ML20046A094
| ML20046A094 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 07/13/1993 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20046A092 | List: |
| References | |
| NUDOCS 9307260181 | |
| Download: ML20046A094 (65) | |
Text
{{#Wiki_filter:. - _ . . . . _ . , .. . . . . _ . - .- . . . , . _ , . e 1 s ATTACHMENT 31 - 9 DUKE POWER COMPANY MCGUIRE NUCLEAR STATION- i MCGUIRE. UNIT 2 CYCLE 9 RELOAD' REPORT y i t 4 o k
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'i RELOAD REPORT McGuire Unit 2 Cycle 9 a
i l F 1 Duke. Power Company Nuclear Generation' Department'
- Nuclear Services Division - "
Nuclear ~ Engineering Section
, Charlotte, North. Carolina g . '[
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- 1. INTRODUCTION AND
SUMMARY
.............. .................... 1-1
- 2. OPERATING-HISTORY......................................... 2-1
- 3. GENERAL DESCRIPTION....................................... 3-1
- 4. FUEL SYSTEM DESIGN........................................ 4-1 4.1. Fuel Assembly Mechanical Design................... 4-1 4.2. Fuel Rod Design................................... 4-1 4.2.1. Fuel Rod Cladding Collapse................ 4-1 4.2.2. Fuel Rod Cladding Stress.................. 4-1 4.2.3. Fuel Rod Cladding Strain.................. 4-2 4.3. Thermal Design.................................... 4-2 4.4. Material Design................................... 4-2 4.5. Operating Experience.............................. 4-2 a
- 5. NUCLEAR DESIGN............................................ 5-1 5.1. Physics Characteristics........................... 5-1 ,
5.2. Nuclear Design Methodology........................ 5-1
- 6. THERMAL-HYDRAULIC DESI GN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 -1
- 7. ACCIDENT ANALYSIS......................................... 7-1
- 8. PROPOSED MODIFICATIONS TO LICENSING BASIS DOCUMENTS....... 8-1 8.1 Changes to Technical Specifications . . . . . . . . . . . . . . . 8-3 8.2 Changes to the Final Safety Analysis Report....... 8-23
- 9. STARTUP PHYSICS TESTING................................... 9-1 10 . REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 - 1 F
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List of Tables , l Table Page 4-1 Fuel Design Parameters and Dimensions . . . . . . . . . . . . . . . . . . 4-3 5-1 Physics Parameters, McGuire 2 Cycles 8 and 9 . . . . . . . . . . 5-2 5-2 Shutdown Margin Calculation for McGuire 2 Cycle 9 . . . . . . 5-4
- 6-1 System Uncertainties Included in the Statistical Core Design Analysis ................... ................ 6-2 6-2 Nominal Thermal-Hydraulic Design Conditions, -
McGuire 2 Cycle 9 .............'........................ 6-3 6-3 DNBR Penalties ........................................ 6-4 8-1 Technical Specification Changes . . . . . . . . . . . . . . . . . . . . . . . . 8-2 l List of Ficures '. Figure Page 3-1 Core Loading Pattern for McGuire Unit 2 Cycle 9 ..... .. 3-2 3-2 Enrichment and BOC Burnup Distribution for McGuire Unit 2 Cycle 9 ........................................3-3 3-3 McGuire Unit 2 Cycle 9 Rod Cluster Control Assembly Locations and Bank Designations ........................ 3-4 3-4 McGuire Unit 2 Cycle 9 B'irnable Absorber and Source Assembly Locations ...................................... 3-5 5-1 BOC (4 EFPD) , Cycle 9 Twc-Dimensional Relative Power Distribution - HFP, Equila brium Xenon . . . . . . . . . . . . . . . . . . 5-5
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'1. INTRODUCTION AND SUM'1ARY This report justifies the operation of-the ninth cycle of McGuire Nuclear Station, Unit 2 at the rated core power level'of 3411 MW th.
Included are the required analyses as outlined in the USNRC document.
" Guidance for Proposed License Amendments Relating to Refueling," July 1975.
The incoming Mark-DW fuel for Cycle 9 is the second McGuire Unit 2 reload batch supplied by B&W Fuel Company (BWFC). To support implementation of Mark-BW fuel in the McGuire and Catawba nuclear stations, Duke Power Company (DPC) developed methods and.models are used to analyze the plants during normal and off-normal operation. The thermal-hydraulic analytical models are documented in topical reports DPC-NE-3000 (Reference 11) and DPC-NE-3002-A (Reference 16) for non-LOCA transients and BAW-10174-A (Reference 13) for LOCA. Portions of the analytical methodology are documented in topical reports DPC-NE-3001-PA (Reference 12) and DPC-NE-2004P-A (Reference 8) . Section 2 of this report describes the operating history for fuel ir McGuire Unit 2. Section 3 is a general description of the reactor core, and the fuel system design is provided in Section 4. Reactor and system parameters and conditions are summarized in Sections 5, 6, and
- 7. Changes to the Technical Specifications, Core Operating Limits Report (COLR), and Final Safety Analysis Report (FSAR) are provided in Section 8. The scope of Startup Physics Testing for McGuire Unit 2, .
Cycle 9 is provided in Section 9. All of the accidents analyzed in the FSAR (Reference 1) have been reviewed'and are applicable for Cycle 9 operation. In those cases l where Cycle.9 characteristics were conservative compared to those t' analyzed for previous cycles, new analyses were not performed. With the exception of the post-LOCA'suberiticality and boron dilution analyses,'the cycle 9 thermal-hydraulic and physics parameters are bounded by the existing MNS FSAR Chapter 15 analyses. The results of c reanalyzed accidents for McGuire Unit 2 Cycle 9 are included in' L Sections 7 and 8. Amendment Number 105 (Unit 1) and Amendment Number 87 (Unit 2) to the' McGuire Nuclear Station Facility Operating License allow the removal of cycle-specific core parameter limits from Technical Specifications and require that these limits be included in'a Core Operating Limits Report (COLR). The Core Operating Limits Report is submitted to the NRC upon issuance and does not require approval prior to' implementation. Changes to the core operating limits are made via the Core Operating Limits' Report. The Technical Specifications have been reviewed, and modifications 7 required for Cycle 9 are given in section 8. Based on the analyses l performed, it has been concluded that McGuire Unit 2 Cycle 9 can be safely operated up to the rated core power level of 3411 MNth- I
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- 2. OPERATING HISTORY The current operating cycle for McGuire, Unit 2 is Cycle 8, which achieved criticality on March 15, 1992 and reached 100% full power on-March 25, 1992. Cycle 8 is scheduled to shut down in June 1993 after 400 +6/-10 EFPD. No operating anomalies have occurred during cycle 8 operations that would adversely affect fuel performance in Cycle 9.
McGuire, Unit 2, Cycle 9 is scheduled to start up in September 1993 at a rated power level of 3411 MWt and has a design cycle length of 395 +/- 1-0 EFPD. 2-1
p 3, GErERAL DESCRIPTIO11 The McGuire Unit 2 reactor core is described in detail in Chapter 4 of the FSAR (Reference 1). The core consists of 193 assemblies, each of
-which is a 17X17 array containing 264 fuel rods, 24 guide tubes, and 1 incore instrument tube. The McGuire 2 Cycle 9 core has 117 burned assemblies and 76 fresh assemblies. The fuel rod outside diameters are 0.360 and 0.374 inch, and the wall thicknesses are 0.0225 and 0,024 inch for the OFA and Mark-BW designs, respectively. The Mark-BW fuel consists of dished end, cylindrical pellets of uranium dioxide, (See Table 4-1 for data). The design loadings are 423.5 and 456.3 kg of -uranium per assembly for OFA and Mark-BW fuel, respectively. The i initial design enrichments of batches 8A, 9A, 10A, and 10B were 3.50 wt % '"U , 3.50 wt% 2"U, 3.95 wt% 2"U, and 3. 80 wt% 2"U. The design enrichment of the fresh batch 11A is 3.65 wt% 2"U.
Figure 3-1 gives the full core loading pattern for Cycle 9. 40 batch ; 9A, 72 batch 10A, and 4 batch 10B assemblies from Cycle 8 will be shuffled to new locations. One batch 8A assembly will be reinserted into the core from the spent fuel pool. The 76 fresh batch 11A assemblies will be loaded into the core in a basically symmetric checkerboard pattern. Figure 3-2 is a quarter core map showing the burnup and region reference number of each assembly at the beginning of Cycle 9. It also provides batch average enrichment and burnup. Cycle 9 will be operated in a-feed-and-bleed mode. Core reactivity is controlled by 53 rod cluster control assemblies (RCCAs), 864 Mark-BW burnable absorbers, and soluble boron shim. The Cycle 9 locations of the 53 rod cluster control assemblies with their respective designations are indicated in Figure 3-3. The Cycle 9 locations of Mark-BW BPRA clusters and number of pins enriched.to 2.5 and 3.0 wt% B4 C-Al O2 3 are shown in Figure 3-4. i i l l 3-1 i 1
FIGURE 3-1 i CORE LCADING PATTERN FOR MCGUIRE UNIT 2 CYCLE 9 '
) 10A 10A 10A 11A 10A 10A 10A D-13 F-15 M-05 F D-05 K-15 M-13 2 10A 10A 11A 10A 11A 10A 11A 10A 11A 10A 10A M-09 J-14 F J-10 F H-09 F G-10 F- G-14 D-09 3 10A 10B 11A 10A 11A 10A 10A 10A 11A 10A 11A 10B 10A G-04 A-08 F L-02 F K-13 H-11 F-13 F E-02 F H-15 J-04 4 10A 11A 9A 11A 9A 11A 9A 11A 9A 11A 9A 11A 10A B-07 F K-11 F L-15 F G-15 F E-15 F L-06 F P-07 5 10A 11A 10A 11) 9A 11A 9A 11A 9A 11A 9A 11A 10A 11A 10A C-12 F P-05 r A-09 F B-12 F P-12 F J-15 F B-05 F N-12 6 10A 10A 11A 9A 11A 9A 11A 9A 11A 9A 11A 9A 11A 10A 10A A-10 F-07 F A-05 F K-05 F N-14 F E-06 F R-05 F K-07 R-10 7 10A 11A 10A 11A 9A 11A 9A 11A 9A 11A 9A 11A 10A 11A 10A L-04 F C-06 F D-14 F N-02 F B-03 F M-14 F N-06 F E-04 e
8 11A 10A 10A 9A 11A 9A 11A BA 11A 9A 11A 9A ICA 10A 11A F G-08 E-08 A-07 F B-13 F J-10 F P-03 F R-09 L-08 J-08 F 9 10A 11A 10A 11A 9A 11A 9A 11A 9A 11A 9A 11A 10A 11A 10A L-12 F C-10 F D-02 F P-13 F C-14 F M-02 F N-10 F E-12 e 10 10A ICA 11A 9A 11A 9A 11A 9A 11A 9A 11A 9A 11A 10A 10A A-06 F-09 F A-11 F L-10 F C-02 F F-11 F R-11 F K-09 R-06 11 10A 11A 10A 11A 9A 11A 9A 11A 9A 11A 9A 11A 10A 11A 10A C-04 F P-11 F G-01 F B-04 F P-04 F R-07 F B-11 F N-04 l-12 10A 11A 9A 11A 9A '11A 9A 11A 9A 11A 9A 11A 10A B-09 F E-10 F L-01 F J-01 F E-01 F F-05 F P-09 13 10A 10B 11A 10A 11A 10A 10A 10A 11A 10A 11A 10B 10A G-12 H-01 F L-14 F K-03 H-05 F-03 F E-14 F R-08 J-12 14 10A 10A 11A 10A 11A 10A 11A 10A 11A 10A 10A M-07 J-02 F J-06 -F H-07 F G-06 F C-02 0-07 15 10A 10A 10A -11A 10A 10A 10A D-03 F-01 M-11 F D-11 K-01 M-03 R P N M L K J H G F e D C B A contains natU rod # Cyc;e ~/ Reinsert YY Region Numter Z-ZZ Cycle 8 locations 2 7 3-2
a a . - as .#z 4.u :.. - ._ 4 >.u.. i FIGURE 3-2 ENRICHMENT AND BOC BURNUP DISTRIBUTION FOR MCGUIRE 2 CYCLE 9 H G F E D C B A 28655.8 0 23088.0 0 30220.0 19559.4 19871.2 0 g- 26797.2 0 20572.5 0 24453.5 18739.5 19057.5 0 29940.8 0 24768.6 0 33955 8 20899.7 21179.1 0 1 2 3 4 5 6 7 8 0 23091.7 0 28661.3 .0 19111.7 0 19314.8 ! g 0 20121.7 0 23020.1 0 18013.9 0 17902.9 0 25054.1 0 32135.3 0 21209.3 0 21211.1 9 10 11 12 13 14- 15 16 > 23088.0 0 30709.1 0 25759.0 0' 19823.0 12797.9 10 20572 5 0 24539.2 0 20696.5 0 18934.8 5693.5 24768.6 0 34543.9 0 29146.6 0 21214.2 17459.7 17 18 19 20 21 22 23 24 O 28657.6 0 30253.0 0 17486.9 0 18278.3 11 0 23008.8 0 24475.7 0 10832.1 0 14685.3 0 32156.5 0 34008.0 0 20708.5 0 20786.0 25 26 27 28 29 30 31 32 30220.0 0 25760.6 0 30697.0 0 18121.3 12 24453.5 0 20674.6 0 24537.7 0 15943.8 33955.8 0 29163.7 0 34518.9 0 20855.0 33 34 35 36 37 38 39 19721.6 19117.8 -0 17487.8 0 14298.3 19186.6 13 18913.2 18021.2 0 10832.6 0 6764.4 17609.3 21041.1 21208.1 0 20706.2 0 18707.0 21165.4 40 41 42 43 44 45 46 19950.9 0 19730.7 0 18138.3 19207.2 ! 14 19150.3 0 18866.4 0 15954.5 17641.8 ' 21217.7 0 21167.0 0 20070.5 21159.3 47 48 49 50 51 52 0 19308.8 12794.2 18295.2 Assembly Avg BU 15 0 17310.1 5692.3 14696.0 Minimum BU 0 21209.0 17451.9 20808.8 Maximum BU ' 53 54 55 56 FA # 5 ENRICHMENT CYCLES NUMBER OF BOC BURNUP REGION w/o U-235 BURNED ASSEMBLIES MWD /MTU ; BA 3.50 (OFA) 2 1 28656 9A 3.50 (OFA)- 2 40 27690 10A 3.95 (MkBW) 1 72 18208 ' 10B 3.80 (OFA) 1 4 14299 11A 3.65 (MkBW) 0 76 0 CORE N/A N/A 193 12976 ( 3-3
i 4-FIGURE 3-3
- MCGUIRE UNIT 2 CYCLE 9 ROD CLUSTER CONTROL ASSEMBLY LOCATIONS & B ANK DESIGNATIONS 2 SA CB CC CD SA r
3 SD SB SB SC 4 SA CD SE CD SA 5 SC SD 6 CB CC CA CC CB 7 SB SU 1 8 CC SE CA CD CA SE CC 9 SB SD t0 CB CC CA CC CB. 11 SD SC 12 SA CD SE - CD SA 13 SC SB SB SD 14 SA CB CC. CB SA e 15 , R P N M L K 1 11 0- F. E D -C B A , XX . ROD BANK IDEN11FIDt 1 3-4
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RGdRE 3-4 MCGUIRE UNIT 2 CYCLE 9 BURNABLE ABSORBER AND SOURCE ASSEMBLY LOCATIONS 1 1 0 2 8 16 16 8 3- 8 16 SS 16 8 4 8 16 12 12 16 8 5 8 16 8 12 8 16 8 ! 6 16 8 12 12 8 16 4 7 16 12 12 12 12 12 16 8 0 12 12 12 12 0 9 16 12 12 12 12 12 16 10 16 8 12 12 8 16 11 8 16 8 12 8 16 8 12 8 16 12 12 16 8 f 13 8 16 SS 16 8 14 8 16 16 8 15 0 R P N M L K J H G F E D C B A NUMBER OF NUMDER OF BL4LNABIF ABSORRFR PINS BACKPI ATES 8 24 12 24 16 24 Total 864 72
- refers to assembly locations with 2.5 w/o bps. All others have 3.0 w/o bps.
