ML100640343

Response to Questions 28 and 30 of Request for Additional Information Regarding a Spent Fuel Pool Criticality License Amendment Request

ML100640343
Person / Time
Site:	Millstone
Issue date:	03/05/2010
From:	Hartz L Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
10-072A, FOIA/PA-2011-0115
Download: ML100640343 (28)

Text

{{#Wiki_filter:)"mlOioll Nuclear Connecticut, Inc. 00 lOIn inion Boulevard, Glen Allen, Virgini.i.' ij(,O \\, -h v.idress: WWVV.dOlll.COnl March 5, 2010 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Serial No.: NLOSIWDC Docket No.: License No.: 10-072A RO 50-423 NPF-49 DOMINION NUCLEAR CONNECTICUT, INC. MILLSTONE POWER STATION UNIT 3 RESPONSE TO QUESTIONS 28 AND 30 OF REQUEST FOR ADDITIONAL INFORMATION REGARDING A SPENT FUEL POOL CRITICALITY LICENSE AMENDMENT REQUEST Dominion Nuclear Connecticut, Inc. (DNC) submitted a stretch power uprate (SPU) license amendment request (LAR) for Millstone Power Station Unit 3 (MPS3) in letters dated July 13, 2007 (Serial Nos, 07-0450 and 07-0450A). The SPU license amendment request was supplemented in a letter dated December 13, 2007 (Serial No. 07-0450C). The SPU LAR included a revised spent fuel pool (SFP) criticality analysis with proposed changes in technical specification (TS) requirements. DNC separated the MPS3 SFP TS change request from the MPS3 SPU request via letter dated March 5, 2008 (Serial No. 07-04500). In a letter dated August 8, 2008, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) regarding the SFP TS. DNC responded to RAI questions 1 through 19 in a letter dated September 30, 2008 (Serial No. 08-0511A). In a letter dated February 2, 2009, the NRC requested additional information. DNC responded to RAI questions 20, 22, 23, and 25 in a letter dated March 5, 2009 (Serial No. 09-084) and to RAI questions 21 and 24 in a letter dated March 23, 2009 (Serial No. 09-084A). Subsequently, in a letter dated January 26, 2010, the NRC requested additional information. DNC responded to RAI questions 26, 27 and 29 in a letter dated March 1, 2010 (Serial NO.10-072). contains the responses to RAI questions 28 and 30. contains an updated markup of MPS3 TS pages affected by this SFP TS change request. These updated markups supersede the TS page markups submitted as part of the SPU license amendment request in the letters dated July 13, 2007 (Serial Nos. 07-0450 and 07-0450A). contains a markup of associated Bases changes for information only. Changes to TS Bases are controlled under the provisions of 10 CFR 50.59.

,2010. Serial No. 10-072A Docket No. 50-423 Response to RAI for the MPS3 Spent Fuel Pool Criticality LAR Page 2 of 3 The information provided by this letter does not affect the conclusions of the significant hazards consideration discussion in the December 13, 2007 DNC letter (Serial No. 07-0450C). Should you have any questions in regard to this submittal, please contact Wanda Craft at 804-273-4687. Sincerely, "\\ ) ).,.//'~4.y1r..' i /' /' C./ Leslie N. Hartz Vice President - Nuclear Support Services Commitments made in this letter:

1. None.

COMMONWEALTH OF VIRGINIA ) ) COUNTY OF HENRICO ) The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Leslie N. Hartz, who is Vice President - Nuclear Support Services of Dominion Nuclear Connecticut, Inc. She has affirmed before me that she is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of her knowledge and belief. 1:: 771 /II'!. Acknowledged before me this J::: day of \\UraA c& My Commission Expires: J1 aa 3/, cJ0 10 10/u' i j/l<.ft VICKI L. HULL Notary Public Commonwealth of Virginia ~ 140542 ~ My Commission Expires May 31. 2010 Notary Public

Serial No. 10-072A Docket No. 50-423 Response to RAI for the MPS3 Spent Fuel Pool Criticality LAR Page 3 of 3 Attachments:

1. Attachment 1: Response to Request for Additional Information (RAI)

Questions 28 and 30 Regarding the Spent Fuel Pool Criticality License Amendment Request

2. Attachment 2: Response to Request for Additional Information (RAI)

Questions 28 and 30 Regarding the Spent Fuel Pool Criticality License Amendment Request, Updated Markup of Technical Specifications Pages

3. Attachment 3: Response to Request for Additional Information (RAI)

Questions 28 and 30 Regarding the Spent Fuel Pool Criticality License Amendment Request, Markup of Technical Specifications Bases Pages For Information Only cc: U.S. Nuclear Regulatory Commission Region I Regional Administrator 475 Allendale Road King of Prussia, PA 19406-1415 Ms. C. J. Sanders Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 08-B3 Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No. 10-072A Docket No. 50-423 ATTACHMENT 1 RESPONSE TO RAI QUESTIONS 28 AND 30 REGARDING THE SPENT FUEL POOL CRITICALITY LICENSE AMENDMENT REQUEST DOMINION NUCLEAR CONNECTICUT, INC. MILLSTONE POWER STATION UNIT :3

Serial No. 10-072A Docket No. 50-423, Page 1 of 15 RESPONSE TO RAI QUESTIONS 28 AND 30 REGARDING THE SPENT FUEL POOL CRITICALITY LICENSE AMENDMENT REQUEST Question 28 In ONC's response to RAI #5, ONC claims that ignoring the lnteqrated Fuel Burnable Absorber (IFBA) is conservative. In that response ONC claims that it is conservative to ignore the presence of IFBA when performing the depletion portion of a spent nuclear fuel criticality analysis. ONC's submittal states they performed two sets of analyses; one in which all residuallFBA was artificially removed after the depletion, and one in which all residuallFBA was retained after the depletion. ONC's submittal indicates that when the residuallFBA is artificially removed the effect of neutron spectral hardening is shown, but when the residuallFBA is left in the fuel assembly, the residuallFBA overcomes the neutron spectral hardening with a conservative result. There is no indication in NUREG/CR-6760 that any residuallFBA was artificially removed in reaching its conclusions. The information presented in NUREG/CR-6760 indicates that any residual IFBA was left in the fuel assembly when determining the effect. ONC response is inconsistent with NUREG/CR-6760 and has not been accepted by NRC staff in the past. There is currently insufficient information in ONC's submittal to reconcile the different conclusions. Please provide a detailed comparison of ONC's analysis and the analysis performed in NUREG/CR-6760 and justify the differences.

