EPU.
As stated in Section 2.8.2.5, the GE14 bundle R-factors are consistent with Monticello hot channel axial void conditions for EPU operation.
As stated in Section 2.8.5.6.2, the small and large ECCS-LOCA analyses conducted for Monticello EPU includes top-peaked and mid-peaked power shapes in establishing the MAPLHGR and determining both the licensing bases and upper bound PCTs. The limiting small and large break licensing basis and upper bound PCTs are reported in PUSAR Section 2.8.5.6.2 and Table 2.8-5.
-Non Proprietaw Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 NRC RAI SNPB-4 Section 2.8.2.5 provides the hot channel outlet void fractions at several operating points in the proposed MNGP expanded operating domain. The section specifies that the R-factor was generated using a void profile with an average void fraction of 60 percent and a maximum of 86 percent. Please sample several hot channels from the EPU cycle PANACI I analysis that are predicted to have low critical power ratios (CPRs) and provide a figure or table demonstrating that the generic R-factor profile conditions are representative of the limiting conditions predicted by PANACI I.
Please refer to a similar RAI on the ESBWR docket regarding this concern (RAI 4.4-68) and the accepted General Electric Hitachi Nuclear Energy Americas (GEH) response.
NSPM Response Figure SNPB-4-1 shows bundle average void fractions corresponding to hot channels with low critical power ratios (MCPRs) from the Monticello EPU/MELLLA+ core used in NEDC-33322, Rev 3. The figure demonstrates that the generic R-factor profile, with an average void fraction of 0.60, is representative of the MCPR-limiting void conditions predicted by PANACI 1.
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-Non Proprietaw Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 Figure SNPB-4-1: Bundle Average Void Fractions for Bundles Having Low CPRs Monti cello EPUIMELLLA+ Core 1.go ---------. 1 - - - - - - - - -, - - - - - - - - - I - - - - - - - - - - - - - - - -, - - - - - - - - -, - - - - - - - - - r - - - - - - - -,
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-Non Proprietaw Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 MRC RBI SNPB-5 Condition 7 of the NRC SER for the IMLTR requires that the emergency core cooling system - loss of coolant accident (ECCS-LOCA) performance analyses consider both top-peaked and mid-peaked power distributions. Section 2.8.5.6.2 states that these two power shapes were considered in the analysis. Consistent with the Condition 7, the power uprate safety analysis report (PUSAR) provides both the upper bound and licensing basis peak cladding temperatures (PCTs). Section 2.8.5.6.2 provides discussion regarding the maximum average planar linear heat generation rate (MAPLHGR) analyses that are consistent with the condition. However, to assist the staff in its review of the ECCS-LOCA performance please provide additional clarification regarding Table 2.8-5 in the PUSAR. Specifically, please clarify the power shape for each PCT listed in the table. Please also provide the accompanying PCT predicted for the other power shape to confirm that the PCTs quoted in the table are appropriate.
NSPM Response Table SNPB-5-1 provides the power shape for each PCT listed in Table 2.8-5 of the PUSAR.
Table SNPB-5-2 provides additional analysis details of other case specific power shapes and PCTs to assist the staff in its review of the ECCS-LOCA performance.
Page 21 of 29
-Non Proprietaw Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 Table SNPB-5-1 Page 22 of 29 Parameter Method Thermal Power (MWt)
Licensing Basis PCT (OF)
Cladding Oxidation (%
Original Clad Thickness)
Hydrogen Generation, Core wide Metal-Water Reaction (%)
Coolable Geometry Core Long Term Cooling Upper Bound PCT (OF)
Limiting Appendix K Large Break PCT (OF)
Limiting Nominal Large Break PCT (OF)
Limiting Appendix K Small Break PCT (OF)
Limiting Nominal Small Break PCT (OF)
Power CLTP
((
EPU SAFERIGESTR 2004
< 2140
< 9.0
< 0.2 OK OK
< 1670 2123 1310 1607 11 79 CLTP SAFERIGESTR 1775
< 2140
< 9.0
< 0.2 OK OK
< 1670 2123 1275 Bounded by EPU Bounded by EPU Shape EPU 11 90 CFR 50.46 Limit
< 2200
< 17
< 1.0 PCT< 2200 OF, and Local Oxidation <I 7%
Core flooded to TAF or Core flooded to jet pump suction elevation and at least one core spray system is operating at rated flow.
