ML20107E154
| ML20107E154 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 02/22/1985 |
| From: | BALTIMORE GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20107E151 | List: |
| References | |
| NUDOCS 8502250398 | |
| Download: ML20107E154 (125) | |
Text
{{#Wiki_filter:e.a_.. l l ENCLOSURE TO CALVERT CLIFFS UNIT 1 CYCLE 8 RELOAD LICENSE SUBMITTAL 4 d 8502250398 DR 850222ADOCK0500g7
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Colvtrt Cliffs Unit 1 Cycle 8 } Reload License Submittal Table of Contents Section
- 1. Introduction and Summary
, 2. Operating History of the Previous Cycle
- 3. General Description 4 Fuel System Design
- 5. Nuclear Design *
, 6. Thermal-Hydraulic Design
- 7. Transient Analysis
- 8. ECCS Performance Analysis
- 9. Technical Specifications
- 10. Startup Testing
- 11. References f
e e h e
- e 9
1.0 INTRODUCTION
AND
SUMMARY
This report provides an evaluation of design and performance for the operation of Calvert Cliffs Unit 1 during its eighth fuel cycle, at full rated power of 2700 MWt. All planned operating conditions remain the same as those for Cycle 7. The core will consist of 141 presently operating Batch F, G, H, and J assemblies, 72 fresh Satch K assemblies, and 4 Batch E assemblies previously discharged from Cycle 4 of Calvert Cliffs Unit 2. Plant operating requirements have created a need for flexibility in the Cycle 7 termination point, ranging from 12,900 MWD /T to 13,900 MWD /T. In performing analyses of design basis events, determining limiting safety settings and -establishing limiting conditions for operation, limiting values of key parameters were chosen to assure that expected Cycle 8 conditions would be enveloped, provided the Cycle 7 termination point falls within the above cycle burnup range. The analysis presented herein will accommodate a Cycle 8 length of up to 14,400 MWD /T. The evaluations of the reload core characteristics have been conducted with respect to the Calvert Cliffs Unit 1 Cycle 7 safety analysis described in Reference 1, hereafter referred to as the " reference cycle" in this report unless otherwise noted. This is an appropriate reference cycle because of the similarity in the basic system characteristics of the two reload cores. l Specific core differences have been accounted for in the present analysis. In all cases, it has been concluded that either the referenes cycle analyses envelope the new conditions or the revised analyses presented herein continue to show acceptable results. Where dictated by variations from the reference cycle, proposed modifications to the plant Technical Specifications are provided and are justified by the analyses reported herein. All Cycle 8 analyses address fuel exposure explicitly. The performance of , Combustion Engineering 14x14 fuel at extended burnup is discussed in i Reference 2. Fuel performance for Cycle 8 has been evaluated with the l FATES 3 computer code (References 3 and 4) as approved by the NRC in Reference 5. j \ i o e I ! 1-1 l I
2'.0 OPERATING HISTORY OF THE PHEVIOUS CYCLE Calvert Cliffs Unit 1 is presently operating in its seventh fuel cycle utilizing Batch J, H, G, F, E, D and B fuel assemblies (including twenty-four Batch D and B assemblies from Unit 2). Calvert Cliffs Unit 1 Cycle 7 began operation on November 30, 1983 and reached full power on December 22, 1983 The Cycle 7 startup testing was reported to the NRC in Reference 6. The reactor has operated up to the present time with the core reactivity, power distributions and peaking factors closely following the calculated predictions. It is presently estimated that Cycle 7 will terminate on or about April 5, 1985. The Cycle 7 termination point can vary between 12,900 MWD /T and 13,900 MWD /T to accommodate the plant schedule and still be within the assumptions of the Cycle 8 analyses. As of February 12, 1985, the Cycle 7 burnup had reached 11,554 MWD /T. e l l e l
^
l 2-1
- ^
l
