ML20095D624
| ML20095D624 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/16/1992 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20095D609 | List: |
| References | |
| NUDOCS 9204270157 | |
| Download: ML20095D624 (25) | |
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I Docket No. 50-315
. f 4102 Attachment 1 Millstone Nuclear Power Station, Unit-No. 2 Proposed Revision to Technical Specifications Spent Fuel Pool Reactivity Proposed Revised Pages April-1992 9204270157 920416 PD9 ADOCK 05000336 P PDR
IllDEX LIMITING COND]TIONS FOR OPERATION AND SURVEllLANCE REQUIREMENTS SECTION EAG[ 3/4.9 REFUEllNG OPERATIONS 3/4.9.1 BORON CONCENTRATION ..................................... 3/4 9-1 3/4.9.2 INSTRUMENTATION ......................................... 3/4 9-2 3/4.9.3 DECAY TIME .............................................. 3/4 9 3 3/4.9.4 CONT AINMENT PENET RAT IONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-4 3/4.9.5 COMMUNICATIONS .......................................... 3/4 9-5 3/4.9.6 CRANE OPERABillTY - CONTAINMENT BUILDING ................ 3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P00L BUILDING. . . . . . . . . . 3/4 9-7 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION . . . . . . . . . . . . . . . 3/4 9-8 3/4.9.9 CONTAINMENT RADIATION MONITORING ........................ 3/4 9-9 3/4.9.10 CONTAINMENT PURGE VALVE ISOLATION SYSTEM ................ 3/4 9-10 3/4.9.11 WATER LEVEL - REACTOR VESSEL ...................... ..... 3/4 9-11 3/4.9.12 STORAGE P0OL WATER LEVEL ................................ 3/4 9-12 3/4.9.13 STORAGE POOL RADIATION MONITORING ....................... 3/4 9-13 3/4.9.14 STORAGE P00L AREA VENTILATION SYSTEM - FUEL MOVEMENT .... 3/4 9-14 3/4.9.15 STORAGE P00L AREA VENTILATION SYSTEM - FUEL STORAGE . . . . 3/4 9-16 3/4.9.16 SHIELDED CASK ........................................... 3/4 9-19 3/4.9.17 MOVEMENT OF FUEL IN SPENT FUEL P0OL ..................... 3/4 9-21 - 3/4.9.18 SPENT FUEL P00L - REACTIVITY CONDITION ................., 3/4 9-22 3/4.9.19 SPENT FUEL POOL - STORAGE PATTERN ....................... 3/4 9-26 3/4.9.20 SPENT FUEL P0OL - CONSOLIDATION ......................... 3/4 9-27 3/4.10 SPECIAL TEST fXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN ........................................ 3/4 10-1 3/4.10.2 GROUP HEIGHT AND INSERTION LIMITS ...................... 3/4 10-2 3/4.10.3 PRESSURE /TEMPLRATURE LIMITATION - REACTOR CRITICALITY .. 3/4 10-3 MILLSTONE - UNIT 2 IX Amendment No. E9, Jp4, Jp9,
**** JJ7, J57
i
.. INDEX BASES i
SECTION PAq[ 3/4.9.9 and 3/4.9.10 CONTAINMENT AND RADIATION MONITORING AND CONTAINMENT PURGE VALVE ISOLATION SYSTEM ................. B 3/4 9-2 , 3/4.9.11 and 3/4.9.12 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL WATER LEVEL ................................. B 3/4 9-2 3/4.9.13 STORAGE POOL RADIATION MONITORING ................... B 3/4 9-3 l 3/4.9.14 and 3/4.9.15 STORAGE POOL AREA VENTILATION SYSTEM ... B 3/4 9-3 3/4.9.16 SHIELDED CASK ....................................... B 3/4 9-3 3/4.9.17 MOVEMENT OF FUEL IN SPENT FUEL POOL ................. B 3/4 9-3 3/4.9.18- SPENT FUEL POOL REACTIVITY CONDITION .............. B 3/4 9-3 3/4.9.19 SPENT FUEL P0OL - STORAGE PATTERN ................... B 3/4 9-4 3/4.9.20 SPENT FUEL P00L - CONSOLIDATION ..................... B 3/4 9-4 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN-MARGIN ..................................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT AND INSERTION LIMITS ................... B 3/4 10-1 3/4.10.3 PRESSURE / TEMPERATURE LIMITATION - REACTOR CRITICALITY ............................... B 3/4 10-1 3/4.10.4 PHYSICS TESTS ....................................... B 3/4 10-1 3/4.10.S CENTER CEA MISALIGNMENT ............................. B 3/4 10-1 3/4.11- RADI0 ACTIVE EFFLUENTS 3/4.'11.1 LIQUID EFFLUENTS .................................... B 3/4 11-1 3/4.11.2 GASE0US EFFLUENTS .......... ........................ B 3/4 11-2 3/4.11.3 TOTAL DOSE-.......................................... B 3/4 11'-4 MILLSTONE - UNIT 2 XIV Amendment No. E9, Jp/, J99, JJ7 ooss Jyy