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- 4. -FUEL SYSTEM DESIGN 4.1- Fuel Assembly Mechanical Desian The McGuire 2 Cycle 9 core will include 76 fresh Mark-BW fuel-assemblies with an enrichment of 3.65 wt %U235 The re-inserted fuel assemblies in Cycle 9 will be Westinghouse Optimized fuel assemblies (45) and Mark-BW fuel assemblies (72). The Mark-BW 17 x 17 Zircaloy spacer grid fuel assembly is similar in design to the Westinghouse standard fuel assembly, Reference 2. The fuel rod outer diameter.and guide tube top section, dashpot diameters, and instrument tube diameter are the same as the Westinghouse standard 17 x 17 design.- The unique ~ !
features of the Mark-BW design include the Zircaloy intermediate spacer grids, the spacer grid restraint system, and the use of Zircaloy grids with the standard lattice design. Mark-BW fuel design dimensions and parameters for McGuire 2 Cycle 9 are listed in Table 4 . 4.2 Fuel Rod Desian Duke Power Company has performed generic Mark-BW mechanical analyses using the approved methodologies described in Reference 3. The generic ' analyses envelope the McGuire 2 Cycle 9 reinsert fuel. Critical Cycle 9 fresh fuel as-built parameters will be compared against values assumed in the generic analyses prior to cycle startup. This will determine the applicability of the analyses to the fresh fuel. The cladding collapse and minimum LHRTM limits in Table 4-1 are based upon ' these generic analyses. 4.2.1 Puel Rod Cladding Collapse The fuel rods.were analyzed for creep collapse using the-CROV computer code, Reference 4, and the methodology described in Reference 3. ' Internal pin pressures and' clad temperatures used in CROV were calculated using the TACO 2 computer code, Reference'5. A conservative power history which envelopes the predicted peaking for. the McGuire 2 Cycle 9 fuel was analyzed. The collapse time was conservatively
- determined to be greater than the maximum-predicted residence time for the Mark-BW fuel (Table 4-1) .
c 4.2.2- Fuel Rod Cladding. Stress As described in Reference 3, Duke Power Company has performed a conservative generic fuel rod cladding stress analysis using the ASME-pressure vessel stress intensity limits as guidelines. 'The maximum cladding stress intensities were shown to be within the ASME limits under all loading conditions. The generic Mark-BW cladding stress analysis ~ includes the following conservatisms: Conservative cladding dimensions. High external pressure. ' Low internal pin pressure. High radial temperature gradient through the clad. 1
4.2.3 Fuel Rod Cladding Strain Diametral Cladding strain resulting from a local power transient is limited to 1.0 %. A generic cladding strain analysis was performed using TACO 2 to determine the maximum allowable local power change that the fuel.could experience without exceeding the 1.0 % limit. The maximum calculated local power change resulting from a worst case core f maneuvering scenario was compared with the maximum allowable power change. This comparison demonstrated that margin exists to the 1% strain limit. 4.3 Thermal Desion The thermal performance'of the Mark-BW fuel assemblies was evaluated using TACO 2 with the methodology given in Reference 3. The nominal fuel parameters used to determine the generic linear heat' rate to centerline melt (LHRTM) limits are given in Table 4-1. The LHRTM i analysis included the following bounding conservatisms: Maximum gap based on as-fabricated pellet and clad data. Maximum incore densification based on resinter test results.
- The maximum predicted Mark-BW assembly burnup at EOC 9 (in Batch 10; is 37,218 MWD /MTU and the maximum predicted fuel rod burnup (in Batch 10) is'38,483 MWD /MTU. The fuel rod internal pressure has been evaluated for the highest burnup rod using TACO 2 and a conservative pin power history. The maximum internal pin pressure is less than the core exit pressure of 2280 psia.
4.4 Material Desion The fresh Mark-BW fuel is not unique in concept, nor does it utilize different component materials. Thus, the chemical compatibility of all possible fuel-cladding-coolant-assembly interactions for the fresh fuel is identical to that of existing Westinghouse OFA and Mark-BW fuel' types. 4.5 Ooeratina Exnerience Experience with the Mark-BW 17 x 17 fuel assembly design started with 3 the irradiation of four lead assemblies in McGuire 1 Cycle 5. .Three l assemblies from this program completed irradiation in McGuire 1 Cycle 7 i with a maximum assembly burnup of 42,756 MWD /MTU. The lead assemblies were examined after each cycle and the fuel assembly bow, twist, growth, and holddown spring set were all within nominal bounds. McGuire 2 Cycle 9 will be the seventh reload batch of Mark-BW fuel supplied to Duke Power Company. l 4-2 f
Table 4-1. Fuel Design Parameters and Dimensions Mark-BW Batch 10 Batch 11 Nominal fuel rod OD, in. 0.374 0.374 Nominal fuel rod ID, in. 0.326 0.326-Nominal active fuel length, in. 144.0 144.0 Nominal fuel pellet OD, in. 0.3195 0.3195 Fuel pellet initial density, % TD 96.0 96.0 Initial fuel enrichment, wt. %U235 3.95 3.65 Estimated fuel assembly average burnup (mwd /mtU) 37,659 20,164 Cladding collapse burnup, (mwd /mtU) >38,700 >38,700 Nominal linear heat rate (LHR), kW/ft 5.43 5.43 Ave, fuel temperature 8 nom. LHR, deg F 1360 1360 Minimum LHR to melt, kW/ft 0-1000 MWD /MTU 21.5 21.5
> 1000 MWD /MTU 21.8 21.8 4-3 E
- 5. . NUCLEAR DESIGN
-5.1 Physics Characteristics Table 5-1 provides the core physics parameters for Cycles 8 and 9 The values were' generated using the methodology described in DPC-NF-2010A (Reference 6) and DPC-NE-3001A (Reference 12). Cycle 9 values are valid for the design cycle length (395 EFPD 1 10 EFPD). Figure 5-1 illustrates a representative relative power distribution for the beginning of Cycle 9 at full power. This case was calculated as part of the design depletion using the PDQ07 methodology as described in DPC-NF-2010A (Reference 6) . This case contained equilibrium xenon and rods in the all rods out (ARO) position. During verification of the control rod insertion limits specified in the COLR, calculated ejected rod worths and their adherence to acceptance criteria were considered. The adequacy of the shutdown margin is demonstrated in Table 5-2. The shutdown margin calculations-include a 10% uncertainty in the available all. rods in.(ARI) position-minus the most reactive stuck rod worth at HZP. The shutdown calculation at the end of Cycle 9 was analyzed at 405 EFPD (395 EFPD + 10 EFPD window). 5.2. Nuclear Desian Methodoloav The Cycle 9 physics parameters appearing in this report were calculated with the PDQ07 and EPRI-NODE-P codes. These codes and methods were approved by the NRC as documented in Reference 6. The PDQ07 calculations were performed in two dimensions; the EPRI-NODE-P calculations were performed in three dimensions. The Reactor. , Protection System (RPS). limits and operational limits for the core were verified by analyses for this fuel cycle using methodology approved by the NRC in Reference 7 The operational limits.are provided in the COLR. I
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l 7 i Table 5-1 Physics Parameters McGuire 2 cycles 8 and 9 i Cycle 8 Cycle 9 i Design nominal' cycle length (a) , EFPD 400 395 Design cycle burnup, MWD /MTU '16226 15560 ,
-Design average core burnup, MWD /MTU 28628 28630 ,
Design initial core loading, MTU. 84.0899 86.5899-Critical boron - BOC,ppmb, no XeIDI HZP, ARO 1696 1750 HFP, ARO 1548 1572 i Critical boron - EOC,ppmb 571 630 HZP, ARO, No Xe HFP, ARO, Eq.~Xe 0 9 ; Total control rod worths --HZP, pcm t BOC 6702 ' 6829 1 EOC(c) 7327 6921 Max ejected rod worth (d) - HZP,pcm
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BOC (D12), Peak Xe 327 302 EOC (D12), Peak Xe 470 445 Max stuck rod worth - HZP, pcm BOC (F-10) 843 ' 821 .. EOC(C}-(F-10) 813 932 d 1 Power defect - HZP to HFP, pcm - . BOC, Eq Xe -1605 -1662 EOC(c), Eq Xe -2952 -2898-Doppler coeff -fHFP, pcm/0F BOC, no Xe -1.17 -1.19 EOC ( c ) ~, eq Xe -1.45 -1.48 Moderator coeff - HFP,'pcm/'F BOC, no'Xe -2.70 -- 4 :. 4 2 - EOC(c), eq Xe, O PPMB -31.51 -33~.22 ;i Boron worth - HFP, pcm/ppmb ' BOC, Eq Xe -7.88 -7.50 1 EOC (c) -9.23 -8.65 9
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I t Table 5-1 Physics Parameters McGuire 2 Cycles 8 and 9 (cont) . Cycle B- Cycle 9 - Equilibrium Xenon worth - HFP, pcm
'BOC (4.EFPD) 2616 2585 EOC 2940 2934 Effective delayed neutron fraction HFP .