3.0 GENERAL DESCRIPTION a. The Cycle 8 core will consist of the number and types of assemblies and fuel batches as described in Table 3-1. The primary change to the core in Cycle 8 is the removal of 76 assemblies (52 Unit 1 assemblies: 36 Batch G, 4 Batch F, 12 Batch E/; 24 Unit 2 assemblies: 12 Batch D/ and 12 Batch B). These assemblies will be replaced by 48 fresh unshimmed Batch K assemblies (4.03 wt% U-235 enrichment), 24 fresh unshimmed Batch K' assemblies (3.43 wt% U-235 enrichment) and 4 Batch E assemblies (3.03 wt% U-235) discharged from Unit 2 Cycle 4 Figure 3-1 shows the fuel management pattern to be employed in Cycle 8. Figure 3-2 shows the locations of the poison pins within the lattice of twice-burned Batch H/ assemblies and the fuel rod locations in unshimmed assemblies. This fuel management pattern will accommodate Cycle 7 termination burnups frca 12,900 MWD /T to 13,900 MWD /T. The Cycle 8 core loading pattern is 90 rotationally symmetric. That is, if one quadrant of the core were rotated 90 0 inte its neighboring quadrant, each assembly would be aligned with a similar assembly. This similarity includes batch type, number of fuel rods, initial enrichment and burnup. Figure 3-3 shows the beginning of Cycle 8 assembly burnup distribution for a Cycle 7 termination burnup of 13,900 MWD /T. The initial enrichment of the fuel assemblies is also shown in Figure 3-3. Figure 3-4 shows the end of Cycle 8 assembly burnup distribution. The end of Cycle 8 core average exposure is approximately 29,400 MWD /T and the average discharge exposure
,is approximately 42,200 MWD /T. The end of cycle burnups are based on
.. Cycle 7 and Cycle .8 lengths of 13,900 MWD /T and 14,400 MWD /T, respectively.
3.1 SCOUT Demonstration Assembly The original configuration of the SCOUT demonstration assembly was - described in Reference 7. It is a Batch F test assembly which was initially inserted in Cycle 4 Changes, similar to those described in Reference 8, were made to this assembly prior to its third cycle of irradiation in Cycle 6. Before returning the assembly to the core for its fourth cycle of irradiation in Cycle 7, 2 segmented test rods were removed from the assembly and replaced with 2 stainless steel rods. The Secut assembly will be reinserted in the core for its fifth cycle. of irradiation in Cycle 8 without further replacement of fuel rods. 3.2 PROTOTYPE Demonstration Assemblies s The e iginal configuration of the PROTOTYPE demonstration assemblies was described in Reference 9. These are Batch G demonstration assemblies which were initially inserted in Cycle 5. Before returning the assemblies to the core for their third cycle of irradiation in Cycle 7, 2 segmented test rods were removed from one of these assemblies and replaced with 2 stainless steel rods. The Prototype assemblies will be reinserted in the core for their fourth cycle of irradiation in Cycle 8 without further replacement of fuel rods. 3-1
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3.3 CEA Patterns The composition of nine CEAs, the configurations of two CEA banks and, j consequently, the overall' CEA bank pattern are being changed for Cycle 8. 1 These changes are being made to support the expansion of the negative MTC ! Tech Spec. (Section 9.0). This support comes in the form of increased net ! available scram worth which is being used in the Steam Line Rupture l Analysis (Section 7 3 2) to compensate for the more adverse reactivity cooldown data that results from the MTC change. ! This increase in scram worth is being brought about by fully strengthening the weak CEA in the vicinity of the worst stuck CEA. In addition to changing the strength of this particular CEA, the compositions of other . CEAs and the configurations of both the lead bank and another CEA bank are being altered. Such additional changes are being made because the weak CEA in the vicinity of the worst stuck CEA is presently part of the lead bank and strengthening .this CEA without changing the composition of the lead bank and, consequently, other banks would lead to an undersirable increase in rodded peaking factors for the lead bank. The specific changes that will be made to the compositions of individual " CEAs, to the configurations of CEA banks and to the overall CEA bank pattern during the Cycle 7 to Cycle 8 outage are summarized below:
- 1. All eight weak CEAs which are presently part of the lead bank, Bank 5, (Figure 3-5) will be converted to full strength CEAs.
- 2. The center CEA which is presently part of Bank 5 and is presently a full strength CEA (Figure 3-5) will be converted to a very weak CEA, i.e., it will be composed of only Al 0 CEA fingers.
23
- 3. The present Bank 5 configuration of 9 CEAs, eight weak and one full strength (Figure 3-5), will be changed to 5 CEAs, four full strength CEAs and one very weak CEA (Figure 3-8).
