ML092090219

Monticello Extended Power Uprate: Response to NRC Reactor Systems Review Branch and Nuclear Code and Performance Review Branch Request for Additional Information (RAI) Dated March 23, 2009 and Nuclear Code and Performance Review Branch Requ

ML092090219
Person / Time
Site:	Monticello
Issue date:	07/23/2009
From:	O'Connor T J Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-09-049, TAC MD9990
Download: ML092090219 (169)

Text

{{#Wiki_filter:XcelEnergy-WITHHOLD ENCLOSURE 5 FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 and 9.17 July 23, 2009 L-MT-09-049 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 Review Branch and Nuclear Code and Performance Review Branch Request for Additional Information (RAI) dated March 23, 2009 and Nuclear Code and Performance Review Branch Request for Additional Information dated April 27, 2009 (TAC No. MD9990)

References:

1. NSPM letter to NRC, License Amendment Request: Extended Power Uprate (L-MT-08-052) dated November 5, 2008, Accession No. ML0832301 11 2. Email P. Tam (NRC) to G. Salamon, K. Pointer (NSPM) dated March 23, 2009, "Monticello

-Draft RAI from Balance of Plant re: proposed EPU amendment (TAC MD9990)" Accession No. ML090820145

3. NSPM letter to NRC, Monticello Extended Power Uprate: Response to NRC Reactor Systems Branch and Nuclear Performance

& Code Review Branch Request for Additional Information (RAI), L-MT-09-017, dated January 16, 2009, (TAC MD9990) Accession No. ML090790388

4. NSPM letter to NRC, Response to NRC Reactor Systems Branch and Nuclear Performance

& Code Review Branch Request for Additional Information (RAI), L-MT-09-025, dated February 23, 2009 (TAC No.MD9990) Accession No. ML091130636

5. Email P. Tam (NRC) to G. Salamon, K. Pointer (NSPM) dated April 27, 2009, "Additional Draft Questions from the Nuclear Performance and Code Review Branch re: EPU Amendment (TAC MD9990)" Accession No. ML091170480 Monticello Nuclear Generating Plant 2807 West County Road 75 9 Monticello MN 55362 M U.S. Nuclear Regulatory Commission L-MT-09-049 Page 2 of 3 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.On March 23, 2009 the U. S. Nuclear Regulatory Commission (NRC) Reactor Systems Branch (SRXB) and Nuclear Code and Performance Review Branch (SNPB) provided sixty RAIs in Reference

2. NSPM had previously addressed ten SRXB RAIs and nine SNPB RAIs listed in Reference 2 in NSPM letter L-MT-09-017 (Reference

3) dated March 19, 2009 (Accession No. ML090790388).

Five additional SRXB RAls listed in Reference 2 were addressed in Reference 4 (Accession No. ML091130636). On April 27, 2009, the SNPB provided three additional RAIs in Reference 5 Enclosure 1 contains the non proprietary version of NSPM's responses to the SXRB RAIs listed in Reference 2 not addressed by NSPM previous correspondence. Enclosure 2 provides the NSPM response to the additional RAIs provided by the SNPB contained in Reference 5.Enclosure 3 contains the following reference documents:

• Supplemental Reload Licensing Report (SRLR) For Monticello Reload 24 Cycle 25, Rev. 1* NEDO -23842: Continuous Control Rod Withdrawal Transient In Startup Range Enclosure 5 contains the proprietary version of information provided in Enclosure

1. GEH requests this proprietary information to 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 4.In accordance with 10 CFR 50.91, a copy of this letter is being provided to the designated Minnesota Official without the proprietary version.Summary of Commitments This letter makes no new commitments and does not affect any existing commitments. U.S. Nuclear Regulatory Commission L-MT-09-049 Page 3 of 3 I declare under penalty of perjury that the foregoing is true and correct.Executed on July,2_, 2009.Timothy J. O'Connor Site Vice President, Monticello Nuclear Generating Plant Northern States Power Company -Minnesota Enclosures cc: Administrator, Region III, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC Minnesota Department of Commerce ENCLOSURE 4 GENERAL ELECTRIC HITACHI (GEH)AFFIDAVIT (3 PAGES FOLLOW) GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, James F. Harrison, state as follows: (1) I am Vice President, Fuel 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-AEP-1314 RI, GEH Responses to NRC RAIs -Reactor System Branch, Round 2, dated June 19, 2009.The proprietary information in Enclosure 1 entitled, GEH Responses to NRC RAIs -Reactor System Branch, Round 2, is identified by a dotted underline inside double square brackets, This sentence is 'a 3.. In each case, the superscript notation ý3) refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination. (3) In making 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-(DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC 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 GEH's competitors without license from 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 GE-MNGP-AEP-1314 R1 Affidavit Page I 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 information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GEH, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties, 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 information, and the subsequent steps taken 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 acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GEH. Access to such documents within GEH is limited on a"need to know" basis.(7) The procedure for approval of external 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, suppliers, 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 information 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 supporting analysis techniques and information, and their application to the design, modification, and processes were achieved at a significant cost to GEH.The development of the evaluation methodology along with the interpretation and application of the analytical results is derived from 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 harm to GEH's competitive position and foreclose or reduce the availability of profit-making 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 GE-MNGP-AEP-1314 R1 Affidavit Page 2 of 3 the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.The research, 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 are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive 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 information 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 opportunity to exercise its competitive advantage to seek an adequate return 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 true and correct to the best of my knowledge, information, and belief.Executed on this 19th day of June 2009.James F. Harrison Vice President, Fuel Licensing GE-Hitachi Nuclear Energy Americas LLC Affidavit for GE-MNGP-AEP-1314 Ri Affidavit Page 3 of 3 ENCLOSURE I NSPM RESPONSE TO SRXB RAIs DATED MARCH 23, 2009 NON PROPRIETARY L-MT-09-049 Enclosure 1 Non Proprietary Page 1 of 48 SRXB RAI No. 2.8.1-1 Address compliance with fuel-dependent limitations on discharge burnup. Provide a comparison of EPU-predicted discharge burnups to those for the current licensed thermal power level and the applicable licensing limits.NSPM Response The End of Cycle (EOC) 24 average discharge exposure is predicted to be 44904.5 MWd/MT. The EPU core average discharge exposure is predicted to be 45693.9 MWd/MT. GE14 fuel is required to have Batch Average Discharge Exposure less than 50000.0 MWd/MT. This is in compliance with the fuel-dependent limitations on discharge burnup. L-MT-09-049 Enclosure 1 Non Proprietary Page 2 of 48 SRXB RAI No. 2.8.1-2 Address compliance with fuel-dependent and Technical Specification limits on U-235 isotopic enrichment. Provide a comparison of EPU-predicted fuel enrichment to that for a fuel loading at the current licensed thermal power level.NSPM Response Cycle 24 weighted average fresh bundle enrichment is predicted to be 3.92%. The predicted Cycle 25 weighted average fresh bundle enrichment was 3.90%. The EPU core weighted average fresh bundle enrichment is predicted to be 4.11%. Maximum licensed GE14 bundle enrichment is 5.0%. This is in compliance with fuel-dependent and Technical Specifications limits on U-235 isotopic enrichment. L-MT-09-049 Enclosure 1 Non Proprietary Page 3 of 48 SRXB RAI No. 2.8.2-1 Provide the predicted reload batch fractions for each of the EPU transition cycle and compare to the most recent operating cycle and the EPU equilibrium design.NSPM Response The reload batch fraction for Cycle 24, Cycle 25, and the EPU core is 0.31, 0.34, and 0.36 respectively. L-MT-09-049 Enclosure 1 Non Proprietary Page 4 of 48 SRXB RAI No. 2.8.2-2 Please provide additional information to support NSPM's invocation of the disposition regarding the thermal limits margin monitoring threshold in light of the fact that the referenced plant is likely of a different design class and vessel size than MNGP.Particularly, address the impact of not monitoring margin to thermal limits during transients or accidents, if any, that have limiting consequences when initiated in the 20-25-percent power range. For example, how would the consequences of a rod withdrawal error in this power range be evaluated? NSPM Response The fuel thermal limits margin monitoring threshold is a fuel bundle requirement that is based on the core average bundle power so the plant design class and vessel size have no impact on the monitoring threshold. The original plant operating licenses set this monitoring threshold at a typical value of 25 percent of rated power. In order to determine the maximum bundle average power allowed by the monitoring threshold, the highest power density plant operating at original licensed thermal power was chosen.Proprietary Information Withheld 1]The NRC staff agreed with this method in the Constant Pressure Power Uprate LTR, NEDC-33004P-A. For MNGP the calculated average power density at 25 percent of the uprated power level is 1.0 MWt. The basis for not monitoring thermal limits below this threshold is the very large margin to critical power as described in the Technical Specification bases, Section 2.1 Safety Limits. Therefore, with these large margins, there are no transients that have limiting consequences when initiated from the 20 -25 percent power range. L-MT-09-049 Enclosure 1 Non Proprietary Page 5 of 48 SRXB RAI No. 2.8.2-3 The EPU license amendment request proposes to eliminate a 1600°F limitation on the Upper Bound Peak Cladding Temperature. Address the changes that the methodology relaxation will have on the MAPLHGR limits, and what impact these changes will have on other operating limits as power distribution limit relaxations propagate through the AOO analyses.NSPM Response In previous operation, a maximum average planar linear heat generation rate (MAPLHGR) setdown was imposed on the MNGP Plant core to comply with the 1600°F limitation on the Upper Bound Peak Cladding Temperature. With the removal of the limitation, as allowed by the NRC via Reference 2.8.2-3-1, this MAPLHGR setdown is no longer required, allowing the MAPLHGR value to be set as determined by fuel operation limits. Application of MAPLHGR limits used by other safety analyses (such as Anticipated Operational Occurrence (AOO) analyses, noted in the RAI statement) would similarly reference from MAPLHGR value as determined by fuel limits, and be consistently applied.Reference 2.8.2-3-1 Stuart A. Richards (NRC) to James F. Klapproth (GENE), Review of NEDE-23785P, Vol. III, Supplement 1, Revision 1, "GESTR-LOCA and SAFER Models for Evaluation of Loss-of-Coolant Accident Volume III, Supplement 1, Additional Information for Upper Bound PCT Calculation," (TAC No. MB2774), February 1, 2002. L-MT-09-049 Enclosure 1 Non Proprietary Page 6 of 48 SRXB RAI No. 2.8.2-4 Provide or otherwise make available to the NRC staff confirmation of the CLTR dispositions relative to the first EPU transition cycle regarding the OLMCPR, MAPLHGR and LHGR operating limits by providing supplemental reload licensing reports for the current cycle and the first EPU implementation cycle.NSPM Response The current Cycle 25 Supplemental Reload Licensing Support (SRLR) is included in Enclosure 3, however, the EPU SRLR will not be available prior to November 2009. In order to demonstrate the effect EPU may have on the limiting transients, analyses have been performed with the EPU equilibrium core. The uncorrected ACPR values are available for various transients from the EPU analysis for the equilibrium core, but are limited to EPU MELLLA conditions, which are most like the Hard Bottom Burn (HBB)power shape strategy. Since the EPU MELLLA flow rate is 99% of rated flow, the closest comparison between Cycle 25 and EPU would be between the Standard (HBB)case reported in the Cycle 25 SRLR which is at 100% rated flow and the EPU MELLLA case. The comparison is shown below and it supports the CLTR dispositions that the effect on transients is small.Table 2.8.2-4-1 GE14C, Cycle 25, GE14C, EPU Equilibrium STANDARD (HBB) Cycle, MELLLA 100% CLTP, 100% EPU, 100% Rated Flow 99% Rated Flow Event Uncorrected ACPR Uncorrected ACPR FW Controller Failure 0.34 0.35 Turbine Trip with Bypass* 0.35 0.35 Turbine Trip w/o Bypass 0.31 0.32 Load Rejection w/o Bypass 0.30 0.28 Inadvertent HPCI /L8 0.36 0.36* The Turbine Trip with Bypass is based on a degraded scram speed compared to the other transients. See Section 2.8.2.3 of the EPU PUSAR (NEDC 33322P, Rev. 3) for additional information on thermal limits. L-MT-09-049 Enclosure 1 Non Proprietary Page 7 of 48 SRXB RAI No. 2.8.3-1 NEDC-33322P refers to a demonstration analysis performed to determine the limiting OPRM setpoint that results in a non-OLMCPR setting stability transient. Discuss whether the initial conditions in this analysis are the same as those used in any analyses supporting the Option IIl license amendment request. Address any differences. NSPM Response The Option III license amendment analysis was performed at EPU conditions. The oscillation power range monitor (OPRM) amplitude setpoint calculation consists of three components, Hot Channel Oscillation Magnitude (HCOM), Delta CPR over Initial CPR vs Oscillation Magnitude (DIVOM) and Two Pump Trip Delta CPR over Initial CPR (DIRPT) analyses. HCOM is the same for both CLTP and EPU. The DIVOM analysis is performed at the same off-rated condition so it is only mildly impacted by differences in the power shape. The DIRPT will increase slightly going from CLTP to EPU due to a lower initial minimum critical power ratio (MCPR) at EPU conditions. L-MT-09-049 Enclosure 1 Non Proprietary Page 8 of 48 SRXB RAI No. 2.8.3-2 What version of emergency operating procedures is currently implemented at MNGP?Provide a short description of the process used to ensure that the EPG variables (e.g.Hot Shutdown Boron Weight (HSBW) and Heat Capacity Temperature Limit (HCTL))are adequate under CPPU conditions. NSPM Response The Emergency Operating Procedures (EOPs) have been implemented in accordance with BWROG Emergency Procedure Guidelines (EPG)/Severe Accident Guidelines (SAG) Rev. 2.The EPG variables are verified to be adequate by EOP Calculations. EOP action levels and limits are calculated in accordance with the BWROG methodology provided in EPG Appendix C. Calculations are performed in accordance with the MNGP process defined by Fleet Procedure FP-E-CAL-01 (CALCULATIONS). Cycle-specific calculations are performed for each cycle of operation to account for the fuel design and loading.Changes to EPG variables are evaluated and implemented with changes to the EOP procedures as appropriate. L-MT-09-049 Enclosure 1 Non Proprietary Page 9 of 48 SRXB RAI No. 2.8.3-3 Provide a short description of how the Stability Mitigation Actions (e.g. immediate water level reduction and early boron injection) are implemented at MNGP. Does operation at CPPU conditions require modification of any operator instructions? NSPM Response The stability mitigation actions are implemented in accordance with the Emergency Operating Procedures. These include C.5-1100, "RPV Control" and C.5-2007, "Failure to Scram." A copy of C.5-2007 follows this RAI response.A short description of the key EOP actions follows. It is assumed that the basic condition is a core instability event with power oscillations that is not accompanied by complicating factors such as a commensurate loss of RPV inventory or RPV water level indication or a condition that causes a large drywell pressure increase.Level Control* Reduce RPV water level to -33" (Prompt level reduction, ADS will be inhibited)" If necessary to reduce power further, allow RPV water level to decrease to x, where x >-126"(Top of Active Fuel) and x < -33" Power Control (These actions are done concurrently with Level Control.)* Run back recirculation flow and then trip the recirculation pumps* Actuate ATWS* If peak to peak power oscillations are above the large oscillation threshold of 25%, start Standby Liquid Control (manual boron injection) Pressure Control (These actions are done concurrently with Level Control)* Open MSIVs and stabilize reactor pressure via the turbine bypass valves" Use the SRVs or other pressure control methods if necessary Assuming the level decrease and boron injection are successful in reducing power to below 3% or terminating the torus heatup (SRVs closed and drywell pressure below the scram setpoint), the following actions would occur.