BOC .0.006228 0.006254 - EOC 0.005244 0.005238 (a) EOC corresponds to 400 EFPD for cycle 8 and 395 EFPD for cycle 9, except where noted. t ( b) ' HZP denotes hot zero power (core average 557 F Tavg); HFP denotes hot full power (core average 5920F Tavg). (c) EOC physics parameters calculated at design EOC plus window. The cycle 8 positive burnup window is 6 EFPD and the cycle 9 window is 10 EFPD. * (d) Ejected rod worth for banks D, C, and B inserted to HZP RIL. b I i P h l l q
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Table.5-2. Shutdown Margin Calculation for McGuire 2 Cycle 9 , Control Rod Worth BOC'(PCM) EOC i ( PCM)
- 1. All rods inserted (ARI), HZP 6633 6921
- 2. ARI less most reactive stuck rod, HZP 5812~ 5989 [
- 3. Less 10% uncertainty 5231 5390 Total available rod worth 5231 5390 Recuired Rod Worth
- 4. Rod insertion allowance (RIA) (b) 243 343
- 5. Power defect, HFP to HZp(b) 1952 3188 Shutdown margin (total available worth less 3036 1859 total required worth) ,
i NOTE: Required shutdown margin is 1300 PCM. t (a) BOC physics parameters are calculated at 4 EFPD. EOC physics , parameters are calculated at 405 EFPD, i.e., design.EOC plus 10 , EFPD. (b) The rod insertion allowances and power defects used in the shutdown margin calculation account for the effects of transient xenon condition.7 5-4
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Figure 5-1 BOC (4 EFPD), Cycle 9 Two Dimensional Relative Power Distribution HFP, Equilibrium Xenon H G F E D C B A 1.0010 1.2589 1.0807 1.2197 .9762 1.1859 1.1068 .8439 8 1.0235 1.3528 1.1552 1.3170 1.0384 1.2654 1.2264 1.1422 1.0225 1.0746 1.0689 1.0798 1.0637 1.0670 1.1080 1.3535 M N Q Q D M A Q C M L N C M C M 1.2566 1.0806 1.2291 .9915 1.2410 1.2355 1.0945 .6711 9 1.3487 1.1443 1.3502 1.0590 1,3372 1.3569 1.2893 .9670 1.0733 1.0590 1.0985 1.0681 1.0776 1.0982 1.1779 1.4410 Q Q N E A A E D P I D M B I C E 1.0760 1.2220 .9485 1.2575 1.0349 1.2155 1.0607 .6448 10 1.1506 1.3462 1.0058 1.3430 1.0995 1.3352 1.2220 .9875 1.0694 1.1016 1.0604 1.0680 1.0624 1.0984 1.1521 1.5315-M D A A O E P I D M I B C E A A 1 2144 .9863 1.2557 .9408 1.1377 1.1993 1.0493 .4582 11 1.3104 1.0534 1.3447 1.0327 1.2762 1.3102 1.3003 .8405 1.0791 1.0680 1.0709 1.0977 1.1218 1.0925 1.2392 1.8342 I B D E I P B F Q A N E C G A A
.9729 1.2380 1.0357 1.1381 .9111 1.1542 .7174 12 1.0347 1.3345 1.1032 1.2763 .9836 1.3139 1.0716 1.0635 1.0779 1.0652 1.1214 1.0795 1.1383 1.4937 M C I P M D A Q Q A H C A A 1.1807 1.2332 1.2154 1.1988 1.1589 .8551 .4093 13 1.2594 1.3548 1.3343 1.3108 1.3131 1.1653 .7668 1.0667 1.0986 1.0979- 1.0934 1.1331 1.3628 1.8737 N L M D B I E N C H A A A A 1.1034 1.0933 1.0622 1.0513 .7233 .4156 P(AVG) 14 1.2215 1.2875 1.2237 1.3019 1.0747 .7830 Peak Pin 1.1070 1.1776 1.1520 1.2384 1.4859 1.8838 Peak / Ass M C I B E C G C A A A A Pin Loc.
.8416 .6707 .6454 .4590 15 1.1403 .9659 .9880 .8415 1.3533 1.4402 1.5309 1.8333 M C E C A A l A A The maximum assembly power is 1.2589 at location G-8.
The maximum pin power is 1.3569 at location C-9. The maximum pin to assembly factor is 1.8838 at location C-14. 5-5
- 6. THERMAL-HYDRAULIC DESIGN The generic and cycle-specific analyses supporting cycle 9 operation were performed by Duke Power Company using the methodology described in Reference 8. Cycle 9 was analyzed using Duke's Statistical Core Design (SCD) methodology. Uncertainties on parameters that affect DtB performance are statistically combined to determine a Statistical DNBR limit (SDL). Using the BWCMV correlation. Reference 9, a generic SDL of 1.40 was calculated using a set of generic uncertainties given in Reference 8'. The system parameter uncertainties used in Reference 8.
and given in Table 6-1 bound the uncertainties specifically calculated for McGuire. Reactor core. safety limits for Cycle 9 are based on a full Mark-BW core and a design FAH of 1.50. The Cycle 9 nominal thermal-hydraulic design conditions are given in Table 6-2.
- The Mark-BW fuel assembly was designed to be hydraulically compatible with Westinghouse optimized fuel (OFA). BWFC has performed a series of flow tests to verify the compatibility of the two designs. The tests showed that the total pressure drop across the OFA fuel is 2.4% higher than the pressure drop across the Mark-BW fuel, Reference 2. A generic transition core analysis was performed to determine the DIGR impact of this difference.
Since the Mark-BW fuel has a. lower overall pressure drop than the OFA design, a Mark-BW assembly in a mixed core will tend to have more flow through it and consequently more DNB margin than the same assembly in an all Mark-BW core. Conversely, flow will be forced out of the OFA fuel in a mixed core; thus, the need to calculate a DIGR penalty for . the OFA fuel. A generic transition core DNBR penalty was determined by . ' modeling a conservative core configuration with'one OFA assembly as the hot assembly. The rest of the core was modeled as Mark-BW fuel. A number of statepoints and peaking conditions were analyzed, yielding a maximum DNBR penalty of 3.8% for the OFA fuel. To provide ~ design' flexibility, margin is added to the SDL to determine a design DNBR limit (DDL). For the generic Mark-BW and McGuire 2 Cycle 9 analyses, the DDL is 1.55 (10.7% margin above the SDL). =The DNBR penalties, such as the OFA transition core penalty, that must be assessed aga' inst the margin are given in Table 6-3. l 1 1 6-1 .1 i p
?
e Table 6-1 System Uncertainties Included in the
' Statistical Core' Design Analysis Reference 8 L
Parameter Uncertainty Distribution Core power +/- 2 % Normal RCS flow +/- 2.2 % Normal Core bypass flow +/- 1.5 % Uniform Pressure +/- 30 psi Uniform Inlet temperature +/- 4 deg F Uniform l s 9 i
?
I w l 6-2
'i
p, Table 6-2. [ c. Nominal Thermal-Hydraulic Design Conditions McGuire 2 Cycle 9 Core power, MWt 3411 Core exit pressure, psia 2280 Vessel ave, temperature, Deg F 588.2 RCS flow, gpm 385,000 Core bypass flow, % 7.5 Reference design FAH 1.50 Reference _ design axial shape 1.55 Cosine CHF correlation BWCMV Statistical DNBR limit 1.40, Design DNBR limit 1.55 6-3 I I
i Table 6-3. - DNBR Penalties l t Statistical DNBR limit 1.40 Design DNBR limit 1.55 , DNBR margin 10.7 % DNBR Penalty Mark-BW OEA Transition core 0% 3.8 + Instrumentation / hardware 4.7 % 1.9-% Rod bow 0% L.5_1 Total DNBR penalty 4.7 % 9.2 % Available DNBR Margin 6.0 % 1.5 % I I 9 0 h l i 6-4 -l 1
- 7. ACCIDENT ANALYSIS Safety Analysis Each.FSAR accident listed below has been examined with respect to changes in Cycle 8 parameters to determine the effect of the Cycle 9 reload and to ensure that thermal performance during hypothetical transients is not degraded.
. Increase in feedwater flow
. Excessive load increases
. Steam system piping failure
. Turbine trip
. Feedwater system pipe break
. Partial loss of forced reactor coolant flow
. Complete loss of forced reactor coolant flow
. Reactor coolant pump shaft seizure (locked rotor)
. Uncontrolled rod bank withdrawal from subcritical or low power startup condition
. Uncontrolled rod bank withdrawal at power
. Dropped rod / rod bank
. Statically misaligned rod
. Single rod withdrawal ,
. Startup of an inactive reactor coolant pump
. Boron dilution
. Rod ejection
. Steam generator tube failure
. Loss-of-coolant accidents With the exception of the post-LOCA subcriticality and boron precipitation analyses, the cycle 9 thermal-hydraulic and physics parameters are bounded by the existing MNS FSAR Chapter 15 analyses.
The post-LOCA suberiticality' evaluation for Cycle 9 fails the acceptance criteria'with the existing refueling water storage tank (RWST) and cold leg accumulator (CLA) minimum boron concentration limits. The post-LOCA subcriticality analysis for Unit 2 is reanalyzed-using a RWST minimum boron concentration of 2175 ppm.and a CLA minimum boron concentration of 2000 ppm. The results of this reanalysis are acceptable for Unit 2 Cycle 9 operation. Thus, the RWST minimum boron concentration limit is to be increased from 2000 ppm to 2175 ppm and the CLA minimum boron concentration limit is to be increased from 1900 ppm to 2000 ppm. In addition, the RWST and CLA maximum boron concentration limits are to be increased from 2100 ppm to 2275 ppm to preserve operating margin. Using the proposed RWST and CLA maximum boron concentration limits, the boron precipitation analysis is reanalyzed to ensure hot leg recirculation is initiated such that post-LOCA boron precipitation in the core is avoided. The hot leg recirculation initiation time is reduced from 9 hours to 7 hours in order.to support the increased maximum boron concentration limits. The allowable range for post-LOCA sump pH is given in the Technical Specification Bases for the RWST boron concentration limits. Given the , proposed changes in CLA and RWST boron concentration limits, the allowable range for post-LOCA containment sump pH is revised from B.5 to 10.5 to an allowable pH range of 7.5 to 10.5. This change is necessary since the acidity'of the boric acid solutions contained in the CLAs and the RWST is being increased. The MNS Technical-Specification and FSAR sections affected by these proposed changes are' provided in'Section 8 of this document. 7-1
- 8. PROPOSED MODIFICATIONS TO' LICENSING BASIS DOCUMENTS Revisions to the Technical Specifications have been proposed for Cycle 9 operation to accomodate the influence of the Cycle 9 core design on the post-LOCA-suberiticality and boron dilution ~ analyses. Table 8 lists the Technical. Specification' changes required for Cycle 9 '
operation. Revisions to the Core Operating Limits Report (COLR) are limited to numerical values and do not involve any changes to the list-of parameters reported. Also included in this section are the McGuire > FSAR changes that are_a result of the post-LOCA subcriticality and the baron precipitation reanalyses. , i a f f i a j
'1 i
8-1 ! I
'I
l Table 8-1 -Technical Specification Changes Snecification 'Descriotion of Chance 3.1.2.5 Changed minimum RWST boron concentration limit. Changed the minimum RWST boron concentration and the post-LOCA containment sump pH discussed in the Bases. 3.1.2.6 Changed the minimum and maximum RWST' boron concentration limits. .; 3.5.1.1 Changed minimum and maximum CLA boron concentration limits. Changed volume weighted average boron concentration required for Action c and discussed in the Bases. 3.5.5 Changed minimum and maximum RWST > boron concentration limits.
- Chan ed the allowable range on pest-LOCA containment sump pH ,
didcussed in the Bases. 3.9.1 Changed the minimum boron concentration required for refueling. This change is also made in the Bases. 3/4.9.12 Changed the minimum spent fuel pool boron concentration limit. ' This change is also made in the , Bases. k r F
.6 8-2 .,
- i. ,
s 8.1 Changes to Technical Specifications . t F I 1 I F F e t 9 k k t i
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' REACTIVITY CONTROL SYSTEMS N M[ 2 80 RATED WATER SOURCE - SHUTDOWN ONL/ '
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LIMITING CONDITION FOR OPERATION
- s. ,
3.1.2.5. As a minimum, and of the following borated water sources shall'be OPERABLE: a. A Boric with: System Acid Storage Syctam and at'least one associated Heat Tracing '
- 1) A minimt.m contained borated water volume of 6132 gallons, l .i
- 2) Between 7000 and 7700 ppm of boron, and '
- 3) A minimum solution. temperature of 65*F.
- b. The refueling water storage tank with:
- 1) A minimum contained borated water volume of 26,000 gallons, 2.17 5
- 2) A minimum baron concentration of-2000 ppm, and
. 3) A minimum solution temperature of 70*F. -
APPLICABILITY: MODES 5 and 6. ACTION: ' With no borated' water source OPERABLE, suspend all operations involving CORE- l ALTERATIONS or positive reactivity changes.- '_ SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall' be demonstrated OPERABLE:
- a. At least once per 7 days by: .
- 1) Verifying the boron concentration of the water, 4 2)
~
Verifying'the contained borated water volume, and -
- 3) Verifying the boric acid storage -tank solution temperature when it is the source of borated water. j
- b. At least'once per 24 hours by verifying-the RWST. temperature when it'- ~'
is the source of' borated water and the outside air temperature is
'less than 70*F. ,
McGUIRE - UNITS 1 and 2- 3/4 1 ' Amendment No. 80 (dnit 1) y Amendment No. 61 (Unit 2) 4 w - .n --
,- e- w
, u NL 5" .