4 The present Bank 4 configuration of four full strength CEAs (Figure 3-6) will: be changed to eight full strength CEAs (Figure 3-9) by the addition of the four CEAs which will be removed from Bank 5.
- 5. The configuration of all other CEA banks will remain unchanged. The present and future overall CEA bank patterns are shown in Figures 3-7 and 3-10, respectively.
t 3-2
TABLE 3-1 CALVERT CLIFFS UNIT 1 CYCLE 8 CORE LOADING Total Initial Number Total Initial Poison Poison of Poison Number Assembly Number of Enrichment Bate Rods Per Loading and Non-Fuel of Fuel yurnup(HW((Tj) Dasignation Assemblies (wt% U-235) BOC8 EOC8
- Assembly (wt% B4C) Rods Rods K 48 4.05 0 13,300 0 0 0 8448 Ka 24 3.40 0 17,800 0 0 0 4224 J III 11 8 4.G5 12,700 29,000 0 0 0 8448 J8III 16 3.40 17,300 31,800 0 0 0 2816 II II) 11 0 4.00 27,400 41,100 0 0 0 7040 II/ III 32 3.55 30,700 43,300 8 3.03 256 5376 G III li 3.65 38,700 50,400 0 0 2 702 Il} 44,200 1 3.03 52,100 0 0 3 173 F(2) 3.03 25,600 36,800 0 0 0 E ,1 704 TOTAL 217 15,000 29,400 261 37,931 (1) Carried over from Cycle 7 to Cycle 8 of Unit 1 (2) Twice burned Batch E fuel discharged from unit 2 Cycle 4.
(3) Cycle 7 burnup of 13,900 MWD /T (ii) Cycle 8 buroup of 14,8100 MWD /T F82025 s t e
1 2 K K
. 3 4 5 6 7 K K K H J 8 9 10 11 12 13 K J H J' H/ H 14 15 16 17 18 19 20 K J J' J H/ K' H 21 22 23 24 25 26 27 28 K J J' J H/ J H J 29 30 31 32 33 34 35 -
36 +t K H J H/ K' H/ K' G 37 38 39 40 41 42 43 44 K J* H/ J H/ J H J 45 . 46 47 48 49 50 51 52 53 ! H H/ K* H K' H K* E g I K 55 56 57 58 59 ++ 60 61 62 + J H H J G J E F i + LOCATION OF DEMONSTRATION ASSEMBLY (SCOUT) , ! ++LOCA'. .'ON OF PROTOTYPE ASSEMBLIES 1 i l l BALTIMORE 9"'" GAS & ELECTRIC CO. CALVERT CLIFFS UNIT 1 CYCLE 8 Calvert Cliffs CORE MAP 3-1 Nuclear Power Plant ,-
-. .. -. . . . - . _ - . . . - - - . -. '-4.-.-.---.- . . - . - . - . - . - _ - _ . - . _ - . ..
UNSHIMMED ASSEMBLY I l f H/ 8 POISON ROD ASSEMBLY
'X iX X X 1
IX X X X l l l FUEL ROD LOCATION ,
@ POISON ROD LOCATION
~
BALTIMORE "9
- GAS & ELECTRIC CO. CALVERT CLIFFS UNIT 1 CYCLE 8 Calvert Cliffs ASSEMBLY FUEL AND OTHER ROD LOCATIONS 32 Nuclear Power Plant 3-3