* Maintain RPV water level between a control band of x and -149 inches (Minimum Steam Cooling RPV Water Level) using only preferred ATWS injection systems" Inject the Hot Shutdown Boron Weight L-MT-09-049 Enclosure 1 Non Proprietary Page 10 of 48* After injection of the Hot Shutdown Boron Weight, restore RPV water level 9 to 48 inches and avoid rapid changes in flow rate After the Cold Shutdown Boron Weight is injected, a controlled depressurization can occur and the reactor taken to cold shutdown conditions using Shutdown Cooling.EPU does not require a modification to any of these actions. THIS PAGE IS AN ,OVERSIZED DRAWING OR FIGURE, THAT CAN BE VIEWED AT THE RECORD TITLED: "FAILURE TO SCRAM, C.5-2007,REV. 15." WITHIN THIS PACKAGE...OR BY SEARCHING USING THE DOCUMENT/REPORT NO.D-01 L-MT-09-049 Enclosure 1 Non Proprietary Page 11 of 48 SRXB RAI No.2.8.3-4 What is the current status of LTSS Option III implementation? When will it be armed in the plant?NSPM Response The OPRM-based Option III LTS equipment was installed in the plant as part of the PRNMS modification at MNGP. Both OPRM trip outputs will be disabled during the OPRM monitoring and evaluation period. The period extends from the startup following PRNM system installation to 90 days of steady-state operation after reaching full power.The monitoring period is described in Section 5.1.2 of Enclosure 1 ofthe MNGP PRNM licensing amendment request dated February 6, 2008 and in NSPM letter dated November 6, 2008. It is currently scheduled to be armed on August 31, 2009.The OPRM arming (not bypassed) requirements are included in Section 3.3.1.1 of the MNGP Technical Specifications. For additional information, please see the SER for PRNMS dated January 30, 2009 (TAC No. MD8064). L-MT-09-049 Enclosure 1 Non Proprietary Page 12 of 48 SRXB RAI No. 2.8.3-5 Is Option III hardware implemented in the Monticello simulator? What are the plans and overall schedule for operator training?NSPM Response The Option III hardware has been implemented in the MNGP simulator, and all the operating crews have been trained on OPRM operation. L-MT-09-049 Enclosure 1 Non Proprietary Page 13 of 48 SRXB RAI No. 2.8.3-6 Will the Option III hardware implemented in Monticello have the DSS/CD software installed for testing purposes? What are the testing plans?NSPM Response All five stability algorithms are currently operating in all four APRMs and providing indications and alarms. The CDA software has been installed, tested, and providing data, but the associated stability trips to the Reactor Protection System (RPS) have been disabled until MNGP is licensed for EPU and MELLLA+ operation. See the response to RAI No. 2.8.3-4 for the OPRM monitoring period.The algorithms were tested by MNGP during Factory Acceptance Testing using a full core simulator provided by GEH. GEH also performed software verification and validation testing. The PRNM test plan at the MNGP included construction, pre-operation, and operational testing including technical specification surveillances. L-MT-09-049 Enclosure 1 Non Proprietary Page 14 of 48 SRXB RAI No. 2.8.3-7 Will the DIVOM curve be implemented as cycle-specific in Monticello? If the generic DIVOM slope will not be used, provide a reference to the DIVOM analysis methodology that will be used.NSPM Response The DIVOM slope will be evaluated on a cycle-specific basis per the BWROG Regional DIVOM Guideline (GE-NE-0000-0028-9714-R1, "Plant-Specific Regional Mode DIVOM Procedure Guideline," June 2005). It will be limited to not less than the generic DIVOM slope of 0.45 as prescribed in NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996. L-MT-09-049 Enclosure 1 Non Proprietary Page 15 of 48 SRXB RAI No. 2.8.3-8 The Nine Mile Point 2 Instability Event showed some reduced sensitivity to low-level oscillations if the Option III parameters were set at minimum sensitivity settings. Have the lessons-learned from this event been incorporated in Monticello? NSPM Response The Nine Mile Point 2 (NMP2) event showed that while the OPRM system resulted in an effective scram, there were a large number of successive confirmation count (SCC)resets. At NMP2, the adjustable period confirmation variables were set at 50 msec for the period tolerance and 3.0 Hz for the cutoff frequency. An evaluation by GEH concluded that the second-order Butterworth filter with a cutoff frequency of 3 Hz was ineffective in removing some residual high-frequency noise from the oscillation signal and lead to numerous SCC resets. This evaluation resulted in a safety communication, SC03-20, "Stability Option III Period Based Detection Algorithm Allowable Settings", October 2003, in which GEH recommended that a period tolerance of 100 msec or higher and a cutoff frequency of 1.0 Hz be adopted. These settings were incorporated in the MNGP EPU stability analysis and were implemented at MNGP. L-MT-09-049 Enclosure 1 Non Proprietary Page 16 of 48 SRXB RAI No. 2.8.3-9 In September 2006, the Hope Creek plant experienced a half-scram indication from the Option III hardware while withdrawing peripheral control rods in low-power bundles.Hope Creek implemented recommendations for speed of rod withdrawal inside the armed region. Have these recommendations been incorporated in the Monticello operator training?NSPM Response The Hope Creek (HC) event is captured in INPO OE23808 and has been recently evaluated by MNGP. The MNGP evaluation determined that no rod withdrawal operator training is necessary. The HC OPRM setpoint of 1.06 as discussed in the Operating Event description is significantly below the MNGP amplitude setpoint of 1.15, and sufficient trip avoidance margin exists for the MNGP instrument. In addition, cognizant HC personnel have indicated to MNGP that the HC setpoint has been raised to 1.11, and the rod withdrawal speed recommendations of the subject OE are no longer performed. L-MT-09-049 Enclosure 1 Non Proprietary Page 17 of 48 SRXB RAI No. 2.8.3-10 Assuming a conservative OPRM setpoint of 1.15, provide the hot-spot fuel temperature as function of time before the scram. Evaluate this fuel temperature oscillation against pellet-clad interaction (PCI) limits. Assume the steady-state fuel conditions before the oscillations are those of point C of Figure 2.8.3-20 of NED33322P (the highest power point in the BSP scram region).NSPM Response The current licensing criteria applicable to RAI No. 2.8.5.1-2 are [[ Proprietary Information Withheldg]. These criteria also apply to RAI No. 2.8.3-10. Additionally, the current licensing criterion that cladding fatigue life usage be less than or equal to 1.0 applies to RAI 2.8.3-10. These criteria are addressed in this response. This response also addresses-the issue of the potential for increased pellet-cladding interaction (PCI)raised in RAI Nos. 2.8.3-10 and 2.8.5.1-2. Since design or licensing criteria for PCI currently do not exist, the issue is addressed qualitatively in terms of impact on reliability. A thermal-mechanical based power-exposure limits envelope is specified Proprietary Information Withheld The LHGR limits are specified to assure compliance with several primary fuel rod thermal-mechanical licensing criteria; these criteria address fuel centerline temperature, [[Proprietary Information Withhe/ld], and fuel rod internal pressure. [[Proprietary Information Withheld A major GNF fuel rod design objective is to specify the LHGR limits curves to achieve balanced margins and a balanced design with high reliability over the rod lifetime.During the core design process, a specified margin is typically maintained between the LHGR limits and the anticipated operation for each bundle. Operation under power uprate conditions will result in more rods in some bundles operating near the specified margin for a larger fraction of the bundle lifetime, thus increasing the potential for fuel failure. The potential for increased failure under power uprate conditions specified in Reference 2.8.3-10-1 is assessed in terms of available GNF operational experience and experimental information below. The impact of power uprate on thermal-mechanical licensing analyses for the GE14 fuel design is also discussed below. L-MT-09-049 Enclosure 1 Non Proprietary Page 18 of 48 Proprietary Information Withheld]] The results from this Severe Power Ramp testing, as compared to the LHGR limits curves for the fuel designs noted above, are also provided in Figure 2.8.3-10--i. It is observed from Figure 2.8.3-10--1 that significant margin exists to the apparent failure threshold represented by the available ramp test results. In addition to barrier fuel's resistance to ramping, ramp rates at power uprate conditions versus non-power uprate conditions are not appreciably different. Thus it is judged that the possible increased cladding mechanical duty associated with operation under power uprate conditions will have negligible impact on the reliability of GNF fuel. It is further noted that the margin to failure is reasonably well-balanced over the entire exposure range, consistent with the design objective noted above.In addition to possible increased fuel duty, other possible effects of power uprate are small changes in core conditions such as increased coolant pressure (and temperature) and changes in flow conditions. [[ Proprietary Information Withheld I]Specific questions relative to the MNGP loss of feedwater heating (LFWH) transient and instability oscillations for EPU are addressed below.The current licensing criteria Proprietary Information Withheld are satisfied for the LFWH transient indicated in RAI No. 2.8.5.1-2. These criteria are based upon preventing wide spread cladding failures during the transient. For instability oscillations indicated in RAI No. 2.8.3-10, the incremental fatigue usage due to the oscillations is negligible in an absolute sense and relative to the margin to the limit (1.0) calculated for the cyclic loading assumed in the fuel rod thermal-mechanical licensing analyses. This criterion is based upon preventing wide spread cladding fatigue failures during normal operation. For MNGP, the peak maximum fraction of linear power density (MFLPD) for the LFWH transient is [[ Proprietary Information Withheld] ]], which is below the apparent failure threshold in Figure 2.8.3-10--1. This value also bounds the peak MFLPD experienced during instability oscillations. These results indicate that the LFWH transient and instability oscillations will have negligible impact on fuel reliability. L-MT-09-049 Enclosure 1 Non Proprietary Page 19 of 48 In summary, on the basis of the generic licensing analyses and the specific analyses to address operation under power uprate conditions summarized above, it is concluded that the [[Proprietary Information Withheldl] fuel designs are fully compliant with existing licensing requirements for operation under power uprate conditions. Based upon available operational experience and experimental data, it is also concluded that operation under power uprate conditions will not significantly affect GNF fuel reliability. References 2.8.3-10-1. 2.8.3-10-2. 'Dresden and Quad Cities Extended Power Uprate', GE-NE-A22-00103-02 Revision 0, August 2000.H. Sakurai, et. al., 'Irradiation Characteristics of High Burnup BWR Fuels', paper presented at the ANS Light Water Reactor Fuel Performance Conference held at Park City, Utah, April 10-13, 2000. L-MT-09-049 Enclosure 1 Non Proprietary Page 20 of 48[[Figure 2.8.3-10-1 LHGR Limits and Severe Ramp Test Failure Data Proorietarv Information Withheld 1] L-MT-09-049 Enclosure 1 Non Proprietary Page 21 of 48 SRXB RAI No. 2.8.3-11 Provide the following information relevant to ATWS-stability: (1) turbine bypass capacity; (2) percent of feedwater (FW) flow that is driven by electric or steam turbine pumps; (3) location of the extraction steam that feeds the feedwater heaters, (4)location of the extraction steam that feeds the FW steam- driven pumps (if any); (5) FW sparger elevation with respect to top of active fuel; (6) location of the SLC injection point in the vessel.NSPM Response 1) The absolute bypass flow remains at 967,440 Ibm/hr as stated in PUSAR Section 2.4.1.2. The bypass capacity is 11.6% of EPU rated reactor steam flow of 8,335,000 Ibm/hr.2) 100% of Feedwater flow is driven by electric motor driven pumps. MNGP does not have steam turbine pumps.3) The extraction points for feedwater heaters are shown below.FW Heater Extraction Point 11A/B 12th stage- LP turbine 12A/B 10th stage- LP turbine 13A/B 8 th stage- LP turbine 14A/B 7 th stage- LP Turbine 15A/B 6 th stage-HP Turbine exhaust 4) Not Applicable, See response to 2) above.5) The center line of the FW sparger is at vessel elevation 466 inches (instrument level -11.5") and the top of active fuel is at 351.5 inches (instrument level -126 inches) for a difference of 114.5 inches. Instrument 0 corresponds to vessel elevation 477.5". See attached drawing.6) Standby Liquid Control injects through the core plate DP nozzle which is 6 ft-3 inches above the inside bottom of the vessel. See attached Figure 9 from the MNGP Operations Manual. THIS PAGE IS AN OVERSIZED DRAWING OR FIGURE, THAT CAN BE VIEWED AT THE RECORD TITLED: "MONTICELLO NUCLEAR GENERATING PLANT REACTOR VESSEL & INTERNALS, DRAWING NO. NX-7831-197-1" WITHIN THIS PACKAGE...OR BY SEARCHING USING THE DOCUMENT/REPORT NO.D-02 L-MT-09-049 Enclosure 1 Non Proprietary Page 22 of 48 MONTICELLO NUCLEAR GENERATING PLANT Ops Man B.03.05-06 Revision 3 Page 11 of 11 Figure 9 Standby Liquid Control And Differential Piping AV'TAP PrnnrIaas ThrOUG**~ C~ARG PI.51 1,0 I a, C..O"L C'fu.RKw.Ap C..C ,m, I/icc L-MT-09-049 Enclosure 1 Non Proprietary Page 23 of 48 SRXB RAI No. 2.8.3-12 Following a turbine trip with full bypass and failure to scram, provide (1) the maximum FW flow that the available pumps can deliver and (2) the ultimate FW temperature after the FW heaters reach equilibrium with the new steam extraction conditions. NSPM Response Response to Part I In these conditions, the maximum FW flow would be less than rated EPU FW flow. The pressure increase from the turbine trip will reduce the FW flow capability. Also, the recirculation pump trip that occurs reduces power and steam flow rate, which also reduces the FW flow demand.Response to Part 2 For this event, it is assumed that all FW heating is lost and that the FW temperature would ultimately end up near the condenser temperature of approximately 80 0 F. L-MT-09-049 Enclosure 1 Non Proprietary Page 24 of 48 SRXB RAI No. 2.8.3-13 Discuss any control system actions that are relevant to ATWS-stability events.Examples are: automatic switching of extraction [steam] for steam driven pumps, flow runback on high pressure.NSPM Response The MNGP control system actions relevant to ATWS instability events include the pressure control system response to maintain turbine inlet pressure and the feedwater control system response to control reactor water level. The key automatic actions include the two recirculation pump trip upon vessel dome pressure reaching the high pressure ATWS RPT setpoint and opening of the safety/relief valves to control vessel pressure. Alternate Rod Insertion (ARI) is not credited in the ATwS analysis. MNGP does not have steam driven feedwater pumps or automatic feedwater runback on high pressure.As stated in Section 9.3.3 of NEDC-33004P, "Constant Pressure Power Uprate" the operator actions contained in NEDO-32047-A, "ATWS Rule Issues Relative to BWR Core Thermal-Hydraulic Stability" and NEDO-32164, "Mitigation of BWR Core Thermal Hydraulic Instabilities in ATWS" effectively mitigate an ATWS instability event at EPU conditions. These operator actions are applicable to MNGP. At MNGP, operators would promptly reduce feedwater flow, lower reactor water level and initiate SLCS injection to mitigate an ATWS instability event. L-MT-09-049 Enclosure 1 Non Proprietary Page 25 of 48 SRXB RAI No. 2.8.4.5-1 Provide plots for the limiting ATWS event (for both EPU and CLTP conditions) similar to the MSIVF plots (Figure 2.8-21 of the NEDC-33322P, Revision 1) which show the bottom vessel pressure and indicate when the Standby Liquid Control System starts injecting. NSPM Response For MNGP, the limiting ATVVS event in terms of vessel overpressure is the Pressure Regulator Failure Open (PRFO) at Beginning-Of-Cycle (BOC) exposure for CLTP and EPU conditions. The plant response to this event, including the bottom vessel pressure response, is shown in Figure 1 for CLTP and Figure 2 for EPU.In Figures 2.8.4.5-1-1 and 2.8.4.5-1-2, SLCS is initiated at the time of the high pressure ATWS RPT plus 120 seconds operator action time. SLCS is initiated at 154 seconds for CLTP (Figure 2.8.4.5-1) and 165 seconds for EPU (Figure 2.8.4.5-2). Following SLCS initiation, the SLCS injection flow reaches the lower plenum following a 60-second transport time. L-MT-09-049 Enclosure 1 Non Proprietary Page 26 of 48 Proprietary Information Withheld Figure 2.8.4.5-1 -1a: Plant Response to PRFO at CLTP (106% OLTP 182% Core Flow at BOC) L-MT-09-049 Enclosure 1 Non Proprietary Page 27 of 48[[Proorietarv Information Withheld Figure 2.8.4.5-1-1b: Plant Response to PRFO at CLTP (106% OLTP /82% Core Flow at BOC) L-MT-09-049 Enclosure 1 Non Proprietary Page 28 of 48[[Proprietary Information Withheld Figure 2.8.4.5-1-1c: Plant Response to PRFO at CLTP (106% OLTP / 82% Core Flow at BOC) L-MT-09-049 Enclosure 1 Non Proprietary Page 29 of 48 Proprietary Information Withheld 1]Figure 2.8.4.5-1-2a: Plant Response to PRFO at EPU (122% OLTP / 99% Core Flow at BOC) L-MT-09-049 Enclosure 1 Non Proprietary Page 30 of 48[[ProDrietarv Information Withheld Figure 2.8.4.5-1-2b: Plant Response to PRFO at EPU (122% OLTP 199% Core Flow at BOC) L-MT-09-049 Enclosure 1 Non Proprietary Page 31 of 48 ProDrietarv Information Withheld Figure 2.8.4.5-1-2c: Plant Response to PRFO at EPU (122% OLTP 1 99% Core Flow at BOC) L-MT-09-049 Enclosure 1 Non Proprietary Page 32 of 48 SRXB RAI No. 2.8.5.1-1 What methods and analytic codes will be used to evaluate the loss of feedwater heating for EPU reload licensing analyses at MNGP?NSPM Response EPU loss of feedwater heating cycle-specific reload licensing analyses will be performed with the NRC approved methods described in the GESTAR II LTR. The computer code used to evaluate the loss of feedwater heating is PANACEA, as documented in NEDE-30130P-A. L-MT-09-049 Enclosure 1 Non Proprietary Page 33 of 48 NRC RAI No. 2.8.5.1-2 Please evaluate the loss of feedwater heating transient at EPU conditions at MNGP to demonstrate conformance to fuel thermal-mechanical acceptance criteria. Provide transient results including those pertaining to fuel thermal-mechanical performance, and specifically address the potential for pellet-cladding interaction and pellet-cladding mechanical interaction. NSPM Response The loss of feedwater heating transient was generically dispositioned for EPU in the Constant Pressure Power Uprate LTR (NEDC-33004P-A). Cycle-specific reload licensing analyses will be performed for MNGP. However, an analysis was performed at EPU conditions for the EPU representative core and the results show acceptable margin to the fuel centerline melt and 1 % cladding strain limits. There is more than 10%margin to the fuel centerline melt and cladding strain limits.Please see response to NRC RAI No. 2.8.3-10 for information on the potential for pellet-cladding interaction. L-MT-09-049 Enclosure 1 Non Proprietary Page 34 of 48 SRXB RAI No. 2.8.5.2-1 Please identify the analytic methods used to analyze the LOFW event for uprated conditions. NSPM Response See response to NRC RAI No. 2.8.4.3-4 (see NSPM Letter L-MT-09-017, Accession No. ML090790388). L-MT-09-049 Enclosure 1 Non Proprietary Page 35 of 48 SRXB RAI No. 2.8.5.2-2 Please provide a more specific description of the EPU conditions assumed for the LOFW transient. For example, did the analysis use "typical" MNGP EPU conditions, is it a bounding EPU core design, or is it based on an actual core design?NSPM Response See response to NRC RAI No. 2.8.4.3-4 (see NSPM Letter L-MT-09-017, Accession No. ML090790388). L-MT-09-049 Enclosure 1 Non Proprietary Page 36 of 48 SRXB RAI No. 2.8.5.3-1 The technical evaluation for the RCP Rotor Seizure and Shaft Break contains a disposition referring to a previously analyzed event and the development of specified acceptable fuel design limits based on this analysis. Please explain: a) Is this analysis generic for all BWRs. MNGP-specific, EPU-specific, or specific to the MNGP EPU request?b) Why is this information presented in the MNGP PUSAR, and not discussed in GEH's response to RAI Set 9 Number 14?c) Chapter 14, Sections 7 and A imply that this accident sequence would have an EPU disposition that is different than the PUSAR passage implies. Please provide clarifying information. NSPM Response Response to Part a The RCP Rotor Seizure in single loop operation is analyzed when a new fuel product line is introduced at a plant. This plant specific analysis establishes a cycle-independent MCPR limit, which is reviewed on a cycle-by-cycle basis to ensure that the limit continues to be met. It was most recently performed for MNGP at CLTP for the transition to GE-14 fuel for Cycle 22.Response to Part b The additional information provided in the PUSAR was meant to supplement the information discussed in GEH's response to RAI No. Set 9, Number 14. It does not change the conclusion for Decrease in Reactor Coolant Flow Rate events stated in GEH's response to RAI No. Set 9, Number 14.Response to Part c The pump seizure event is performed for single loop operation (SLO). The allowed single loop operation is restricted to the MELLLA boundary and the existing maximum SLO core flow and the MELLLA boundary are unchanged for EPU in terms of absolute power and flow. Thus, there is no change to the inputs or results of the pump seizure event for EPU. L-MT-09-049 Enclosure 1 Non Proprietary Page 37 of 48 SRXB No. 2.8.5.4-1 MNGP states that the Rod Withdrawal Error at low-power event analysis is based upon information contained within a BWROG report regarding stability interim corrective action (Reference 30 of the EPU Application). Is this the correct reference regarding the basis for this event evaluation? If so, please make this report available for staff review. If not, please provide the staff with the correct reference. NSPM Response The proper reference for the Rod Withdrawal Error at low-power event analysis is NEDO-23842 "Continuous Control Rod Withdrawal Transient in the Startup Range", April 1978 contained in Enclosure