Response

The analysis presented in response to request for additional information (RAJ) 5 was performed using site specific information for the fuel assembly design, core operating parameters, and Integral Fuel Boron Absorber (IFBA) designs used at Millstone Power Station Unit 3 (MPS3). The calculations reported in response to RAI 5 were performed with 5.0 wt% enriched fuel with IFBA modeled over the entire axial length of the fuel. The generic analysis of 17x17 fuel presented in NUREG/CR-67130, Reference 1, considered both full-length and part-length IFBA. The conclusion presented for full length IFBA in Reference 1 agrees with the conclusion presented in RAI 5 for full-length IFBA. The following discussion identifies the apparent differences between the MPS3's analysis and the analysis performed in NUREG/CR-6760. The results presented in response to RAI 5 and additional results below are applicable to the MPS3 SFP criticality safety analysis. The results presented here are not necessarily representative of other fuel lattices, assembly designs, IFBA designs or core operating parameters outside of MPS3.

Serial No. 10-072A Docket No. 50-423, Page 2 of 15 Tables 6 and 7 contain modeling information from Section 3.3 of Reference 1. A column is added to each table to show the values used in the MPS3 analysis. Table 6 - Summary of Parameters Used for the Depletion Calculations (Table 1 of Section 3.3 of NUREG/CR-6760 Parameter NUREG/CR-6760 Response to RAI 5 Moderator Temperature (OK) 600 585.65 Fuel Temperature (OK) 1000 1029.93 Fuel Density (a/cm3) 10.44 (U02) 10.69 (U02) Clad Temperature (OK) 600 614.93 Clad Density (g/cm 3) 5.78 (Zr) 6.58 (Zircaloy-4) Power Density (MW/MTU) 60 41.5 Moderator Boron Concentration (ppm) 650 1000 Most of the calculations in Reference 1 were performed assuming a uniform burnup profile. The MPS3 analysis performed in response to RAI 5 used the axial power distribution shown in Figure 3-5 ofWCAP-16721-P; the fuel, moderator, and cladding temperatures varied as a function of power. The values shown in Table 6 correspond to the values at a relative power of 1.0. The footnote to Table 1 of Reference 1 states that cases were aliso calculated using a power density of 30 MW/MTU and that the change in effective multiplication factor (~k) results were not sensitive to variations in power density. Table 7 - Fuel Assembly Specifications (Table 2 of Section 3.3 of NUREG/CR-6760) Parameter NUREG/CR-6760 MPS3 Analysis Rod Pitch (ern) 1.260 1.260 Assembly Pitch (ern) 21.5 21.5 Claddina Outside Diameter (ern) 0.8898 0.950 Cladding Inside Diameter (ern) 0.8001 0.836 Pellet Outside Diameter (ern) 0.7840 0.819 Guide/Instrument Tube Outside Diameter (ern) 1.204 1.224 Guide/Instrument Tube Inside Diameter (cm) 1.124 1.143 Array Size 17x17 17x17 Number of Fuel Rods 264 264 Number of Guide/Instrument Tubes 25 25 Reference 1 examines 17x17 assemblies containing 80, 104, and 156 IFBA rods with boron loadings of 1.57 and 2.355 mg10B/inch. The MPS3 analysis uses 17x17 assemblies containing 156 IFBA rods with an IFBA loading similar to the maximum Reference 1 loading. As mentioned above, the calculations reported in response to HAl 5 were performed with 5.0 wt% enriched fuel with IFBA modeled over the entire axial length of the fuel.

Serial No. 10-072A Docket No. 50-423, Page 3 of 15 Section 3.3.5.5 of Reference 1 states that modeling a shorter IFBA stack can result in larger differences in calculated eigenvalues between cases depleted with and without IFBA present. Tables 10 and 11 of Reference 1 show the ~k effects when a non-uniform axial burnup profile is modeled with IFBA modeled over the entire axial length of the fuel and with IFBA modeled over 120 inches, centered axially in the fuel rod. The more realistic IFBA model, 120 inches centered axially, is more limiting. MPS3 uses 120 inch IFBA centered axially in the fuel rod. Therefore, the MPS3 specific study was re-performed modeling a 120 inch IFBA region centered axially in the fuel assembly, which is representative of the actual IFBA length used in the core. Results of this new study are presented in Tables 8 and 9. Tables 8 and 9 show the difference between a model accounting for IFBA during the depletion subtracted from a model that did not account for IFBA during the depletion. Both models use the RAI 21 axial burnup profiles, isotopic number densities from PARAGON depletions, and 3D SCALE Version 4.4 SFP models to determine SFP Keff. A positive value indicates that the residuallFBA is more important than the spectral hardening effect. A negative value indicates that the spectral hardening caused by the IFBA has a higher reactivity worth than the residual l'B. ResiduallFBA is explicitly modeled in the IFBA model. The depletion parameters used in this study are discussed in response to RAI 30 (Table 11). bl 8 k i R 2 f 3 0 d 5 0 0' F Ta e -~ In emon or. an wt~o ue I BU Decay ~k 3.0 wt% BU Decay ~k 5.0 wt% (MWd/MTU) (years) [krno IFBA\\-kIlFBA\\l (MWd/MTU) (years) [krno IFBA\\-kIlFBA\\l 15,000 0 +0.00720 40,000 0 -0.00160 15,000 10 +0.00838 40,000 10 -0.00112 20,000 0 -0.00033 45,000 0 -0.00173 20,000 10 +0.00018 45,000 10 -0.00175 f 30 d 5 0 to' F Table 9 - ~k in Region 3 or. an . w '10 ue I BU Decay ~k 3.0 wt% BU Decay ~k 5.0 wt% (MWd/MTU) (years) [krno IFBA\\-kIlFBA\\l (MWd/MTU) (years) [krno IFBA\\-kIlFBA\\l 25,000 0 -0.00298 55,000 I) -0.00194 25,000 25 -0.00192 55,000 2:5 -0.00127 30,000 0 -0.00343 60,000 I) -0.00221 30,000 25 -0.00235 60,000 ~:5 -0.00172 At low burnups the residuallFBA is enough to overcome the increase in reactivity due to spectral hardening. However, once the IFBA has burned out, the WCAP-16721-P analysis is potentially non-conservative. Rather than attempt to identify off-setting conservatisms in the existing analysis, a burnup penalty is applied to the final burnup credit curves for Regions 2 and 3 as part of the response to RAI 30. The burnup limits in Region 1 are low enough that the residual IFBA overcomes the increase in reactivity due to spectral hardening. Decay times greater than zero were investigated to ensure that the effect did not increase with increasing decay times. The limiting reactivity difference occurs at zero decay time for all but one case (Region 2 at 45,000 MWd/MTU). In that case the difference is very small and statistically insignificant.