REACTIVITY CONTROL SYSTEMS (/( h(( 2. f . , .,,, BORATED WATER SOURCES - OPERATING ONL)' LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2:
- a. A Boric Acid Storage System and at least one associated Heat Tracing System with:
- 1) A minimum contained borated water volume of 20,453 gallons, {
- 2) Between 7000 and 7700 ppm of boron, and
- 3) A minimum solution' temperature of 65 F.
- b. The refueling water storage tank with:
- 1) A contained barated water volume of at least 372,100 gallons, 2175 2/Z.75
- 2) Between -fMHW and eMO ppm of boron,
- 3) A minimum solution temperature of 70*F, and
- 4) A maximum solution temperature of 100 F.
APPLICABILITY: MODES 1, 2, 3 and 4. ACT:0N:
- a. With the Boric Acid Storage System inoperable .and being. used as one of the above required borated water sources, restore the storage system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUT 00'/H MARGIN equivalent to at least 1% delta k/k at 200*F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in.
COLD SHUTDOWN within the next 30 hours.
- b. With the- refueling water storage tank inoperable, restore the tank to OPERABLE status within 1 hour or be in at.least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the f.ollowing 30 hours. -
l McGUIRE - UNITS 1 and 2 3/4 1-12 Amendment No.80-(Unit 1) Amendment No.61 (Unit 2) 8-5 1
l hl45f 6 , REACTIVITY CONTROL SYSTEMS hfl ( BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION
;3.1.2.6 As a minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2:
- a. A Baric Acid Storage System and at least one associated Heat Tracing System with:
- 1) A minimum contained borated wate,r volume of 20,453 gallons, {
- 2) Between 7000 and 7700 ppm of boron, and
- 3) A minimum solution' temperature of 65*F.
- b. The refueling water storage tank with:
~
- 1) A contained borated water volume of at least 372,100 gallons,
- 2) Between 2000 and ppm of boron,
- 3) A minimum solution temperature of.70*F, and 4)- A maximum solution temperature of 100*F.
APPLICABILITY; MODES 1, 2, 3 and 4. ACTION:
- a. With the Boric Acid Storage System inoperable and being. used as one of the above required borated water sources. restore the storage system to OPERABLE'~ status within 72 hours or be in at'least HOT STANOBY within the next 6 hours'and borated to'a. SHUTDOWN MARGIN equivalent to at least 1% delta k/k at 200*F; ~ restore the Boric Acid - y Storage System to OPERABLE status,within the next 7 days _or_be'in. .
COLD SHUT 00WN within the next 30 hours. ' i
- b. With the refueling water stdrage tank inoperable, restore the tank i to OPERABLE status within 1 hour or be in at least HOT STANOBY within the next 6 hours and in-COLD SHUTDOWN within the f.ollowing. )
30 hours. - 1 i 4
. McGUIRE:- UNITS 1 and 2- 3/4 1-12 Amendment'Ho.80 (Unit 1) 8-6 Amendment No.61 (Unit 2)
-l
=
MC d uif C ' 3/4.5 EMERGENCY CORE COOLING SYSTEMS ~ (/ pted 1 3/4.5.1 ACCUMULATORS , COLO LEG INJECTION LIMITING CONDITION FOR OPERATION 3.5.1.1 Each-cold leg injection accumulator shall be OPERABLE with:
- a. The isolation valve open,
- b. A contained borated water volume of between 6870 and 7342 gallons, >
- c. A boron concentration of between and ppm,
- d. A nitrogen cover pressure of between 585 and 639 psig, and .
- e. A water level and pressure channel OPERABLE.
APPLICABILITY: MODES 1, 2, and 3*. ACTION:
- a. With one accumulator inoperable, except as a result of a closed isolation valve or baron ccncentration less than -MOO ppm, restore the inoperable accumulator to OPERABLE status within 1 hour or be in at least HOT STANOBY within the next 6 hours and reduce Reactor Coolant! System pressure to less than 1000 psig within the following 6 hours.
- b. With one accumulator inoperable due to the isolation valve being closed, either immediately .open the isolation valve or be in at least HOT STANOBY within 6 hours and reduce Reactor Coolant System pressure l to less than 1000 psig within the following 6 hours.
- c. With one accumulator inoperable due to boron concentration less than
-MO& ppm and:
2.000
- 1) ThevkuYeweightedaverageboronconcentrationoftheaccumula- -
tors -Mee ppm or greater, restore the inoperable accumulator to OPERABLE status within 24 hours of the low boron determination or be in at least HOT STANOBY within the next 6 hours and reduce Reactor Coolant System pressure to less than 1000 psig within - l-the following 6 hours. 2 coo 1900. '
- 2) The volume weighted average Doron concentration of the accumula-tors less than M00 ppm but greater than -MOG ppm, restore the l inoperable accumulator to OPERABLE status or return the volume weighted average boron concentration of the three limiting ac-cumulators to_ greater than/t900 ppm and enter ACTION c.1 within 2000 6 hours of the low boron determination or be in HOT STANDBY within the next 6 hours and reduce Reactor Coolant System l pressure to less than 1000 psig'within the following 6 hours.
- Reactor Coolant System pressure above 1000 psig. _ {
McGUIRE - UNITS 1 AND 2 3/4 5-1 Amendment No.128(Unit 1) Amendment No.110(Unit 2) 87
ver a v v EMERGENCY CORE COOLING SYSTEMS M c Guire Unik 2. ONL)! ; LIMITING CONDITION FOR OPERATION (Continued) 1900 3) The volume weighted average boron concentration of the accumula- i tors -1900 ppm or less, return the volume weighted average baron } concentration of the three limiting accumulator to greater than 1900 4600 ppm and enter ACTION c.2 within 1 hour of the low boron j determination or be in HOT STAND 8Y within the next 6 hours and reduce Reactor Coolant System pressure to less than 1000 psig within the following 6 hours. l SURVEILLANCE REQUIREMENTS 4.5.1.1.1 Each cold leg injection accumulator shall be demonstrated OPERABLE:
- a. At least once per 12 hours by:
- 1) Verifying the contained borated water volume and nitrogen .
cover pressure in the tanks, and
- 2) Verifying that each cold leg injection accumulator isolation valve is open.
- b. At least once per 31 days and within 6 hours after each solution volume increase of greater than or equal to 1% of tank volume not resulting from normal makeup by verifying the boron concentration of the accumulator solution;
(
- c. At least once per 31 days when the RCS pressure is above 2000 psig by verifying that power to the isolation valve operator is disconnected; and
- d. At least once per 18 months by verifying proper operation of the power disconnect circuit.
4.5.1.1.2 Each cold leg injection accumulator water level and pressure channel shall be demonstrated OPERABLE:
- a. At least'once per 31 days by the performance of an ANALOG CHANNEL OPERATIONAL TEST, and
- b. At least once per 18 months by the performance of a CHANNEL CALIBRATION.
( McGUIRE - UNITS 1 AND 2 3/4 5-2 - Amendment No.128 (Unit 1) ~ Amendment No.110 (Unit 2) 88
M L drtA 6'rt EMERGENCY' CORE COOLING SYSTEMS MN1 . 3/4.5.5 REFUELING WATER STORAGE TANK Qf/kf ') LIMITING CONDITION FOR OPERATION * , 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE'with: ;
- a. A contained borated water volume of at least 372,100 gallons, it75 u75 !
- b. A boron concentration of between-0000- and eMe ppm of boron, l
- c. A minimum solution temperature of 70*F, and !
- d. A maximum solution temperature of 100*F. !
APPLICABILITY: MODES 1, 2, 3, and 4. .! ACTION: l With the RWST inoperable, restore the tank to OPERABLE status within.1 hour or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
. 1) Verifying the contained borated water volume in the tank, and
- 2) Verifying the boron concentration of the water, b, At least once per 24 hours by verifying the RWST temperature when the outside air temperature. is either less than 70*F or greater than 100*F.
a r \ McGUIRE - UNITS 1 and 2 3/4 5-12
b 6 bit C, 3/4.5 EMERGENCY CORE COOLING SYSTEMS NhN l. ~ 3/4.5.1 ACCUMULATORS COLD LEG INJECTION oML'/ LIMITING CONDITION FOR OPERATION 3.5.1.1 Each cold leg injection accumulator shall be OPERABLE with;
- a. The isolation valve open,
- b. A contained borated water volume of between 6870 and 7342 gallons, 2.275
- c. A boron concentration of between 1900 and-eMO. ppm,
- d. A nitrogen cover pressure of between 585 and 639 psig, and
- e. A water level and pressure channel OPERABLE.
APPLICABILITY: MODES 1, 2, and 3". ACTION:
- a. With one accumulator inoperable, except as a result of a closed isolation valve or boron concentration less than 1900 ppm, restore the inoperable accumulator to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and reduce Reactor Coolant System pressure to less than 1000 psig within the following 6 hours,
- b. With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANOBY within 6 hours and reduce Reactor Coolant System pressure l to less than 1000 psig within the following 6 hours,
- c. With one accumulator inoperable due to boron concentration less than 1900 ppm and:
- 1) The volume weighted average boron concentration of the accumula-tors 1900 ppm or greater,. restore the. inoperable accumulator to OPERABLE status within 24 hours of the low boron determination or be in at least HOT STANDBY within the next 6 hours and reduce Reactor Coolant System pressure to less than 1000 psig within l the followir.g 6 hours.
- 2) The volume weighted average boron concentration of the accumula-tors less than 1900 ppm but greater than 1800 ppm, restore the l inoperable accumulator to OPERABLE status or return the volume weighted average boron concentration of the three limiting ac- ;
cumulators.to greater than 1900 ppm and enter ACTION c.1 within j 6 hours of the low boron determination or be in HOT STANDBY l within the next 6 hours and reduce Reactor Coolant System l ; pressure to less than 1000 psig within the following 6 hours, j
- Reactor Coolant System pressure above 1000 psig. _.
k l
~
McGUIRE - UNITS 1 AND 2 3/4 5-1 Amendment No. 128(Unit 1) Amendment No. 110(Unit 2) Sto
M(,,b-uifC . . EMERGENCY CORE COOLING SYSTEMS' L( es(4- I - () ff [;)/ 3/4.5.5 REFUELING WATER STORAGE TANK i {
- LIMITING CONDITION FOR OPERATION *:
+
9 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with: ' i . a. A contained borated water volume of at least 372,100 gallons,
- 22. r 5
- b. A boron co'ncentration of between 2000 and ete4r ppm of boron,
- c. A minimum solution temperature of 70*F, and I
- d. -A maximum solution' temperature of 100*F.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- With the RWST inoperable, restore the tank to OPERABLE status within 1 hour or be in at-least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS
- 4.5.5 The-RWST shall be demonstrated OPERABLE:
- a. At least once per 7 days by: q
. 1) Verifying the contained borated water volume in the tank, and [
- 2) Verifying the boron concentration of the water.
i
- b. At least once per 24 hours-by verifying the RWST temperature when i
~;
the outside air temperat,ure is either less than 70*F or greater than 100*F. 1 b
, -i: i
. 's - :j
. . .r McGUIRE . UNITS 1 and 2 -3/4.5-12 !
11 .
,. . , y
c . M(_ h(Air 6 3/4.9 REFUELING OPERATIONS b' 1 3/4.9.1' BORON CONCENTRATION bbb LIMITING CONDITION F0k OPERATION 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure met: that the more restrictive of the following reactivity conditions is
- a. Either a K,77 of 0.95 or less, or
- b. 2. f 75 A boron concentration of greater than or equal to-2000 ppm.
APPLICABILITY: MODE 6*, fully tensioned or with thewithheadthe reactor vessel head closure bolts less than removed. ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive react'vity i changes and initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its equivalent isuntil concentration r$ K ((ored to greater than or equal to-0000
, whichever is ppmis reduce the more restrictive. 2r7S SURVEILLANCE REQUIREMENTS
- 4. 9.1.1 The more restrictive of the above two reactivity conditions shall be datermined prior to:
- a. Removing or unbolting the reactor vessel head, and
- b. Withdrawal of any full length control rod in excess of 3 feet from its fully inserted position within the reactor vessel.
4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling '72 hours. canal shall be determined by chemical analysis at least once per 4.9.1.3 NV-250 shall be verified closed under administrative control at least once per 72 hours; or, NV-131, NV-140, NV-176, NV-468, NV-808, and-either NV-132 or NV-1026 shall be verified closed under administrative control at least once per 12 hours when necessary to makeup to the RWST during refueling operations.
"The reactor shall be maintained in MODE 6 whenever fuel-is in the reactor vessel with the vessel head closure bolts less than fully tension'ed or with the head removed.-
McGUIRE - UNITS 1 and 2 3/4 9-1 Amendment No.106(Unit 1) 8 12 Amendment No. 88 (Unit 2).
MC uire REFUELING OPERATIONS i 2 3/4.9.12 FUEL STORAGE - SPENT FUEL. STORAGE POOL bb , LIMITING C0tlDITIO'1,, FOR OPERATION 3.9.12 Fuel is to be stored in the spent storage pool with: -
- a. The baron concentration in the spent fuel pool maintained at greater than or equal to-6000-ppm; and 2.17 S b.