3.

L-MT-09-049 Enclosure 1 Non Proprietary Page 38 of 48 SRXB RAI No. 2.8.5.4-2 What methods and analytical codes will be used to analyze the Rod Withdrawal Error event both at low-power and at power for the MNGP EPU reload licensing analysis?NSPM Response The RWE at startup analysis is based on a generic study, NEDO-23842" Continuous Control Rod Withdrawal Transient in the Startup Range", April 1978, that concludes it is a non-limiting event. The methods applied in the generic study are consistent with those for the CRDA Rod Drop Accident Analysis for Large Boiling Water Reactors, Licensing Topical Report, March 1972 (NEDO-10527) including Supplements 1 and 2. There was no updated analysis performed since no change in peak fuel enthalpy is expected due to EPU because the RWE at startup is a localized low-power event. However, indirectly, EPU fuel and core designs can lead to generally higher rod worth distribution and therefore higher peak fuel enthalpy at low power. Proprietary Information Withheld If the peak fuel rod enthalpy is conservatively increased by a factor of 1.2, the RWE at startup peak fuel enthalpy at EPU will be 72 cal/gram. This enthalpy is well below the acceptance criterion of 170 cal/gram.The RWE at power evaluation is performed using the computer code PANACEA version 11. This RWE event and analysis is described in the MNGP USAR Section 14.3. L-MT-09-049 Enclosure 1 Non Proprietary Page 39 of 48 SRXB RAI No. 2.8.5.4-3 Clarify which peak fuel rod enthalpy was used as the basis for the predicted peak fuel rod enthalpyof 72 cal/gram at EPU conditions. Was the peak fuel rod enthalpy predicted at originally licensed power or at the current licensed power used to calculate the peak fuel rod enthalpy at EPU conditions? NSPM Response A value of 60 cal/gram is obtained from NEDO-23842 "Continuous Control Rod Withdrawal Transient in the Startup Range", April 1978, which is a generic report on RWE at startup. No change in peak fuel enthalpy is expected due to EPU since this is a localized low-power event, but it was increased conservatively by the ratio of the power increase, a factor of 1.2 (72 cal/gram) which is significantly larger than any indirect effects from EPU fuel and core design. Proprietary Information Withheld L-MT-09-049 Enclosure 1 Non Proprietary Page 40 of 48 SRXB RAI No. 2.8.5.4-4 Please refer to RAI No. 2.8.5.4-1 and provide similar information concerning the same reference in PUSAR Section that discusses the Control Rod Drop Accident.NSPM Response The proper reference for the CRDA analysis is NEDO-21231, "Banked Position Withdrawal Sequence," January 1977. L-MT-09-049 Enclosure 1 Non Proprietary Page 41 of 48 SRXB RAI No. 2.8.5.4-5 What methods and analytical codes will be used in the reload licensing analysis to predict the fuel and system response to a Control Rod Drop Accident at EPU conditions? NSPM Response The CRDA evaluation is based on a generic study, NEDO-21231, "Banked Position Withdrawal Sequence," January, 1977, that concludes it is a non-limiting event. The methods applied in the generic study are consistent with those for the CRDA Rod Drop Accident Analysis for Large Boiling Water Reactors, Licensing Topical Report, March 1972 (NEDO-1 0527) including Supplements 1 and 2. There was no updated analysis performed since no change in peak fuel enthalpy is expected due to EPU because the CRDA is a localized low-power event. However, indirectly, EPU fuel and core designs can lead to generally higher rod worth distributions and therefore higher peak fuel enthalpy at low power. Proprietary Information Withheld If the peak fuel rod enthalpy is conservatively increased by a factor of 1.2, the CRDA peak fuel enthalpy at EPU will be 162 cal/gram. This enthalpy is well below the acceptance criterion of 280 cal/gram. L-MT-09-049 Enclosure 1 Non Proprietary Page 42 of 48 SRXB RAI No. 2.8.5.4-6 Please discuss the basis for concluding that a 120% multiplier on peak fuel enthalpy for the Control Rod Drop Accident is conservative. NSPM Response This factor is conservative because it is expected that EPU by itself, does not increase peak fuel enthalpy for this localized low-power event. However, indirectly, EPU fuel and core designs can lead to generally higher rod worth distribution and therefore higher peak fuel enthalpy at low power. Proprietary Information Withheld L-MT-09-049 Enclosure 1 Non Proprietary Page 43 of 48 SRXB RAI No. 2.8.5.4-7 In the Technical Evaluation of the Control Rod Drop Accident (CRDA), it is stated that, if a conservative multiplier of 1.2 is applied to the peak fuel enthalpy at the current licensed thermal power, the peak fuel enthalpy at EPU conditions will be 162 cal/gram.However, the Updated Safety Analysis Report (USAR) for MNGP, Section 14.7.1,"Control Rod Drop Accident Evaluation," states that the maximum peak fuel enthalpy for a CRDA is 158 cal/gram. This translates to a peak fuel enthalpy of 190 cal/gram at EPU conditions using the conservative multiplier of 1.2 approach. Explain the discrepancy between the USAR value of peak fuel enthalpy and the value found in the MNGP EPU application. NSPM Response The reference, NEDO-21231, "Banked Position Withdrawal Sequence," January 1977, calculates a peak fuel enthalpy of 135 cal/gram based on the use of the Banked Position Withdrawal Sequence. Using the 1.2 approach, the EPU value is 162 cal/gram. The MNGP USAR CRDA maximum peak fuel enthalpy of 158 cal/gram represents the upper bound enthalpy of a limiting rod worth derived from many CRDA calculations with rod worths exceeding those acceptable from the BPWS and is therefore very conservative. L-MT-09-049 Enclosure 1 Non Proprietary Page 44 of 48 SRXB RAI No. 2.8.5.4-8 The CLTR/PUSAR disposition for the CRDA considers uprate performance against fuel enthalpy only. Please consider the performance of this transient against fuel thermal-mechanical criteria to demonstrate that the EPU does not leave the plant unacceptably susceptible to, for instance, pellet-cladding interaction or pellet-cladding mechanical interaction. NSPM Response The CRDA is not considered a transient, but an accident. The design limit for this accident is set at 280 cal/gram locally to limit the number of failed rods. It is not intended to prevent any failed rods. The CRDA accident and analysis is described in the MNGP USAR Section 14.7. L-MT-09-049 Enclosure 1 Non Proprietary Page 45 of 48 SRXB RAI No. 2.8.5.6-7 The disposition for Topic 4.2.6, ECCS Net Positive Suction Head, contained in the CLTR appears to be unaddressed for the MNGP PUSAR. Please explain.NSPM Response The CLTR requires a plant specific evaluation of EGGS pump Net Positive Suction Head (NPSH). The EGGS NPSH evaluation is included in PUSAR Section 2.6.5, Containment Heat Removal. This includes a MNGP specific evaluation of the effect of the increased wetwell temperature on NPSH. The evaluation demonstrates that the EGGS NPSH requirements at EPU conditions are met without any changes to the maximum amplitude of overpressure credited in the current design basis. L-MT-09-049 Enclosure 1 Non Proprietary Page 46 of 48 SRXB RAI No. 2.8.5.6-8 a) Do the calculated cladding oxidation levels account for pre-existing oxidation? b) If not, what amount of pre-existing oxidation will exist on the limiting bundle?c) How much pre-existing oxidation will exist on the more highly exposed bundles?d) Does the transient oxidation result presented consider oxidation on both surfaces of the fuel cladding?NSPM Response The Information Notice 98-29 addresses concerns regarding pre-transient oxidation in the ECCS-LOCA Evaluation Model. This matter was reviewed with respect to the GEH ECCS-LOCA evaluation model during the GNF Technology Update of April 29-30, 2008. No NRC review was requested since the conservative treatment presented would not fundamentally change the approved methodology. The conclusion of that review was that GEH would not include pre-transient oxidation in the ECCS-LOCA calculation unless there was real prospect that the acceptance criterion (17% local oxidation limit) could be challenged. (This vulnerability would be principally seen on BWR/2 plants, only; these plants now include pre-transient oxidation on ECCS-LOCA analysis on a forward fit basis.)So, in response: Response Part a No, pre-existing oxidation is not considered in the calculated cladding oxidation levels reported from the analysis for MNGP EPU.Response to Part b The amount of pre-existing oxidation that will exist on the limiting bundle will be [[Proprietary Information Withheld equivalent cladding reacted.Response to Part c The generic assessment for GE14 fuel shows no more pre-existing oxidation will exist on the more highly exposed bundles than Proprietary Information Withheld. This value bounds the pre-existing oxidation for the most highly exposed bundle for MNGP. Considering the calculated transient oxidation reported in the analysis, one can conclude ample margin remains to the 17%Acceptance Criterion, [[Proprietary Information Withheldl] L-MT-09-049 Enclosure 1 Non Proprietary Page 47 of 48 Response to Part d Yes, the inside surface of the rod cladding is considered in the total calculation of transient oxidation. Oxidation on the inside surface of rod cladding is considered for times after perforation of the fuel rod. This treatment is consistent with the general ECCS-LOCA analysis methodology as coolant water becomes capable of ingress and contact with the fuel rod inside surface.SRXB RAI No. 2.8.6-1 The PUSAR indicates that the capability of the new/spent (as appropriate) fuel storage facility is evaluated whenever a change to fuel design is introduced. Although there is no new fuel design being introduced for the uprate, the characteristics of both the new fuel and the discharged fuel will change. Demonstrate that the power uprate fuel, both new and spent, remains within the bounds of the fuel pool safety analyses.NSPM Response As stated in Section 2.8.1 of the PUSAR,"Monticello transitioned to GEl4 fuel in Cycle 21 and will continue to use only GE fuel types through EPU implementation. No new fuel product line designs are introduced and there are no changes to fuel design limits required by EPU. The fuel design limits are established for all new fuel product line designs as a part of the fuel introduction and reload analyses. Therefore, no additional fuel and core design evaluations are required for EPU and the generic evaluation in the CLTR is acceptable." At EPU conditions, the decay heat produced by any given fuel bundle is increased for both the normal and emergency core offload scenarios. Notwithstanding the above, the actual spent fuel pool decay heat load can readily be controlled by the fuel discharge rate and other administrative controls. An evaluation of expected EPU conditions has been performed that demonstrated the normal and emergency spent fuel pool heat removal capacity continues to bound the expected EPU heat loads without reducing the existing SFP cooling margins.Existing plant instrumentation and procedures provide adequate indications and controls for monitoring SFP temperature and level during normal batch offloads and the unexpected case of the limiting full core offload.Existing administrative controls will continue to be used to conservatively manage core offloads such that neither the normal SFPCC nor emergency cooling capabilities with the RHR System in fuel pool cooling assist mode are exceeded at EPU conditions. L-MT-09-049 Enclosure 1 Non Proprietary Page 48 of 48 These controls evaluate actual outage conditions in order to optimize maintenance activities while maintaining adequate cooling margins. Depending on the actual decay heat conditions, spent fuel pool characteristics, and environmental conditions during the particular EPU outage under evaluation, discharge time periods may be extended or other controls conservatively applied to assure cooling margins. No modifications to the SFPCC System are required to support EPU.SFP crud activity and corrosion products will increase slightly due to EPU conditions. However, the increase is insignificant and the SFP water quality will be maintained by the SFPCC system. The slight increase in crud under EPU conditions may result in slightly more frequent backwashes of the filter-demineralizers or replacement of the resins. This is acceptable because the liquid rad waste system is currently processing at about 53% of its design capacity.The spent fuel pool temperature is normally maintained at 125 deg F or less. The SFPCC is designed to maintain a maximum spent fuel storage pool temperature of 140 deg F. The High Density Fuel Storage System racks are designed for a maximum spent fuel pool temperature of 150 deg F during all normal and abnormal conditions. Therefore, the spent fuel racks are maintained at a temperature that is below the maximum design temperature. These values do not change between CLTP and EPU. ENCLOSURE 2 NSPM RESPONSE TO SNPB RAIs DATED 4-27-2009 L-MT-09-049 Enclosure 2 Page 1 of 3 SNPB RAI No. 10 Please update the Power Uprate Safety Analysis Report to reflect the use of an additional 350-psi critical pressure margin.NSPM Response In response to SNPB-6 (NSPM Letter L-MT-09-017 Accession No.ML090790388), NSPM documented the basis for not including the use of an additional 350-psi critical pressure margin.The impact of the 350-psi reduction in Pcritical on the GE14 thermal mechanical limit curves has subsequently been evaluated using the same methodology that was used to derive the original GE14 thermal mechanical limits. The evaluation resulted in revised Linear Heat Generation Rate (LHGR) versus exposure limit curves, which ensure that the rod pressure is maintained in conformance with the reduced Pcritical criteria. The revised LHGR limit curve for plants referencing NEDC-33173P is defined in Appendix C of NEDC-32868P, GE14 Compliance With Amendment 22 of NEDE-2401 1-P-A (GESTAR II), Revision 3, April 2009.The EPU core design used for the PUSAR can accommodate the slight modification in the Linear Heat Generation Rate (LHGR) limit curve. Future GE14 reload core designs at MNGP will use the LHGR limit curve in Appendix C of NEDC-32868P, GE14 Compliance With Amendment 22 of NEDE-240 11-P-A (GESTAR II), Revision 3, April 2009, which incorporates the additional 350-psi critical pressure margin. L-MT-09-049 Enclosure 2 Page 2 of 3 SNPB RAI No. 11 Please provide specific details of the actions taken by NSPM to address the penalty.NSPM Response Please see the response to NRC SNPB RAI No. 10. L-MT-09-049 Enclosure 2 Page 3 of 3 SNPB RAI No. 12 Please provide information that describes how NSPM ensures that subsequent cycle safety analyses address the NRC staff's limitations regarding fuel thermal-mechanical analysis methods, specifically limitations 12 and 14 from the safety evaluation report for the IMLTR.NSPM Response Limitation 12 states, in part, "Once the PRIME LTR and its application are approved, future license applications for EPU and MELLLA+ referencing LTR NEDC-33173P must utilize the PRIME T-M methods." PRIME is currently in NRC review and has not been approved, and PRIME is not the basis for the evaluation of upcoming EPU application. By letter dated February 27, 2009 (MFN 09-143), GEH committed to supplement the GEH Licensing Topical Report, NEDC-33173P, describing the implementation of the PRIME code models into the downstream safety analysis codes. The transmittal letter for the supplement is to include a schedule to upgrade to the downstream codes.Regarding Limitation 14, please see the response to NRC SNPB RAI No. 10. ENCLOSURE 3 REFERENCE DOCUMENTS: SUPPLEMENTAL RELOAD LICENSING REPORT (SRLR) FOR MONTICELLO RELOAD 24 CYCLE 25, Rev. 1 NEDO -23842: CONTINUOUS CONTROL ROD WITHDRAWAL TRANSIENT IN STARTUP RANGE GNF Global Nuclear Fuel A Joint Venture of GE, Toshiba, & Hitachi 0000-0083-9607-SRLR Revision 1 Class I March 2009 Supplemental Reload Licensing Report for Monticello Reload 24 Cycle 25 MONTICELLO 0000-0083-9607-SRLR Reload 24 Revision I Important Notice Regarding Contents of This Report Please Read Carefully This report was prepared by Global Nuclear Fuel -Americas, LLC (GNF-A) solely for use by Northern States Power Company (NSP) ("Recipient") in support of the operating license for MONTICELLO (the"Nuclear Plant"). The information contained in this report (the "Information") is believed by GNF-A to be an accurate and true representation of the facts knownby, obtained by or provided to GNF-A at the time this report was prepared.The only undertakings of GNF-A respecting the Information are contained in the contract between Recipient and GNF-A for nuclear fuel and related services for the Nuclear Plant (the "Fuel Contract") and nothing contained in this document shall be construed as amending or modifying the Fuel Contract. The use of the Information for any purpose other than that for which it was intended under the Fuel Contract, is not authorized by GNF-A. In the event of any such unauthorized use, GNF-A neither (a) makes any representation or warranty (either expressed or implied) as to the completeness, accuracy or usefulness of the Information or that such unauthorized use may not infringe privately owned rights, nor (b) assumes any responsibility for liability or damage of any kind which may result from such use of such information. Page 2 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision I Acknowledgement The engineering and reload licensing analyses, which form the technical basis of this Supplemental Reload Licensing Report, were performed by GNF-A/GEH Nuclear Analysis personnel. The Supplemental Reload Licensing Report was prepared by Richard McCord. This document has been verified by Dave Knepper.Page 3 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision *The basis for this report is General Electric Standard Application for Reactor Fuel, NEDE-2401 1-P-A-16, October 2007; and the U.S. Supplement, NEDE-2401 I-P-A- 16-US, October 2007.1. Plant-unique Items Appendix A: Analysis Conditions Appendix B: Decrease in Core Coolant Temperature Events Appendix C: ARTS Off-Rated Limits Curves Appendix D: Expanded Operating Domain Analyses Appendix E: SLO Pump Seizure Operating Limit Appendix F: Mislocated Fuel Loading Error Appendix G: Thermal Mechanical Compliance Appendix H: Turbine Trip with Bypass and Degraded Scram Appendix I: Monticello Non-Standard SRLR Items Appendix J: List of Acronyms 2. Reload Fuel Bundles FuelType Cycle Number Loaded Irradiated: GE14-PlODNAB391-14GZ-IOOT-145-T6-2480 (GE14C) 21 1 GE 14-P 1ODNAB393-17GZ-1OOT- 145-T6-2598 (GE14C) 22 8 GE14-PlODNAB393-17GZ-100T-145-T6-2599 (GE14C) 22 12 GE 14-P 1ODNAB393-17GZ-IOOT- 145-T6-2599 (GE 14C) 23 32 GE14-P 1ODNAB392-16GZ-1OOT- 145-T6-2824 (GE 14C) 23 120 GE 14-P 1ODNAB392-16GZ-1OOT- 145-T6-2931 (GE 14C) 24 104 GE 14-P 1ODNAB392-17GZ-1OOT- 145-T6-2932 (GE 14C) 24 43 New: GE14-P1ODNAB391-12GZ-IOOT-145-T6-3103 (GE14C) 25 16 GE 14-P 1ODNAB392-16GZ-1OOT- 145-T6-3102 (GE 14C) 25 40 GE14-P1ODNAB375-16GZ-100T-145-T6-3101 (GE14C) 25 52 GE 14-P 1ODNAB424-14GZ-1OOT- 145-T6-3100 (GE 14C) 25 16 GE 14-P 1ODNAB392-16GZ-1OOT- 145-T6-2931 (GE 14C) 25 40 Total: 484 Page 4 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 3. Reference Core Loading Pattern Core Average Cycle Exposure Exposure 33784 MWd/MT 14199 MWd/MT Nominal previous end-of-cycle exposure: (30648 MWd/ST) (12881 MWd/ST)Minimum previous end-of-cycle exposure (for cold 33334 MWd/MT 13749 MWd/MT shutdown considerations): (30240 MWd/ST) (12473 MWd/ST)18596 MWd/MT 0 MWd/MT Assumed reload beginning-of-cycle exposure: (16870 MWd/ST) (0 MWd/ST)Assumed reload end-of-cycle exposure (rated 33091 MWd/MT 14495 MWd/MT conditions): (30020 MWd/ST) (13150 MWd/ST)Reference core loading pattern: Figure 1 4. Calculated Core Effective Multiplication and Control System Worth -No Voids, 201C Beginning of Cycle, keffective Uncontrolled 1.113 Fully controlled 0.958 Strongest control rod out 0.990 R, Maximum increase in strongest rod out reactivity during the cycle (Ak) 0.000 Cycle exposure at which R occurs 0 MWd/MT_______________________________________________ (0 MWd/ST)5. Standby Liquid Control System Shutdown Capability Boron (ppm) Shutdown Margin (Ak)(at 20 0 C) (at 160 0 C, Xenon Free)Analytical Requirement Achieved 660 >_0.010 0.019 Page 5 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 6. Reload Unique GETAB Anticipated Operational Occurrences (AOO) Analysis Initial Condition Parameters 1 Operating domain: STANDARD (HBB)Exposure range : BOC to EOC (Application Condition: 1)Peaking Factors Fuel Bundle Bundle Initial Local Radial Axial R-Factor Power Flow Design (MWt) (1000 lb/hr) MCPR GE14C 1.45 1.61 1.41 1.040 5.749 97.9 1.46 Operating domain: MELLLA (HBB)Exposure range : BOC to EOC (Application Condition: 1)Peaking Factors Fuel Bundle Bundle Initial Local Radial Axial R-Factor Power Flow Design (MWt) (1000 lb/hr) MCPR GE14C 1.45 1.53 1.37 1.040 5.449 80.4 1.46 Operating domain: STANDARD (UB)Exposure range : BOC to EOC (Application Condition: 1 )Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow Design (MWt) (1000 lb/hr) MCPR GE14C 1.45 1.66 1.21 1.040 5.923 95.2 1.50 Operating domain: MELLLA (UB)Exposure range : BOC to EOC (Application Condition: 1 )Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow Design___ _(MWt) (1000 lb/hr) MCPR GE14C 1.45 1.61 1.24 1.040 5.735 77.5 1.45 1 Exposure range designation is defined in Table 7-1. Application condition number is defined in Section 11.Page 6 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 7. Selected Margin Improvement Options 2 Recirculation pump trip: Rod withdrawal limiter: Thermal power monitor: Improved scram time: Measured scram time: Exposure dependent limits: Exposure points analyzed: No No Yes Yes (ODYN Option B)No No I Table 7-1 Cycle Exposure Range Designation Name Exposure Range 3 BOC to EOC BOC25 to EOC25 2 Refer to the GESTAR basis document identified at the beginning of this report for the margin improvement options currently supported therein.3 End of Rated (EOR) is defined as the cycle exposure corresponding to all rods out, 100% power/100% flow, and normal feedwater temperature. For plants without mid-cycle OLMCPR points, EOR is not applicable. Page 7 MONTICELLO 0000-0083-9607-SRLR Reload 24 Revision 1 8. Operating Flexibility Options 45 6 The following information presents the operational domains and flexibility options which are supported by the reload licensing analysis.Extended Operating Domain (EOD): Yes EOD type: Maximum Extended Load Line Limit (MELLLA)Minimum core flow at rated power: 82.4 %Increased Core Flow: Yes Flow point analyzed throughout cycle: 105.0 %Feedwater Temperature Reduction: No ARTS Program: Yes Single Loop Operation: Yes Equipment Out of Service: Safety/relief valves Out of Service: Yes (credit taken for 5 valves)PROOS Yes 9. Core-wide AOO Analysis Results 7 Methods used: GEMINI, GEXL-PLUS 4 Refer to the GESTAR basis document identified at the beginning of this report for the operating flexibility options currently supported therein.5 ICF is not licensed above reactor power of 1670 MWt.6 The Reload Analysis for SRVOOS evaluated thermal margins and over-pressurization only.7 Exposure range designation is defined in Table 7-1. Application condition number is defined in Section 11.Page 8 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 Operating domain: STANDARD (HBB)Exposure range : BOC to EOC (Application Condition: 1)Uncorrected ACPR Event Flux Q/A Event_____________(% rated) (% rated) GE14C Fig.FW Controller Failure 648 140 0.34 2 Turbine Trip with Bypass 669 138 0.35 3 Turbine Trip w/o Bypass 618 135 0.31 4 Load Rejection w/o Bypass 499 128 0.30 5 Inadvertent HPCI /L8 614 144 0.36 6 Operating domain: MELLLA (HBB)Exposure range : BOC to EOC (Application Condition: 1)Uncorrected ACPR Event Flux Q/A Event_____________(% rated) (% rated) GE14C Fig.FW Controller Failure 579 138 0.35 7 Turbine Trip with Bypass 578 135 0.32 8 Turbine Trip w/o Bypass 536 132 0.32 9 Load Rejection w/o Bypass 416 124 0.30 10 Inadvertent HPCI/L8 535 139 0.36 11 Operating domain: STANDARD (UB)Exposure range : BOC to EOC (Application Condition: 1)Uncorrected ACPR Event Flux Q/A Event_____________(% rated) (% rated) GE14C Fig.FW Controller Failure 456 131 0.35 12 Turbine Trip with Bypass 493 132 0.40 13 Turbine Trip w/o Bypass 446 127 0.33 14 Load Rejection w/o Bypass 371 119 0.31 15 Inadvertent HPCI/L8 435 134 0.37 16 Page 9 MONTICELLO 0000-0083-9607-SRLR D 1 Operating domain: MELLLA (UB)Exposure range : BOC to EOC (Application Condition: 1)Uncorrected ACPR Event Flux Q/A (% rated) (% rated) GE14C Fig.FW Controller Failure 300 121 0.32 17 Turbine Trip with Bypass 361 124 0.36 18 Turbine Trip w/o Bypass 296 117 0.30 19 Load Rejection w/o Bypass 243 109 0.25 20 Inadvertent HPCI/L8 289 123 0.34 21 10. Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary The ARTS Rod withdrawal error (RWE) event was originally analyzed in the document: Average Power Range Monitor, Rod Block Monitor, and Technical Specification Improvement (ARTS) Program for Monticello Nuclear Generating Plant, NEDC-30492-P, April 1984.The control Rod Withdrawal Error (RWE) analysis was evaluated for Cycle 25. The conclusion from the analysis is that the calculated ARTS minimum delta critical power ratio (ACPR) is 0.25. The resulting operating limit critical power ratio is reported in the Section 11.The RI3M operability requirements specified in Section 4.5 of NEDC-30492-P have been evaluated and shown to be sufficient to ensure that the Safety Limit MCPR and cladding 1% plastic strain criteria will not be exceeded in the event of an unblocked RWE event.Page 10 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 11. Cycle MCPR Values 8 9 Two loop operation safety limit: Single loop operation safety limit: Stability MCPR Design Basis: ECCS MCPR Design Basis: SLO Pump Seizure OLMCPR: 1.10 1.12 See Section 15 See Section 16 (Initial MCPR)See Pump Seizure Appendix E Non-pressurization Events: Exposure range: BOC to EOC GE14C Fuel Loading Error (misoriented) 1.29 Fuel Loading Error (mislocated) Not Limiting Control Rod Withdrawal Error (RBM setpoint at 114%) 1.35 Loss of Feedwater Heating (See Appendix B) 1.23 Limiting Pressurization Events OLMCPR Summary Table: 10 AppI.Cond. Exposure Range Option A Option B Cond.GE14C GE14C 1 Base Case BOC to EOC 1.70 1.59' Exposure range designation is defined in Table 7-1.9 For single loop operation, the MCPR operating limit is 0.02 greater than the two loop value.10 Each application condition (Appl. Cond.) covers the entire range of licensed flow and feedwater temperature unless specified otherwise. The OLMCPR values presented apply to rated power operation based on the two loop operation safety limit MCPR.Page 11 MONTICELLO R oln~r 9dL 0000-0083-9607-SRLR Revision 1 RPI-1 24 Pressurization Events: " Operating domain: STANDARD (H131)Exposure range : BOC to EOC (Application Condition: 1 )Option A Option B G6i6E14C GE14C FW Controller Failure 1.67 1.50 Turbine Trip with Bypass 1.53 1.53 Turbine Trip w/o Bypass' 1.64 1.47 Load Rejection w/o Bypass 1.62 1.45 Inadvertent HPCI /L8 1.69 1.52 Operating domain: MELLLA (HBB)Exposure range : BOC to EOC (Application Condition: 1)Option A Option B GE14C GE14C FW Controller Failure 1.68 1.51 Turbine Trip with Bypass 1.50 1.50 Turbine Trip w/o Bypass 1.65 1.48 Load Rejection w/o Bypass 1.63 1.46 Inadvertent HPCI /L8 1.69 1.52 Operating domain: STANDARD (UB)Exposure range : BOC to EOC ,(Application Condition: I Option A Option B 6 iGE14C GE14C FW Controller Failure 1.69 1.52 Turbine Trip with Bypass 1.59 1.59 Turbine Trip w/o Bypass 1.66 1.49 Load Rejection w/o Bypass 1.63 1.46 Inadvertent HPCI /L8 1.70 1.53" Application condition numbers shown for each of the following pressurization events represent the application conditions for which this event contributed in the determination of the limiting OLMCPR value.Page 12 MONTICELLO Reload ?4 0000-0083-9607-SRLR Re~viqion 1 Operating domain: MELLLA (UB)Exposure range : BOC to EOC (Application Condition: 1I)Option A Option B GiiE14C GE14C FW Controller Failure 1.65 1.48 Turbine Trip with Bypass 1.54 1.54 Turbine Trip w/o Bypass 1.63 1.46 Load Rejection w/o Bypass 1.57 1.40 Inadvertent HPCI /L8 1.67 1.50 12. Overpressurization Analysis Summary Event Psi Pdome Pv Plant (psig) (psig) (psig) Response MSIV Closure (Flux Scram) -1279 1282 1301 Figure 22 STANDARD (HBB)MSIV Closure (Flux Scram) -MELLLA 1271 1273 1291 Figure 23 (HBB) 1271___ 1273___ 1291__ Figure_23 13. Loading Error Results Variable water gap misoriented bundle analysis: Yes 12 Misoriented Fuel Bundle ACPR GE 14-P 1ODNAB392-16GZ-1OOT- 145-T6-2931 (GE14C) 0.15 GE 14-P 1ODNAB392-17GZ-100T- 145-T6-2932 (GE 14C) 0.07 GE 14-P ODNAB424-14GZ-1 OOT- 145-T6-3 100 (GE 14C) 0.12 GE14-PIODNAB375-16GZ-100T-145-T6-3101 (GE14C) 0.19 GE 14-PI ODNAB392-16GZ-1OOT- 145-T6-3102 (GE 14C) 0.15 GE14-PIODNAB391-12GZ-100T-145-T6-3103 (GE14C) 0.16 12 Includes a 0.02 penalty due to variable water gap R-factor uncertainty. Page 13 MONTICELLO 0000-0083-9607-SRLR Reload 24 Revision 1 14. Control Rod Drop Analysis Results This is a banked position withdrawal sequence plant, therefore, the control rod drop accident analysis is not required. NRC approval is documented in NEDE-2401 I-P-A-US.15. Stability Analysis Results 15.1 Stability Option III Solution MONTICELLO has implemented BWROG Long Term Stability Solution Option III using the Oscillation Power Range Monitor (OPRM) as described in Reference 1 in Section 15.4. The plant specific Hot Channel Oscillation Magnitude (HCOM) (Reference 2 in Section 15.4) and other cycle specific stability parameters are used in the Cycle 25 Option III stability evaluation. A Backup Stability Protection (BSP)evaluation is provided in the event that the Option III OPRM system is declared inoperable. The following Option III OPRM stability setpoint determination described in Section 15.2 and the implementation of the associated BSP Regions described in Section 15.3 provide the stability licensing bases for MONTICELLO Cycle 25.15.2 Detect and Suppress Evaluation A reload Option III evaluation has been performed in accordance with the licensing methodology described in Reference 3 in Section 15.4. The stability based Operating Limit Minimum Critical Power Ratio (OLMCPR) is determined for two conditions as a function of OPRM amplitude setpoint. The two conditions evaluated are: (1) a postulated oscillation at 45% rated core flow quasi steady-state operation (SS), and (2) a postulated oscillation following a two recirculation pump trip (2PT) from the limiting rated power operation state point.The OPRM-setpoint-dependent OLMCPR(SS) and OLMCPR(2PT) values are calculated for Cycle 25 in accordance with the BWROG regional mode DIVOM guidelines described in Reference 4 in Section 15.4. The Cycle 25 Option III evaluation provides adequate protection against violation of the Safety Limit MCPR (SLMCPR) for the two postulated reactor instability events as long as the plant OLMCPR is equal to or greater than OLMCPR(SS) and OLMCPR(2PT) for the selected OPRM setpoint in Table 15-2.The relationship between the OPRM Successive Confirmation Count Setpoint and the OPRM Amplitude Setpoint is provided in Reference 3 in Section 15.4 and Table 15-1. For intermediate OPRM Amplitude Setpoints, the corresponding OPRM Successive Confirmation Count Setpoints have been obtained by using linear interpolation. The OPRM setpoints for Two Loop Operation (TLO) are conservative relative to Single Loop Operation (SLO) and are, therefore, bounding.Page 14 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 Table 15-1 Relationship between OPRM Successive Confirmation Count Setpoint and OPRM Amplitude Setpoint Successive OPRM Confirmation Amplitude Count Setpoint Setpoint 6 >1.04 8 >1.05 9 >1.06 10 >1.07 11 >1.08 12 >1.09 13 >1.10 14 >1.11 15 >1.13 16 >1.14 17 >1.16 18 >1.18 19 >1.21 20 >1.24 Page 15 MONTICELLO 0000-0083-9607-SRLR D.-; ; -1 Table 15-2 OPRM Setpoint Versus OLMCPR OPRM Amplitude OLMCPR(SS) OLMCPR(2PT) Setpoint 1.05 1.202 1.108 1.06 1.223 1.128 1.07 1.245 1.149 1.08 1.269 1.170 1.09 1.293 1.192 1.10 1.318 1.215 1.11 1.342 1.238 1.12 1.367 1.261 1.13 1.394 1.285 1.14 1.421 1.311 1.15 1.450 1.337 OLMCPR Off-rated Rated Power Acceptance OLMCPR OLMCPR (see Criteria @45% flow Section 11)15.3 Backup Stability Protection The BSP region boundaries were calculated for MONTICELLO Cycle 25 for normal feedwater temperature operation. The endpoints of the regions are defined in Table 15-3. The region boundaries, shown in Figure 24, are defined using the Generic Shape Function (GSF). See Reference 5 in Section 15.4.Page 16 MONTICELLO Reload 24 0000-0083:9607-SRLR Revision 1 Table 15-3 BSP Region Intercepts for Normal Feedwater Temperature Region Boundary Power Flow Core Highest Intercept (%) (%) DR DR DR Al 63.8 40.0 < 0.798 < 0.255 BI 48.1 33.8 < 0.799 < 0.279 A2 72.8 50.0 < 0.798 < 0.241 B2 32.3 31.2 < 0.799 < 0.081 15.4 References