1 INITIAL ENRICHMENT W/O U-235 , BOC 8 BURNUP (MWD /T), EOC 7 = 13,900 MWD /T 1 K 2 K 4.05 4.05 0 0 3 K 4 K 5 K 6 H 7 J 4.05 4.05 4.05 4.00 4.05 0 0 0 27,500 13,400 8 K 9 J 10 H 11 J' 12 H/ 13 H 4.05 4.05 4.00 3.40 3.55 4.00 0 10,500 26,300 17,500 29,600 26,900 14 K 15 J 16 J' 17 J 18 H/ 19 K* 20 H 4.05 4.05 3.40 4.05 3.55 3.40 4.00 0 13,900 16,700 10,500 31,300 0 27,000 21 K 22 J 23 J' 24 J 25 H/ 26 J 27 H 28 J 4.05 4.05 3.40 4.05 3.55 4.05 4.00 4.05 0 10,500 17,600 13,900 30,300 12,200 27.700 13,400 29 K 30 H 31 J 32 H/ 33 K* 34 Hi 35 K' 36 G 4.05 4.00 4.05 3.55 3.40 3.55 3.40 3.65 0 26,400 10,500 31,100 0 31,400 0 38,700 37 K 38 J' 39 H/ 40 J 41 H/ 42 J 43 H 44 J 4.05 3.40 3.55 4.05 3.55 4.05 4.00 4.05 45- K 0 17,500 31,300 12,200 31,400 15,800 27,700 15,900- . 4.05 o 46 H 47 H/ 48 K* 49 H 50 K* 51 H 52 K' 53 E 4.00 3.55 3.40 4.00 3.40 4.00 3.40 3.03 4 54 K 27,500 29,500 0 27,700 0 28,900 0 25,600 4.05 0 55 J 56 H 57 H 58 J 59 G 60 J 61 E 62 F 4.05 4.00 4.00 4.05 3.65 4.05 ' 3,o 3.03 13,400 26,900 27,000 13,400 38,700 15,900 25,60b 44,200 4 BALTIMORE CALVERT CLIFFS UNIT 1 CYCLE 8 Figure GAS & ELECTRIC CO. ASSEMBLY AVERAGE BURNUP AT BOC Calvert Cliffs AND INITIAL ENRICHMENT DISTRIBUTION 33 Nuclear Power Plant .- 34 .-----..__-:--.--,
I 1 K 2 K 11,000 14,000 i 3 K 4 K 5 K 6 H 7 J 11,300 14,800 16,400 40,200 28,800 8 K 9 J 10 H 11 J' 12 H/ 13 H 12,500 26,500 40,500 31,300 41,500 39,600 14 K 15 J 16 J' 17 J 18 H/ 19 K+ 20 H 12,500 29,600 32,000 27,600 43,900 17,900 41,200 i 21 K 22 J 23 J' 24 J 25 H/ 26 J 27 H 28 J 11,300 26,500 32,600 30,600 43,300 28,800 42,200 30,200 29 K 30 H 31 J 32 H/ 33 K* 34 H/ 35 K' 36 G 14,700 40,400 27,400 43,900 17,700 44,200 18,100 50,400 37 K 38 J' 39 H/ 40 J 41 H/ 42 J 43 H 44 J 45 K 16,400 31,300 43,800 28,700 M,200 31,700 41,800 31,300 , 11,000 H 47 K* 49 K* 53 46 H/ 48 H 50 K* 51 H 52 E 40,200 41,400 17,900 42,200 18,000 42,800 17,200 36,800 14,100 55 J 56 H 57 H 58 J 59 G 60 J 61 E 62 F 28,800 39,600 41,200 30,200 50,400 31,300 36,800 52,100 BALTIMORE Rgure GAS & ELECTRIC CO, CALVERT CLIFFS UNIT 1 CYCLE 8 Calvert Cliffs ASSEMBLY AVERAGE BURNUP AT EOC (MWD /T) 34 Nuclecr Power Plant . 2-7
Figura 3-5 CALVERT CLITES UNIT 1 CYCLE 7 BANK-5 CONFIGURATION I a 3 4 5 6 7 V 9 /c // /A, /3
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Figure 3-6 CALVERT CLIFFS UNIT 1 CYCLE 7 3xat-4 cohTICURATIon i G Q 4 6' 4 7 ? 9 /0 // /A, /3
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l Figura 3-7 CALVERT CLIFFS UNIT 1 CYCLE 7 CEA BANK PATTERN " t 2 3 4 l [ b 1 V 4 /0 // /L 13 C 1 1, C .
/4 /$ 14 /7 /R If 40 at as 23 i4 l A C 5 C A j af ad a7 af 29 Jo 31 Ja, 3J 34l Jr 36 37 l
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( C 1 1 C 314 aif 216 at? 3-10 .