Serial No. 10-072A Docket No. 50-423, Page4 of 15 Question 30 Please verify the average core exit temperature for nominal reactor coolant system (RCS) flow. Please verify the maximum core exit temperature for the minimum TS allowed RCS flow. Please provide the maximum fuel assembly exit temperature for the minimum TS allowed RCS flow. Please provide the moderator temperature used for each case in the RAI #21 responses.

Response

As stated in response to RAI 4, the use of the word nominal in Table 2.8.3-1 of the stretch power uprate (SPU) license amendment request (LAR) refers only to the nominal value used in the analysis, and does not refer to the plant actual operating nominal temperature. The reported inlet temperature is a nominal value. Outlet temperature is maximized by assuming minimum RCS flow. ThE! average core exit temperature at uprated conditions with nominal reactor coolant system (RCS) flow is 620.4 of. The maximum core average exit temperature for the minimum technical specification (TS) allowed RCS flow is 628 of. Calculations supporting the response to RAI 21 used an exit temperature of 628 of for blanketed fuel. The depletion calculations for no blanket fuel depleted in pre-uprate cycles used an outlet temperature of 620.6 OF. This temperature results from considering the minimum RCS flow with the pre-uprate power level of 3411 MWt. Maximum fuel assembly exit temperature is dependent on assembly power and is higher than maximum core average temperature. Assembly power and exit temperature depend on fuel management strategy and vary with fuel burnup, burnable poison loading, and the number of fuel cycles in which the assembly is resident. For the calculation of isotopic number densities, it is not practical to simulate all possible power histories in detail. Rather, a conservative but constant value for assembly exit temperature is needed. At any point in the lifetime of a fuel assembly, a lifetime average exit temperature can be determined from the lifetime average assembly relative power. The lifetime average assembly relative power is simply the assembly burnup divided by the total cycle burnup for the cycles the assembly has been used in up to that point in the lifetime of the assembly. The examples below will illustrate this calculation: A) An assembly midway through its first cycle has a burnup of 12,000 MWd/MTU when the cycle burnup is 10,000 MWd/MTU. The lifetime average assembly relative power is 12,000/10,000 or 1.2. Using the core power (3650 MWt) x 1.2, the MPS3 core minimum flow, and the MPS3 nominal inlet temperature, the lifetime average assembly exit temperature can be calculated with a heat

Serial No. 10-072A Docket No. 50-423, Page 5 of 15 balance. For this example, the lifetime average assembly exit temperature at 12,000 MWd/MTU assembly burnup is 639.7 OF. B) An assembly at the end of its third cycle has a burnup of 60,000 MWd/MTU. The total cycle burnup of the three cycles the assembly was used in is 60,000 MWd/MTU. The lifetime average assembly relative power is 60,000 160,000 or 1.0. Using the core power (3650 MWt) x 1.0, the MPS3 core minimum flow, and the MPS3 nominal inlet temperature, the lifetime average assembly exit temperature is 628 OF. This assembly would very likely have experienced higher than lifetime average exit temperatures in the first two cycles (similar to example A) and lower than lifetime average in the final cycle. Using this method, a survey of the fuel assemblies in the four most recent MPS3 reload cores was performed to identify a bounding lifetime average exit temperature as a function of assembly burnup. Individual assembly values were calculated at mid and end of cycle for first and second burn fuel and at end of cycle for third burn fuel. A composite maximum power history was determined that bounds the fuel assemblies in the survey. The curve (linear between points) defined by these values does not represent the history of any individual assembly, but bounds the maximum lifetime average values of the assemblies surveyed at the assembly burnups. Table 10 shows the results of this survey: fA sembly Burnup f F t I mg empera ure as a unc Ion 0 Si Assembly Burnup Maximum Exit (MWd/MTU) Temperature (OF) 0 648.7 28,000 648.7 50,000 643.6 60,000 635.8 Table 10 - Limiti T It is not credible to assume lifetime average assembly exit temperatures significantly higher than those shown in Table 10. WCAP-16721-P depletion calculations were performed using the maximum core average exit temperature (628 OF). In order to quantify the effect of assembly exit temperatures above core average, Table 10 bounding temperatures were combined with the limiting burnup profiles and axial nodalization identified in response to RAI 21, and the parameters in Table 11. Depletion calculations were performed using PARAGON, and discharged assembly reactivities were determined using SCALE Version 4.4 SFP models for all cases.

Serial No. 10-072A Docket No. 50-423, Page 6 of 15 d F IA T bl 11 D I a e ep etlon an ue ssembly Parameters RJ\\ls 28 & 30, Parameter WCAP-16721-P mid-enriched blankets Profile 5 from DOE Limiting profiles Axial Burnup Distribution Topical Report identified in RAI 21 Number of axial zones modeled 4 24 Tin 556.4 OF 556.4 OF Tout 628 OF SI~e Table 10 Soluble Boron present during depletion constant 1000 ppm constant 1000 ppm Power 3650 MWt

3650 MWt Theoretical Density of fuel 97.5%

97.5% solid, right cylinder solid, right cylinder (i.e., no dishing or (i.e... no dishing or Fuel pellet shape chamfering) chamfering) Design Basis Fuel Westinghouse 17x17 Westinghouse 17x17 Assembly STD STD Fuel initial enrichments 3.0, 4.0, and 5.0 wt% 3.0, 4.0, and 5.0 wt% Blankets modeled? No Yes Because the response to RAI 21 demonstrated that fuel containing mid-enriched blankets is more limiting than fuel containing natural blankets, only mid-enriched blanket fuel is considered in this penalty calculation. The same conditions summarized in Table 11 were also used for the IFBA penalty calculation presented in response to RAI 28. The results of this combined exit temperature and burnup profile calculation are summarized in Table 12 for Region 2 and Region 3. Note that in Table 12 a negative value denotes an increase in reactivity.