1. BWR Owners' Group Long-term Stability Solutions Licensing Methodology, NEDO-31960-A, November 1995 (including Supplement 1).2. Reactor Long-Term Stability Solution Option III: Licensing Basis Hot Channel Oscillation Magnitude for Monticello Nuclear Generating Plant, GHNE-0000-0073-4167-R2, December 2007.3. Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications, Licensing Topical Report, NEDO-32465-A, August 1996.4. Plant-Specific Regional Mode DIVOM Procedure Guideline, GE-NE-0000-0028-9714-R 1, June 2005.5. Backup Stability Protection (BSP) for Inoperable Option III Solution, OG-02-0119-260, July 2002.16. Loss-of-Coolant Accident Results 13 16.1 10CFR50.46 Licensing Results The ECCS-LOCA analysis is based on the SAFER/GESTR methodology.

The licensing results applicable to each fuel type in the new cycle are summarized in the following table.13 Lattice numbers are defined in the Fuel Bundle Information Report Page 17 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision I Table 16.1-1 Licensing Results Core-Wide Licensing Local Me-Wate Fuel Type Basis PCT Oxidation Re atio (OF) (%)Reaction (0)(%) (%)GE14C 1960 < 5.00 < 0.10 The Monticello SAFER/GESTR-LOCA analysis results for the GE14C fuel type are documented in Reference 1 for GE14C in Section 16.4.The RHR intertie open line analyses for GE14C are documented in Reference 2 for GE14C in Section 16.4. The analyses indicate that plant operation up to 20% of 1880 MWt power with the RHR intertie line open is acceptable from an ECCS performance standpoint, provided a MAPLHGR multiplier of 0.75 is implemented or that the peak bundle power does not exceed 3.9 MW for GE14C fuel.16.2 10CFR50.46 Error Evaluation The 1OCFR5 0.46 errors applicable to the Licensing Basis PCT are shown in the following table(s).Table 16.2-1 Impact on Licensing Basis Peak Cladding Temperature for GE14C 1OCFR50.46 Error Notifications Number Subject PCT Impact (OF)Error in WEVOL calculation of downcomer free 2002-05 voue0 volume 2003-01 Impact of SAFER level/volume table error on PCT -15 2003-03 Impact of SAFER initial separator pressure drop error 0 on PCT 2003-05 Impact of Postulated Hydrogen-Oxygen Recombination 0 on PCT 2006-01 Impact of Top Peaked Power Shape for Small Break +30 LOCA Analysis Total PCT Adder (4F) +15 The GE 14C Licensing Basis PCT remains below the 1OCFR50.46 limit of 2200'F.The 1OCFR50.46 errors applicable to the Upper Bound PCT are shown in the table below.Page 18 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision I Table 16.2-2 Impact on Upper Bound Peak Cladding Temperature for GE14C 10CFR50.46 Error Notifications PCT Impact Number Subject (OF)Error in WEVOL calculation of downcomer free 2002-05 voue0 volume 2003-01 Impact of SAFER level/volume table error on PCT -15 2003-03 Impact of SAFER initial separator pressure drop error 0 on PCT 2003-05 Impact of Postulated Hydrogen-Oxygen Recombination 0 on PCT 2006-01 Impact of Top Peaked Power Shape for Small Break +30 LOCA Analysis Total PCT Adder (IF) +15 The GE14C Upper Bound PCT remains below the 1600'F limit with a LHGR and MAPLHGR setdown of 15%.16.3 ECCS-LOCA Operating Limits The composite MAPLHGR operating limits for the core fuel bundles in this cycle are shown in the following table(s).Page 19 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision I Table 16.3-1 MAPLHGR Limits Bundle Type: GE 14-P 1ODNAB393-17GZ-IOOT-145-T6-2598 (GE 14C)Average Planar Exposure MAPLHGR Limit (kW/ft)Lat. Lat. Lat. Lat. Lat. Lat. Lat.GWd/MT (GWd/ST) 5719 5720 5721 5722 5723 5724 5725 0.00 (0.00) 9.22 8.21 8.41 8.21 8.14 9.61 10.44 0.22 (0.20) 9.14 8.26 8.45 8.27 8.20 9.56 10.42 1.10(1.00) 8.94 8.35 8.53 8.40 8.34 9.42 10.34 2.20 (2.00) 8.88 8.47 8.64 8.59 8.53 9.39 10.33 3.31 (3.00) 8.90 8.59 8.75 8.74 8.72 9.42 10.35 4.41 (4.00) 8.93 8.71 8.87 8.90 8.91 9.46 10.38 5.51 (5.00) 8.97 8.84 8.99 9.05 9.10 9.51 10.40 6.61 (6.00) 9.01 8.97 9.10 9.20 9.25 9.55 10.43 7.72 (7.00) 9.04 9.11 9.21 9.37 9.44 9.58 10.45 8.82 (8.00) 9.07 9.23 9.34 9.55 9.63 9.61 10.47 9.92 (9.00) 9.09 9.35 9.47 9.73 9.82 9.64 10.48 11.02 (10.00) 9.11 9.47 9.60 9.90 9.94 9.65 10.49 12.13 (11.00) 9.12 9.59 9.71 10.05 10.01 9.66 10.49 13.23 (12.00) 9.13 9.70 9.82 10.17 10.07 9.67 10.49 14.33 (13.00) 9.13 9.80 9.90 10.14 10.06 9.67 10.49 15.43 (14.00) 9.13 9.90 9.97 10.08 10.01 9.67 10.48 16.53 (15.00) 9.13 9.98 10.02 10.06 9.99 9.66 10.48 18.74 (17.00) 9.12 10.10 10.11 10.04 9.98 9.65 10.47 22.05 (20.00) 9.11 10.23 10.22 10.02 9.97 9.64 10.45 27.56 (25.00) 8.95 10.30 10.42 10.00 9.95 9.61 10.43 33.07 (30.00) 8.30 10.07 10.08 9.89 9.83 9.01 10.13 38.58 (35.00) 7.65 9.55 9.56 9.28 9.23 8.36 9.50 41.33 (37.49) 7.34 9.28 9.30 9.00 8.95 8.05 9.18 44.09 (40.00) 7.02 9.02 9.03 8.71 8.66 7.73 8.86 49.60 (45.00) 6.38 8.48 8.50 8.17 8.13 7.09 8.23 54.78 (49.70) 3.98 -- -- -- -- -- --55.12 (50.00) -- 7.90 7.93 7.66 7.61 5.63 7.60 58.14 (52.75) -- -- -- -- 4.21 --60.63 (55.00) -- 6.26 6.30 6.52 6.40 -- 5.93 63.50 (57.61) -- 4.94 4.99 5.30 5.27 -- 4.59 63.55 (57.65) ............. 4.57 63.59 (57.68) -- 4.90 ........-- Page 20 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 Average Planar Exposure MAPLHGR Limit (kW/ft)Lat. Lat. Lat. Lat. Lat. Lat. Lat.GWd/MT (GWd/ST) 5719 5720 5721 5722 5723 5724 5725 63.72 (57.81). -- -- 4.89 -- -- -- -64.44 (58.46) .-- -- 4.90 --64.47 (58.49) .... 4.89 --..Page 21 MONTICELLO 0000-0083-9607-SRLR D 1 3."1 V1.V 8~l Table 16.3-2 MAPLHGR Limits Bundle Type: GE 14-P 1ODNAB393-17GZ-100T- 145-T6-2599 (GE 14C)Average Planar Exposure MAPLIIGR Limit (kW/ft)Lat. Lat. Lat. Lat. Lat. Lat. Lat.GWd/MT (GWd/ST) 5719 5720 5721 5726 5727 5724 5725 0.00(0.00) 9.22 8.21 8.41 8.42 8.36 9.61 10.44 0.22(0.20) 9.14 8.26 8.45 8.47 8.41 9.56 10.42 1.10(1.00) 8.94 8.35 8.53 8.59 8.54 9.42 10.34 2.20 (2.00) 8.88 8.47 8.64 8.77 8.72 9.39 10.33 3.31 (3.00) 8.90 8.59 8.75 8.92 8.91 9.42 10.35 4.41 (4.00) 8.93 8.71 8.87 9.06 9.10 9.46 10.38 5.51 (5.00) 8.97 8.84 8.99 9.21 9.26 9.51 10.40 6.61 (6.00) 9.01 8.97 9.10 9.34 9.40 9.55 10.43 7.72(7.00) 9.04 9.11 9.21 9.49 9.56 9.58 10.45 8.82 (8.00) 9.07 9.23 9.34 9.65 9.73 9.61 i0.47 9.92 (9.00) 9.09 9.35 9.47 9.81 9.90 9.64 10.48 11.02 (10.00) 9.11 9.47 9.60 9.96 9.98 9.65 10.49 12.13 (11.00) 9.12 9.59 9.71 10.09 10.03 9.66 10.49 13.23 (12.00) 9.13 9.70 9.82 10.18 10.08 9.67 10.49.14.33 (13.00) 9.13 9.80 9.90 10.16 10.07 9.67 10.49 15.43 (14.00) 9.13 9.90 9.97 10.08 10.01 9.67 10.48 16.53 (15.00) 9.13 9.98 10.02 10.05 9.99 9.66 10.48 18.74 (17.00) 9.12 10.10 10.11 10.03 9.97 9.65 10.47 22.05 (20.00) 9.11 10.23 10.22 10.02 9.96 9.64 10.45 27.56 (25.00) 8.95 10.30 10.42 9.99 .9.94 9.61 10.43 33.07 (30.00) 8.30 10.07 10.08 9.89 9.84 9.01 10.13 38.58 (35.00) 7.65 9.55 9.56 9.29 9.24 8.36 9.50 41.33 (37.49) 7.34 9.28 9.30 9.00 8.95 8.05 9.18 44.09 (40.00) 7.02 9.02 9.03 8.71 8.67 7.73 8.86 49.60 (45.00) 6.38 8.48 8.50 8.18 8.13 7.09 8.23 54.78 (49.70) 3.98 -- -- -- -- -- --55.12 (50.00) -- 7.90 7.93 7.66 7.62 5.63 7.60 58.14 (52.75) .-- -- -- -- 4.21 --60.63 (55.00) -- 6.26 6.30 6.55 .6.44 -- 5.93 63.50 (57.61) -- 4.94 4.99 5.31 5.28 -- 4.59 63.55 (57.65) -- -- -- --. 4.57 63.59 (57.68) -- 4.90 -- I.-- -- --Page 22 MONTICELLO Q.,1-A IA 0000-0083-9607-SRLR I? --Average Planar Exposure MAPLHGR Limit (kW/ft)Lat. Lat. Lat. Lat. Lat. Lat. Lat.GWd/MT (GWd/ST) 5719 5720 5721 5726 5727 5724 5725 63.72 (57.81) -- -- 4.89 -- -- -- --64.46 (58.48) .-- -- 4.90 ....64.49 (58.50) -- -- -- 4.89 -- -- --Page 23 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 Table 16.3-3 MAPLHGR Limits Bundle Type: GE 14-P 1 ODNAB392-16GZ-1 OOT- 1 45-T6-2824 (GE 14C)Average Planar Exposure MAPLHGR Limit (kW/ft)Lat. Lat. Lat. Lat. Lat. Lat. Lat.GWd/MT (GWd/ST) 6817 6818 6819 6820 6821 6822 6823 0.00 (0.00) 9.22 8.33 8.55 8.64 8.59 9.62 10.22 0.22(0.20) 9.14 8.37 8.58 8.68 8.63 9.56 10.20 1.10(1.00) 8.94 8.44 8.64 8.75 8.73 9.42 10.13 2.20 (2.00) 8.88 8.54 8.72 8.85 8.85 9.39 10.13 3.31 (3.00) 8.90 8.63 8.81 8.95 8.97 9.42 10.17 4.41 (4.00) 8.93 8.73 8.89 9.06 9.08 9.46 10.21 5.51 (5.00) 8.97 8.84 8.98 9.14 9.21 9.51 10.25 6.61 (6.00) 9.01 8.93 9.05 9.23 9.31 9.55 10.29 7.72 (7.00) 9.04 9.02 9.13 9.33 9.42 9.59 10.32 8.82 (8.00) 9.07 9.11 9.22 9.46 9.55 9.61 10.35 9.92 (9.00) 9.09 9.21 9.33 9.61 9.70 9.64 10.36 11.02 (10.00) 9.11 9.32 9.45 9.76 9.86 9.65 10.38 12.13 (11.00) 9.12 9.44 9.57 9.91 9.85 9.66 10.39 13.23 (12.00) 9.13 9.56 9.69 9.75 9.59 9.67 10.39 14.33 (13.00) 9.13 9.69 9.80 9.65 9.50 9.67 10.39 15.43 (14.00) 9.13 9.80 9.89 9.64 9.50 9.67 10.39 16.53 (15.00) 9.13 9.90 9.97 9.65 9.52 9.67 10.39 18.74 (17.00) 9.13 10.00 10.09 9.67 9.54 9.66 10.38 22.05 (20.00) 9.11 10.00 10.08 9.69 9.56 9.64 10.36 27.56 (25.00) 8.95 9.97 10.04 9.70 9.60 9.61 10.33 33.07 (30.0,0) 8.30 9.94 10.00 9.55 9.41 9.01 9.99 38.58 (35.00) 7.65 9.38 9.38 9.02 8.91 8.37 9.35 41.33 (37.49) 7.34 9.08 9.09 8.74 8.67 8.05 9.04 44.09 (40.00) 7.02 8.79 8.79 8.45 8.43 7.73 8.72 49.60 (45.00) 6.38 8.23 8.23 7.91 7.90 7.10 8.09 54.79 (49.71) 3.98 -- -- -- -- -- --55.12 (50.00) -- 7.70 7.69 7.39 7.39 5.64 7.46 58.16 (52.77) -- -- -- -- 4.21 --60.63 (55.00) -- 6.26 6.32 6.09 5.79 -- 5.56 62.85 (57.01) -- -- -- --. 4.53 63.16 (57.30) -- --... 4.68 ..--63.50 (57.61) -- 4.94 5.00 4.84 ......Page 24 MONTICELLO R1nA~d 94 0000-0083-9607-SRLR Average Planar Exposure MAPLHGR Limit (kW/ft)Lat. Lat. Lat. Lat. Lat. Lat. Lat.GWd/MT (GWd/ST) 6817 6818 6819 6820 6821 6822 6823 63.59 (57.68) -- 4.91 -- -- -- -- --63.72 (57.81) ..-- 4.90 --......63.76 (57.85) --.... 4.73 I. -- .Page 25 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 Table 16.3-4 MAPLHGR Limits Bundle Type: GE 14-P 1ODNAB3 92-16GZ- 1OOT- 145-T6-2931 (GE 14C)Average Planar Exposure MAPLHGR Limit (kW/ft)Lat. Lat. Lat. Lat. Lat. Lat. Lat.GWd/MT (GWd/ST) 7357 7358 7359 7360 7361 7362 7363 0.00 (0.00) 9.22 8.32 8.53 8.62 8.58 9.62 10.22 0.22 (0.20) 9.14 8.36 8.56 8.66 8.62 9.56 10.20 1.10(1.00) 8.94 8.43 8.62 8.73 8.70 9.42 10.13 2.20 (2.00) 8.88 8.52 8.71 8.83 8.82 9.39 10.13 3.31 (3.00) 8.90 8.62 8.79 8.94 8.93 9.42 10.17 4.41 (4.00) 8.93 8.72 8.88 9.04 9.05 9.46 10.21 5.51 (5.00) 8.97 8.82 8.96 9.12 9.18 9.51 10.25 6.61 (6.00) 9.01 8.92 9.03 9.21 9.29 9.55 10.29 7.72(7.00) 9.04 9.00 9.11 9.32 9.40 9.59 10.32 8.82 (8.00) 9.07 9.09 9.21 9.45 9.54 9.61 10.35 9.92 (9.00) 9.09 9.19 9.32 9.59 9.69 9.64 10.36 11.02 (10.00) 9.11 9.31 9.44 9.75 9.85 9.65 10.38 12.13 (11.00) 9.12 9.43 9.56 9.90 9.84 9.66 10.39 13.23 (12.00) 9.13 9.55 9.68 9.74 9.58 9.67 10.39 14.33 (13.00) 9.13 9.68 9.79 9.64 9.50 9.67 10.39 15.43 (14.00) 9.13 9.79 9.89 9.63 9.49 9.67 10.39'16.53 (15.00) 9.13 9.89 9.97 9.64 9.51 9.67 10.39 18.74 (17.00) 9.13 9.99 10.08 9.67 9.53 9.66 10.38 22.05 (20.00) 9.11 10.00 10.07 9.68 9.55 9.64 10.36 27.56 (25.00) 8.95 9.96 10.03 9.70 9.60 9.61 10.33 33.07 (30.00) 8.30 9.93 10.00 9.55 9.41 9.01 9.99 38.58 (35.00) 7.65 9.37 9.38 9.02 8.90 8.37 9.35 41.33 (37.49) 7.34 9.08 9.08 8.73 8.67 8.05 9.04 44.09 (40.00) 7.02 8.79 8.78 8.44 8.43 7.73 8.72 49.60 (45.00) 6.38 8.23 8.22 7.90 7.90 7.10 8.09 54.79 (49.71) 3.98 -- -- -- -- -- --55.12 (50.00) -- 7.70 7.69 7.38 7.39 5.64 7.46 58.16 (52.77) -- -- -- -- 4.21 --60.63 (55.00) -- 6.25 6.31 6.07 5.78 -- 5.56 62.85 (57.01) -- -- -- -- 4.53 63.13 (57.27) ..- -- -4.68 --63.50 (57.61) -- 4.93 4.99 4.83 --Page 26 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 Average Planar Exposure MAPLHGR Limit (kW/ft)Lat. Lat. Lat. Lat. Lat. Lat. Lat.GWd/MT (GWd/ST) 7357 7358 7359 7360 7361 7362 7363 63.54 (57.65) -- 4.91 -- -- -- -- --63.68 (57.77) .... 4.91 --......63.73 (57.82) ...... 1 4.73 ......Page 27 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision I Table 16.3-5 MAPLHGR Limits Bundle Type: GE 14-PI ODNAB392-17GZ-IOOT- 145-T6-2932 (GE 14C)Average Planar Exposure MAPLHGR Limit (kW/ft)Lat. Lat. Lat. Lat. Lat. Lat. Lat.GWd/MT (GWd/ST) 7357 7364 7365 7366 7367 7368 7369 0.00(0.00) 9.22 8.20 8.39 8.44 8.35 9.62 10.31 0.22(0.20) 9.14 8.24 8.44 8.49 8.40 9.56 10.29 1.10 (1.00) 8.94 8.34 8.52 8.59 8.53 9.42 10.23 2.20 (2.00) 8.88 8.45 8.63 8.73 8.71 9.39 10.23 3.31 (3.00) 8.90 8.58 8.74 8.87 8.89 9.42 10.26 4.41 (4.00) 8.93 8.70 8.86 9.02 9.07 9.46 10.30 5.51 (5.00) 8.97 8,83 8.98 9.17 9.25 9.51 10.34 6.61 (6.00) 9.01 8.96 9.08 9.30 9.38 9.55 10.37 7.72 ( 7.00) 9.04 9.10 9.20 9.45 9.55 9.59 10.40 8.82 (8.00) 9.07 9.21 9.32 9.62 9.72 9.61 10.43 9.92 (9.00) 9.09 9.33 9.46 9.78 9.89 9.64 10.45 11.02 (10.00) 9.11 9.46 9.58 9.93 9.98 9.65 10.46 12.13 (11.00) 9.12 9.58 9.70 10.06 10.02 9.66 10.47 13.23 (12.00) 9.13 9.69 9.81 10.16 10.07 9.67 10.47 14.33 (13.00) 9.13 9.80 9.89 10.14 10.05 9.67 10.47 15.43 (14.00) 9.13 9.89 9.96 10.06 9.99 9.67 10.47 16.53 (15.00) 9.13 9.97 10.01 10.03 9.97 9.67 10.47 18.74 (17.00) 9.13 10.10 10.10 10.01 9.96 9.66 10.46 22.05 (20.00) 9.11 10.22 10.21 9.99 9.95 9.64 10.44 27.56 (25.00) 8.95 10.30 10.41 9.96 9;93 9.61 10.41 33.07 (30.00) 8.30 10.07 10.08 9.85 9.82 9.01 10.11 38.58 (35.00) 7.65 9.54 9.55 9.24 9.22 8.37 9.47 41.33 (37.49) 7.34 9.28 9.29 8.96 8.94 8.05 9.15 44.09 (40.00) 7.02 9.02 9.03 8.67 8.65 7.73 8.84 49.60 (45.00) 6.38 8.47 8.50 8.12 8.12 7.10 8.21 54.79 (49.71) 3.98 -- -- -- -- -- --55.12 (50.00) -- 7.90 7.93 7.59 7.60 5.64 7.58 58.16 (52.77) -- -- -- -- 4.21 --60.63 (55.00) -- 6.25 6.30 6.41 6.40 -- 5.86 63.40 (57.52) -- -- -- -- 4.57 63.50 (57.61) -- 4.93 4.98 5.28 5.27 --63.57 (57.67) -- 4.90 -- -- -- -- '--Page 28 MONTICELLO Reload 24 0000-0083-9607-SRLR Reviqinn I Average Planar Exposure MAPLHGR Limit (kW/ft)GWd/MT (GWd/ST) Lat. Lat. Lat. Lat. Lat. Lat. Lat.7357 7364 7365 7366 7367 7368 7369 63.70 (57.79) -- -- 4.89 -- -- -- --64.44 (58.46) .....--. 4.90 ....64.48 (58.50) ...... 4.89 ......Page 29 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 Table 16.3-6 MAPLHGR Limits Bundle Type: GE 14-P 10DNAB424-14GZ-IOOT- 145-T6-3100 (GE 14C)Average Planar Exposure MAPLHGR Limit (kW/ft)Lat. Lat. Lat. Lat. Lat. Lat.. Lat.GWd/MT (GWd/ST) 8482 8483 8484 8485 8486 8487 8488 0.00 (0.00) 9.23 7.71 7.75 7.73 7.65 9.62 10.29 0.22 (0.20) 9.14 7.77 7.81 7.80 7.72 9.56 10.26 1.10(1.00) 8.94 -- -- 7.90 7.81 9.42 10.17 2.20(2.00) 8.88 7.95 8.00 8.02 7.93 9.39 10.15 3.31 (3.00) 8.89 -- -- 8.14 8.06 9.42 10.17 4.41 (4.00) 8.92 -- 8.22 8.27 8.19 9.46 10.20 5.51 (5.00) 8.95 8.25 8.33 8.40 8.33 9.50 10.24 6.61 (6.00) 8.99 8.36 8.44 8.53 8.47 9.54 10.27 7.72 (7.00) 9.02 8.46 8.56 8.67 8.62 9.58 10.29 8.82 (8.00) 9.05 8.57 8.68 8.82 8.77 9.61 10.32 9.92 (9.00) 9.07 8.69 8.80 8.97 8.94 9.63 10.33 11.02 (10.00) 9.09 8.80 8.92 9.13 9.11 9.65 10.34 12,13 (11.00) 9.10 8.92 9.03 9.26 9.29 9.66 10.35 13.23 (12.00) 9.11 9.04 9.09 9.34 9.40 9.66 10.35 14.33 (13.00) 9.11 9.11 9.17 9.44 9.50 9.67 10.35 15.43 (14.00) 9.11 9.17 9.25 -- 9.61 9.66 10.35 16.53 (15.00) 9.11 9.25 -- 9.65 9.68 9.66 10.35 17.64 (16.00) 9.11 9.33 9.42 9.74 9.72 9.66 10.34 18.74 (17.00) 9.11 -- 9.51 9.82 9.76 9.65 10.34 19.84 (18.00) 9.10 9.51 9.58 9.89 9.78 9.65 --20.94 (19.00) 9.10 9.59 9.65 9.93 9.80 --..22.05 (20.00) -- 9.67 9.71 9.94 9.81 -- --23.15 (21.00) 9.09 9.74 9.76 9.95 -- 9.62 10.31 24.25 (22.00) -- 9.80 9.82 -- -- -- --25.35 (23.00) 9.08 9.86 -- .9.83 9.61 10.30 26.46 (24.00) 9.06 9.91 9.92 -- -- 9.61 10.29 27.56 (25.00) 8.93 9.86 9.88 9.96. 9.84 9.60 10.29 33.07 (30.00) 8.28 9.49 9.49 9.84 9.74 9.00 9.94 38.58 (35.00) 7.64 9.09 9.10 9.34 9.21 8.35 9.30 41.33 (37.49) 7.32 8.87 8.88 9.09 8.96 8.04 8.99 44.09 (40.00) 7.00 8.66 8.67 8.84 8.71 7.72 8.67 49.60 (45.00) 6.36 8.18 8.19 8.35 -- 7.09 8.04 Page 30 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision I Average Planar Exposure MAPLHGR Limit (kW/ft)Lat. Lat. Lat. Lat. Lat. Lat. Lat.GWd/MT (GWd/ST) 8482 8483 8484 8485 8486 8487 8488 54.68 (49.61) 3.98 -- -- -- -- -- --55.12 (50.00) -- 7.45 7.48 7.87 7.75 5.60 7.41 58.10 (52.71) -- -- -- -- 4.21 --60.63 (55.00) -- 4.97 5.00 5.92 6.11 5.44 60.90 (55.25) -- 4.85 -- -- ----60.96 (55.30) .... 4.85 ........62.63 (56.82) ............ 4.51 62.80 (56.97) ...... 4.93 .-- -.63.16 (57.30) ......