DEFINITIONS VENTING 1.35 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during venting. Vent, used in system names, does not imply a VENTING process. MEMBER (S) 0F THE PUBllC 1.36 HEMBER(S) 0F THE PU3LIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the pl ant. The term "REAL MEMBER OF THE PUBLIC" means an individual who is exposed to existing dose pathways at one particular location. SITE BOUNDARY 1.37 The SITE B0UNDARY shall be that line beyond which the land is not owned, leased or otherwise controlled by the licensee. HNRESTRICTED AREA 1.38 An UNRESTRICTED AREA shall be any area at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection of individuals from expcsure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industrial, commercial institutional and/or recreational purposes. STORAGE PATTEkN 1.39 The Region B and C spent fuel racks contain a cell blocking device in every 4th rack lochtion for administrative control. This 4th location will be referred to as the blocked location. A STORAGE PATTERN refers to a blocked location and all adjacent and diagonal cell locations surrounding the blocked location within the respective region. MILLSTONE - UNIT 2 1-8 Amendment No. JE,4, JJJ 0060
. REFUELING OPERATIONS MOVEMENT OF FUEL IN SPENT FUEL POOL 11MITING CONDITION FOR OPERATION __
3.9.17 Prior to movement of a fuel assembly, or a consolidated fuel storage box, in the spent fuel pool, the boron concentration of the pool shall be maintained uniform and sufficient to maintain a boron concentration of greater than or equal to 800 ppm. APPLICABIllTY: Whenever a fuel assembly, or a consolidated fuel storage box, is moved in the spent fuel pool. ACTION: With the boron concentration less than 800 ppm, suspend the mo tement of all fuel in the spent fuel pool. SURVElllANCE RE001REMENT 4.9.17 Verify that the boron concentration is greater than or equal to 800 ppm within 24 hours prior to any movement of a fuel assembly, or a consolidated fuel storage box, in the spent fuel pool and every 72 hours thereafter. l l l MILLSTONE - UNIT 2 3/4 9-21 Amendment No. Jp9, JJ7 0061
MFUELING OPERATIONS SPENT FUEL P0OL--REACTIVITY CONDITION (LMITING CONDITION FOR OPERATION 3.9.18 The Reactivity Condition of the spent fuel pool shall be such that K eff is less-than-or-equal-to 0.95 at all times. APPLICABillTY: Whenever fuel is in the spent fuel pool. ACTION: ! Borate until K eff f .95 is reached. \ SURVEILLANCE REQUIREMENT 4.9.18.1 Ensure that all fuel assemblies to be placed in Region C (as shown in Figure 3.9-2) of the spent fuel pool are within the enrichment and burn-up limits of Figure 3.9.1 by checking the assembly's design and burn-up documen-tation. 4.9.18.2 Ensure that the contents of each consolidated fuel storage box to be ~ placed in Region C (as shown in Figure 3.9-2) of the spent fuel pool are - within the enrichment and burn-up limits of Figure 3.9-3 by checking the design and burn-up documentation for storage box contents. 4.9.18.3 Ensure that all fuel assemblies to be placed in Region A (as shown in Figure 3.9-2) of the spent fuel pool are within the enrichment and burnup limits of figure 3.9-4 by checking the assembly's design and burnun documentation. MILLSTONE - UNIT 2 3/4 9-22 Amendment No. JM , JJ/
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. 0.0 3.3 3.5 3.7 3.9 4.1 4.3 4.5 FUEL ASSEMBLY INITIAL ENRICHMENT, WT. % U 235 FIG. 3.9-4 ' MINIMUM 'AEQUIRED FUEL- ASSEMBLY EXPOSURE AS A FUNCTION OF INITIAL ENRICHMENT TO PERMIT STORAGE IN REGION A l
l MILLSTONE - UNIT 2 3/4 9 25a 0061
4 REFVELING OPERATIONS SPENT FUEL POOL - STORAGE PATTERN LIMITING CONDITION FOR OPERATION 3.9.19.1 Each STORAGE PATTERN of the Region C spent fuel pool racks shall require either that: (1) A cell blocking device is installed in those cell locations shown in Figure 3.9 2; or (2) If a cell blocking device has been removed, all cells of the STORAGE PATTERN must have consolidated fuel in them, including the formerly blocked location; or (3) Meet both (a) and ('): b (a) lf a cell blocking device has been removed, all cells of the STORAGE PATTERN must have consolidated fuel in them except the formerly blocked location. , (b) The formerly blocked location is vacant and a consolidated fuel box or cell blocking device is immediately being placed into
-the formerly blocked cell.