Less than LBPCT
-Non Proprietary Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16, 2009 Table SNPB-5-2 Page 23 of 29
-Non Proprietaw Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 NRC RAI SNPB-6 The disposition of Condition 14 from the NRC SER for the IMLTR is not consistent with the SER dated January 17, 2008. This SER incorporates Appendix F, which discusses the findings of the staff review of GE's Part 21 evaluation of non-conservatisms in the GSTRM thermal mechanical (T-M) methodology. The staff concludes in Appendix F that an additional margin of 350 pounds per square inch (psi) is required in the critical pressure analysis. Please confirm that this 350 psi additional margin in the critical pressure is included in the T-M safety analysis.
NSPM Response In References 1 and 2, GEH addressed the potential non-conservatism in GE Thermal Mechanical Methodology (GESTRM). Subsequent communication with the NRC staff indicated that the Part 21 concern was sufficiently addressed such that the additional 350-psi margin was no longer warranted. That position was reflected in a recent NRC approval of an EPU application that referenced the use of the IMLTR but did not apply the additional margin of 350 psi. GEH anticipates a revision to the referenced Appendix F to remove the additional margin.
Therefore, the Monticello PUSAR does not reflect the use of the additional 350-psi margin.
References:
RAI SNPB-6, Part 21 Evaluation of GSTRM
- 1. GEH letter (MFN 07-040), Part 21 Notification: Adequacy of GEH Thermal-Mechanical Methodology, GESTR-M, dated January 21,2007.
- 2. GEH letter (MFN 07-040), Part 21 Notification: Adequacy of GEH Thermal-Mechanical Methodology, GESTR-M - Supplement I, dated January 4, 2008.
Page 24 of 29
-Non Proprietary Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 NRC RAI SNPB-7 The disposition of Condition 17 is consistent with the NRC SER for the IMLTR, however, the predicted steady state bypass void fraction has not been provided in the PUSAR Appendix A. Please provide the analysis for the EPU reference core to provide reasonable assurance that this condition will be met on a cycle specific basis.
NSPM Response During discussions on January 10, 2009, the NRC clarified that the intent of the RAI is to determine the process that would be used to implement Condition 17 on a cycle-specific basis.
The best-estimate means of determining 4-channel bypass void fraction is with TRACG.
TRACG was applied in response to Methods RAI 14 (Reference I). TRACG is capable of accurately modeling bypass heating and cross flow.
A conservative approach (ISCOR) was discussed in Reference 2 and 3. ISCOR conservatively calculates hot.bypass channel voiding using its direct moderator-heating model and providing no credit for cross flow while applying additional conservatism with bounding 4-bundle peaking. The use of ISCOR is a more simplified and efficient process to implement compared to the use of TRACG and typically demonstrates margin to the 5% bypass void fraction requirement at the LPRM D Level.
For Monticello's reload core prior to EPU implementation, a calculation will be performed with the conservative ISCOR process at licensed EPU core power and minimum core flow (e.g. 120 % CLTR, 99% flow). The purpose of the calculation is to confirm that the bypass void fraction remains below 5 percent at all LPRM levels when operating at steady-state conditions within the licensed operating domain consistent with Reference 4.
If the resulting bypass void fraction is found to exceed the 5% requirement, it is acceptable to relax the conservative ISCOR input assumptions as long as the overall approach can be demonstrated to remain conservative relative to TRACG. It is also acceptable to perform a cycle-specific TRACG analysis with consideration of assumptions that will tend to maximize bypass void fraction (e.g. bypass flow and 4-bundle peaking).
The highest calculated bypass voiding at any LPRM level will be provided with the plant-specific SRLR.
Page 25 of 29
9 Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16, 2009
References:
NRC RAI SNPB-7, Steady-State Bypass Void
- Methods Interim Process," (MFN 05-038) dated May 3, 2005.
- 2. NRC letter, T. Blount (NRC) to R. Brown (GEH), Final Safety Evaluation for GE-Hitachi Nuclear Energy Americas, LLC (GHNE) Topical Report (TR) NEDC-33006P, "Maximum Load Line Limit Analysis PluslW(MFN 07-517) dated September 17, 2007.
- 3. GEH Letter, G. Stramback (GEH) to NRC, " Completion of Responses to MELLLA Plus A 0 0 RAls," (MFN 04-026) dated March 4, 2004.