-- 4.95 ....Page 31 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision I Table 16.3-7 MAPLHGR Limits Bundle Type: GE 14-P 1 ODNAB375-16GZ-IOOT-1 45-T6-3 101 (GE 1 4C)Average Planar Exposure MAPLHGR Limit (kW/ft)Lat. Lat. Lat. Lat. Lat. Lat. Lat.GWd/MT (GWd/ST) 8482 8489 8490 8491 8492 8493 8494 0.00 (0.00) 9.23 8.23 8.23 8.28 8.18 9.62 10.15 0.22 (0.20) 9.14 8.27 8.28 8.33 8.24 9.56 10.13 1.10(1.00) 8.94 8.32 8.36 8.42 8.33 9.42 10.05 2.20 (2.00) 8.88 -- 8.43 -- 8.45 9.39 10.05 3.31 (3.00) 8.89 -- 8.64 8.57 9.42 10.09 4.41 (4.00) 8.92 -- 8.55 8.71 8.70 9.46 10.13 5.51 (5.00) 8.95 8.58 8.61 8.78 8.83 9.50 10.17 6.61 (6.00) 8.99 8.65 8.68 8.86 8.94 9.54 10.21 7.72 (7.00) 9.02 8.72 8.77 8.98 9.06 9.58 10.24 8.82(8.00) 9.05 8.80 8.87 9.11 9.19 9.61 10.27 9.92 (9.00) 9.07 8.89 8.99 9.26 9.34 9.63 10.29 11.02 (10.00) 9.09 9.00 9.11 9.40 9.49 9.65 10.30 12.13 (11.00) 9.10 9.11 9.23 9.55 9.49 9.66 10.31 13.23 (12.00) 9.11 9.23 9.35 9.43 9.26 9.66 10.32 14.33 (13.00) 9.11 9.35 9.46 9.36 9.21 9.67 10.32 15.43 (14.00) 9.11 9.46 9.56 9.36 9.22 9.66 10.31 16.53 (15.00) 9.11 9.57 9.65 9.39 9.26 9.66 10.31 17.64 (16.00) 9.11 9.66 9.73 9.42 9.29 9.66 10.30 18.74 (17.00) 9.11 9.64 9.73 9.44 9.31 9.65 10.30 19.84 (18.00) 9.10 9.64 9.74 9.45 9.32 9.65 --20.94 (19.00) 9.10 9.64 9.74 9.46 -- -- 10.28 22.05 (20.00) -- 9.65 9.75 -- -- --23.15 (21.00) 9.09 9.66 9.75 -- .9.62 10.27 24.25 (22.00) -- 9.67 -- 9.49 9.37 -- --25.35 (23.00) 9.08 9.67 -- 9.50 -- 9.61 10.25 26.46 (24.00) 9.06 -- -- 9.51 9.39 9.61 --27.56 (25.00) 8.93 9.68 9.75 9.53 9.41 9.60 10.24 33.07 (30.00) 8.28 9.60 9.61 9.29 9.15 9.00 9.88 38.58 (35.00) 7.64 9.02 9.02 -- 8.66 8.35 9.24 41.33 (37.49) 7.32 8.75 8.75 8.54 8.44 8.04 8.92 44.09 (40.00) 7.00 8.47 8.47 8.29 8.21 7.72 8.60 49.60 (45.00) 6.36 7.95 7.95 7.81 7.77 7.09 7.97 Page 32 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision I Average Planar Exposure MAPLHGR Limit (kW/ft)Lat. Lat. Lat. Lat. Lat. Lat. Lat.GWd/MT (GWd/ST) 8482 8489 8490 8491 8492 8493 8494 54.68 (49.61) 3.98 -- -- -- -- --55.12 (50.00) -- 7.45 7.44 7.35 7.32 5.60 7.34 58.10 (52.71) -- -- -- -- 4.21 --60.63 (55.00) -- 5.46 5.50 5.52 5.19 -- 5.27 61.88 (56.14) -- -- -- 4.64 ..--61.93 (56.19) -- 4.87 ...-- --..62.00 (56.24) .-- 4.87 ......--62.29 (56.51) ......-- --. 4.49 62.53 (56.72) ...... 4.69 ......Page 33 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 Table 16.3-8 MAPLHGR Limits Bundle Type: GE 14-P 1ODNAB392-16GZ-1OOT- 145-T6-3102 (GE 14C)Average Planar Exposure MAPLHGR Limit (kW/ft)Lat. Lat. Lat. Lat. Lat. Lat. Lat.GWd/MT (GWd/ST) 8482 8495 8496 8497 8498 8499 8500 0.00 (0.00) 9.23 8.33 8.53 8.62 8.57 9.62 10.22 0.22(0.20) 9.14 8.37 8.56 8.66 8.62 9.56 10.20 1.10(1.00) 8.94 8.44 8.62 8.73 -- 9.42 10.12 2.20 (2.00) 8.88 8.53 8.70 8.83 8.81 9.39 10.12 3.31 (3.00) 8.89 -- -- 8.92 8.92 9.42 10.15 4.41 (4.00) 8.92 8.73 8.87 9.03 9.03 9.46 10.19 5.51 (5.00) 8.95 8.83 8.94 9.10 9.15 9.50 10.23 6.61 (6.00) 8.99 8.92 9.01 9.18 9.26 9.54 10.27 7.72 (7.00) 9.02 9.00 9.08 9.28 9.36 9.58 10.30 8.82 (8.00) 9.05 9.09 9.17 9.40 9.49 9.61 10.33 9.92 (9.00) 9.07 9.19 9.28 9.54 9.64 9.63 10.35 11.02 (10.00) 9.09 9.30 9.39 9.69 9.79 9.65 10.36 12.13 (11.00) 9.10 -- 9.51 9.84 9.75 9.66 10.37 13.23 (12.00) 9.11 9.54 9.63 9.68 9.51 9.66 10.37 14.33 (13.00) 9.11 9.66 9.73 9.58 9.44 9.67 10.37 15.43 (14.00) 9.11 9.77 9.83 9.58 9.44 9.66 10.37 16.53 (15.00) 9.11 9.87 9.91 -- 9.47 9.66 10.36 17.64 (16.00) 9.11 9.96 9.98 9.63 9.50 9.66 10.36 18.74 (17.00) 9.11 9.98 10.04 9.65 9.52 9.65 10.35 19.84 (18.00) 9.10 9.98 10.05 9.66 9.53 9.65 --20.94 (19.00) 9.10 9.98 10.06 9.67 9.54 -- --23.15 (21.00) 9.09 9.99 10.06 -- 9.56 9.62 10.33 24.25 (22.00) -- -- -- 9.69 -- -- --25.35 (23.00) 9.08 9.99 -- -- 9.57 9.61 10.31 26.46 (24.00) 9.06 9.98 10.06 9.71 -- 9.61 10.31 27.56 (25.00) 8.93 9.97 10.05 9.72 9.60 9.60 10.30 33.07 (30.00) 8.28 9.94 9.99 9.54 9.39 9.00 9.96 38.58 (35.00) 7.64 9.38 9.38 9.03 8.89 8.35 9.32 41.33 (37.49) 7.32 9.08 9.08 8.77 8.65 8.04 9.00 44.09 (40.00) 7.00 8.78 8.78 8.52 8.41 7.72 8.69 49.60 (45.00) 6.36 8.22 8.21 7.98 7.95 7.09 8.06 54.68 (49.61) 3.98 ...-- -- -- -- --Page 34 MONTICELLO Reload 24 0000-0083-9607-SRLR Revisinn 1 Average Planar Exposure MAPLHGR Limit (kW/ft)Lat. Lat. Lat. Lat. Lat. Lat. Lat.GWd/MT (GWd/ST) 8482 8495 8496 8497 8498 8499 8500 55.12 (50.00) -- 7.68 7.67 7.47 7.46 5.60 7.43 58.10 (52.71) ----- -- 4.21 --60.63 (55.00) -- 6.18 6.23 6.05 5.71 -- 5.48 62.69 (56.87) ......-- --. 4.52 63.00 (57.16) .. --... 4.67 ..--63.41 (57.53) -- 4.90 ......--.. 63.50 (57.61) .... 4.91 4.79 ......63.52 (57.62) .... 4.90 --.......63.66 (57.75) ....-- 4.72 ......Page 35 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 Table 16.3-9 MAPLHGR Limits Bundle Type: GE14-PI0DNAB391-12GZ-1OOT-145-T6-3103 (GE14C)Average Planar Exposure MAPLIIGR Limit (kW/ft)GWdIMT (GWd/ST) Lat. Lat. Lat. Lat. Lat. Lat. Lat.8482 8501 8502 8503 8504 8505. 8506 0.00 (0.00) 9.23 8.72 8.73 8.85 8.78 9.62 9.95 0.22 (0.20) 9.14 8.75 8.75 8.88 8.81 9.56 9.92 1.10( 1.00) 8.94 8.79 8.79 8.93 8.90 9.42 9.83 2.20 (2.00) 8.88 8.85 8.86 9.01 9.02 9.39 9.82 3.31 (3.00) 8.89 -- 8.92 9.09 9.14 9.42 9.85 4.41 (4.00) 8.92 8.98 8.99 -- 9.23 9.46 9.90 5.51 (5.00) 8.95 9.05 9.06 9.25 9.33 9.50 9.95 6.61 (6.00) 8.99 9.11 9.11 9.31 9.39 9.54 9.99 7.72 (7.00) 9.02 9.18 9.17 9.38 9.47 9.58 10.02 8.82 (8.00) 9.05 9.24 9.24 9.48 9.57 9.61 10.05 9.92 (9.00) 9.07 9.30 9.33 9.60 9.69 9.63 10.07 11.02 (10.00) 9.09 9.37 9.43 -- 9.83 9.65 10.09 12.13 (11.00) 9.10 9.46 -- 9.85 9.72 9.66 10.09 13.23 (12.00) 9.11 9.56 9.63 9.66 9.49 9.66 10.10 14.33 (13.00) 9.11 -- 9.73 9.58 9.43 9.67 10.10 15.43 (14.00) 9.11 9.76 9.82 9.59 9.45 9.66 10.10 16.53 (15.00) 9.11 9.86 9.90 9.61 9.48 9.66 10.09 17.64 (16.00) 9.11 9.94 9.97 9.64 9.51 9.66 10.09 18.74 (17.00) 9.11 9.98 10.03 9.66 9.52 9.65 10.08 19.84 (18.00) 9.10 9.98 10.06 9.67 9.53 9.65 --20.94 (19.00) 9.10 9.98 10.06 9.68 -- --.22.05 (20.00) -- 9.98 10.06 -- -- --23.15 (21.00) 9.09 -- -- -- 9.56 9.62 10.05 24.25 (22.00) -- -- 9.70 -- -- 10.04 25.35 (23.00) 9.08 9.98 -- 9.71 9.58 9.61 --26.46 (24.00) 9.06 9.98 10.06 9.72 -- 9.61 10.03 27.56 (25.00) 8.93 9.97 10.05 9.73 9.60 9.60 10.02 33.07 (30.00) 8.28 9.94 9.99 9.54 9.40 9.00 9.58 38.58 (35.00) 7.64 9.38 9.38 9.03 8.89 8.35 8.94 41.33 (37.49) 7.32 9.09 9.08 8.78 8.65, 8.04 8.62 44.09 (40.00) 7.00 8.79 8.78 8.52 8.42 7.72 8.30 49.60 (45.00) 6.36 8.23 8.22 7.98 7.96 7.09 7.67 Page 36 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 Average Planar Exposure MAPLHGR Limit (kW/ft)GWd/MT (GWd/ST) Lat. Lat. Lat. Lat. Lat. Lat. Lat.8482 8501 8502 8503 8504 8505 8506 54.68 (49.61) 3.98 -- -- -- -- -- --55.12 (50.00) -- 7.69 7.68 7.47 7.46 5.60 7.04 58.10 (52.71) ...-- -... 4.21 --60.63 (55.00) -- 6.23 6.24 6.04 5.71 -- 4.50 60.86 (55.21) -- -- -- --. 4.40 62.99 (57.14) -- --... 4.67 ..--63.50 (57.61) -- 4.92 4.93 4.79 --....63.60 (57.70) -- 4.88 --......... 63.61 (57.71) ..-- 4.88 --......63.64 (57.73) ....-- 4.73 ......Page 37 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 Table 16.3-10 MAPLHGR Limits Bundle Type: GE 14-P 1 ODNAB391-14GZ-I OOT- 145-T6-2480 (GE 1 4C)Average Planar Exposure MAPLHGR Limit (kW/ft)Lat. Lat. Lat. Lat. Lat. Lat. Lat.GWd/MT(GWd/ST) 5129 5130 5131 5132 5133 5134 5135 0.00 (0.00) 9.22 8.41 8.63 8.42 8.37 9.61 10.28 0.22 (0.20) 9.14 8.45 8.66 8.47 8.43 9.56 10.26 1.10 (1.00) 8.94 8.54 8.74 8.58 8.56 9.42 10.17 2.20 (2.00) 8.88 8.65 8.84 8.73 8.74 9.39 10.16 3.31 (3.00) 8.90 8.77 8.96 8.89 8.90 9.42 10.18 4.41 (4.00) 8.93 8.90 9.08 9.05 9.08 9.46 10.21 5.51 (5.00) 8.97 9.03 9.18 9.23 9.26 9.51 10.24 6.61 (6.00) 9.01 9.16 9.27 9.40 9.45 9.55 10.27 7.72 (7.00) 9.04 9.27 9.38 9.55 9.63 9.58 10.29 8.82 (8.00) 9.07 9.39 9.49 9.72 9.80 9.61 10.31 9.92 (9.00) 9.09 9.51 9.60 9.88 9.98 9.64 10.33 11.02 (10.00) 9.11 9.63 9.71 10.03 10.13 9.65 10.34 12.13 (11.00) 9.12 9.75 9.80 10.14 10.24 9.66 10.34 13.23 (12.00) 9.13 9.84 9.88 10.22 10.24 9.67 10.34 14.33 (13.00) 9.13 9.92 9.94 10.28 10.23 9.67 10.34 15.43 (14.00) 9.13 9.98 9.99 10.29 10.23 9.67 10.34 16.53 (15.00) 9.13 10.03 10.03 10.29 10.23 9.66 10.33 18.74 (17.00) 9.12 10.10 10.10 10.28 10.22 9.65 10.32 22.05 (20.00) 9.11 10.21 10.20 10.26 10.21 9.64 10.30 2756 (25.00) 8.95 10.40 10.39 10.23 10.19 9.61 10.27 33.07 (30.00) 8.30 10.07 10.08 10.10 10.04 9.01 9.93 38.58 (35.00) 7.65 9.54 9.55 9.49 9.44 8.36 9.29 41.33 (37.49) 7.34 9.28 9.29 9.20 9.15 8.05 8.97 44.09 (40.00) 7.02 9.02 9.03 8.91 8.87 7.73 8.65 49.60 (45.00) 6.38 8.49 8.50 8.37 8.33 7.09 8.02 54.78 (49.70) 3.98 -- -- -- -- -- --55.12 (50.00) -- 7.92 7.93 7.85 7.81 5.63 7.39 58.14 (52.75) --- -- -- 4.21 --60.63 (55.00) 6.26 6.29 6.60 6.60 -- 5.40 62.55 (56.74) -- -- -- -- 4.51 63.50 (57.61) 4.95 4.98 5.29 5.29 --63.72 (57.81) 4.85 -- -- -- -- --Page 38 MONTICELLO RP1~ lnt"94 0000-0083-9607-SRLR 1p n 1 Average Planar Exposure MAPLHGR Limit (kW/ft)Lat. Lat. Lat. Lat. Lat. Lat. Lat.GWd/MT (GWd/ST) 5129 5130 5131 5132 5133 5134 5135 63.79 (57.87) -- -- 4.85 -- -- -- --64.37 (58.40) ...-- -- 4.90 ....64.39 (58.42) ..-- 1 4.89 .-- I --The single loop operation multiplier on MAPLHGR, and the ECCS analytical initial MCPR values applicable to each fuel type in the new cycle core, are shown in the following table. The ECCS MAPLHGR multiplier for core flow rates at or below 80% of rated core flow is 0.94 for GE14C.Table 16.3-11 Initial MCPR and Single Loop Operation Multiplier on MAPLHGR Fuel Type Initial MCPR Single Loop Operation Multiplier on MAPLHGR GE14C 1.350 0.90 16.4 References The SAFER/GESTR-LOCA analysis base reports applicable to the new cycle core are listed below.References for GE14C 1. Monticello ECCS-LOCA Evaluation for GEJ4, GE-NE-J1 103878-09-02P, August 2001 2. Monticello Nuclear Plant GEJ4 ECCS-LOCA Evaluation with the RHR Intertie Line Open, NSA 01-459, October 10, 2001 Page 39 MONTICELLO 0000-0083-9607-SRLR Reload 24 Revision I 52 E] E] AM 50 48 [E]__IF [DF]_ _ f] [ E__E] E] [E] [IE] [E]_rq[E]46 BE~~~~] FLF] E]IFI ] IEI--EIEl E 46 44 D][ _I.E" E El__.._rE] DD E[1] DOl..r EIE._I.K" 10 [] []E 42 MEl DIE]IDID M-E'RO'E] ENJD ElQ[0E 40 El E] I El BE 38 .I lF] DIEI M E]E 36 DO [ RE] D] c'.Ul"F2 '._DO DID E10 FRE DIJ-l].D E]iE] ][[ ]34 Emm Fm 32 E] [Do ED] EK] EF] E] F2__] ITLD ] Q] F2 ] FL_] TL]_E [E] F2[] F2][] E+][[] r QLEC ]__I 30 [E 2 E l19E iFE] JLll[] DD1 II[ MEli] DMEFE lF2 lEl Fal FI 28 FRL] [EI.5FI D QU]1 1-I] +0 DO E11 DOZC 0_0I- 10 El D] El[][16 F2 EY] ET] E__1'0 E]lE [D E] [] EF] [.1 EqE] [LF] El F] E] E ME]24 B 10 El E R LUD []__L-El~rI ME [][__FLIED E1_21 Eil_2- END D] 19 111]E 22K 20 F IE [Es] E [LE QLFR Eli 0 DIEF III- [El El (El ElDE l l1 18 I lE E 1 1 IEIFI M IFIE] U]EFEEEIE1 14 ED ]1] M ] E__.] E]_2 F EE EI ]1F1E1 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 Fuel Type A=GE I 4-P10DNAB393-I7GZ-I 00T- 145-T6-2598 (Cycle 22) G=GE14-P IODNAB392-I6GZ-I00T-145-T6-2931 (Cycle 25)B=GEI4-P1ODNAB393-17GZ-100T-145-T6-2599 (Cycle 22) H=GE14-PIODNAB424-14GZ-100T-145-T6-3 100 (Cycle 25)C=GE14-PIODNAB393-17GZ-100T-145-T6-2599 (Cycle 23) I=GE14-P10DNAB375-16GZ-100T-145-T6-3 101 (Cycle 25)D=GE14-PIODNAB392-16GZ-100T-145-T6-2824 (Cycle 23) J=GE14-P10DNAB392-16GZ-100T-145-T6-3102 (Cycle 25)E=GE14-PI0DNAB392-16GZ-IOOT-145-T6-2931 (Cycle 24) K=GE14-PIODNAB391-12GZ-1OOT-145-T6-3103 (Cycle 25)F=GE14-PI0DNAB392-17GZ-100T-145-T6-2932 (Cycle 24) L=GE14-PIODNAB391-14GZ-100T-145-T6-2480 (Cycle 21)Figure 1 Reference Core Loading Pattern Page 40 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 300.0 2bU..-a- Dome Press Rise (psi)Safety Valve Raw A Relief Vahte Raow--Bypass Valve Flow 200.0-150.0-100.0-50.0]U.O ." -" : .3 0.0 10.0 20.0 30.0 40.0 Time (sec)0.0 10.0 20.0 30.0 Time (see)40.0 150.0 C C 2 0 Q¶0 U U 50.0 0.0 10.0 20.0 30.0 40.0 Time (sec)0.0 10.0 20.0 30.0 Tine (sec)40.0 Figure 2 Plant Response to FW Controller Failure (EOC STANDARD (HBB))Page 41 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 30U 2W.0 1003.0 0.0 0.0 30 6.0 0.0 3.0 Time (sec) Tine (sec)6.0 100.0n 6 6 0 2 0 U 0.0 3.0'ulia (sec)6.0 0.0 3.0 T-ne (sec)6.0 Figure 3 Plant Response to Turbine Trip with Bypass (EOC STANDARD (HBB))Page 42 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision I 150.0 100.0 50.0 0.0 30 6.0 0.0 3.0 6.0 Tine (sec) Time (sec)r 5 0.0 3.0 6.0 0.0 3.0 Time (sec) Tine (sec)6.0 Figure 4 Plant Response to Turbine Trip w/o Bypass (EOC STANDARD (HBB))Page 43 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision I 0.0 3.0 6.0 0.0 3.0 Time (sec) Time (sec)6.0 r 0.U 0.0 3.0 Tine (sec)6.0 0.0 3.0 Time (sec)6.0 Figure 5 Plant Response to Load Rejection w/o Bypass (EOC STANDARD (HBB))Page 44 MONTICELLO 0000-0083-9607-SRLR Reload 24 Revision I 0.0 I -0.0 10.0 20.0 30.0 0.0 10.0 20.0 30.0 Tune (sec) Tite (Sec)150.0 100.0 50.0 a Level(inch-REF-SEP-SKRT) -)- Vessel Steam Row a; Turbine Steam Flow o Feedwater Row rJ B e^^ý I .. ..nn1 I-- --- jfi! A 9 ...0.0 10.0 20.0 Tine (sec)30.0 0.0 10.0 20.0 Tine (sec)30,0 Figure 6 Plant Response to Inadvertent HPCI /L8 (EOC STANDARD (HBB))Page 45 MONTICELLO Pp~ln~- 0000-0083-9607-SRLR D 1 1 Reload 24 v a U11 0.0 10.0 20.0 30.0 40.0 Time (sec)0.0 10.0 20.0 30.0 40.0 Time (sec)-E- LeveI(inch-REF-SEP-SKRT)-x- Vessel Steam Row 150.0 -a-- Turbine Steam Flow s Feedwater Row 50.0 U S U.U I i i : A, 96 0.0 10.0 20.0 30.0 40.0 Time (see)0.0 10.0 20.0 30.0 Time (sec)40.0 Figure 7 Plant Response to FW Controller Failure ( EOC MELLLA (HBB))Page 46 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 0.0 3.0 6.0 0.0 30 Tinr (sec) Time (sec)6.0 200.0 100.0 0.0-100.0 0 5 0.0 3.0 Time (Sec)6.0 0.0 3.0 Time (sec)6.0 Figure 8 Plant Response to Turbine Trip with Bypass (EOC MELLLA (HBB))Page 47 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 150.0 100.0 50.0 0.0 0.0 3.0 6.0 0.0 3.0 Time (sec) Time (sec)6.0 U U 0.0 3.0 6.0 0.0 3.0 Time (sec) Tine (see)6.0 Figure 9 Plant Response to Turbine Trip w/o Bypass (EOC MELLLA (HBB))Page 48 MONTICELLO PP1lnnd 9L1.0000-0083-9607-SRLR Revision I Reload 24 150.0 100.0 5100 0.0 3.0 6.0 0.0 3.0 6.0 Tine (sec) Tife (sec)4)4)C 0 S 0 U (4 44 4)0.0 3.0 Time (sec)6.0 0.0 3.0 6.0 Time (sec)Figure 10 Plant Response to Load Rejection w/o Bypass I (EOC MELLLA (HBB))Page 49 MONTICELLO R~ln~id Lt 0000-0083-9607-SRLR D Reload 24 11 ý -350.0 300.250.0.0200.0 150.0 100.0 50.0 0.0 0 150.0-100.0 50.0.0 10.0 20.0 T-T (sec)30.0 0.0 10.0 20.0 30.0 Tinx (see)r 0 0.0 10.0 20.0 30.0 Time (sec)0.0 10.0 20.0 Time (sec)30.0 Figure 11 Plant Response to Inadvertent HPCI /L8 (EOC MELLLA (HBB))Page 50 MONTICELLO Rplnnci 94 0000-0083-9607-SRLR Reload 24 200.0 150.0 100.0 50.0 0.0-a- Dome Press Rise (psi)---- Safety Valve Raw A. Relief Valve Row--- Bypass Valve Flow Z 1oEI ..... ... '0.0 10.0 20.0 30.0 40.0 Tune (sec)0.0 10.0 20.0 30.0 Time (sec)40.0 1.0 0.0 E 09 -1.0-2.0 0.0 10.0 20.0 30.0 40.0 0.0 10.0 20.0 30.0 Time (sec) Tine (sec)Figure 12 Plant Response to FW Controller Failure (EOC STANDARD (UB))40.0 Page 51 MONTICELLO 0000-0083-9607-SRLR Revision 1 Rel-l 24 150,0 100.0 50.0 0.0 0.0 3.0 6.0 0.0 3,0 Time (sec) Time (sec)6.0 100.0 E 0 U U 0.0 3.0 Tkne (sec)6.0 0.0* 3.0 Time (sec)6.0 Figure 13 Plant Response to Turbine Trip with Bypass (EOC STANDARD (UB))Page 52 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 150.0 100.0-a- Dome Press Rise (psi)--- Safety Va1ve Row--Relief Valve Row Bypass Valve Raw 200.0-d 50.0 0.0 100.0+U.U +0.0 3.0 Time (see)6.0 0.0 3T 0 Tnne (sec)6.0 r U 0.0 3.0 6.0 0.0 3.0 Tine (sec) TiMe (sec)6.0 Figure 14 Plant Response to Turbine Trip w/o Bypass (EOC STANDARD (UB))Page 53 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 0.0 3.0 6.0 0.0 3.0 Tine (sec) Time (sec)6.0 U 0 E 0 U 0.0 3.0 6.0 0.0 3.0 Ti-e (see) Ti-e (sec)6.0 Figure 15 Plant Response to Load Rejection w/o Bypass (EOC STANDARD (UB))Page 54 MONTICELLO Reload 24..0000-0083-9607-SRLR Revision 1 S1,50.0 0.0 10.0 20.0 30.0 Tine (sec)0.0 100 20.0 30.0 Tint (sec)150.0 100.0=1 U U CS 0.0 "10.0 20.0 30.0 0.0 10.0 20.0 30.0 Tine (5ec) Tinm (sec)Figure 16 Plant Response to Inadvertent HPCI /L8 (EOC STANDARD (UB))Page 55 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 200.0 150.0 100.0 50.0 0.0 0.0 10.0 20.0 30.0 40.0 Tine (sec), 0.0 10.0 20.0 30.0 Time (sec)40.0 150.0 100.0 50.0 0.0 1.0-0.010 (J U 5,-a- Void Reactivity