Serial No. 10-072A Docket No. 50-423, Page 7 of 15 Table 12 - Temperature and Burnup Profile Effects (~k) for 3.0, 4.0, and 5.0 wt% Fuel BU Decay Region 2 BU Decay Region 3 (MWd/MTU) Time (yrs) ~k (MWd/MTU) Time (yrs) ~k 15,000 0 -0.00244 25,000 0 -0.01364 3.0 15,000 10 -0.00290 3.0 25,000 25 -0.01092 wt% 20,000 0 -0.00723 wt% 30,000 0 -0.01500 20,000 10 -0.00932 30,000 25 -0.01118 25,000 0 -0.00273 40,000 0 -0.00125 4.0 25,000 10 -0.00124 4.0 40,000 25 +0.00918 wt% 30,000 0 -0.00313 wt% 45,000 0 -0.00072 30,000 10 -0.00030 45,000 25 +0.01290 40,000 0 -0.00243 55,000 0 +0.01049 5.0 40,000 10 +0.00138 5.0 55,000 25 +0.02602 wt% 45,000 0 -0.00199 wt% 60,000 0 +0.01453 45,000 10 +0.00374 60,000 25 +0.03114 RAI 21 burnup shape and nodalization results indicated that some of the WCAP-16721-P Keff results were non-conservative in the 20,000-30,000 MWd/MTU burnup range (Table 21-6, RAI 21). Combining increased moderator exit temperature with the conservative axial nodalization and burnup profiles results in larqer reactivity penalties in the 20,000-30,000 MWd/MTU as shown in Table 12. At high burnups, the conservatism inherent in WCAP-16721-P burnup profile ("ProfilE! 5", RAI 21) becomes much larger than the effect of increased moderator temperature. The burnup limits in Region 1 are too low to be significantly affected by these factors for three reasons. WCAP-16721-P calculations were performed with a uniform profile. As indicated in RAI question 3, a uniform axial shape is considered conservative for burnup < 10,000 MWd/MTU. The maximum calculated burnup requirement in WCAP-16721-P for Region 1 is 5,743 MWd/MTU (WCAP Table 5-1). Therefore, there are no end effect concerns for Region 1. Depletion history effects accumulate gradually with increasing assembly burnup, and the burnup requirement in Region 1 is very low. Region 1 T8 Figure 3.9-1, which is being retained, requires over 2,200 MWd/MTU more burnup at 5 wt% than required by the \\NCAP-16721-P analysis at 5 wt%. The additional burnup specified by T8 Figure :3.9-1 over and above the WCAP-16721-P calculated minimum fuel assembly burnup provides margin to further illustrate that no burnup penalty is needed for Reqion 1. Rather than attempt to identify off-setting conservatisms in the existinq analysis for Regions 2 and 3, burnup penalties calculated from reactivity penalties are applied to the final burnup credit curves. Prior to this RAI response, the response to RAI 21 and RAI 26 included a reactivity penalty. In this response, reactivity penalties are also identified for RAI 28 and RAI 30. The penalty in RAI 30 includes and supersedes the penalty

Serial No. 10-072A Docket No. 50-423, Page 8 of 15 identified in RAI 21. The reactivity penalties identified in RAls 26,28, and 30 are summarized in Tables 13 and 14. An IFBA penalty was not explicitly determined for 4.0 wt% fuel, so a generic penalty is applied. The penalty chosen is 0.00450 ~k and is larger than the penalties calculated for either 3.0 or 5.0 wt% fuel. In addition to the penalties applied in response to the RAls, the unallocated administrative margin is increased from 0.001 ~keff used in WCAP-16721-P to 0.005 ~keff. An additional 0.004 ~keff of unallocated administrative margin is included in Tables 13 and 14. Table 13 - Summary of Region 2 Penalties (~k) I Exit Temp. & i Decay Operating Burnup Additional BU Time History IFBA RAI Profile Admin. (MWd/MTU) (yrs) RAI26 28 RAI30 Margin Total 15,000 0 -0.00200 0 -0.00244 -0.00400 -0.00844 3.0 15,000 10 -0.00200 0 -0.00290 -0.00400 -0.00890 i wt% 20,000 0 -0.00200 -0.00033 -0.00723 -0.00400 -0.01356 I 20,000 10 -0.00200 0 -0.00932 -0.00400 -0.01532 25,000 0 -0.00200 -0.00450 -0.00273 -0.00400 -0.01323 4.0 25,000 10 -0.00200 -0.00450 -0.00124 -0.00400 -0.01174 wt% 30,000 0 -0.00200 -0.00450 -0.00313 -0.00400 -0.01363 30,000 10 -0.00200 -0.00450 -0.00030 -0.00400 -0.01080 40,000 0 -0.00200 -0.00160 -0.00243 -0.00400 -0.01003 5.0 40,000 10 -0.00200 -0.00112 +0.00138 -0.00400 -0.00574 wt% 45,000 0 -0.00200 -0.00173 -0.00199 -0.00400 -0.00972 I 45,000 10 -0.00200 -0.00175 +0.00374 -0.00400 -0.00401 Table 14 - Summary of Region 3 Penalties (~k) r--- Exit Temp. & I i Decay Operating Burnup Additional t BU Time History IFBA RAI Profile Admin. I (MWd/MTU) (yrs) RAI26 28 RAI30 Margin Total 25,000 0 -0.00200 -0.00298 -0.01364 -0.00400 -0.02262 I 3.0 25,000 25 -0.00200 -0.00192 -0.01092 -0.00400 -0.01884 wt% 30,000 0 -0.00200 -0.00343 -0.01500 -0.00400 -0.02443 30,000 25 -0.00200 -0.00235 -0.01118 -0.00400 -0.01953 40,000 0 -0.00200 -0.00450 -0.00125 -0.00400 -0.01175 4.0 40,000 25 -0.00200 -0.00450 +0.00918 -0.00400 -0.00132 wt% 45,000 0 -0.00200 -0.00450 -0.00072 -0.00400 -0.01122 45,000 25 -0.00200 -0.00450 +0.01290 -0.00400 +0.00240 55,000 0 -0.00200 -0.00194 +0.01049 -0.00400 +0.00255 5.0 55,000 25 -0.00200 -0.00127 +0.02602 -0.00400 +0.01875 wt% 60,000 0 -0.00200 -0.00221 +0.01453 -0.00400 +0.00632 60,000 25 -0.00200 -0.00172 +0.03114 -0.00400 +0.02342