Storage in Region 2 restricted to irradiated fuel which has decayed at least 16 days and one of the following:
- 1) fuel which has been qualified in accordance with Table 3.9-1; or
- 2) Fuel which has been qualified by means of an analysis using NRC approved methodology to assure with a 95 percent probability at a 95 percent confidence level that k is no greater than 0.95 Scluding all uncertainties; or eff
- 3) Unqualified fuel stored in a checkerboard configuration. In the event checkerboard storage is used, one row between normal storage locations and checkerboard storage locations will be vacant.
APPLICABILITY: [A During storage of fuel in the spent fuel pool. ACTION: ,
- a. Suspend all actions involving the movement of fuel in the spent fuel pool if it is determined a fuel assembly has been placed in the incorrect Region until such time as the correct storage location is determined. Move the assembly to its correct location before resumption of any other fuel movement.
- b. Suspend all actions involving the movement of fuel in the spent fuel pool if it is determined the~ pool boron concentration is less than ppm, until such time as the boron concentration is increased to ppm or greater.
St.17 G
- c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.12a. Verify all fuel assemblies to be placed in Region 2 of the spent fuel pool are within the enrichment and burnup limits of Table 3.9-1 or that keff < 0.95 by checking the assemblies' design and burnup documentation or the assemblies' qualifying analysis documentation respectively. d
- b. Verify at least once per 31 days That the spent fuel pool boron' con-centration is greater than-9660 ppm. ,
2.l*76 McGUIRE - UNITS 1 and 2 3/4 9-16 Amendment No. 68 (Unit 2) 8-13 Amendment No. 87 (Unit 1)
~
g Mc(nice ' s 2. REACTIVITY CONTROL SYSTEMS BASES (' MODERATOR TEMPERATURE COEFFICIENT (Continued) and near the end of the fuel cycle are adequate to confirm th to the reduction in RCf boron concentration associated w 3/4.1.1.4 MINIMUM TEMPERATURI FOR CRITICALITY with the Reactor Coolant System average temperature limitation is required to ensure: fhis less tha - is within it analyzed temperature range, (2) the trip instrumentation is(1) th within its temperature. minimum RT normal operating range, (3) the pressurizer is cap NDT 3/4 '.2 B0 RATION SYSTEMS available during each mode of facility operation.The Boron Injection Sys The com (A perform this function include: (1) borated water sources,ponents required to(2) chargin (3) separate ~ flow paths, (4) boric acid transfer pumps, (5) associated Heat Tracing. Systems, and (6) an emergency power supply from OPERABLE diesel generatbrs. ih the RCS average temperature above 200 F, a minimum of two boron the event an assumed failure renders one of the The flow paths i boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN and cooldown from expected to 200 F. operating conditions of 1.3% delta k/k after. xenon decay { The maximum expected boration capability requirement occurs at E0L from full power equilibrium xenon conditions and requires 16,321 gallons of 7000 ppm borated water from the boric acid storage tanks or 75,000 (RWST). gallons of 2000-ppm borated water from the refueling water storage tank
~4175 With the RCS temperature below 200 F, one Baron Injection System is acceptable without single failure consideration on the basis of the stable CORE ALTERATIONS and positive reactivity changes in the e Injection System becomes inoperable.
The limitation for a maximum of one centrifugal charging pump to be ' OPERABLE and the Surveillance Requirement to verify all charging pumps except ' the required OPERABLE pump to be inoperable below 300*F provides-assurance that aPORV. single mass addition pressure ' transient can'be relieved by the operation of a (~ McGUIRE - UNITS 1 and 2 B 3/4 1-2 Amendment No. 42 (Unit 1) Amendment No. 23 (Unit 2) 8-14
M C htAire. m it I i REACTIVITY CONTROL SYSTEMS ONQ ,., BASES ; BORATION SYSTEMS (Continued) The boron capability. required below 200 F is sufficient to provide a SHUT 140 F. 00WN MARGIN of 1% delta k/k after xenon decay and cooldown from 200 F to This condition requires either 2000 gallons of 7000 ppm borated water from the boric acid storage tanks or 10,000 gallons of 2000 ppm borated water from the refueling water storage tank. The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics. The limits on contained water lume and boron concentration of the RWST also ensure a pH value of between 0.5 and 10.5 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and m:chanical minimizes systems the effect and components. of chloride and caustic stress corrosion on The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6. 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) accident analyses thelimited. ar.e potential effects of rod misalignment on associated OPERABILITY of the control rod position indicatdrs is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. The control rod insertion limit and shutdown rod insertion limits are specified in the CORE OPERATING LIMITS REPORT per specification 6.9.1.9. The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of- peaking factors and 'a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation. In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the r'esults remain valid during future operation.
- The maximum *od drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with T greater than or equalto551*Fandwithallreactorcoolantpumpsoperatin6VEnsuresthatthe-m asured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.
Control rod positions and OPERABILITY of the rod position indicators are r: quired to be verified on a nominal basis of once per 12 hours with more frequent inoperable. verifications required if an automatic monitoring channel is applicable LCO'sThese areverification satisfied. frequencies are adequate for assuring that the - McGUIRE - UNITS 1 and 2 0 3/4 1-3 Amendment No.105 (Unit 1) Amendment No. 87 8-15 (Unit 2)
b C ht4 i rd. h&2 REACTIVITY CONTROL SYSTEMS Ob BASES B0 RATION SYSTEMS (Continued) The boron capability required below 200 F is sufficient to provide a SHUT 140 F. 00WN MARGIN of 1% delta k/k after xenon decay and cooldown from 200*F to This condition requires either 2000 gallons of 7000 ppm borated water from the boric acid storage tanks or 10,000 gallons of 2000 ppm borated water from the refueling water storage tank. 2175 The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics. 75 The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 0-5 and 10.5 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and m:chanical minimizes systems the effect and components. of chloride and caustic stress corrosion on The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6. 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is caintained, and (3)ar.e accident analyses thelimited. potential effects of rod misalignment on associated OPERABILITY of the control rod position indicatdrs is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. The control rod insertion limit and shutdown rod insertion limits are specified in the CORE OPERATING LIMITS REPORT per specification 6.9.1.9. The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER.' These restrictions provide assurance of fuel rod integrity during continued operation. In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation. The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with T greater than or. equal to 551 F and with all reactor coolant pumps operatint)V8nsures that.the 3 =asured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions. Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours with more frequent inoperable. verifications required if an automatic monitoring channel is applicable LCO's Theseareverification satisfied. frequencies are adequate for_ assuring that the - McGUIRE - UNITS 1 and 2 B 3/4 1-3 Amendment No.105 (Unit 1) Amendment No. 87 (Unit 2) 8-16
N f. b*tM f C 3/4.5 EMERGENCY CORE COOLING SYSTEMS ON LY BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) Cold Leg Accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pres-sure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe rup-tures. The limits on accumulator volume, bo'ron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met. The allowed down time for the accumulators are variable based upon boron concentration to ensure that the reactor is shutdown following a LOCA and that any problems. are corrected in a tiraely manner. Subcriticality is assured when g4 baron concentration is above 4600 ppm, so additional down time is allowed when ( concentration is above p 09 ppm. A concentration of less than 4909: ppm in any _ single accumular.or or as a volume weighted average may be indicative of a pro g blem, such as valve leakage, but since reactor shutdown is assured, additional time is allowed to restore boron concentration in the accumulators.
) The accumulator power operated isolation valves are considered to be
" operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met. In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required.
The limits for operation with an accumulator inopr able for any ' reason ex-cept an isolation valve closed minimizes the time expo mre of the plant to a LOCA event occurring concurrent with failure of an M P tional accumulator which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the rccctor'in a mode where this capability is not required. .
. The original licensing bases of McGuire assumes both the UHI system and f the Cold Leg Accumulators function to mitigate postulated' accidents. Subse- .
quent analyses, documented in "McGuire Nuclear Station, Safety Analysis for UHI Elimination" dated September 1985, and docketed by Duke letter dated October 2, 1985, support the determination that UHI is no longer required pro-vided the Cold Leg Accumulator volume is adjusted to be consistent with that ; assumed in the Safety Analysis. McGUIRE'- UNITS 1 and 2 B 3/4 5-1 Amendment No. 82-(Unit 1) 8-17 Amendment No. 63 (Unit 2)
N t btAirL L;4- 1 3/4.5 EMERGENCY CORE COOLING SYSTEMS * , ONLY
/
, BASES
~
3/4.5.1 ACCUMULAT0RS The OPERABILITY of each Reactor Coolant System (RCS) Cold Leg Accumulator ensures that a sufficient volume of borated water will be immediately. forced into the reactor core through each of the cold legs in the event the RCS pres- i sure falls below the pressure of the accumulators. This initial surge of water , into the core provides the initial cooling mechanism during large RCS pipe rup- ! tures. l The limits on accumulator volume, bo'ron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are i met. The allowed down time for the accumulators are variable based upon boron : concentration to ensure that'the reactor is shutdown following a LOCA and that any problems are corrected in a timely manner. Subcriticality is assured when ($k boron concentration is abovt4509 ppm, so additional down time is allowed when L concentration is aboveg+6ee hpm. A concentration of less than 1900 ppm in any single accumulator or as a volume weighted average may be indicative of a pro- i blem, such as valve leakage, but since reactor shutdown is assured, additional i time is allowed to restore boron concentration in the accumulators. !
) The accumulator power operated isolation valves are considered to be
" operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met. In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required.
The limits for operation with an accumulator inoperable for any reason ex-cept an isolation. valve closed minimizes the time exposure of the plant to a LOCA event occurrino concurrent with failure of an additional accumulator which
. may result in unaccentable peak cladding temperatures. If a closed isolation valve cannot be immediately opcned, the full capability of one accumulator is not available.and prompt action is required to place the reactor in a mode where this c@ ability is not required. .
The original licensing bases of McGuire assumes both the UHI system and the Cold leg Accumulators function to mitigate postulated accidents. Subse-' . quent analyses, documented in "McGuire Nuclear Station, Safety Analysis for UHI Elimination" dated September 1985, and docketed by Duke letter dated . October 2, 1985, support the determination that UHI is no longer required pro-vided the Cold Leg Accumulator volume is adjusted to _ be consistent with that assumed in-the Safety Analysis. l , : McGUIRE - UNITS'1 and 2 B 3/4 5-1 Amendment No. 82-(Unit 1)- , 8-18 Amendment No. 63 (Unit 2)
Fo r Mo ony EMERGENCY CORE COOLING SYSTEMS / BASES 3/4.5.2 and 3/4.5 3 ECCS SUBSYSTEMS The OPERABI'LITY of two independent ECCS subsystems. ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration. Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling.to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long-term core cooling capability in the - recirculation mode during the accident recovery period. With the RCS temperature below 350 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable. reactivity condition of the reactor and the limited core cooling requirements. The limitation for a maximum of one centrifugal charging pump and one Safety Injection pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and Safety Injection pumps except the required OPERABLE charging pump to be inoper Q1e below 300 F provides assurance that a mass addition pressure transient can be relieved by the operation of a. single
- PORV. .
The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for throttle valve position stops and flow balance testing , provide assurance that proper ECCS flows will be maintained in th'e event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow
'from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in ,
accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide j an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses. 3/4.5.4 [ Deleted] 4 3/4.5.5 REFUELING WATER STORAGE TANK The OPERABILITY of the refueling water storage tank (RWST) as part of the ECCS' ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWST minimum volume and baron concentration ensure that: (1) sufficient water is available within containment to permit recirculation cooling flow to the core, and , (2) the reactor will remain subcritical in the cold condition following mixing )> of the RWST and the RCS water volumes with all control rods inserted except McGUIRE - UNITS 1 and 2 B 3/4 5-2 8-19 Amendment No.82 (Unit 1) { Amendment No.63 (Unit 2) l
.- - _ ~ . _ _ _ _ . _ __ . _ - -. _
' . N ( h Y i Y f.,,
- EMERGENCY CORE COOLING SYSTEMS Sifs .
l4-2. l
.I' -
- BASES .
, REFUELING WATER STORAGE TANK (Continued) ,
for the most-reactive control assembly. These assumptions-are consistent with the LOCA analyses. , . - '
.The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical !
characteristics. - 75
~The limits on contained water volume and boron concentration of the RWST-also ensure a pH value of. between Oh and 10.5 for the solution recirculated 1
within containment after a LOCA. This pH band minimizes the evolution of
-iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. .