• -- Doppler ReactMty* Scram Reactivity-s- Total Reactivity 0.0 10.0 20.0 30.0 Tine (sec)40.0 0.0 10.0 20.0 30.0 Time (sec)40.0 Figure 17 Plant Response to FW Controller Failure (EOC MELLLA (UB))Page 56 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision I 3130.0-3W0.200.0 --E-- Dome Press Rise (psi)--- Safety Valve Flow Relief Valve Flow---Bypass Valve Flow 100.0 I 0.0 3.0 Tiue (ec)6.0 0.0 1.0 0.--1.0 t 3.0 Tnne (sec)6.0 0.0 3.0 6.0 0.0 3.0 Time (sec) Tune (see)Figure 18 Plant Response to Turbine Trip with Bypass (EOC MELLLA (UB))6.0 Page 57 MONTICELLO Reload 24 OOOO-0083-9607-SRLR Rpvisqinn 1 0.0 3.0 6.0 0.0 3.0 Timne (sec) Tmie (sec)6.0 200.0 100.0 0.0 C 2 C r~)U U-100.0 0.0 3.0 Time (sec)6.0 0.0 3.0 Time (sec)6.0 Figure 19 Plant Response to Turbine Trip w/o Bypass (EOC MELLLA (UB))Page 58 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 0.0 3.0 6.0 0.0 3.0 Tine (sec) Tine (sec)6.0"o 10.0 1.0 0- 0.0 E-10-2.0 0.0 3.0 Ti.m (see)6.0 0.0 3.0 Time (sec)6.0 Figure-20 Plant Response to Load Rejection w/o Bypass (EOC MELLLA (UB))Page 59 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 0.0 10.0 20.0 30.0 0.0 10.0 20.0 30.0 Time (sec) Tine (sec)150.0 100.0 50.0 0.0 0 0 2 0 U C-, 0.0 10.0 20.0 30.0 0.0 "10.0 20.0 Time (sec) Tine (sec)Figure 21 Plant Response to Inadvertent HPCI /L8 (EOC MELLLA (UB))30.0 Page 60 MONTICELLO PplnnA ')A 0000-0083-9607-SRLR D 1ýv a M-0.0 4.0 8.0 0.0 4.0 Time (c0 Time (c0 8.0 200.0 100.0-E- LeveI(inch-REF-SEP-SKRT)-x- Vesse Steam Row-Turbine Steam Row Feedwater Flow 1.0 ro 0.0 -E-2.0o-.0 0.0 0.04 8 0.0 4.0 Time (sec)4.0 limne (sec)8.0 Figure 22 Plant Response to MSIV Closure (Flux Scram) -STANDARD (HBB)Page 61 MONTICELLO V1 1A 0000-0083-9607-SRLR D 1 flL1~JU try 1w1~1u v I 0.0 4.0 .0 0.0 4.0 Tine (sec) Time (sec)8.0 200.0 100.0 0.0 0 0 C E C Q U U-100.0 0.0 4.0 Time (see)8.0 0.0 4.0 T'nh (see)8.0 Figure 23 Plant Response to MSIV Closure (Flux Scram) -MELLLA (HBB)Page 62 MONTICELLO 0000-0083-9607-SRLR Reload 24 Revision 1 110.00 II--e-- Natural Circulation Line , I III 6 Extended Operating Domain 10 0 .0 0- ------------ --------- BSP Scram Region Boundary BSP Controlled Entry Region Boundary 80.00 ,- -, T o II I Ii I i 7 0 .0 0 ---------------------------------------- L J ------LI ------40.00 --..... -------------------