Serial No. 10-072A Docket No. 50-423, Page 9 of 15 Table 15 shows eigenvalues used to calculate the burnup worth, the largest (most negative) applicable reactivity penalty identified in Tables 13 and 14, the burnup penalty applied, and the worth of the burnup penalty applied. The eigenvalues used to calculate the burnup worth are taken from Tables 4-9 and 4-10 ofWCAP-16721-P. For each enrichment/region/burnup combination, the penalty is conservatively chosen as zero or the most negative of the 0 decay time value or the decay time value from Tables 13 and 14. The Table 15 burnup worth was calculated by assuming a linear reactivity change over an interval of burnup near the point of potential non-conservatism. The difference in two SCALE Version 4.4 keff values was divided by the associated burnup difference to determine the reactivity worth associated with an increase in assembly average burnup. For example, the Region 2, 3.0 wt% burnup worth was calculated as: (0.87720 - 0.94421)/(25 -15) =-0.00670 ~k/GWd/MTU The "BU Penalty Applied" column in Table 15 shows the total burnup penalty that will be applied to the burnup limit for each initial enrichment at all decay times. Finally, the "Worth of BU Penalty" is the burnup penalty multiplied by the burnup worth to demonstrate that the reduction in reactivity due to the penalty is greater than the maximum identified non-conservatism. Net burnup credit is not taken in Table 15, even though net credit is indicated in Table 14 for the 5 wt% cases in Region 3. Monte Carlo uncertainties were not included in the determination of the burn up worth or the reactivity of the penalty because, as explained in response to RAI 16, the Monte Carlo uncertainty is already accounted for in the calculation of the total biases and uncertainties. Additionally, the conditions used to determine the reactivity penalty and the penalty selection method are both very conservative.

Serial No. 10-072A Docket No. 50-423, Page 10 of 15 d P If W rth T bl 15 B a e urnup 0 san ena res i Region 2 Worth of Max BU Penalty BU BU BU Worth Reactivity Applied Penalty (GWd/MTU) keff a (L\\k/GWd/MTU) (flk) (GWd/MTU) (flk) 3.0 15 0.94421 0.00034 -0.00670 -0.01532 2.30 -0.01541 wt% 25 0.87720 0.00029 4.0 25 0.95012 0.00030 -0.00519 -0.01363 2.65 wt% 35 0.89821 0.00030 -0.01375 5.0 35 0.95615 0.00032 -0.00442 -0.01003 2.30 wt% 45 0.91198 0.00031 -0.01017 i Region 3 Worth of Max BU Penalty BU BU BU Worth Reactivity Applied Penalty (GWd/MTU) keff a (L\\k/GWd/MTU) (flk) (GWd/MTU) (flk) I 3.0 25 0.91738 0.00033 -0.00539 -0.02443 4.70 -0.02533 wt% 35 0.86343 0.00032 4.0 35 0.93691 0.00032 -0.00460 -0.01175 2.60 -0.01196 wt% 45 0.89088 0.00027 5.0 45 0.94932 0.00032 -0.00422 0.0 0 0 ! wt% 55 0.90713 0.00030 The only penalty identified in Tables 13 and 14 that would affect the maximum fresh fuel enrichment is the additional administrative margin. The other penalties are not applicable because the reactivity of fresh fuel is not impacted by depletion conditions. In order to increase the amount of administrative margin associated with the determination of the maximum allowable fresh fuel enrichment, the enrichment uncertainties reported in Tables 4-6 and 4-7 ofWCAP-16721-P were used. The enrichment uncertainties reported assumed a 0.05 wt% increase in initial enrichment. To find the enrichment associated with a change in reactivity of 0.004 L\\k, 0.004 L\\k was divided by the enrichment uncertainty and multiplied by 0.05 wt%. Maximum fresh fuel enrichments for Regions 2 and 3 are shown in Tables 16 and 17, respectively. The calculated Region 1 maximum fresh fuel enrichment of 3.79 wt% is bounded by the existing TS Figure 3.9-1 value (3.7 wt%). The burnup versus enrichment curves provided in response to this RAI completely supersede the burnup versus enrichment curves previously provided. Tables 16 and 17 and Figures 2 and 3 show the new burnup versus enrichment curves. The calculations supporting these curves have been performed with conservatisms including high temperature, high soluble boron, maximum IFBA loading, higher than credible fuel density, and with the most conservative applicable burnup profile. In addition, allowances for depletion power history uncertainty, burnup worth uncertainty, and measured burnup uncertainty are conservatively treated as biases, and administrative

Serial No. 10-072A Docket No. 50-423, Page 11 of 15 margin is increased to 0.5% ~k. These factors provide assurance that the storage requirements presented in Tables 16 and 17 and Figures 2 and 3 are conservative. Table 16 Minimum Required Assembly-Average Burnup versus Initial 235U Enrichment and Decay Time for the "Region 2" Storage Configuration Initial Assembly Average Burnup Enrichment (MWd/MTU) (wt% 235U) oyr Decay 5 yr Decay 10 yr Decay 1.79 0 0 0 3.00 19191 18107 17378 4.00 31811 29849 28808 5.00 44638 41952 40290 The required assembly burnup as a function of 235U enrichment in the "Region 2" storage configuration is described by the following polynomials: Assembly Burnup (0 yr decay) = +489.007 e3 -5764.586 e2 +34878.836 e -46767.429 Assembly Burnup (5 yr decay) = +510.476 e3 -5945.212 e2 +34470.872 e -45581.560 AssemblyBurnup(10yrdecay)= +421.399 e3 -5030.783 e2 +31053.732 e -41883.913

Serial No. 10-072A Docket No. 50-423, Page 12 of 15 10 years oyears 5 years 5.00 4.50 4.00 3.50 3.00 2.50 2.00 i ... i..*** r .. ~- ..~ - i I I i I I ",.l-I Vl I-I -..-"1-- i/I I -1..- i-' /i) I i I J I J I I i 'y 7 1-I I / 7 _ / -;/ ~ACCEPTABLEl I -~...- 7 1-- I -r- / / I I / i I I I I I I*H- --.-.- -- -~ i 'V J... j I r I / vv I ~ 1/V 1-11 t.. J 1/1/ I..j.._ + I I

J 1/

I I ____ 1_ 1*..****- .. +- - -v' / J J: c--_.::_>-_ i 0-_".. I* IV / 1.-I-I* tT""" I* I ,--f--- L I '-1 / v I I I f* ..... -+ I I V. --l i '. +-. I / V V i IIUNACCEPTABLE I

1

/ //' I .L. I ~J'V I t /r>> I I / V)'i I i I I 'j', i I I

.. L I...