1 . i
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j 8 3/4 5-3
= McGUIRE - UNITS.1:and 2 . Amendment No.~ 82 (Unit: 1)'-
l l~ 8-20 ' Amendment' No. 63 (Unit .2); ; i e r -
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%+2 3/4.9 REFUELING OPERATIONS gg[j/
BASES
==-
3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that: (1) the reactor will remain subcritical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent withanalyses. in the accident the initial conditions assumed for the boron dilution incident The value of 0.95 or less for K,ff includes a 1% delta k/k conservative allowance for uncertainties. Similarly, the boron concentration value of-2000 ppm or greater includes a conservative uncertainty allowance of 50 ppm boron. 2,gyg The Reactor Makeup Water Supply to the Chemical and Volume Control (N'V) System is normally isolated duririg refueling to prevent diluting the Reactor Coolant System boron concentration. Isolation is normally accomplished by closing valve NV-250. However, isolation may be accomplished by closing valves NV-131, NV-140, NV-176, NV-468, NV-808, and either NV-132 or NV-1026, when it is necessary to makeup water to the Refueling Water Storage Tank during l refueling operations. ! 3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core. 3/4.9.3 DECAY TTME
+
The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products. This decay time.is consistent with the assumptions used in the accident analyses. 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration closure and OPERABILITY of the Reactor Building Containment Purge Exhaust System HEPA filters and charcoal adsorbers ensure that a release of radioactive material within con-tainment will be restricted from leakage to the environment or filtered through ths HEPA filters and charcoal adsorbers prior to discharge.to the atmosphere. The OPERABILITY and closure restrictions are sufficient to restrict radio-active material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE. Operation of the Reactor Building Containment Purge Exhaust System HEPA filters and charcoal adsorbers and the resulting iodine removal capacity are consistent-with the assumptions of the accident analysis. The methyl iodide penetration test criteria for the carbon samples have been made.more restictive than required for the assumed iodine removal in the accident analysis because the humidity to be seen by the charcoal adsorbers'may be graater than 70% under normal operating conditions,. McGUIRE - UNITS 1 and 2 8 3/4 9-1 Amendment No.106 (Unit 1) Amendment No.88 (Unit 2)
f NC 6Ldre k unN 2. BASES _
%~
3/4.9.9 and 3/4.9.10 WATER LEVEL - REACTOR VESSEL and STORAGE POOL The restrictions on minimum water level ensure that sufficient water d:pth is available to remove 99% of the assumed 10% iodine gap activity raleased from the rupture of an irradiated fuel assembly. The minimum water d:pth is consistent with the assumptions of the accident analysis. 3/4.9.11 FUEL HANDLING VENTILATION EXHAUST SYSTEM The limitations on the Fuel Handling Ventilation Exhaust System ensure that all radioactive material released from an irradiated fuel assembly will b3 filtered through the HEPA filters and charcoal adsorbers 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. ANSI H510-1975 will be used as a procedural-guide for surveillance testing. The tethyl iodide penetration test criteria for the carbon samples have been mads more restrictive than required for the assumed iodine removal in the cccident analysis because the humidity to be seen by the charcoal adsorbers try be greater than 70% under normal operating conditions. 3/4.9.12 FUEL STORAGE - SPENT FUEL STORAGE POOL The requirements for fuel storage in the spent-fuel pool on 3.9.12 (a) and (b) ensure that: (1) the spent fuel pool will remain subcritical during fuel stor:ge; and (2) a uniform boron concentration is maintained in the water volume.in the spent fuel pool for reactivity control. The value of 0.95 or less for K6ff which includes all uncertainties at the 95/95 probability / ccnfidence level as described in Section 9.1.2.3.1 of the FSAR is the accept-Enca criteria for fuel storage in the spent fuel pool. Table 3.9-1 is con-ssrvatively developed in accordance with the acceptance criteria and methodology referenced in Section 5.6 of the Technical Specifications. Storage in a checker-board configuration in Region 2 meets all the acceptance criteria referenced in S ction 5.6 of the Technical Specifications and is verified in a semi-annual b: sis after initial verification through administrative controls. Ths Action Statement applicable to fuel storage in the spent fuel pool ~ ensures that: (1) the spent fuel pool is protected from d stortion in the fuel storage ptttern that could result in a critical array during the movement of fuel; and (2) the boron concentration is maintained at-Beee ppm during all actions involv-ing movement of fuel in the spent fuel pool. 2.l75 The Surveillance Requirements applicable to fuel storage in the spent fuel pool ensure that: (1) fuel stored in Region 2 meets the enrichment and burnup limits of Table 3.9-1 or the Keff < 0.95 acceptance criteria of an analysis using NRC approved methodology; and (2) the boron concentration meets the 1900 ppm limit. ROS , McGUIRE - UNITS 1 and 2 B 3/4 9-3 Amendment No.69(Unit 1) g Amendment No.50(Unit 2)
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e r 8.2 Changes to the Final Safety Analysis Report
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McGuire Nucicar Station '6.3 Emergency Core Cooling System - The accumulator relief valves am sized to pass nitrogen gas at a rate in excess of the accumulator gas fill
,line delivery rate. The relief valves also pass water in excess of the expected accumulator in-leakage rate, ' ' but this is not considered to be necessary, because the time required to fill the gas space gives the operator ample opportunity'to correct the situation. Other relief valves are installed in various sections of the Emergency Core Cooling System to protect lines which have a lower design pressure than the Reactor Coolant System. The valves are of the enclosed bonnet type which provides an additional barrier for isolating system fluids.
Butterfly Valves Each main residual heat removal line has an air-operated butter 0y valve which is nonnally open and is designed to fail in the open position. These valves are left in the full open position during normal operation to manmin flow from this system to the Reactor Coolant system during the injection mode of the Emergency Core Cooling System operation. ' Piping ' The piping layouts for the ECCS are shown in Figure 6182. All piping joints are welded except for the pump and butterfly valve flange connections. Weld connections for pipes sized 2-1/2 inches and larger are butt welded. Reducing tees are used where the branch size exceeds one half of the header size. Branch connections of sizes that are equal to or less than one half of the header size conform to the ANSI code. Branch connections 1/2 inch through 2 inches are attached to the header by means of full penetration welds, using pre-engineered integrally reinforced branch connections. Minimum piping and fitting wall thicknesses as determined by ANSI B31.1.0-1967 edition formula are increased to account for the manufacturer's pennissible tolerance of minus 12-1/2 percent on the nominal wcll and an appropriate allowance for wall thinning on the extemal radius during any pipe bending operations in the shop fabrication of the subassemblies. , he ECCS water is borated to a nominal concentration of 9:00& ppm. Hence, no boron crystallimtion is , expected to occur so long as the water is maintained at a temperature ahve freezmg. - 6.3.2.2.2 System Operadon The operation of the Emergency Core Cooling System, following a loss of coolant accident, can be divided into two distinct modes: -
- 1. The injection mode in which any reactivity increase following the postulated accidents is tenninated, initial cooling of the core is accomplished, and coolant lost from the primary system in the case of a loss of coolant accident is replenished, and
- 2. He recirculation mode in which long term core cooling is provided during the accident recovery period.
Discussion of these modes follows.
; Break Socetrum Coverage _ I 1
The principal mechanical components of the Emergency Core Cooling System which provide core cooling immediately following e loss of coolant accident are the accumulators, the safety injection pumps, the'
- centrifugal charging pumps, the residual heat removal. pumps, refueling water storage tank, and the ossociated valves, and piping.
1 i., i (ol MAY 1992). 6-145 ! 8 24 l l
.i
En Id o[l y t McGuire Nucicar Statio 6.3 Emergency Core Cooling System j i Components such as remote motor operated valves and flow and pressure transmitters have been shown ,
. capable of operating for the required post-accident periods, when exposed to post-loss-of-coolant l environmental conditions. All other Emergency Core Cooling System components are located outside of t the Containment. !
l The speciEcation of individual parameters as given in Table 6123 and Table 6-124 includes due l consideration of allowances for margins over and above the required performance value (e.g., pump flow ; and net positive suction head), and the most severe conditions to which the component could be : subjected (e.g., pressure, temperature, and flow). f
.i This consideration ensures that the Emergency Core Cooling System is capable of meeting its minimum !
required level of functional performance. ; 6.3.3.9 Use of RHR Spray , No earlier than 50 minutes after initiation of the LOCA, the low head RHR flow is diverted from the core low head injection path to the auxiliary spray headers. For muumum safeguards, one high head safety injection pump and one centrifugal charging pump would supply the coolant to the core after realignment of a portion of the RHR pump discharge to the auxiliary spray headers. The amount of water which should be supplied to the core at a reactor coolant system pressure of 10 psig (which is approximately the peak containment pressure) is approximately 110 lbm/sec.' At 50- minutes after i 1 hypothetical LOCA the core has been quenched so that effluent carryover has been terminated. The time 1, that effluent carryover or entrainment from the core ends is conservatively assumed to occur when the core mixture height reaches the 10 foot elevation (at approximately 150 seconds). At 50 minutes, the thin ; and thick metal sensible heat has been removed and temperatures reduced to the saturating temperature I for the containment pressure. The only heat generation at saturation time is decay heat; The decay heat mass boiloff at 50 minutes, which is the muumum time specified in the operating procedures that the RHR low head flow can be diverted to the RHR spray is 61.5 lbm/sec based on the following assumptions: 1.102 percent of engineered safeguards design rating of 3579 Mwt.
- 2. ANS infinite decay heat with 20 percent margin (10CFR 50.96 Appendix K).
- 3. Coolant entering the core is subcooled by 60 Btu /lbm.
Therefore, the coolant entering the Reactor Coolant System piping is about 200 percent of that requised by the decay heat mass boiloff, calculated with conservative assumptions. It should be noted that the minimum time given above for diversion of RHR, low head flow to the containment spray syste.m is consistent with the peak containment pressure analysis presented in Section 6.2.1.1.3.1, " Loss of Coolant Accident" on page 6-9. Boron precir>itation Evaluation An analysis has been performed to determine the maximum boron concentration in' the reactor vessel following a hypothetical LOCA. This analysis used the method and assumptions described in Reference 5 on page 6-173 with the principal input parameters given in Table 6-137. The analysis considers the increase in boric acid concentration in the reactor vessel during the long term cooling phase of_ a LOCA, assuming a conservatively small effective vessel volume including only the free volumes of the reactor core - and the upper plenum below the bottom of the hot leg nozzles. This assumption conservatively neglects ') . the mixing of the boric acid solution with directly connected volumes, such as the reactor vessel lower plenum. The calculation of boric acid concentration in the reactor vessel considers a cold leg break of the
-(01 MAY 1992) 8 25 6-165
6.3 Emergency Core Cooling System McGuire Nuclear Station Reactor Coolant System in which steam is generated in' the core from decay heat while the boron cssociated with the boric acid solution is completely separated from the steam and remains in the effective vessel volume. The results of the analysis show that the maximum allowable boric acid concentration established by the NRC, which is the boric acid solubility limit minus 4 weight percent, will not be exceeded in the vessel if het leg injection is iritiated hours after the LOCA occurs. The safety injection flow to the Reactor Coo t System hot legs will exceed (assuming failure of one ECCS train) the decay heat mass boil off (g/fibm/sec) at this time. This hot leg flow will dilute the reactor vessel boron concer.tration by passing relatively dilute boron solution from the hot leg through the
- vessel to the cold leg break location. Centrifugal charging pump flow will continue to be provided to the Reactor Coolant System cold legs and will preclude any boron concentration buildup in the vessel for
, breaks in the hot leg. Potential boron precipitation di="~I in Section 6.3.2.5, "Desigu Pressures and Temperatures" on page 6-151. A single active failure analysis is presented in Table 6-133. A passive failure analysis is presented in Table 6-134. Since the ECCS is designed to meet the single failure criterion, no back up means is required to be provi'ded to prevent the buildup of boron concentration. All components of the ECCS are ANS Safety Class 2 and Seismic Category 1. ECCS testing is discussed in Section 6.3.4, " Tests and Inspections." 6.3.4 TESTS AND INSPECTIONS [ In order to demonstrate the readiness and operability of the Emergency Core Cooling System .the components are subjected to periodic tests and inspections. Performance tests of the components are performed in the manufacturer's shop. An initial pre-operational system flow test is performed to demonstrate the proper functioning of all of the components. A 1/3 scale model test is performed to determme the acceptability of the ECCS recirculation intake design. The main objective of this test is to determine if vortices or air ingestion will occur, and to alleviate the problem. This test includes all equipment, piping, cable trays, etc., which may affect the flow characteristics. The recirculation screen is tested with portions of the screen blocked to determine if any flow irregularities occur. If any problems other than vortices or air ingestion are present, this test will determine the problem and be an aid in producing a solution. Quality Control Tests and inspections are carri5 out during fabrication of each of the Emergency Core Cooling System components. These tests are conducted and documented in accordance with the Quality Assurance program discussed in Chapter 17. " Quality Assurance" on page 17-1. These tests are intended to evaluate the hydraulic and mechanical performance of the passive and active components involved in the injection mode by demonstrating that they have been installed and adjusted so they operate in accordance with the design intent. These tests are divided into several individual sections that may be performed as conditions allow without compromising the integrity of the tests.