Ei 4 0u.00 ,- ----" -- ------------ ----------------:-I I I 0.00 ............ ... +. ,_ .0 10 20 30 40 50 60 70 80 90 100 110 Core Flow (0/)Figure 24 BSP Region Boundaries Page 63 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision I Appendix A Analysis Conditions The reactor operating conditions used in the reload licensing analysis presented in Table A-1. The pressure relief and safety valve configuration Table A-2. Additionally, the operating flexibility options listed in Section licensing analysis.for this plant and cycle are for this plant are presented in 8 are supported by the reload Table A-1 Reactor Operating Conditions Analysis Value Parameter RCF LCF NFWT NFWT Thermal power, MWt 1775.0 1775.0 Core flow, Mlb/hr 57.6 47.5 Reactor pressure (core mid-plane), psia 1040.0 1036.8 Inlet enthalpy, Btu/lb 523.7 518.5 Non-fuel power fraction 0.036 0.036 Steam flow, Mlb/hr 7.27 7.26 Dome pressure, psig 1010.0 1010.0 Turbine pressure, psig 960.2 960.3 Table A-2 Pressure Relief and Safety Valve Configuration Valve Type Number of Lowest Setpoint Valves (psig)Safety/Relief Valve 8 1170.0 Page 64 MONTICELLO V 1 -1 ')A 0000-0083-9607-SRLR MOTCEL 00008-67SL Appendix B Decrease in Core Coolant Temperature Events The Loss-of-Feedwater Heater event was analyzed at 100% rated power using the BWR Simulator Code.The use of this code is permitted in GESTAR II. The transient plots, neutron flux and heat flux values normally reported in Section 9 are not an output of the BWR Simulator Code; therefore, those items are not included in this document. The OLMCPR result is shown in Section 11.Page 65 MONTICELLO 0000-0083-9607-SRLR Reload 24 Revision 1 Appendix C ARTS Off-Rated Limits Curves The ARTS off-rated limits have been evaluated for Cycle 25. This evaluation concluded that no changes to any of the off-rated limits are necessary for Cycle 25. The Kp limits developed in Cycle 23 (Reference C-i) to accommodate the potential use of Option B operating limits are applicable to Cycle 25. All other limits, as defined in Reference C-2, are not changed and still apply.A pressure regulator OOS analysis was performed for Monticello in Reference C-3. The PROOS limits contained in Reference C-3 are applicable between 45% and 100% power and are independent of flow.The existing limits below Pbypass (45% power) are bounding for a PROOS. The MAPFACp and LHGRFACp limits are bounded by the current limits, i.e., unique PROOS MAPFACp and LHGRFACp limits are not required. The limits determination, for both OLMCPR and MAPFACp and LHGRFACp, contained sufficient conservatism such that they are generic for GE 14 fuel and do not need to be analyzed on a cycle specific basis unless plant design changes invalidate key analysis inputs.References C-1. Supplemental Reload Licensing Report for Monticello Nuclear Generating Plant Reload 22 Cycle 23, 0000-0029-6441-SRLR, Rev. 0, January 2005.C-2. GEJ4 Fuel Design, Cycle Independent Transient Analyses for Monticello Generating Plant, GE-NE-0000-0014-7048-01P, Rev. 0, March 2003.C-3. Nuclear Management Company Monticello Nuclear Generating Plant Pressure Regulator Downscale Failure Analysis, GE-NE-0000-0051-2643-RO, Rev. 0, September 2007.Page 66 MONTICELLO 0000-0083-9607-SRLR Reload 24 Revision 1 Appendix D Expanded Operating Domain Analyses To provide Monticello Nuclear Generating Plant with operating improvements, expanded operating domain analyses were performed for maximum extended load line limit (MELLL) operation and for increased core flow (ICF) operation up to 105% of rated flow at 94.1% power.Coastdown operation beyond full power exposure to 42.3% power under conditions bounded by 112%core flow is conservatively bounded by the MCPR operating limits given in Section 11 of this document at the applicable core flow and feedwater temperature conditions in the expanded operating domain.(Reference D-3).100% core flow (Standard Domain)The standard (rated) operating domain has been calculated and results reported in Section 11.Maximum Extended Load Line Limit: The operation domain MELLL was established for Monticello Nuclear Generating Plant in Reference D-2.Increased Core Flow: Operation with ICF throughout the operating cycle was justified for Monticello Nuclear Generating Plant in Reference D-1. With the introduction of power uprate the ICF capability was not licensed at current rated power. Therefore, for Cycle 25 it was established that the rated operating limits calculated as part of this report when multiplied by the appropriate off-rated multiplier (i.e. Kp) at powers between 94.1% and 42.3% operation with ICF throughout the cycle is bounded.Turbine Trip with Bypass and Degraded Scram See Appendix H for a discussion on this event.Rod Block Monitor (RBM) Considerations The ARTS Rod Withdrawal Error (RWE) analysis validated that the following MCPR values provide the required margin for full withdrawal of any control rod during Monticello Nuclear Generating Plant Cycle 25: For power < 90%: MCPR > 1.75 For power > 90%: MCPR > 1.44 Below is summarized the MCPR limits based on a 1.10 safety limit.Page 67 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision I Trip Set Pt.Without RBM Filter Required Cycle 25 Rated MCPR Limit based on ARTS Generic MCPR for a SLMCPR=1.10 1.35 Bounded by Kp Bounded by Kp HTSP -114.0%ITSP -119.0%LTSP -124.0%References D- 1. Safety Review of Monticello Nuclear Generating Plant Increased Flow Operation Throughout Cycle 14, NEDC-31778P, December 1989.D-2. Maximum Extended Load Line Limit Analysis for Monticello Nuclear Generating Plant Cycle 15, NEDC-31849P-1, Supplement 1, June 1992.D-3. General Electric Standard Application for Reactor Fuel, GESTAR II, NEDE-2401 1-P-A-15, September 2005; U. S. Supplement, NEDE-240 11-P-A-I5-US, September 2005.D-4. GE BWR Licensing Report: Average Power Range Monitor, Rod Block Monitor, and Technical Specification Improvement (ARTS) Program for Monticello Nuclear Generating Plant, NEDC-30492-P, April 1984.Page 68 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision I Appendix E SLO Pump Seizure Operating Limit The cycle-independent OLMCPR calculated for a recirculation pump seizure event when operating in Single Loop Operation (SLO) is 1.58 (Reference E-I). When adjusted for the off-rated power/flow conditions for SLO, (i.e., cycle specific Kp), this limit corresponds to a rated power/flow OLMCPR of 1.45 for Cycle 25, assuming a SLO SLMCPR _< 1.12. To ensure that the Reference analysis is still bounding for the current operating cycle, Monticello C25, the rated OLMCPR must be set > 1.45 when operating in SLO.References E-1. GEl4 Fuel Design, Cycle Independent Transient Analyses for Monticello Generating Plant, GE-NE-0000-0014-7048-OIP, Rev. 0, March 2003.Page 69 MONTICELLO 0000-0083-9607-SRLR Reload 24 Revision 1 Appendix F Mislocated Fuel Loading Error The Monticello Nuclear Generating Plant Cycle 25 Mislocated Fuel Loading Error analysis was evaluated. The event is non-limiting for fuel types through GE14 if the following condition is satisfied: OLMCPRpat,,,cycI e >_ 1.28 x (SLMCPR,,,,,,c /,1.07)This criterion has been demonstrated to be generically applicable to GEl4 reloads.The minimum OLMCPR calculated for Monticello Cycle 25 is 1.59 (shown in Section 11 for GE14 fuel from BOC25 to EOC25) while the plant/cycle specific SLMCPR is 1.10. Using 1.10 in the equation yields 1.32 on the right side.Using these values the above equation would yield 1.59 > 1.32.Therefore, the Mislocated Fuel Loading Error is non-limiting for Monticello Cycle 25.Page 70 MONTICELLO"P o -,' )A ", 0000-0083-9607-SRLR ID 1 Appendix G Thermal Mechanical Compliance A thermal-mechanical compliance check is performed for all analyzed transients to assure that the fuel will operate without violating the thermal-mechanical design limits. These limits are designed such that reactor operation within these limits provides assurance that the fuel will not exceed any thermal-mechanical design or licensing limits during all modes of operation. The fuel thermal-mechanical limits are met for Monticello Cycle 25.Page 71 MONTICELLO 0000-0083-9607-SRLR Reload 24 Revision 1 Appendix H Turbine Trip with Bypass and Degraded Scram The Turbine Trip with Bypass (TTWBP) event was analyzed with the postulated Option A degraded scram and an OLMCPR value was determined. No Option B analysis was performed for the TTWBP.The Option A calculated OLMCPR for the TTWBP is used for Option B and this value sets the OLMCPR limit for Option B because it is higher than the most limiting OLMCPR calculated for a pressurization event, the HPCIL8 transient. Therefore, if the cycle average scram time does not satisfy the criterion provided in Reference H-1 and Monticello Nuclear Generating Plant decides to interpolate between Option A and Option B scram times, this can be accomplished by using the procedure provided in Reference H-I with the following modification to Equation 4 of Reference H-I1: The modified equation to establish the new operating limit for pressurization events is given below: OLMCPRNow = MAX OLMCPROPo + T- TB AOLMCPR, OLMCPRrwBP TA -TB (4)where: rave and TB are defined in Equations I and 3 of Reference H-i, respectively; TA =the technical specification limit on core average scram time to the 20 percent insertion position OLMCPR Option B = the most limiting OLMCPR calculated for a pressurization event actually analyzed for Option B AOLMCPR= the difference between OLMCPR Option A and OLMCPR Option B For Monticello Cycle 25 the OLMCPRs for the HPCIL8 event are 1.70 for Option A and 1.53 for Option B. Therefore, the AOLMCPR for the HPCIL8 event is 0.17. The OLMCPR for the TTWBP event is 1.59.This approach is cycle independent with the TTWBP analyzed in this manner as long as the cycle specific OLMCPR Option B and AOLMCPR values are used in the calculation. References H-1. Monticello Option B Licensing Basis, LRC03.040, March 24, 2003 from L. R. Conner to Rick Rohrer.Page 72 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 Appendix I Monticello Non-Standard SRLR Items This appendix contains Monticello non-standard SRLR items that are being provided at the request of Xcel Energy.The following table summarizes the cycle rated power and flow MCPR values for .the events reported in this SRLR. If the event's Option A or Option B limit are merged together in a single column, then the event can not be interpolated based on scram times. For a description of how to implement Option B scram times see Appendix H.Cycle MCPR values Exposure range: BOC25 to EOC25 Option A Option B GEI4C GE14C FW Controller Failure 1.69 1.52 Load Reject w/o Bypass 1.63 1.46 Loss of Feedwater Heating 1.23 Fuel Loading Error (misoriented) 1.29 Fuel Loading Error (mislocated) Determined to be non-limiting SLO Pump Seizure 1.45 Turbine Trip with Bypass 1.59 Control Rod Withdrawal Error (RBM setpoint at 114%) 1.35 Load Rejection with Bypass'4 Determined to be non-limiting Turbine Trip w/o Bypass 1.66 1.49 Inadvertent HPCI /L8 Turb Trip 1.70 1.53 LOCA Analysis Limit MCPR 1.35 Stability OLMCPR Determined per Table 15-2 based on OPRM Amplitude Setpoint 14 This event corresponds to "Single Turbine Control Valve Slow Closure (GESIL 502)". Since Cycle 22 results for this event were far from limiting and no significant changes have occurred that would significantly increase this event's results for this cycle, this event was determined to be non-limiting. Page 73 MONTICELLO R XI 0000-0083-9607-SRLR Revision 1 Reload 24 Additional Section 9 Information For the inadvertent HPCI event, the level 8 trip was modeled as being lower than the OPL-3 setpoint ,value in order to trip the turbine on a level 8 signal during the event. This was done since confirmation could not be obtained that during this event that a level 8 trip would not occur.Additional Section 12 Information The Dome Pressure Safety Limit, provided via the OPL-3, of 1332.0 PSIG is satisfied. Additional Section 16 Information No single-loop operation multiplier on PLHGR is required.A study has been performed which shows bundle power of 3.9 MW is not exceeded at or below 20% power.Maximum Subcritical Banked Withdrawal Position (MSBWP)The Maximum Subcritical Banked Withdrawal Position analysis confirmed that the reference core loading pattern satisfied cold shutdown margin requirements including bank position 04.Page 74 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 Appendix J List of Acronyms Acronym Description ACPR Delta Critical Power Ratio Ak Delta k-effective 2RPT Two Recirculation Pump Trip ADS Automatic Depressurization System ADSOOS Automatic Depressurization System Out of Service AOO Anticipated Operational Occurrence APRM Average Power Range Monitor ARTS APRM, Rod Block and Technical Specification Improvement Program BOC Beginning of Cycle BSP Backup Stability Protection Btu British thermal unit BWROG Boiling Water Reactor Owners Group COLR Core Operating Limits Report CPR Critical Power Ratio DIVOM Delta CPR over Initial MCPR vs. Oscillation Magnitude DR Decay Ratio DS/RV Dual Mode Safety/Relief Valve ECCS Emergency Core Cooling System ELLLA Extended Load Line Limit Analysis EOC End of Cycle (including all planned cycle extensions) EOR End of Rated (All Rods Out 100%Power / 100%Flow / NFWT)EPU Extended Power Uprate ER Exclusion Region FFWTR Final Feedwater Temperature Reduction FMCPR Final MCPR FOM Figure of Merit FWCF Feedwater Controller Failure FWHOOS Feedwater Heaters Out of Service FWTR Feedwater Temperature Reduction GDC General Design Criterion GESTAR General Electric Standard Application for Reactor Fuel GETAB General Electric Thermal Analysis Basis GSF Generic Shape Function HAL Haling Burn HBB Hard Bottom Bum HBOM Hot Bundle Oscillation Magnitude HCOM Hot Channel Oscillation Magnitude HFCL High Flow Control Line HPCI High Pressure Coolant Injection ICA Interim Corrective Action Page 75 MONTICELLO Reload 24 0000-0083-9607-SRLR Revision 1 Acronym Description ICF Increased Core Flow IMCPR Initial MCPR IVM Initial Validation Matrix Kf Off-rated flow dependent OLMCPR multiplier Kp Off-rated power dependent OLMCPR multiplier L8 Turbine Trip on high water level (Level 8)LCF Low Core Flow LHGR Linear Heat Generation Rate LHGRFACf Off-rated flow dependent LHGR multiplier LHGRFACp Off-rated power dependent LHGR multiplier LOCA Loss of Coolant Accident LPRM Local Power Range Monitor LRHBP Load Rejectionwith Half Bypass LRNBP Load Rejection without Bypass LTR Licensing Topical Report MAPFACf Off-rated flow dependent MAPLHGR multiplier MAPFACp Off-rated power dependent MAPLHGR multiplier MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCPR Minimum Critical Power Ratio MCPRf Off-rated flow dependent OLMCPR MCPRp Off-rated power dependent OLMCPR MELLLA Maximum Extended Load Line Limit Analysis MELLLA+ MELLLA Plus MOC Middle of Cycle MRB Maximal Region Boundaries MSIV Main Steam Isolation Valve MSIVOOS Main Steam Isolation Valve Out of Service MSR Moisture Separator Reheater MSROOS Moisture Separator Reheater Out of Service MTU Metric Ton Uranium MWd Megawatt day MWd/ST Megawatt days per Standard Ton MWd/MT Megawatt days per Metric Ton MWt Megawatt Thermal N/A Not Applicable NBP No Bypass NCL Natural Circulation Line NFWT Normal Feedwater Temperature NOM Nominal Burn NTR Normal Trip Reference OLMCPR Operating Limit MCPR OOS Out of Service OPRM Oscillation Power Range Monitor Pbypass Reactor power level below which the TSV position and the TCV fast closure scrams are bypassed Page 76 MONTICELLO RPInnd 94.0000-0083-9607-SRLR Revision 1 Reload 24 Acronym Description Pdome Peak Dome Pressure PsI Peak Steam Line Pressure Pv Peak Vessel Pressure PCT Peak Clad Temperature PHE Peak Hot Excess PLHGR Peak Linear Heat Generation Rate PLU Power Load Unbalance PLUOOS Power Load Unbalance Out of Service PRFDS Pressure Regulator Failure Downscale PROOS Pressure Regulator Out of Service Q/A Heat Flux RBM Rod Block Monitor RC Reference Cycle RCF Rated Core Flow RFWT Reduced Feedwater Temperature RPS Reactor Protection System RPT Recirculation Pump Trip RPTOOS Recirculation Pump Trip Out of Service RV Relief Valve RVM Reload Validation Matrix RWE Rod Withdrawal Error SC Standard Cycle SL Safety Limit SLMCPR Safety Limit Minimum Critical Power Ratio SLO Single Loop Operation SRLR Supplemental Reload Licensing Report S/RV Safety/Relief Valve SRVOOS Safety/Relief Valve(s) Out of Service SS Steady State SSV Spring Safety Valve STU Short Tons (or Standard Tons) of Uranium TBV Turbine Bypass Valve TBVOOS Turbine Bypass Valves Out of Service TCV Turbine Control Valve TCVOOS Turbine Control Valve Out of Service TCVSC Turbine Control Valve Slow Closure TLO Two Loop Operation TRF Trip Reference Function TSIP Technical Specifications Improvement Program TSV Turbine Stop Valve.TSVOOS Turbine Stop Valve Out of Service TT Turbine Trip TTHBP Turbine Trip with Half Bypass TTNBP Turbine Trip without Bypass UB Under Burn Page 77 1IED01-23842 CLASS I CONTINUOUS CONTROL ROD WITHDRAWAL TRANSIENT IN THE STARTUP RANGE April 18, 1978 BY: R. C. Stirn J. F. Klapproth Approved By: E. P. Stroupe, Manager Systems Design and Specifications Nuclear Engineering /15,6 -C.I NED*.2 3342 CLASS I TABLE OF CONTENTS DISCLAIMER OF RESPONSIBILITY ...... ii ABSTRACT ....... .......... .1+/-INTRODUCTION ......... .................... 1 SEQUENCE OF EVENTS ............ ............. 2 METHOD OF ANALYSIS ........... ............... 3 SCRAM TRIP SETTING ............. ............. 4 MODIFICATION OF ROD SHAPE CURVE ...... ........ 9 RESULTS .................... ................. 10 CONCLUSIONS ............ .................. .. 12 REFERENCES ........... ................... ...15-i-NEDl-.2384 2 CLASS I DISCLAIMER OF RESPONSIBILrTY This document was prepared by or for the General Electric Company. Neither the General Electric Company nor any of the contributors to this document: A. Makes any warranty or representation, express or Implied, with respect to the accuracy. completeness, or usefulness of the Informaton contained in this docu-ment, or that the use of any InformatIon disclosed in this document may not Inringe privately owned rights; or B. Assumes any responsibility for liability or damage of any kind which may result from the use of any information disclosed in this document.Lt2 NEtO-23842 CLASS I ABSTRACT This Report documents the Method of Analysis used in performing the Continuoue Control Rod Withdrawal Transient in the Startup Range. The analyses are performed on a generic basis for a typical BWR and a detailed description of the calculational procedures is provided. Also included are the results of these analyses.-iii-NEIM-23842 CLASS I INTRODUCTION The continuous control rod withdrawal transient analysis in the startup range was performed to demonstrate that the licensing basis criteria for fuel failure will not be exceeded when an out-of-sequence control rod is withdrawn at the maximum allowable normal drive speed.The rod sequence control system (RSCS) and the rod worth minimizer (RWM)constraints on roe sequences will prevent the continuous withdrawal of an out-of-sequence rod for BWR/4-5 reactors. In the case of BWR/6, the rod pattern control function of the rod control and information system (RC&IS)performs this function. However, these analyses are performed for BWR/4-5 reactors to demonstrate that, even for the unlikely event where the RWM and RSCS fail to block the continuous withdrawal of an out-of-seauence rod, the licensing basis criteria for fuel failure is still satisfied. For BWR/6 reactors the RC&IS is a dual channel system; therefore, the continuous withdrawal transient in the startup range is not evaluated. The methods and design basis used for performing the detailed analysis for this event, are similar to those previously approved for the control rod drop accfldent (CRDA). (1), (2), (3) Additional simplified point model kinetics calculations were performed to evaluate the dependence of peak fuel enthalpy on the control blade worth. For the detailed calculation, the 50% control rod denuity pattern was selected as the initial starting condition which is consistent with the approved design basis for the CRDA. (1), (2), (3)-I-NEDS-23842 CLASS I The licensing basis criteria for fuel failure is: The contained energy of a fuel pellet located in the peak power region of the core shall not exceed 170 cal/gm-UO 2.SEQUENCE OF EVENTS Time (sec) 3WR12-5 0 1. The reactor is critical and operating in the startup range.>0 2. The operator selects and withdraws an out-of-sequence control rod at the maximum normal drive speed of 3.6 ips.-4 sec 3. (&)Botb the RWM and the RSCS fail to block the selection (selection error) and continuous withdrawal (withdraw error)of the out-of-sequence rod.4-8 tee 4. The reactor scram is initiated by the intermediate range monitor (IRM) system or the average power range monitor system (APRM).5-9 sec 5. The prompt power burst is terminated by a combination of Doppler and/or scram feedback.10 sec 6. The transient is finally terminated by the scram of all rods, including the control rod being withdrawn. (Technical Speci-fication scram insertion times are assumed, 5 sec. to 90%insertion).(a) For BWR/2-3,only the RWM is availabli-for r-" block on self.tion and withdraw errors. NEDO-23542 CLASS I Time V. (sac) :WR/6 0 1. The reactor is critical and operating in the startup range.>0 2. The operator selects an out-of-sequence rod.>O 3. The RC&IS blocks any movement of an out-of-sequence rod.0-4 sec or 4. The operator trys to withdraw an in sequence control rod continuously. 4 sec 5. The RC&IS blocks the continuous withdrawal of an in sequence rod (withdraw error).>1C min. 6, Operator reinserts rod drive to correct position.METHODS OF ANALYSIS Since the rod worth calculations using the approved design basis methods (1),(2),(3) use three dimensional geometry, it is not practical to do a detailed analysis of this event parameterizing control rod worths.Therefore, the methods of analysis employed were to perform a detailed evaluation of this event for a typical BWR and control rod worth (1.6% Ak)and to use a point model calculation to evaluate the results over the expected ranges of out-of-sequence control rod worths. The detailed calculations are performed to (1) demonstrate the consequences of this event over the expected power operating range and (2) demonstrate the validity of the approximate point model calculation. The point model calculation will demonstrate that the licensing criteria for fuel failure is easily satisfied over the range of expected out-of-sequence control rod worths. These methods are described in more detail below.The method used to perform the detailed calculation are identical to those used to perform the design basis control rod drop accident with the following exceptions: NEMD-23842 CLASS I 1. The rod withdrawal rate is 3.6 ips rather than the blade drop velocity of 3.11 fps.2. Scram is initiated either by the IPJi or 15v APR4 scram in the startup range. The IRM system is assumed to be in the worst bypass condition allowed by technical specifications.