I !-Mk"- 'u i 1 i I I I: I $'1-J. +- -f-'" i--~-L -,~- II ~= I I ..~ r -I I " J fj =- --l*~-* I t l~ i 0-'" I +-( y; .~-+- I Iff --t---' i i_. I g I '\\+-r-I) '1=r ,- --I T.~~: I "1 t-tf I i i I 10,000 50,000 35,000 o 1.50 40,000 45,000 5,000 15,000 a.

IEdl 25,000

i5 J

~ 20,000 LL 5 ~ 30,000 i Initial 23*U Enrichment (nominal w/o) Figure 2 - Minimum Required Fuel Assembly Burnup versus Initial 235U Enrichment for the "Region 2" Storage Configuration NOTE: For assemblies from Post-Uprate (3650 MWt) Cores, the nominal fuel enrichment of blankets must be S 2.6 wt% U-235, and nominal blanket length must be at least 6 inches on both ends of the fuel. Fuel batches A, B, C, and 0 may not be stored in Region 2.

Serial No. 10-072A Docket No. 50-423, Page 13 of 15 Table 17 Minimum Required Assembly-Average Burnup versus Initial 235U Enrichment and Decay Time for the "Region 3" Storage Configuration for Post Uprate Cores Initial Assembly Average Burnup (MWd/MTU) Enrichment oyr Decay 5 yr Decay 10 yr 15 yr 20 yr 25 yr (wt% 235U) Decay Decay Decay Decay 1.43 0 0 0 0 0 0 3.00 30216 28304 27172 26413 25730 25374 4.00 43389 40617 38685 37440 36493 35685 5.00 55566 52057 49717 47978 46835 45874 The required assembly burnup as a function of 235U enrichment in the "Region 3" storage configuration is described by the following polynomials: Assembly Burnup (0 yr decay) = +522.404 e3 -6766.842 e2 +41211.965 e -46623.210 Assembly Burnup (5 yr decay) = +500.629 e3 -6444.047 e2 +38898.060 e -43910.737 Assembly Burnup (10 yr decay) = +564.139 e3 -7010.171 e2 +39711.046 e -44101.356 Assembly Burnup (15 yr decay) = +563.298 e3 -7004.075 e2 +39213.501 e -43399.874 Assembly Burnup (20 yr decay) = +554.180 e3 -6860.665 e2 +38282.980 e -42335.826 Assembly Burnup (25 yr decay) = +620.608 e3 -7508.292 e2 +39906.561 e -43527.461

Serial No, 10-072A Docket No, 50-423, Page 14 of 15 o years 10 years 15 years 20 years 25 years 5 years 5.00 4.50 4.00 3.50 3.00 250 2.00 1.50 'r -t-j-T-I +-r -+, I -=( t '1 I I H i I I t., T I 1-/fY7 -- --- ---T--

---

~---

.-t--,--- -~--- ---+- I I i.Ii T.: ~ 1--'---+ -~~ --,}j,i i I I I T iLl T--- /!~ --- +- I /1 i / /r: I ACCEPTABLE / 714 I 1/ / // -t-- --L~t- / / / --;--+- 1-'1-- -- / )1' '/ i i/ T ~ Ii,T !=T 1/ / '/ '/ / 'j j IF~ -t--- i T

/ v I

I IJ I ' 1/ ,t--iF J I: i, 1- / j-t-t---., / '/ ~~--F i J V V -IT' -ft---i I '/ / V 't,r--'- 'J / V-r--___U-I i I 1-j I Z ~' I

C---t

~ T i I-II/ // =H+ UNACCEP~;~LE}rl t I-! j - -j '//0 1= v/. 'I T 1---- - 'T -f ~ ~~*.**_._-# j ~-- --~- i 'i--- ~~ 1, TF-F I t- -1 r I r-- ----- r 1-I I [ J If, i j jh 'f T J Ii' T Ill.. 1=,,***, r I j~ i T I I i .: 1.i ___C !1!~" _I_~_, I I ,----j- -I-I-- - --Ii i I -I,4-II --7. + 1-,' I [11 +_t_ IAI ,-- -i- -- --;-- -r-t-r- ! !Jr...- t j-1--f---- j 1-\\ :+i=~L 1--+--+- i I l.r,--,- r ~7 1=1= i--C.--- i--- I--I-r I-tt-1--it --r I 10,000 20,000 60,000 40,000 45,000 o 1.00 15,000 50,000 5,000 55,000 Q.

lI:

~ 30,000

is E

5\\ Ul ~ 25,000 s u.. i 35,000 I Initial 235U Enrichment (nominal w/o) Figure 3 - Minimum Required Fuel Assembly Burnup versus Initial 235U Enrichment for the "Region 3" Storage Configuration NOTE: For assemblies from Post-Uprate (3650 MWt) Cores, the nominal fuel enrichment of blankets must be :S 2.6 wt% U-235, and nominal blanket length must be at least 6 inches on both ends of the fuel.

Serial No. 10-072A Docket No. 50-423, Page 15 of 15

References:

1. C. E. Sanders and J. C. Wagner, "Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit," NUREG/CR-6760, March 2002.

Serial No. 10-072A Docket No. 50-423 ATTACHMENT 2 RESPONSE TO RAI QUESTIONS 28 AND 30 REGARDING THE SPENT FUEL POOL LICENSE AMENDMENT REQUEST UPDATED MARKUP OF TECNNICAL SPECIFICATIONS PAGES DOMINION NUCLEAR CONNECTICUT, INC. MILLSTONE POWER STATION UNIT 3