APPLICABillTY: Fuel in the Spent fuel Pool ACTION: ! Take immediate action to comply with either 3.9.19.l(1), (2) or (3). l SURVEILLA.NCE RE0VIREMENTS 4.9.19.1 Verify that 3.9.19.1 is satisfied at the following times. (1) Prior to removing a cell blocking device l (2) Prior to removing a consolidated fuel storage box from its Region C storage lscation. l
. MILLSTONE - UNIT 2- 3/4 9-26 Amendment No. JJ7, JEJ 0061
REFUELING OPERAT10fd SPENT FUEL POOL - STORAGE PATTERN ljMITING CONDITION FOR OPERATION 3.9.19.2 Each STORAGE PATTERN of the Region B spent fuel pool racks shall require that: (1) A cell blocking device is installed in those cell locations shown in Figure 3.9-2; or (2) If a cell blocking device has been removed, all cells in the STORAGE PATTERN must be vacant of stored fuel assemblies. APPLICABillTY: Fuel in the spent fuel pool. ACTION: Take immediate action to comply with either 3.9.19.2(1) or (2). EURVEILLANCE RE0VIREMENTS 4.9.19.2 Verify that 3.9.19.2 is satisfied prior to renoving a cell blocking device. MILLSTONE UNIT 2 3/4 9-26a 0061
REFUELING OPERATIONS BASES 3/4.9.13 STORAGE POOL RADIATION MONITORIE The OPERABILITY of the storage pool radiation monitors ensures that sufficient radiation monitoring capability is _ available to detect excessive radiation levels resulting from 1) the inadvertent lowering of the storage pool water level or 2) the release of activity from an irradiated fuel assembly. 3/4.9.14 & 3/4.9.15 STORAGE POOL AREA VENTILATION SYSTEM The limitations on the storage pool area ventilation system ensures that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the- atmosphere. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the accident analyses. 3/4.9.16 SHIELDED CASK The limitations of this specification ensure that in an event of a cask tilt accident 1) the doses from ruptured fuel assemblics will be ithin the assumptions of the safety analyses, 2) K g77 will remain s .95, 3/4.9.17 MOVEMENT OF f_yEL IN SPENT FUEL P0OL The limitations of this specification ensure that, in the event of a fuel assembly or a consolidated fuel storage box drop accident into a Region 8 or C rack location completing a 4-out-of-4 fuel assembly geometry, K eff will remain i s 0.95. 3/4.9.18 SPENT FUEL POOL - kEACTIVITY CONDITION The limitations described by Figures 3.9-1 and 3.9-3 ensure that the reactivity _of fuel assemblies and consolidated fuel storage boxes, . introduced into the Region C spent fuel racks, are conservatively within the assumptions of the safety analysis. The : limitations described by figure 3.9-4 ensure that the reactivity of ! the fuel assemblies, introducted into the Region A_ spent fuel racks, are conservatively within the assumptions of the safety analysis. l r l i l MILLSTONE - UNIT 2 _B 3/4 9-3 Amendment No 39, JOS, JJ7, 00 0 yyy
REFUEllNG OPERATIONS BASES 3/4,9,19 SPENT FVEL P0OL - STORAGE PATTERL1 The limitations of this specification ensure that the reactivity conditions of the Region B and C storage racks and spent fuel pool Keff Will remain less than or equal to 0.95. The Cell Blocking Devices in the 4th location of the Region C storage l racks are designed to prevent inadvertent placement and/or storage of fuel assemblies in the blocked locations. The blocked location remains empty to provide the flux trap to maintain reactivity control for fuel assembly storage in any