3. The blade being withdrawn inserts along with remaining drives at technical specification insertion rates upon initiation of scram signal.SCRAM TRIP SETTING (IRM)The Intermediate Range Honitor (IRM) system consists of two groupings (or channels) of detectors (A and B) as shown in Figure 1 for a typical BWR.The detector closest to the control rod being withdrawn is bypassed on both channel A and B. The power level registered by the second closest detector of both channels must reach the scram trip level before the scram is initiated.

Of the four closest detectors (2 on channel A and 2 on channel B) to the control rod being withdrawn, the scram initiating detector is the furthest away from the control rod.The further away the scram initiating detector is from the rod being withdrawn, the longer the duration of the transient before scram initiation. To be conservative in the estimate of the time to scram, the furthest possible distance between a control rod and a scram initiating IRM detector is used in the analysis. NEDO-23842 CLASS I For the arrangement shown in Figure 1, the furthest distance possible between a withdrawn rod and its scram initiating detector is shown. Rod (34,11)is assumed to be withdrawn, then the scram initiating detector is located on channel B, 156.9 cm from the pulled rod. This is the furthest distance between a withdrawn control rod and its scram initiating detectcr for this example.Below is a table of values for the initial core power level and the corresponding IRM trip set point.Initial Core Power Level (PZERO) IRM Scram Set Point (Fraction of Rated) (Fraction of Rated)10.8 2.4 x 10-5 4 x 10-6 7.6 x 10-5 1.26 x 10-5 2.4 x 10-4 4 x 10-5 7.6 x 10-4 1.26 x 107 4 2.4 x lO-3 4 x 10 4 7.6 x 10 1.26 x 10-3 2.4 x 10-2 4 x 10-3 7.6 x 10-2 If the initial core power level is 10-8 of rated, then an LRM detector is tripped only after the power near the monitor reaches a level of 2.4 x 105 of rated. Thus, the second closest IM detector on strings A and B must register a local power of 2.4 x 10-5 of rated before the monitor trips and initiates the scram. FI GA 1 NEfl3-23842 CLASS I TYPICAL B14R IWM ARREAflDIENT 39 -___37~33 29-27 25 23 mEDA+/- 1AG 21OSoato iW 19 ~ YOS RPE RRSRI CD " -crýRr Tr -cr NE0B-23842 CALSS I PSCRAM is the average core power when the ItR detectors on both strings have been tripped. In the case being evaluated, with PZERO , PSCRAM is the average core power when the local power reaches 2.4 x 10-5 of rated at the location of the second cloest IRM detector on string B.To determine PSCRAM, a preliminary transient calculation (1), (2), (3)must be run with a PSCRAM value around two orders of magnitude larger than the IRM scram set point. Thc kinetics edit of the transient calculation output gives the local power in the core as a function of the coordinates R, Z and time. To find the proper entry, set: R -distance from rod pulled to the last monitor tripped before scram sets in.Z -distance from the top of the core to where the IRM detectors are located.For the case being investigated, the scram trip monitor is 156.9 cm from the rod pulled, and the monitor is b5.73 cm above the center line of the core.Once the proper entry is located, a search through the successive time steps determines when the local power reaches the IRM scram set point.The core average power fraction at this time is the PSCRA?1 value entered into the final transient calculation. NEr§-23842 CLASS I For the case analyzed, with PZERO 8, the following table is constructed to facilitate the determination of PSCRAM.Core Average Power (P/A) for (Fraction of Rated) (R -156.9, Z (Normalized 10-8 .449 3.008E-7 .463 9,185E-6 .468 1.355E-5 .468 l.a64E-5 .46&4.97C'E-05 .469 7.795E-05 .472 1.500E-4 .473 No&. Power in Node-45.73) (R -156.9, Z -45.73)to 1.0) (Fraction of Rated).4496E-8 1.3939E-7 4.2986E-6 6.347E -6 8.735E -6 2.333E -5 3.679E -5 7.094E -5 Time (Seconds)0 6.99 7.82 7.87 7.9]8.03 8.07 8.12 8.25 8.67 6. 747E-4 7. 029E-01.474.475 3.198E 3.339E-4-1 The first, second and fourth columns of the directly from the kinetics output.above table ar 0 obtained The scram trips when the scram initiating IRM reads 2.4 x 10-5 of rated power. The power near the scram initiating monitor reaches this value at time t -8.03 seconds. When t -8.U3 seconds, the core average power 5.11E-5, thus PSCRAM -5.1lE-5.For this analysis, an instrument delay time for the scram system of 90 msec.is used. Thus, the scram bank of rods begins to move .09 seconds after the scram initiating detector registers a power of 2.4 x 10-5 of rated.- NEDb.23842 CLASS I In 8.12 seconds, the rod being withdrawn .s. moved: (8.12 sec. x .3 ft/sec) -2.44 ft Thus, the control rod '3 :ithdrawn only 2.44 feet before the scram starts Inserting all the control rods, Including the control rod being withdrawn. MODIFICATION OF ROD SHAPE CURVE In the preliminary transient calculation used to determine PSCRAIM, the rod shape function represented by the solid line on Figure 7 was used.This curve depicts the control rod being withdrawn at .3 ft/sec until the entire rod is withdrawn. In reality, the rod is withdrawn only 2.44 feet before the scram starts to reinsert the rod. Figure 7 shows the actual rod shape funciton after scram initiation at time TS in the fona of a dashed line.With the new rod shape curve and PSCRAM value, the final transient calculation is run.The point model kinetics calculations use the same equations employed in the Adiabatic Approximation described on page 4-1 of reference

1. The rod reactivity characteristics and scram reactivity functions are input identical to the adiabatic calculations, and the Doppler reactivity is input as a function of core average fuel enthalpy.

The Doppler reactivity feedback function input to the point model calculations was derived from the detailed analysis of the 1.6t rod worth case described above. This is a conservative ( ussumption for higher rod worths since the power peaking and hence spatial Doppler feedback will be larger for higher rod worths. As will be seen in--9-- NEDO-23842 CLASS I the results section, maximum enthalpies resulted from cases initiated at (. 1% of rated power. In this power range the APRM will initiate scram at 15Z of power; hence, the APRM! 15t power scram was used for these calculations thereby eliminating the need to perform the spatial analysis required for the IRM scram. All other inputs are consistent with the detailed transient calculation. The point model kinetics calculations results in core average enthalpies. The peak enthalpias were calculated uaing the following equationi ho + (P/A)T (f -ho)where: S Final peak fuel enthalpy ho R Initial fuel enthalpy IRfi Final core average fuel enthalpy (P/A) Total peaking factor (radial peaking) * (axial peaking) .(local fuel pin peaking)For these calculations the (radial x axial) peaking as a function of rod worth was obtained from the calculations performed in section 3.6 of reference (2). It was conservatively assumed that no power flattening due to Doppler feedback occurred during the course of the transient. RESULTS Figures 2 thru 5 present the results of the continuous control rod with-drawal accident in the startup range for the detailed analyses. Figures 7 and 6 givr the control rod and scram reactivity inputs, respectively. NE.1S-23842 CALSS I Fbr an initial core power of 10-8 of rated, the peak enthalpy generated by the transient is:= Pi(h -ho) + ho-1.184 (16.507 -16.45) + 16.45 -16.52 cal/gm where: Pp E Local fuel pin peaking factor h Peak enthalpy from R-Z adiabatic calculation ho S Initial fuel enthalpy For an initial core power of 1S-2 of rated, the peak enthalpy generated by the transient is:-1.184 (49.98 -16.45) + 16.45 -56.2 cal/gm In neither case did the fuel even remotely approach the 170 cal/gm needed to exceed the dnsign criteria for fuel failure. It is also noted from the above results that the maximum peak enthaply results for the case that was initiated from 1% of rated power; hence, the point model calculations performed to assess impact of control rod worth on peak fuel enthalpy will be initiated from 1% of power.The reactivity insertion resulting from moving the control rod is shown in Figure 8 for the point kinetics calculations. The core average power versus time and the global peaking factors from reference (2) are shown in Figures,9 and 10, respectively. The results of the point kinetics calculation are summarized in Table 1 along with the results of the detailed analysis.From Figure 9 and Table 1, it is shown that the core average energy deposition is insansitive to control rod worth; therefore, the only change in peak enthalpy as a function of rod worth will result from NEDO-23842 CLASS I differences in the global peaking which increases with rod worth. The global peaking factors shown in Figure 10 were obtained from reference (2).Comparing these values to the more exact R-Z value used in the detailed calculations demonstrates that the reference (2) values are reasonable for their application in this study. For all cases the peak fuel enthalpy is well below the licensing design criteria of 170 cal/gm.Cases 4 and 5 of Table I show that the point kinetics calculations give conservative results relative to the detailed evaluations. The primary difference is that the global peaking will flatten during the transient due to Doppler feedback. This is accounted for in the detailed calculation but the point kinetics caluclations conservatively assumed that the peaking remains constant at its initial value.The differences in core average and peak enthalpy between cases 1 and 5 are due to the fact that for case 1 the scram was initiated by the 15%APRM scram set point, whereas, in case 5 the scram was initiated by the IRII's. As seen by Figure 2, this occurred at a core average power of 21%.Since the APR1 trip point will be reached first, it is reasonable to take credit for the APR4 scram.CONCLUSIONS From this study the following conclusions can be stated: 1) The resultant peak fuel enthalpies due to the continuous withdrawal of an out-of-sequence rod in the startup range results in peak fuel enthalpies which are significantly less than the licensing basis criteria of 170 cal/gm. NEDD-23842 CLASS I 2) The point model calculations used to assess the sensitivity of peak enthalpy as a function of control rod worth are in good agreement with, and slightly conservative relative to the more detailed design basis model which is smployed to evaluate the continuous rod with-drawal transient in the startup range. NEDG-23842 CLASS I TABLE 1

SUMMARY

OF RESULTS FOR DETAILED POINT KINETICS EVALUATIONS OF CONTINUOUS IN THE STARTUP RANGE AND ROD WITHDRAWAL CASE 1 2 3 4 5 CONTROL ROD WORTH (Z-6k)1.6 2.0 2.5 1.6 (a)1.6 Md)hf (cal/gm)17.3 17.3 17.2 18.3 18.3 P/A(c)24.2 30.9 46.0 19.7(b)19.7 h (cal/gd)&2.7 50.0 58.5 56.2 59.6 (a) Detailed transient calculation. All other data reported are for point kinetics calculations.(b) The P/A -19.7 is the initial value. For the detailed analysis this value will decrease during the course of the transient since the power shape will flatten due to Doppler feedback.(c) P/A i global peaking factor (Radial x Axial).(d) Point kinetics calculation with IRM initiated scram and 3-D simulator global peaking. NEDI-23842 CLASS I REFERENCES (1) Paone, C. 3., et.al., "Rod Drop Accident Analysis for Large Boiling Water Reactors", NEDO-10527, March 1972.(2) Stirn, R. C. , et. al., "Rod Drop Accident Analysis for Large BWRs", NEDO-10527, Supplement 1, July 1972.(3) Stirn, R. C. , "Rod Drop Accident Analysis for Large Boiling Water Reactors Addendum No. 2 Exposed Cores", NEDO-10527, Supplement 2, January 1973. '3 FIGURE 2 1-6--_.4-4 I rz CONTINUOWS CONTROL ROD HOT STARTUP: (1) 1.58% ak Rod (2) 0.3 fps Withdra (3) IRM Scram for W Condition (4) P 0 i 10-2 of Ra (5) 1967 Product Li CLASS I 4-F W17THRAL FROM wal Velocity orst Bypass ted tne Tech. Spec. ..-GWD/T 4t +/-T t-pr. 1-t-t. A:__*Scram Rate (6 Exposure 0.0-'-'---4-4-4-4 A -.---.--6.-d.. 4 ......J.-L. I I-----Wi--2-T.U9~4Tr~Tz2 I-~tmtW+/- H-H-4-I I'-1-F lii 1-! 11 S _______________ _______________

• .-J -. -4 t C U, rti r4 1 I 7'3-j .=S~I+/-fl44-~" -~.r~. --*t4-C-z.9.--j 73 7~f TTT t~ flAr~r5t .r+/-t..-W I~ 1-i 'z~At~-~ r4r4-rs U -rrrnr -~ ~r __ -rq ~~r-q ~ I ttrt rrr7-~J -r E ý mz _=ffu 7m: a ýT --i_b V~t rli-I--i-N: 2-I I¶0~I-Ii}t, U 7-7-T-7 t4z1t. 2'

-'-1 -'4T'3 4 S 6 7 NEro-23S42 CLASS I-17-Neo*L F16urg.- -3 I s1Sj~SJ *~ fli :5~:~7-~; -we-HS. jqr7 ?u~RO~S ~J3'~aArt~.o;~7i~3Ow~A.-rep Li hi r-t , , i-3 10 i..e 4.4.0.1~.C..t.~; : '3.9-- ý-E--~~~ ~ ---- ----------* ~ ~ ~ ~ 7 .7 ',.,*__.'ILLT-1, 16'44 ai1 10 ..io~ a I IHIi !!lii! ; I i i I-* a --a .-.-I --.-I -"L-5 "1.0 Mms (-secAlýc ) .-LS-NEDO-28342 48- N A~ 28342.2o 18.3S 18.25= 18.15= 18.05S= 17.95 17.85 17. 75 17.6S I 17.SS_ 17.45 17.35-'.17.25 17.15 17.0S 16.95 U 16.85 16.75 16.65 i6.55 3&.45 FIGURE 4 CONTINUOUJS CMTRIL RLU WITHDRAWAL I-.N::1j FROM HOt STARTUP (1) 1.58% Ak Rod (2) 0.3 fps Withdraal Velocity (3) IRM Scram for Worst Bypass.Condition (4) Po ]0 2 of Rated (5) 1967 Product Line Tech.Spec. Scram Rate (6) ] hcposure -0.0 GWD/T III! ¶*-IT.1 i-i--~1~: ' : ' : : : -: ......'z ....' ..: ..: : i " -' ; ....." : ....." ....: .....: ' .." ...., L.......~ ~ ~ .-t .......... ........-.................., ....: ..., : .., : ...~~~~ .ii i .i : .. .... .... .... ... -7: : -.L.:.7 7"; 7:" ... ... ....o,.. ...... ..... .._ -l.....--........................4 ....-. ... ...s.'. .L:, = : 4 -: ::: : : : : : : : : -.. .. ... ...... ...M..--=-=--=);- "- : t: 'l-":--. -:;j:_- : " F:t.: :-t< +!i: ;(: < :.::" L=: ---= = ---- " = -' %- , ...:..: -..... ........ .,.- ..... ....... .........................., ! i i i , i i : -F , l : '.... ..~~ ~ ~~~~~~ ~~ ~~~.... .. ..-........ t ....., .... !, .k ...l .l ...... .. ..... ...:- ........ ... ...... ............. ................... ..... .... ......... .. ..... ..... ... i...... ......I .............l ....L .I .......... ..........! : ' ..: ., ...... ......:. .: ' .....: : .' .=: ; ..: "* " :. --,.. ...., ... .: " ; ... ..... .... .... .. ..... .. ... .....0 4 ..I I' .... '°'.-I"i'.:I-*17I 1 Z 3 4 5 6 7 8 Time, (see) CLASS I-19-FIGURE .5.CONTINUOUS CONROL ROD WIVIH)RAL FROMI.. HOT STARTUP: (1) 1.58% Ak Rod (2) 0.3 fps Withdrawal Velocity (3) IRM Scram for Worst Bypass 2} Condition (4) Po -In-$ of Rated (5) 1967 Product Line Tech. Spec.Scram Rbate:::::::"" -"(6) Exposure 0.7 T ""': " 16.453~hhhk t ~t 7i:1 7 71'1 T L-A-7!I Ni LI--:rr: 4-- 4 --I-4.L I I I-L~~~~ t7=mriprri-I---I A a E L'... ... ......-. ..... .... ... ... ..I I I*..7 .... ... .. ...... ........a ;. ...-..' 7-I --16.452 -.---..... ........ .I: : _1TIME, (sec) !. :i -ea.. ...-......,.-.. ... .. ... ... .. .... ....... ......._ Lim TT-7 1:: :::1 1. 77. .... ' 7-I.r I". NE[1-23842 CLýASS I-a-; FIGURE 6.:" ROD GROUPS 1 -4 EXCLUDING?::! ~CENTER ROD* BEING SCRAMMl~ED ,*-r-SCRAM OF CENTER ROD IS INCLUDED ..IN ROD S.APE FUNCTION FIGURE 8 09.. ... ..~ ~ ~ ~ ~ ~~ ~. .........NE O* EN SRME.... ~~~~~~ ~ ~ ... ......SRMO1ENE O SICUE*~~1 :m ;..r=-_ i': t"- :77 07.06--.... .......04 02 Z01 g0-i........I.. ... 1 .. 7- m v t TI Z:: 7' 71f-1 -77 :.%4~1~ 1-7 i.i ".I : -L:LL. -.1 0 .1 .2 .3 .4 .5 .6 .7 .8 1.0:: 4+o.... ... ..(RA Cr ORE FULLY fl1ROLLED.. ............- 1~:i ft i i i I tT i i 117T-.71711,7. 1;: ;1;;, T I......... .... jý;J: it J CLASS .1 FIGURE 7 i REACTIVIIY INSERTION WIThi TIM~ CAUSED BY{7 THE t4Wv1Eir OF THE CENTER C0OMROL RO)D BOL, Tf -=T = 286*C 4-0=i8 i=~1~~~7V.t-I-"4 m 2 Z.4.l 71.F: 7 9 016T......'.. ....:....-' ..... .. ..01~:!~ ....... ._._" -,, 01 .... "....7 : --: -013~,. ..I....Control Rod Being ed -.....L::: 7.7-0161 1ý .. ........cram Inserts Control R TT~t:L.0 4 8 12 16 20 24 28 32 36- 40: I...................~. .-I I -. I*- -----t --t 1.~iiI~i~.. ..... .... ....-4-." : : .TIME, (sec)I p 0 I 1 I ; I : I I..............WI". M-f 4: .... .... ........ ... .... ... ..IL I :1:: .... .... .I ... ...... NEDO-23842 CLASS I FIGURE 8-Qt...n tok's#r vtc5 NP~z Rasroi> ~erv,'4 I i I I C Jla.&-7II74~--------..... --77 -7* -i--.----I- -____~1~ -j -4 A tj l Jr f I-- I.....- -_ .... .... ... 7 ......_ __ __ __ __+J C 0 alo 0+4=777 =. ....LI .;.:.:.' ..: ..;.:.. ..::. .../7:7.---------------------. .. ..- ..-- --.. ... ... .A.....7-- ---1 T------7 7 7-- 7:: 7 --- --0.01 occ om am 0 1 7---I--[kE~~n-. 100~ -I I I I I o 4.o leo I.o ,,0 Zo.o Z4.o 2so 32.0 A, o 4o-o-Time (sac esL,-%) 384 2 CLASS I FIGURE 9 K I ORO' D -poi AlT t41c61- KIAICT 4 2.45* -Rom) vJOR-).2? 0% RCVD WAORT K 1,9 R~ol>VCRf I ý-1/flI LI'///// /4-/f , /* //---I* ---.- -I.... I-2 40 4'al20 -24s-NEDO-23842 CLASS I FIGURE 10?/.--m ;.:)> kl'Do-osi .i-,,(2Z)(--V SI'[7........ ....* ~ ~ ~ -:- -- ------ ---. .: (I.~ -.443 S.4.it N 0-7 :"-7:r: .tl tn. Af.. .ZO IC.....: ...--...: : ........: ..............- :. ........ ..<- -: .... .... ..... .... ... _ -.. .. .:- :::: L -'-.:'"'i] ~i.................... .- A ---..4-4j, I p -: ("6i 1.0 CONTrZOL 'RJr'%r NUCLEAR ENERGY DIVISIONS a GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNIA 96125 GENERAL 0 ELECTRIC TECHNICAL INFORMATION TITLE PAGE EXCHANGE AUTHOR SUBJECT TIE NUMBER 3. V. Klapproth DATE 4/20/78 TITLE GE CLASS Continuous Control Rod Withdrawal Transient in the Startup Range GOVERNMENT CLASS COPY FILED AT TECHNICAL NUMBER OF PAGES SLJPORT SERVICES. R&UO. SAN JOSE, 28 CALIFORNIA 95121 (Mail Code 211)ISLUMMARY This report documents the Method of Analysis used in performing the Continuous Control Rod Withdrawal Transient in the Startup Range. The analyses are performed on a generic basis for a typical BWR and a detailed description of the calculational procedure is provided. Also included are the results of these analyses.By aitdng out this w-ctangle and folding in half. the above information can be fited into a ndard cad file.DOCUrMNT NU/MER NED&-23842 INFORMATION PREPARED FOR BWR Systems Engineering Department SCTION Nuclear Engineering SUILONG AND ROOM kUMBR Em 1224/1850 MAIL CODE 765 NE"414 ism7)}}