( NovembeI 28, 2000 Y REFUELING OPERATIONS REFUELING OPERATIONS 3/4.9.13 SPENT FUEL POOL - REACTIVITY LIMITING CONDITION FOR OPERATION 3.9.13 The Reactivity Condition ofthe Spent Fuel Pool shall be such that keff is less than or equal to 0.95 at all times. APPLICABILITY: Whenever fuel assemblies are in the spent fuel pool. ACTION: With keff greater than 0.95: a. Borate the Spent Fuel Pool untillcetris less than or equal to 0.95, and b. Initiate immediate action to move anyfuel assembly which does not meet the requirements ofFigures 3.9-1, 3.9-j@t9-4, to a location for which that fuel assembly is allowed. cf; \\..C2!"'ij) SURVEILLANCE REQUIREMENTS 4.9.13.1.1. Ensure that all fuel assemblies to be placed in Region 1 "4-0UT-OF-4" fuel storage are within the enrichment and burnup limits ofFigure 3.9-1 by checking the fuel assembly's des' -up documentation. C"'-IJ.,.,....'~ 4.9.13.1.2. Ensure a el assemblies to be placed in Region 2 fuel storage are within the enric ent d burnup limits ofFigure 3,9-3 by checking the fuel assembly's desi d bum-up documen \\lS* j ~ lCd.. S tV..., iY\\ 6r'-LCI'YI,..... ' 3*W "'lIIt) tu"J,.fI~S r """'410\\ 4 yeo 4.9.13.1.3, Ensure that all fuel assemblies" 0 be placed in Region 3 fuel storage are WI n the enrichment, decay time, and bumup limits of Figure 3.9-4 by checking the fuel assembly's design, decay time, and burn-up documentation. E II SI1~ t~ fo 'lpq F~c9 ~S.st","~I'L:S ~Sc" i.. fost-"I"IIf-c.. (3{Sl>.-.111,.) u.-.J,'H*"'s Wkl'C.~ tH"4 to be f.tl\\c.c~ i" R~l~~ 3 fM,(~ .d',"~' eerL ttll tit,... + I L. A"~ h......,-"'I" q,'",,' ts id* FI;ltrc..1.1*S tk.f. -t~~ic~....eM ,.,l.I,ruk" I J I fl I ,lIS" 04~ h'l't\\<<', Qa1 ~., 'Mc.kl*~ +kc fr.tcP ".sSI"f.,:r I Pot1"'. wt' ~oc",-c~r~ h"l1

• MILLST0NE - UNlT 3.

3/49-16 .Amendment No. 39, 5%,.J:89.,

4.5 , 3.0 ' 3.5 4.0 Initial Fuel Enrichment ( w/o U*235 ) 2.5 15 -r----+-----:~~--~-+--.:..----=y...-----+------l 10 r------:;f--T----I~----+--------I---~--I------I 25 -r-----t----+--:'~~-+-+--:.,£---+-----:--+----~ November 28, 2000 FIG RE 3.9-3 Minimum Fuel Assembly Burnup Versus Nominal Initial Enrichm for Region 2 Storage Configuration Rep./", c-e; W ;tk) ~ tIC. t: ~ 30 t------;---~+-----h----++-~L---+----~ 35 r-----lT---+-----t------+--~--+--+:.--~ 40,---Jr----r-----,-----..----~-----r-.../-----. ~ ~ 3: CJ ~ 20 r----+-----+--~d7L-+_~---+_-----+-------1

sc..

J to G).r MILLSTONE - UNIT 3 3/4 9~20 Amendment No.189

Figure 3.9-3 Minimum Fuel Assembly Burnup and Decay Time Versus Nominal Initial Enrichment for Region 2 Storage Configuration years years o years 5.00 4.50 2.50 3.00 3.50 4.00 Initial 2>1U Enrichment (nominal w/o) 2.00 0 V 5 / / v 1 V ACCEPTABLE / v/ / 1/ / / V v VV iVIV ,I1/ V IV 1/ V. V VV IUNACCEPTABlE I //V rv '/ IV. /. '/ 1/,'(I rh /) H/ '/I if), lV f// h,. IH ~ v 45,000 5,000 50,000 o 1.50 40,000 35,000 10,000 .. 15,000 r'~ g,i 25,000 ~'J !l 20,000 IL NOTE: For assemblies from Post-Uprate (3650 MWt) Cores, the nominal fuel enrichment of blankets must be < 2.6 wlo U-235; and nominal blanket length must be at least 6 inches on both ends of the fuel. Fuel batches A, B, C, and D may not be stored in Region 2. MILLSTONE-UNIT3 3/49-20 Amendment No. 489

Ne'leAlger 28, 2000r FIGURE 3.9-4 Minimum Fuel Assembly Burnup and Decay Time Versus Nominal Initial Enrichment for Region 3 Storage Configuration po.,.. <<ss e..... &,.I'c.r V'DW\\ ~yt -c.e.pV'~"e.. (3~11 "'liJt) Coll:,5 60 -y----,------.' 50 r----i-----'1t__---t----t-----t--+----{] ACCEPTABLE DOMAIN 40 r----+----I-----t---~-___:r_-A<--,,e..---f

J.-

~ 3: CJ ~ 30 +----,...+----f---/---I-r-,"--+t----;-------;

Jc...

Jen Qj

J U.

20 +-----f-:l,e..--r--r-.:I7'----I-----t-----t-----t 10 ~'r_---1----+----t-----It__---t---_1 -.- 0 year decay time 5 year decay time -.lr-10 year decay time -*- 20 year decay time 5.00 4.50 4.00 3.50 3.00 2.50 O-l-----!----4-----I----I------I-----I 2.00 Initial Fuel Enrichment (w/o U-235), MILLSTONE - UNIT 3 3/4 ?-21 Amendment No.~,

Figure 3.9-5 Minimum Fuel Assembly Burnup and Decay Time Versus Nominal Initial Enrichment for Region 3 Storage Configuration for Assemblies from Post-Uprate (3650 Mwt) Gores* a years 15 years 20 years 25 years 5 years a years 1/ 1.1 1/ 1 U V 1..1 1/ 1/1.1 ~ ACCEPTABLE U r/ 1/ 1*/ 17 rj 1/ '~ 1/ 1..1 1.1 r..I 1/ 1/ 'l 1/ ~ v* I '.I '/.'61 1/ ',flU UNACCEPTABLE f'A [I I ~ I'"" 1.4 20,000 10,000 5,000 45,000 50,000 15,000 40,000 60,000 55,000 Iss.ooc ~. Ii! 30,000 ~, .cJ 25,000 ~ u, o 1.00 1.50 2.00 ~!.50 3.00 3.50 4.00 Initial u"u Enrichment (nominal wfo) 4.50 5.00 NOTE: For assemblies from Post-Uprate (3650 MWt) Cores, the nominal fuel enrichment of blankets must be ~ 2.6 w/o U-235, and nominal blanket length must be at least 6 inches on both ends of the fuel. MILLSTONE - UNIT 3 3/4 9-Amendment No.