adjacent locations. Or'y loaded consolidated fuel storage boxes may be placed and/or stored in the 4th location, completing the STORAGE PATTERN, after all adjacent, and diagonal, locations are occupied by loaded consolidated fuel storage boxes. - The Cell Blocking Devices is the 4th location of the Region B storage racks are designed to prevent inadvertent placement and/or storage in the blocked locations. The blocked location remains empty to provide the flux trap in maintain reactivity control for fuel assembly storage in any adjacent locations. Region B is designed for the storage of new assemblies in the spent fuel pool and for fuel assemblies which have not sustained sufficient burnup to be stored in Region A or Region C. 3/4.9.20 SPENT FVEL POOL - CONSOLIDATION The limitations of these specifications ensure that the decay heat rates and radioactive inventory of the candidate fuel assemblies for consolidation are conservatively within the assumptions of the safety analysis. MILLSTONE - UNIT 2 B 3/4 9-4 Amendment No. JJ7, JSJ 0063
4 I . DESIGN FEATURES _ 10_LUML 5.4.2 - The ' total water and steam volume of the reactor coolant system is
- 10,060 + 700/-0 cubic feet.
5.5 EMERGENCY CORE COOLING S.YSTLMS 5.5.1 The emergency core cooling systems are designed and shall be maintained in accordance with the original design provisions contained in Section 6.3 of
~ the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.
5.6 FUEL STORAGE CRITICALITY 5.6.1 a) Ihe new fuel (dry) storage racks are designed and shall be maintained with suffichnt center to center distance between assemblies to
< .95. The maximum nominal fuel enrichment to be stored in ensure a k*N 4.50 weight percent of U-235, there racks b) Region A of the spent fuel storage pool is designed and shall be maintained with a nominal 9.8 inch center to conter distance between storage locations to ensure a K f s .95 with the storage pool filled with unborated water. Fuel assembliesYtored in this region must comply with Figure 3.9-4 to ensure that the design burnup has been sustained, c) Region B of the spent fuel storage pool is designed and shall be maintained with a nominal 9.8 inch center-to-center distance between storage locations to ensure K filled with unborated water. Fuel assemblieY7 s .95inwith stored thisaregion storage poc1 may have a maximum nominal enrichment of 4.5 weight percent U-235. Fuel assemblies stored in this region are placed n a 3 out of-4 STORAGE PATTERN for reactivity control.
d) Region C of the spent fuel storage pool is designed and shall be l maintained with a 9.0 inch center to center distance between storage locations i to ensure a K s .95 with. the storage pool filled with' unborated water. Fuel assemblie 7[tored -in this region must comply with Figure 3.9-1 to ensure that the design burn-up has been sustained. Fuel assemblies stored in this region are placed in a 3 out of 4 STORAGE PATTERN for reactivity control. The contents of- consolidated fuel storage boxes to be stored in this region must
- comply with Figure 3.9-3.
e) Region C of the spent fuel storage pool is designed to permit storage of consolidated fuel in the 4th location of the storage rack - and ensure a-X ~< 0.95. Placement of consolidated fuel in the 4th location is
- only permi$d if all surrounding cells of the STORAGE PATTERN are occupied by
- consolidated fuel.