. One of these individual sections consis(s of system actuation tests to verify the operability of all b Emergency Core Cooling System valves initiated by the safety injection "S" signal, the phase A, 6 166 (of MAY 1992) 8 26
McGuire Nuclear Station Appendix 6. Chapter 6 Tcbles and Figures Table 6-123 (Page 1 of 2). Emergency Core Cooling System Component Parameters Component Parameters Cold Len Iniection Accumulators Number 4 Design Pressure, psig 700 Design Temperature, 'F 300 Operating Temperature *F 60-150 Normal Operating Pressure, psig 600 Minimum Operating Pressure, psig 585 Total Volume, ft S 1363 each Minimum Water Volume, ft' 918 each Maximum Na Gas Volume ft2 445 each Boric Acid Concentration, Nominal, 9009-ppm UAni4 I > DII 1 TAM + 2.) 2.I'5 % Relief Valve Setpoast, psig -700 Centrifucal Charcine Pumps Number 2 Design Pressure, psig 2800 Design Temperature, 'F 300 Design Flow Rate, gpm* 150 Design Head, ft. 5800 Maximum Flow Rate, gpm 550 , Head at Maximum Flow Rate, ft. 1400
. Discharge Head at Shutoff, ft. 6000 Motor Rating, bhp + 600 Safety Iniection Pumps Number 2 Design Pressure, psig 1750 Design Pressure, *F 300 [
Design Flow Rate, gpm 400 Desi@ Head, ft. , 2600 Maximum Flow Rate, spm 650 Head at Maximum Flow Rate, ft. 1950 Discharge Pressure at Shutoff, psig 1520 Motor Rating, bhp + 400
. Besidual Heat Removal Pumps & Refer to Section 5.5.7 for parameter Heat Fxchnneen informntinn 4
(01 MAY 1992) 8-27 l
i~ McGuire Nuclear Station . Appendix 6. Chapter 6 Tables and Figures Table 6-124. Normal Operating status of Emergency Core Cooling System Components for Core Cooling Number of Safety injecdon Pumps Operable 2 Number of Chagng Pumps Operable 2 Number of Residual Heat Removal Pumps Operable 2 Number of Residual Heat Exchangers Operable 2 Refueling Water Storage Tank Volume, Gal., mmimum 372,100 1 Boron Concentration in Refueling Water Storage Tanks, minimum ppm l/nif l) 2,000 Boron Concentration in Cold leg Accumulator, minimum ppm ( Mm4 O I,$o , Number of Accumulators 4 Mmimum Indicated Cold leg Accumulator Pressure, psig 5,85 ; Mmimum Cold leg Accumulator Water Volume, ft* 918 System Valves, Interlocks, and Piping Required for the Above Components All which are Operable
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. (Ol' MAY 1992) I i
i 8-28 i d
McGuire Wuc! car Station Appendix 6. Chapter 6 Tables and Figures -t
- } Table 6-137.' Parameters for Boron Precipitation Analysis Reactor Core Power 3479 MWt 3 Total Inventory of Boric Acid Solution (Includes RCS, SI Accumulators, RWST 6.9 x 10' lbm 2 and Ice Bed) 2 Boron Concentration Measurement Uncertainty 1.0 %
2 Effective Vessel Volume (Core and Upper Plenum Volume to the bottom of hot 972 leg nozzles) Safety Injection Subcooling 55 BTU /lbm Containment Pressure 14.7 psia 2 Ice Condenser Maximum Boron Concentration 2330 ppm Fo r L6 o n (y T l 1 s
) w (Of MAy 3992) 8-29 l i
McGuire Nuclear Station 9.1 Fuel Storage and Ilandling 9.1.3.1.2 Water purification The system demineralizer and filters are designed to maintain adequate purification to permit u access to the spent fuel storage area for plant personnel, provide means for purifying transfer cana refueling pool water during refueling, and provide purification capability for the refueling tank, hetrough, skimmer KF System strainers, also maintains and sbmmer filters. the optical cladty of the spent fuel pool water esurface by use o The Spent Fuel Pool will be sampled on a weekly basis for the following: chlorides, fluoddes turbidity. He expected range for pH is 4.0-8.0. Chloride and fluodde' concentrations will be limited maximum at less than of 1.00.15 FTU.ppm. (Westinghouse Primary Chemistry Specifications.) Turbidity will be maintaine maintained greater than or eaual to 2,000 ppm; the expected range isBaron c 2,000-4,000 ppm Boron. (Boron concentration is set by Technical Specificanons') On a monthly basis, a GeLiisotopic analysis will be performed. No specificat' ns are set for Geli analys(*s. ((Adl- 0 2 IE PP"' b M 2) 9.1.3.1.3 Spent Fuel P I Dewater'ing Protection System piping is arranged so that failure of any pipeline cannot drain the spent fuel pool below the wat level required for radiation shielding. A water level of ten feet or more above the top of the, s fuel assemblies is maintained to limit direct gamma dose. 9.1.3.1.4 Spent Fuel Pool Makeup In order to provide specified shielding and water volumes m Ge fuel pool dudng plant operation piping provides the refueling makeup water storage capabilities. Borated makeup water can be supplied to the spent fuel pool tank. Demmeralized water can be supplied to the pool by the-Reactor Makeup Water Pumps, and emergency makeup water can be supplied to the pool from the Nuclear Service Water System. All means of makeup are manually initiated and manually terminated. 9.1.3.2 System Description An identical Spent Fuel Cooling Systern, as shown in Figure 913, is provided for each unit. The syste consists of two coohng loops, one purification loop, and one skimmer loop. The fuel pool cooling pumps take suction from the spent fuel pool. These pumps circulate the water through the cooling loops and the purification loop in various combinations prior to returning the wate to the spent fuel pool. The spent fuel pool heat load is transfermd to the Component Cooling Syste the fuel pool cooling heat exchangers. The fuel pool cooling pre-fiher, demineralizer, and post-fdter will rdequately operations. remove corrosion and fission products from the p spent fuel ~ ool water during fuel handli The fuel pool skimmer pump taken suction from the skimmer trough, that collects water from the sp!! fuel pool surface. Floating debris is removed by the fuel pool skimmer strainer and fdier. Optically cl water is then discharged below the pool surface at vadous locations. Discharge throttling valves are ! provided for optimizing the spent fuel pool skunmer loop operation. The pool cooling and purification system is manually controlled from a local control panel. High. j temperature and low liquid level in both the fuel pool and the refueling canal, plus high radiation in the i fuci pool area alanns are provided in the Control Room as per Regulatory Guide 1.13. Also alarmed in ; the Control Room are high liquid level in both the fuel pool and the refueling canal. local gages are ~ provided for high differential pressure across each strainer and filter. A (01 MAY 1992) 9-17 8-30
~ 9.2 Water Systems McGuire Nuclear Station
- 3. a sufficient volume of water in the Containment sump to permit .he initiation of recirculation.
he only portion of the Refueling Water System shown in Figure 9-65 that is safety related Seismic Category 1, is th refueling water storage tank and associated NRC Quality Class B piping that connects to the ECCS. A failure analysis of the lines from the refueling water storage tank to the safety injection, centrifugal charging, and residual heat removal pumps is provided in Chapter 6, " Engineered Safety Features" on page 6-1. The failure of non-seismic Category I equipment and piping in the Refueling Water System and interfacing systems does not affect the ability of the Refuehng Water System to perform its intended safety related function.
-Qht c.
He water in the tank is borated to a concentration which assures reactor shutdown by approximately tee percent 6k/k when all rod cluster control assemblies are inserted and the reactor is cooled down for refueling: Upon an Engineered Safety Features signal, the centrifugal charging pumps, the safety injection . pumps, and the residual heat removal pumps take suction from the tank. The Containment Spray pumps can also take suction from the tank. . Automatic heating is'provided to mamtain the tank tempu4me chove 70*F. Refueling water recirculation pumps are used to recirculate heated water from the tank through the transport line as necessary to maintain the line water temperature above 70*F. 9.2.5.4 Tests and Inspections The Refueling Water System design is verified by pre-operational testing. Testing includes verification of , system flow paths used in normal operation and a verification of electrical heater input and associated control interlocks. The instrumentation associated with the refueling water storage tank is checked pre-operationally and during each refueling. 9.2.5.5 Instrumentation Application Refer to Section 6.3.5.4, " Level Indication" on page 6-170 for RWST level instrumentation application
. during Emergency Core Cooling System operation. RWST level indications also aid the operator during the emptying and refilling of the refueling water storage tank during refueling. An overflow line to the spent fuel pool is available to handle overflow. The temperature indication is provided to monitor the taak water temperature so that action can be taken if tank water temperature starts approaching the minimum tank water temperature.
All other instrumentation is provided to aid the operation of the system in its various modes. 9,2.6 TREATED WATER SYSTEMS 9,2.6.1 Design Bases he treated water systems are designed to provide the follouing:
- 1. drmking and sanitary water requirements, 1
- 2. filtered water requirements, and j
- 3. deminerahzed water makeup requurments.
. l l
9.2.6.2 System Description l The treated water systems consist of the following:(See Figure 9-66 through Figure 9-76).
- 1. Filtered Water System,
(- 9 54 (01 M AY 1992) . ! 8-31
McGuire Nuclear Station cN ^ 15.6 Decrease in Reactor Coolant inventory 5 2 -He BEACH computer c Fo< h (o od m FRAP T6-B&W at the end of refill and
. 2 core flooding rates from RELAPS/ MOD 2-B&W to calculate fuel thermal responses from the end of refill 2 to the end of reflood. The results from the BEACH computer code are the same as the fuel thermal 2
response results by the FRAP-T6-B&W code, namely, metal water reaction, hot pin thermal response and 2 peak cladding temperatures. A more detailed description of the BEACH code is available in Reference 27 2 on page 15-115. 2 Post-LOCA Suberiticalitv Evaluation i 2 An analysis has been performed to determme the sump mixed mean boron concentration as a function of - 2 pre-trip Reactor Coolant System (RCS) boron concentration for a postulated large break loss of Coolant 2 Accident (LOCA). Water mass contributions from the RCS, Cold leg Accumulators (CLAs), Refueling 2 Water Storage Tank (FWST) Ice Condenser, and Emergency Core Cooling System (ECCS) and - 2 containment spray piping have been taken into account. This analysis used the principle input parameters 2 provided in Table 15-46. High concentration borated water volumes (e.g., FWST and CIAs) are 2 conservatively minimiwA using Technical Specification mmimum allowed values minus associated 2 measurement uncertamties. A borated water mass contribution from Ice Condenser melt has been 2 evaluated from the start of the LOCA until the initiation of sump recuculation. Potential borated water 2 holdup in upper containment from the initiation of normal containment spray was taken into account. 2 Results of the analysis are compared with the required boron concentrations necessary to keep the core
.2 subcritical, with no credit taken for control rod insertion, during the sump rectreulation mode. The 2
analysis provides an available' sump mixed mean boron concentration curve that must bound the required 2 all rods out (ARO) critical baron concentrations for caeh cycle. The required ARO critical boron ' 2 concentrations are evaluated for each core design as part of the reload safety analysis process. ]2 2 Since much of the ECCS piping is used during normal operation for residual heat removal and normal chemical and volumne control, much of this ECCS piping contains relatively low concentration borated 2 . water. A single failure of an ECCS train would therefore be non-conservative, and was thus not 2 considered in the Post-LOCA Suberiticality Analysis. Small Break LOCA Evaluation Mod.ej ne NOTRUMP computer code is used in the analysis ofloss-of-coolant accidents due to small breaks in the reactor coolant system. The NOTRUMP computer code is a state-of the-an one-dimensional general network code consisting of a number of advanced features. Among these features are the calculation of thermal non-equilibrium in all fluid volumes, flow regime-dependent drift flux calculations with counter current flooding limitations, mixture level tracking logic in multiple stacked fluid nodes, and regime-dependent heat transfer correlations. The NOTRUMP small break LOCA emergency core cooling system (ECCS) evaluation model was developed to determine the RCS response to design basis small break LOCAS and to address the NRC concems expressed in NUREG-0611. " Generic Evaluation of Feedwater Transients *and Small Break less-of-Coolant Accidents in Westinghouse Designed Operating , Plants." I l in NOTRUMP, the RCS is nodied into volumes interconnected by flowpaths. The broken loop is l modeled explicitly with the intact loops lui.'oed into a second loop. The transient behavior.of the system is determined fru,a the goveming conseivdon equations of mass, energy and momentum applied throughout the system. A detailed description of NOTRUMP is given in Reference 11 on page 15-114 and Reference 15 on page 15-114. 1 The use of NOTRUMP in the analysis involves, among other things, the representation of the reactor .) core as heated control volumes with an associated bubble rise model to permit a transient mixture height {
\
l (01 MAY 1992) 15-107
McGuire Nuclear Station Appendix 15 , Chapter 15 Tables and Figures
)
2 Table 15-46. Parameters for Post-LOCA Suberiticality Analysis 2 Volume Grouping Boron Concentration (ppm) 2 Iow Head Safety Injection (LHSI) Discharge A900"
,2 to Intennediate Head Safety Injection (IHSI) 2 aad High Head Safety injection (HHSI) suction Rfg3 2 (Valve nil 36B to Valves NI332A & NI333B) 2 Refueling Water Storage Tank (RWST) to Valve FW28 2 FWST to IHSI suedon 2 FWST to Valve NV223' 2 Normal Containment Spray Discharge 2 Containment Spray suction from RWST 2
2 LHSI Discharge to Aux. Cont. Sprary 2 LHSI Suction from Sump 3502 2 LHSI Suction from loop C Hot Leg 2 Containment Spray Suction from Sump 2 2 RCS I 2 LHSI Discharge to Cold Legs 2 LHSI Discharge to IHSI and HHSI Suction 2 (Valve ND58 to Valves NI332A & NI333B) 2 (LHSI Dir. charge to Valves ND58 & nil 36B) 2 LHSI Discharge to B and C Hot Legs 2 Valve FW28 to LHSI Suction variable 2 2 LHSI Mini-Flow 2 IHSI Discharge to LHSI Discharge 2 IHSI Discharge to Hot legs 2 IHSI Mini Flow
#'2 HHSI Discharge to Cold Legs 2 Valve NV223 to HHSI Suction 2 Note:
~
2 8EOC Mode 4 RCS boron concentration. 2 2' variable" indicates that the associated volume concentration is assumed equal to the RCS boron 2 concentration. which is a fimetion of humun. 4= 8-33 (01 MAY 1992) L
- 9. STARTUP PHYSICS TESTING l
The standard reload startup physics testing program conducted at Duke Power Westinghouse units is summarized below. This program is based on the ANS standard and is documented in Chapter _14 of the FSAR-(Reference i 1). The purpose of the test program is to provide assurance that the 5 reactor core is loaded correctly and can be operated as designed. l Zero Power Physics Testina (ZPPT)
- All Rods Out Critical Boron Concentration (AROCBC)
Isothermal Temperature Coefficient (ITC)
- Control Rod Bank Worth (Reference 15)
Power Escalation Testina (PET)'
- Flux Symmetry Check (Lcw Power, e.g. 30% FP)
Core Power Distribution - CPD (Intermediate Power)
- CPD (High Power) - All Rods Out Critical Boron Concentration - AROCBC (High Power)
The above tests will be performed during each initial startup after refueling. Additional testing may be performed if conditions warrant.