NEDC-33173P," (MFN 08-693) dated September 18, 2008.
Page 26 of 29
-Non Proprietaw Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 NRC RAI SMPB-8 The disposition of Condition 18 is not sufficiently detailed for the staff to ensure, consistency between the disposition and the staff's SER for the IMLTR. The staff reviewed a description of the operating limit minimum CPR (OLMCPR) used in the oscillation power range monitor (OPRM) setpoint analysis, however, requires some additional clarification based on the statements in Section 2.8.3.1. Please clarify the final paragraph of Section 2.8.3.1 regarding the illustrative analysis and confirm that the OPRM setpoints are based on an OLMCPR that does not incorporate the OLMCPR penalty (0.01 adder based on the void fraction uncertainty). Please refer to the guidance in MFN 08-693 regarding disposition of this condition.
NSPM Response The final paragraph of Section 2.8.3.1 provides an illustrative analysis for the OPRM setpoint of the application of the 5% calibration error required by Limitation 18 of the NRC's SER for the NEDC-33193P (IMLTR). As stated in GEH letter dated September 18, 2008, MFN 08-693, the illustrative example for the OPRM setpoint does not incorporate the 0.01 OLMCPR adder for void fraction uncertainty required by IMLTR Limitation 19. The actual OPRM setpoint for Monticello's EPU operating cycle will also not incorporate the OLMCPR adder consistent with MFN 08-693.
Page 27 of 29
-Non Proprietaw Response To Reactor Systems Branch & Nuclear Performance and Code Review Branch RAls dated January 16,2009 NRC SNPB-9 The PUSAR states that, on a cycle specific basis, the SLMCPR is calculated at the rated condition, the minimum flow point along the LTPL, and the SLO statepoint. The staff notes that Monticello is a BWW3. Please provide and justify the values for the feedwater flow and core flow uncertainties that are used for the rated, minimum flow, and SLO SLMCPR analyses. Please refer to a similar RAI on the docket for the staff review of LTR NEDC-33006P regarding this concern (RAI 17) and Figure 17-3 from the response documented in GEH Letter (MFN 07-041) dated January 25,2007.
MSPM Response The feedwater flow and core flow uncertainties are consistent with GEH LTR, NEDC-32601 P-A, "Methodology and Uncertainties for Safety Limit MCPR Evaluations." Table 2.1 of NEDC-32601 P-A provides a summary of the uncertainties used in the MCPR analysis. Table 2.1 of NEDC-32601 P-A also is applicable to feedwater flow uncertainty used in the SLO calculations.
11-A revision to page 2-240 of the PUSAR is attached which eliminates the extraneous information regarding the statepoints associated with SLMCPR calculations.
GEH letter (MFN 07-041), dated January 25, 2007, is related to the NEDC-33006P, Maximum Extended Load Line Limit Plus," (MELLLA+). The minimum-flow stated point for rated power at MELLLA+ conditions might be as low as 80%, compared to 99% for rated power at EPU conditions. Therefore, the response to RAI 17 for MELLLA+ is not applicable to EPU conditions.
Page 28 of 29
-Non Proprietaw NEDC-33322P The Safety Limit MCPR (SLMCPR) can be affected slightly by EPU due to the flatter power distribution inherent in the increased power level. Experience has shown that the power uprate flatter power distribution results in an increase in the SLMCPR of 5 0.01. This effect is not changed by the EPU approach (Reference I).
The SLMCPR analysis reflects the actual plant core-loading pattern and is performed for each plant reload core (see Reference 4). ((
I] The calculated values will be reported in the Supplemental Reload Licensing Report (SRLR) for the EPU core. The SLMCPR will Include a 0.02 adder for SLO as required by Reference 25.
2.8.2.3.2 MCPR Operating Limit
((
Per the CLTR, the EPU operating conditions have only a small effect on the MCPR Operating Limit. The MCPR Operating Limit is calculated by adding the change in MCPR due to the limiting A 0 0 event to the SLMCPR and is determined on a cycle specific basis. EPU does not change the method used to determine this limit. The effect of EPU on A 0 0 events is addressed in Section 2.8.5. Based on experience with previous plant specific power uprate submittals, the effect on the MCPR Operating Limit due to EPU is small and does not significantly affect plant operation. ((