MILLSTONE - UNIT 2 5-5 Amendment No. M , S , R9, "n D7,1%
DESIGN FEATURES DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draitling of the pool below elevation 22'6". CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 224 storage locations in Region A, 160 storage locations in Region B and 962 storage locations in Region C for a total of 1346 storage locations.* ,
*This translates into 1237 storage locations to receive spent fuel and i 109 storage locations to remain blocked. I I
w MILLSTONE - UNIT 2 5-Sa Amendment No. JE, EE, JP9,
ill,1H
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4 4 Docket No. 50-336
- B14102
.1 Attachment 2 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Spent Fuel Pool Reactivity Spent Fuel Pool Criticality Safety Analyses 8
April 1992
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U.S. Nuclear Regulatory Commission Attachment 2/B14102/Page 1 April 16, 1992 Spent Fuel Pool Criticality Safety Analyses This attachment is intended to -document results of our criticality safety analyses of the Millstone Unit No. 2 Region 1 storage cells with observed and postulated gaps present in the Boraflex abarber material. The Boraflex poison degradation has been very conservatively incorporated into the criticality design analysis. To date, approximately half of the poisoned rack - cells have been tested - and characterized - for gap formations. Test data identifies a Boraflex panel defect rate of 16% with the largest observed gaps at a 2% shrinkage rate, With further gap growth anticipated, the mechanical inputs for the criticality analysis assumed 4% gap formations at the observed test locations and a 4% gap formation with a random distribution in all of the other Boraflex panels. These assumptions are considered conservative because EPRI data supports the 4% maximum shrinkage value and the random distribution is supported by the NNECO test data. These analyses are based on the CE design. for the Region 1 Boraflex poisoned racks, as originally licensed for Millstone Unit No. 2 in Amendment #109, dated January 15,- 1986. The calculations utilize a three-dimensional NITAWL-KENO-Sa model with the 27-group SCALE cross-section set. Sections of the old Region I have been redefined as two new regions: Region A- utilizing all of the cells in a 4-of-4 cell arrangement with credit for fuel burnup. Region B using fresh fuel of 4.5% average enrichment in a 3-of-4 arrange-ment (fourth cell empty). Shrinkage -of 4X, was also assumed resulting in 5.65" gaps in every Boraflex panel _(a very conservative assumption). The consequence of various axial distrfutions was also investigated. A shrinkage of 4% in width was conserva-tively assumed although examination of the Boraflex from Cell D9 did not~ show any visible evidence of such shrinkage, Table i summarizes ~ re ults of several calculations (including the original design with fresh fuel in' every location) intended to show the magnitude of the reactivity effects of in the Millstone Unit No. 2 racks. To provide some perspective for the analyses, a calculation was made assuming all- the Boraflex was lost, resulting in a 0.194 6k total reactivity " worth" of - the Boraflex. If 4% is -lost through gap formation, then the order of magni-tude of the- expected reactivity effect due- to gaps is (0.04 0.194) - 0.008 Sk.
U.S. Nuclear Regulatory Commission
. Attachment-2/B14102/Page2 April 16, 1992 Reaion 8: 3 of 4 Cell Arranaement with15% Fresh fuel Calculations for a 3-of-4 arrangement with fresh 4.5% enriched fuel (fourth cell empty) are summarized in Table 2. At the present time, the gaps which have been observed in the Boraflex (as of the most recent Blackness test) have a negligible reactivity effect within the statistical accuracy of KENO-Sa calculations.
Considering that the Boraflex has already seen three fuel cycles, it is not likely that significant fu <her growth would be expected, flowever, for conservatism, it was assumed Rat these gaps increase to 5.65" (equivalent to 4% axial shrinkage) at the locations observed in the- Blackness tests (Case 7, 4 Table 2). To assure very conservative upper bound conditions, further calculations were made assuming that additional gaps of 5.65" appear in all other panels throughout the racks. Based on the fact that the axial distribution of observed gaps is random, a random distribution of these additional gaps in the axial direction was assumed as the reference case. (Gap locations were derived by using a PC random number generator.) The maximum k for the upper bound reference case ' Case 9, Table 2) was calculated to'Ye 0.9179, including width shrinkage, bias and all uncertainties (calculational and manufacturing tolerances, see Table 3). Thus, with the 3-of-4 arrangement, the maximum keff remains substantially below the NRC criterion (0.95 keff)' Westinghouse and CE fuel show a slightly higher reactivity than the ANF fuel used for the primary analyses. For Westinghouse and CE fuel, the maximum.
- reactivities for- 4.5% enriched fuel were calculated to be 0.9252 and 0.9201 respectively.
The temperature and void coefficients of reactivity are negative. Therefore, the calculations were conservatively based on a temperature of 4*C (maximum
- water density) and any temperature increase above 4'C would result in reduced reactivity.
Two accident conditions were also considered, as follows: (Note: Under the accepted single failure criterion, it is not necessary to consider the simul-taneous occurrence of multiple independent accident conditions. Therefore, credit for the presence of soluble poison is allowed under accident condi-
- tions.)