NeV9FRBer 28. 2000 Y DESIGN FEATURES 5.6 FUEL STORAGE CRITICALITY (1) 5.6,,1.1 a. b. The spent fuel storage racks are made up of 3 Regions' whi'ch are designed and shall be maintained to ensure a !<eft less than or equal to 0.95 when fl boded with unbor-atsd water. The storage rack Regions are: Region 1, a nominal 10.0 Inch (North/S9uth) and a nominal lq.455 inch (East/West) center to center dtstance,,credits a fix;ed neut\\'"qn absorber (BORAl) within the rack, and canstote fueli" 2 storage configurations: With credit' Jor-fuel, blirrhji>a~ s*hown in Hgure 3.9~1,,fUel may be stored in a "4':'OUT;.0f:"'4" storage configurat.i()n. (2) With credit for every 4th location blocked and empty' of

fuel, fuel up to,S weight percent nomi"nql~nric:;hm.ent, regardless of fuel burnup, may be stored'ina llj --OUT';;OF'4 "

storage config~ration., Fuel storage i:n this c,onfiguration is subject to the inl~rface resttictionsspecified in Figure 3.9'-2. ",~fo\\ 'FIo\\~.{l clcat.~;h*~c;.. RegiO'ri2, a nomt 9.017 '1ncb.center to center drstanca, credits a fixed neutron bsorber (BORAL) within' the' rack, and with credit for fuel burnup as shown in Figure 3.9-3, fuel may be' stored in all available Region 2 storage locations. I", c. DRAINAGE 't Region 3, a nominal 10.35 inch center to center dis:tancejwith credf t for fuel burnup and fuel decay time as shown in Figure 3.9- ' 4~uel may be stored in all available Region 3 storage locations. Tnel Borafl ex contained ins' tara e rae* FoY' etS-S(..... ',*CrS t.\\s.ul.e.y.c l"f sN f. i... fte-t.t.~y..te. (3ill J1wf) (lI~C.S 00(" F,"~",vc. '3.4l-S' f-py AS.s'KoI.,.,JI~ MSe.J J'~p'$r-L{"tMo~,-;(>oHlf)t) 5.6.2 The spent fuel storage pool is design and shall be maintained to C4;y'c,$ prevent,inadvertent draining of the pool below elevation 45 feet. MlkL.STONE:., UNIT 3 ~.". i!:.

Serial No. 10-072A Docket No. 50-423 ATTACHMENT 3 RESPONSE TO RAI QUESTIONS 28 AND 30 REGARDING THE SPENT FUEL POOL LICENSE AMENDMENT REQUEST MARKUP OF TECNNICAL SPECIFICATIONS BASES PAGE FOR INFORMATION ONLY DOMINION NUCLEAR CONNECTICUT, INC. MILLSTONE POWER STATION UNIT 3

'\\- .Febrnary 20, 2002-BASES 3i4.9. J3 SPENT FUEL POOL - REACTIVITY During normal Spent Fuel Pool operation,the spent fuel racks are capableof maintaining Keff at less than 0.95 in an unboratedwaterenvironment. Maintaining Kerr at less than or equal to 0.95 is accomplished in Region I 3*0UT-OF-4 storageracks by the combination of geometryof the rack spacing, the use offixedneutron absorbers in the racks, a maximum nominal5 weight percent fuel enrichment,andthe use of blockingdevices in certain fuel storage locations, as specified by the interfacerequirements shownin Figure 3.9-2. MaintainingKerr at less than or equal to 0.95 is accomplished in Region 14-0UT-OF-4 storageracks by the combination ofgeometryofthe rack spacing, the use offixedneutron absorbers in the racks, and the limitson fuel enrichment/fuel burnup specitied in Figure3.9-1. Maintaining Keffat less than or equal to 0.95 is accomplished in Region 2 storage racks by the combination of geometry ofthe rack spacing, the use offixed neutronabsorbersin the \\' racks, and the limits on fuel enrichment/fuelbumup specified in Figure,1.l)-.~. II landfuel decay time I Maintaining Keff at less tbanor equal to 0.95 is accomplished in Region3 storage racks by the combination of geometry ofthe rack spacing,and the limits on fuel enrichment/fuel bumup and fuel decay time specified in Figure3.')..4;:Fixed neutron absorbers are notcredited in the Region 3 fuel storage racks. ~ ~, 3.9-4, and 3.9.5 I The limitations described by Figur 3.9-J, 3.9-2,3.9-] nlul-W-4 ensure that the reactivity ofthe fuel assemblies stored in the spent fuel pool are conservatively within the assumptions of the safety analysis. 3.9-4 for assemblies used exclusively Inthe pre-uprate (3411 MWt) cores and Flaure 3.9-5 forasse biles used I the DOSt-U te (3650 wn cores. Administrative controls have been developedand instituted to verify that the fuel enrichment, fuel burnup, fuel decay times, and fuel interfacerestrictions specifiedin Figures 3.9-1, 3.9-1, 3.9-3 afld*.a~9-4 are compliedwith 3.9-4. and 3.9.5 II!. 14dJ liS Tts h"l: "(~S S.,,~ CI'(oI'tel "---------l------~i'" +ke Nt>-t"c. Ok Ft,,,,,,,",-s 3/*t9.J4 SPENTFUELPOOL-STORAGEPATTERN ")-1-1 4"" ~_4l-S The limitations ofthis specificationensure that the reactivity conditionsof the Region 1 3*0UT-OF-4 storage racks and spent fuel pool keffwill remain less than or equal to 0.95. The Cell Blocking Devices in the 4th locationofthe Region 1 3-0UT-OF-4 storage racks are designed to prevent inadvertentplacement andlor storage offuel assembliesin the blocked locations. The blocked location remainsempty to provide the flux trap to maintain reactivity control for fuel assemblies in adjacent and diagonal locations ofthe STORAGE PATIERN. STORAGE PATIERN for the Region 1storage racks will be established and expanded from the walls of the spent fuel pool per Figure 3.9-2to ensure definition and control of the Region 13-0UT-OF-4 boundary to other storageregions and minimize the number ofboundaries Where a fuel misplacement incident can occur. MILLSTONE - UNIT 3 B 3/49-9 .Amendment No. 39, 95, M+, ~, 89, 203}}