.All aspects of the physice testing program are acceptable with respect to implementation'of the Lake Power Company licensing analyses and a complete reload batch of hark-BW fuel assemblies. Therefore, operation with either a mixed Westinghouse and BWFC core or future cores with all BWFC fuel will not require any changes to the' current Duke-startup '
physics testing program. 1 i 9-1.
l l l
- 10. REFERENCES ;
- 1. McGuire Nuclear Station, Final Safety Analysis Report, Docket "i Nos. 50 - 369/370.
- 2. BAW-10172P-A, Mark-BW Mechanical Design Report, Babcock & Wilcox, Lynchburg, Virginia, December 19, 1989.
. 3. DPC-NE-2001PA, Rev.'1, Fuel Mechanical Reload Analysis Methodology for Mark-BW Fuel, Duke Power Company, October 1990.
- 4. BAW-10084A, Rev. 2, Program to Determine In-Reactor Performance -
of-B&W Fuels - Cladding Creep Collapse, Babcock & Wilcox, October { 1978. l S. BAW-10141P-A, Rev. 1, TACO 2 - Fuel Performance Analysis,' Babcock !
& Wilcox, June 1983.
-{
- 6. DPC-NF-2010A, McGuire Nuclear Station / Catawba Nuclear Station Nuclear Physics Methodology for Reload Design, Duke Power ,
Company, June 1985. 7 DPC-NE-2011PA, Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors, Duke Power Company, March 1990. '
- 8. DPC-NE-2004P-A, McGuire and Catawba Nuclear Stations Core Thermal- Hydraulic Methodology using VIPRE-01, Duke Power Company,. December 1991.
- 9. BAW-10159P-A, BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies, Babcock & Wilcox, Ju'y 1990.
- 10. BAW-10173P-A, Mark-BW Reload Safety Analysis for Catawba and McGuire, Babcock & Wilcox,-Revision 2, February 20, 1991.
- 11. DPC-NE-3000P, Duke Power Company, Thermal-Hydraulic Transient Analysis Methodolog;, Revision 1, May 1989.
t
- 12. DPC-NE-3001-PA, Dake Power Company, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology, Revision 1, November 1991.
- 13. BAW-10174-A, Mark-BW Reload LOCA Analysis for the Catawba and' McGuire Units, Babcock & Wilcox, May 1991. ,
~
- 14. .BAW-10168-A,.B&W Loss-of-Coolant Accident Evaluation Model For.
- Recirculating Steam Generator Plants, Babcock & Wilcox, Lynchburg, Virginia, January 1991. i
- 15. DPC-NE-1003A, Revision 1, McGuire Nuclear Station / Catawba Nuclear- ,
Station Rod Swap Methodology Report for Startup Physics Testing, , December 1986. ,
- 16. DPC-NE-3002-A, McGuire Nuclear Station / Catawba Nuclear Station FSAR= Chapter 15 System Transient Analysis Methodology, November l 1991.
t i i 10-1 i f
ATTACHMENT 4 DUKE POWER COMPANY MCGUIRE NUCLEAR STATION NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS j i _j
As required by 10 CFR 50.91, this analysis is provided concerning whether the proposed Technical Specification (TS) ame.ndments involves a no significant hazards consideration, as defined by 10 CFR 50.92. The standards for determining if a proposed amendment involves no significant hazard considerations are; if operation of the facility in accordance with the proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety. For the purpose of this analysis, the proposed TS amendments provided by this submittal are grouped into 3 categories. The following is a brief discussion of each of the 3 categories: ,
- 1) Administrative type changes in nature. This category is comprised of changes to the TS page numbering scheme and adding a designation to the top and bottom of a TS page if the TS page is only applicable to Unit 1 or Unit 2 only. The color of the page has also been changed. If the TS page is applicable to both Units 1 and 2, then the page color remains white. If the TS page is applicable to Unit 1 only, then the page color is changed to '
blue. If the TS page is applicable to Unit 2 only, then the page color is changed to canary. The intent of these changes is to minimize confusion over which unit a particular TS requirement is applicable to, since for certain TS requirements there will be different numerical values specified for certain parameters.
- 2) Changes to the Cold Leg Accumulator (CLA) and to the Refueling Water Storage Tank (RWST) boron concentration limits. The proposed changes that fall into this category are; an increase in the maximum boron concentration limit for the Unit 1 CLA; an-increase in the maximum and the minimum boron concentration limitsEfor the Unit 2 CLA; an increase in the minimum boron concentration limit for the Unit 21RWST while in mode 5 and 6; an increase in the maximum boron concentration limit for the Unit 1 RWST while in modes 1 through 4;'and an increase in the maximum and minimum boron concentration limits for Unit 2 RWST while in modes 1 through 4. Certain numerical values found within the Bases of the McGuire TSs are also revised. Specifically, the changes involves a decrease in the minimum pH value of the sump recirculation water following a Loss Of Coolant Accident (LOCA) and the boron concentration limits were revised to reflect the above changes, where appropriate.
- 3) Changes to the boron concentration limits during refueling operations. The proposed changes that fall into this category are; an increase in the minimum boron concentration limit for the Unit 2 refueling canal; and an increase in the minimum boron concentration limit for the Unit 2 spent fuel pool water. The Bases were also revised to reflect the above changes.
The following discussion is a summary of the evaluation of the proposed amendments against the 10 CFR 50.92(c) requirements to demonstrate that all three standards are satisfied.for each of the 4 categories. First Standard (Amendment would not) involve a significant increase in the probability or consequences of an accident previously evaluated. CATEGORY 1: The changes within this category are administrative in nature and only involve how the TS are structured. These particular changes are not considered to be initiators of any previously evaluated accident. As such, the probability of accidents previously evaluated would not change as a result of the proposed changes within this category. The proposed changes do not contribute in any way to the outcome of an accident that has been previously evaluated, nor do they play a role in the mitigation of any previously evaluated accident because the changes associated with this category only concern how the TSs are structured. Accordingly, the consequences of previously evaluated accidents is not altered by the proposed changes from this category. CATEGORY 2: The proposed amendments provided by this category primarily involve changes to the boron concentration limits for the RWST and the CLA for a particular unit. The changes to the boron concentration limits of the RWST and the CLA-are necessary to support the safe operation of McGuire Unit 2 Cycle 9. The minimum boron concentration limits ensure the reactor will remain subcritical during a LOCA and the limits are determined by the NRC approved methodology described in the Duke Power Topical Report DPC-NF-2010A. For modes 5'and 6, the minimum RWST boron concentration limits ensure that negative reactivity control _is available when the plant is a cold shutdown condition. The requirements are based on ensuring that a 1% delta K/k shutdown margin is maintained at all times during the cycle. The required boron concentration limits are determinedLby the NRC approved methodology discussed in DPC-NF-2010A.
E The maximum RWST and CLA boron concentration limits are necessary to provide adequate operating margin, given the increase in the minimum boron concentration limits. In addition, the maximum boron concentration limits ensure that boron precipitation is precluded following a LOCA. The methods and assumptions utilized to perform the boron precipitation analysis is described in a Westinghouse letter CLC-NS-309 dated April 1, 1975 and is consistent with previous reload submittal approved by the NRC for McGuire. The revision to the minimum allowable value for post-LOCA containment sump pH is necessary due to the proposed increases in the RWST and CLA boron concentration limits. The higher boric acid content could result in a post-LOCA mixed containment sump pH of less than 8.5. As a result the pH band specified within the Bases of the McGuire TSs is revised to be greater than or equal to 7.5 and less than or equal to 10.5 and is consistent with the NRC criteria for sump pH after a LOCA, contained in Branch Technical Position MTEB 6-i. In summary, all proposed changes associated with this category are a result of analysis that have been performed by analytical methods and techniques that have been accepted by the NRC and whose results are clearly within all NRC acceptance criteria. Accordingly, the proposed changes associated with this category do not significantly increase the probability or consequences of any previously evaluated accidents. CATEGORY 3: The proposed changes provided by this category primarily involve changes to the boron concentration limits in the Unit 2 refueling canal and the Unit 2 spent fuel pool water. These changes are intended to be consistent with the proposed change to the RWST minimum boron concentration limit. During refueling, the water in the refueling canal and the spent fuel pool can be mixed during fuel transfer. Raising the refueling canal and spent fuel pool minimum boron concentration limits to the RWST minimum boron concentration requirement will prevent the RWST boron concentration from getting out of specification upon post refueling refill. As such, the proposed change would provide additional shutdown margin during refueling operations. Further, the minimum boron concentration limit in these two areas are not considered to be initiators.of any accidents that have been previously analyzed.
~
Therefore, the proposed changes from this category will not increase the probability of a previously evaluated accident or increase the consequences of the accident.
Second Standard (Amendment would not) create the possibility of a new or different kind of accident from any kind of accident previously evaluated. CATEGORY 1: As discussed above, the changes associated with this category are administrative in nature and only affect how the TSs are structured. Systems, structures or components at McGuire are not affected or changed in any way. Procedures and how the plant is operated and maintained will not be changed. Accordingly, a new or different kind of accident from any accident previously evaluated would not be created. CATEGORY 2: The proposed changes associated with this category are a result of analysis that have been performed by analytical methods and techniques that have been reviewed and approved by the NRC and the results form the analysis are clearly within all NRC acceptance criteria. The changes proposed ensure the safe operation of McGuire 2 Cycle 9 during normal operational situations as well as in response to design base events. The changes also ensure that stress corrosion cracking will not occur for an extended period following a LOCA and iodine retention in the containment sump water is not adversely impacted. Accordingly, the proposed changes of this category will not result in a new or different kind of accident. CATEGORY 3: The proposed changes of this category provide additional shutdown margin beyond what normal would be required during refueling operations. The increase in the minimum boron > concentration within the refueling canal and the spent' fuel pool does not result in any additional operational concerns nor adversely impact any systems, structures or components at McGuire. Accordingly, the proposed changes associated with this category will not create any new or different kinds of accidents than those that have been previously. evaluated. i
l l
-j I
Third Standard (amendment would not) involve a significant reduction in a margin of safety. CATEGORY 1: The changes associated with this category only affect how the TSs will be structured and are considered to be-administrative in nature. No margin of safety are affected in any way by the proposed changes associated with-this category. Accordingly, the' proposed changes do not involve a reduction in a margin of safety. CATEGORY 2: The analytical methods and techniques utilized to determine the proposed changes associated with this category have been reviewed and approved by the NRC. The results of the analyses that were performed confirm that the proposed changes are well within all NRC acceptance criteria. The. proposed TS changes ensure that McGuire 2 Cycle 9 will operate safely and that the consequences-of design base events for all modes of operation are within NRC approved acceptance criteria for McGuire. Accordingly, the proposed changes associated with this category will not involve a. significant reduction in a margin of safety. CATEGORY 3: The proposed changes associated with this category provide an additional shutdown margin beyond what would normally be required. As such, a margin of safety would be enhanced by the changes of this category. Accordingly, there would be no' reduction in a margin of safety as a result of the proposed changes associated with this category. Based on the above and the supporting technical justification, Duke has concluded that there is no significant hazard consideration involved in this request.}}