. Mislocated fuel assembil--For the case of a fresh fuel assembly assumed to be accidentally- installed into one of the empty cells of an otherwise filled Region B array, the maximum K 77 was calculated to be 0.9436, which remains below the NRC criterio7
. Mislocated Consolidated fuel assembiv--This accident assumes that a consolidated fuel bundle is accidentally loaded into one of the empty cells of Region B. Calculations for this case resulted in a
U.S. Nuclear Regulatory Commission Attachment 2/B14102/Page 3 April 16, 1992 maximum k of 0.9364 which is well within the NRC criterion. This is a verfIconservative calculation that assumes a consolidation ratio of 2 with unburned rods of 4.5% enrichment rather than spent fuel rods. Sionificance the Axial Gap Distribution Since the potential effect of the axial gap distribution is of concern, we have calculated the reactivity effect of 5.65" gaps for several assumed distributions. The actual distribution of gaps in the Millstone Unit No. 2 - Boraflex appears to be random or very r.early so. Blackness Tests conducted in many plants generally substantiate the assumption of a random distribution. Calculations for two assumed distributions are summarized in Table 4. On the basis of this evaluation, it is concluded that the distribution of gaps in the axial direction has a comparatively minor impact on the distribution. The observed gap distribution (augmented to 5.65" at all gap locations plus a random distribution of 5.65" gaps in Boraflex panels which did not have gaps) yields the same reactivity as the assumption of a completely random distribu-tion of the same size gaps. For the extreme (and noncredible) case of all gaps assumed to occur only in the central 50% of the rack height, the k was 0.005 Sk above the randomly distributed case, and for an assumed *[bsine distribution of gaps, the reactivity was 0.0028 Sk higher than the reference random distribution. Neither of these hypothetical distributions would result in exceeding the NRC criterion. In addition, we have investigated the consequences of the Boraflex shrinkage resulting in a reduction in length of 5.65" (4%) exposing an unpoisoned zone at the ends of the fuel assemblies. This case resulted in a smaller reactivi- - ty effect than the case of gaps distributed throughout the rack. Reaion A: 4-of-4 Cell Arrangement with Burnup Credit The storage racks are capable at accepting spent fuel utilizing all cells. Calculations have been made for the storage racks loaded in a 4-of-4 arrang:- ment with spent fuel of a specified minimum burnup. Since the required burnup is not large, we selected a conservative value for the design k 0.9317 with fuel of 4.5% enrichment and 8670 MWD /MTU burnup) knowingmoth![7 if (st, not all, of the spent fuel have burnups well in excess of the minimum required. Thus, a conservative value may be used without significant impact on Millstone Unit No. 2 operations. With this design basis reactivity, the misloading of either a fresh fuel assembly or a consolidated fuel bundle will not result in exceeding the NRC criterion.
4 U.S. Nuclear Regulatory Commission Attachment 2/814102/Page 4 April 16, 1992 Table 3 summarizes the uncertainties for Region A, based upon fuel of 4.5% initial enrichment burned to 8670 MWD /MTV. With these uncertainties, the maximum k is 0.9317 (95% probability at the 95% confidence level), Calcu-lations wb also made for oQer assumed initial enrichments and a curve of limiting burnup (for the same reactivity) is presented in Figure 3.9-4 of Technical Specifications. With Westinghouse or CE fuel of 4.5% initial , average enrichrent, the burnup limit curve will be the same although the calculated reactivities will be slightly higher ('0.9381 and 0.9335 for Westinghouse and CE fuel respectively). Discharged f uel would normally be expected to have burnups considerably in excess of the minimum required, resulting in a much lower reactivity. Calculations for Region A were also made to determine the effect of the axial distribution in burnup. At the low design basis burnup for Region A, no effect was expected and calculations showed that the k with axially dis-tributed burnups is less than that of the reference unif h burnup case. (See also Turner, "An Uncertainty Analysis--Axial Burnup Distribution Effects" in Sandia Report SAND 89-018, October 1989.) Interfaces with Other Reaions Calculations were also made to determine--if there might be any adverse reac-tivity effects along the interface between regions. Even without credit for the isolating water-gap between modules, no adverse effects were found for any of the interfaces--Regions A and B, Regions A and C, and Regions B and C (see l Figure 3.9-2 in Technical Specifications). Region C is the old Region II, designed _for burned fuel, Based upon _ the analyses performed, it is concluded that, in the presence of l the conservatively postulated maximum gaps (4% or 5.65") in all Boraflex panels, 4% shrinkage in width, and all uncertainties included, that (1) the Millstone Unit No. 2 spent fuel storage racks can safely accom-
-modate fresh 4.5% enriched fuel in a 3 out of 4 loading pattern with the fourth cell empty.
(2) the Millstone Unit No. 2 spent fuel storage racks can safely accom-modate spent fuel of the burnup-enrichment combinations indicated in Figure 3.9-4-of the Technical Specifications, using all cells in a 4 out of 4 arrangement. (3) no credible accident c;ndition will result in exceeding the regula-eff 5 95. tory reactivity limit of k (4)- the assumed axial gap distribution has only a minor impact on the calculated reactivity of the racks (measured distribution used in
- the reference case analysis), and, for any credible assumption of the distribution of postulated gaps, the maximum keff will remain within NRC criterion, f
b
- - - , --,s. e -
r -+ , c - .,,,n
U.S. Nuclear Regulatory Commission Table 1/B14102/Paga 1 April 16, 1992 Table 1 Background KENO-Sa Calculations (width shrinkage not included) Hai _. Case Max k gpf 1 Original Design (4 of 4), no gaps 0.9812 2 4 of 4 Loading, random 5.65" gaps 0.9879 3 3 of 4 Loading, no gaps 0.9113 4 3 of 4 loading, random 5.65" gaps 0.9163 5 no Boraflex 1.0838 1
8 II.S. iuclear Regula'.ory Commission i cie 2/B:4102/Page 1 April 10. 1992 Table 2 Criticality Calculations for 3 of 4 Loading herangement (4.5% Enriched fuel No Burnop)
!!91 CALe -
M32_l eff 3 3 of 4 loading patte c. no gaps 0.9113 6 With gaps as measured in Blackness 0.9110 Testing 7 Observed Gaps increased to 5.65" 0.9126 I!b 3 Observed Caps increased to 5.65" plus 0.9163 5.65" gaps randomly distributed in all other Boraflex Panels 9* Same as Case 8 but with 4% shrinkage 0.9179* in width of the Boraflex 10 Reference Case (9) with Westinghouse 0.9251 fuel 11 Reference Case (9) with CE fuel 0.9201 12 Accident of a fresh 4.5% assembly 0.9420 installed in an empty cell 13 Accident of 4.5% Consolidated Bundle 0.9348 installed in an empty cell
- Reference Case
U.S. Nuclear Regulatory Commission Table 3/B14102/Page 1 April 16, 1992 Table 3 Calculations Uncertainties and Reactivity Effects of Manufacturing lolerances Reactivity 6k item h aion A kniOIL]) Uncertainty in Blas i 0.0018 KEN 0 Statistics (95%/95%) i0.0019(4f5)) (or i 0.0012 B 10 Loading Tolerance 1 0.0022 i 0.0020 (1 0.003 g B-10/cmt) Boraflex Width (i 1/16") 1 0.0009 i 0.0016 Enrichment Tolerance (.* 0.09%) i 0.0020 1 0.0020 002 Der,.tyTolerance(i2%) i 0.0021 1 0.0021 Lattice Spacing (1 0.09") 1 0.0096 1 0.0113 SS Box ID (i 0.05") i 0.0042 1 0.0073 SS wall thickness ( 0.012") i 0.0015 1 0.0053 _ Uncertainty in Depletion NA i 0.0028 Calculations (5% in burnup) Statistical Average 1 0.0115 1 0.0154 (or i 0.0114) (or 1 0.0153) (4) fcr 1000 generations of 500 neutrons each. (5) for 2500 generations of 500 neutrons each,
U.S. Nuclear Regulatory Commission Table 4/014102/Page 1 April 16, 1992 Table 4 Significance of the Axial Distribution of Gaps locations Nidth shrinkage of the Boraflex not included)
!!L ._
Case But_k,f f 3 3 of 4 arrangement, no gaps 0.9113 8 With coscrved gap locations 0.9163 (5.65" gaps) and a random distribution of 5.66" gaps in all other Boraflex panels 9A With a random distribtition of 5.65" 0.9163 gap. i. all panels 15 With an assumed cosine distribution 0.9191 in gap locations 16 Random distribution '.o gap locations 0.9212 in the central 50i' of the axial height . _ _ _ _ . _ _ _ . . _ _ _ _ _ _ _ __ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , _ _ _ _ _ _ _ _ _ _ _}}