Figura 3-8 CALVERT CLITIS UNIT 1 CYCLE 8 3ANK-5 CONFIGURATION I 2 3 4 S 6 ? V 9 /0 II /A, 13
/4 IS 14 17 ff 96 40 31 dk n3 R4 35 46 a7 RF 39 Jo 31 JR 3J 34 3r 36 37 J3 39 90 41 44 VJ T'f 45 V6 17 Tl 'li 53 f/ #2 S3 59 55 56 47 SB 51 da 61 di 64 64 e5 44 47 61 61 9e 11 72 73 7f 7f 14 17 77 ff VC RI 33 93 84 PS 16 77 FB ?? 70 9/ 14 93 fy 15 f6 97 1F 99 10 0
/08 103 Q /09 /05 /06 /07 /08 O O lie / 11 Ili II) /14 9 ll6 Ill O*O O OO 9*6 9 ll1
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1 51 162 153 154 15 5 156 /37 /51 /Si /60 /4/ /6A 16 3 /61 lis l /66 /67 /69 /69 /7e /7/ /7A /73 /14 /15 /76 177 173 /?f IVe MI /14 113 /31 /15 /76 /37 /18 /RT /70 19/ /1L 113 i l94 /ff /76 /97 /f? G8 a% 20' J/4 603 o?0Y eS
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I l Figura 3-9 CALVERT CLIFFS UNIT 1 CYCLE 8 BANK-4 CONFIGURATION I & 3 4 S' 6 7 ? 9 to // /A. /J
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Figurn 3-10 CALVERT CLIFFS UNIT 1 CYCLE 8 4 CEA BANK PATTERN t I t a 3 4 i 5' 6 7 ? 9 to // /4, /J
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, 4.0 FUEL SYSTEM DESIGN 4.1 Michtnicel Design 4.1.1 Fuel Design The mechanical design for the Batch K reload fuel is identical to that of I the Batch J fuel described in the reference cycle submittal (Calvert Cliffs Unit 1 Cycle 7, Reference 1), with the exception of the design features listed below,
- a. The height of the lower end fitting is shorter. This reduction is !
i achieved by shortening the legs of the lower end fitting assembly,
- b. The overall lengths of the guide tubes are increased to compensate for the shorter lower end fitting described in a. This increase is achieved by increasing the length of the buffer region, i.e., tapered region. The combination of this shorter lower end fitting and the longer guide tubes maintains the same overall assembly length as that of the Batch J fuel.
- c. The elevations of the Inconel grid and the uppermost Zircaloy grid are changed to maintain their same relative elevations with respect to fuel rods as those of the reference cycle fuel design.
The changes described above were analyzed and found to have no significant i adverse effect on the performance of the Batch K fuel relative to that of the Batch J fuel. These changes will result in improved performance by increasing the shoulder gap from 1.400 inches to 1.775 inches. . The mechanical designs of the Calvert Cliffs Unit 1 Satch F, G, and H fuel assemblies were described in References 2, 3, and 4, respectively. Details of the Calvert Cliffs Unit 2 Batch E/ fuel assemblies that will be used in Cycle 8 can be found in Reference 5. 4.1.2 Clad Collapse C-E recently completed an EPHI-sponsored reassessment of the phenomena of interpellet gap formation and clad collapse in modern PWR fuel rods (i.e., nondensifying fuel in prepressurized tubes). The report concluded that the I collapse time for modern fuel is significantly larger than its expected useful life. This conclusion was based upon both empirical data covering several vendors'- fuel and an analytical evaluation of the propensity for clad collapse into a postulated gap of finite length. A draft copy of this report was submitted to the NRC for evaluation as Attachment 5 of Reference 6. A synopsis of this report focusing on C-E manufactured modern fuel was submitted along with this draft copy as Attachment 4 of Reference
- 6. Based upon the conclusion and recommendation of Attachment 4 (Reference
- 6) that cycle specific clad collapse analyses are not necessary for modefn C-E manufactured fuel, a cycle specific calculation has not been prepared for Unit 1 Cycle 8.
There will be four test rods in the SCOUT assembly in Cycle 3 wnien have gap regions in the active core of sufficient length to require evaluation. Such an evaluation has been comoleted wnien shows that the minimum collapse time for these rods (52,000' EFPM) exceeds the cumulative exposure at the end of Cycle 8 (49,600 EFPH). The calculations for these test rods utilized the finite gap version of the CEPAN computer described in Attachment 5 of Reference 6. 4-1 . . .
4.1.3 Dimensional Changes All standard fuel assemblies in Cycle 8 were reviewed for shoulder gap clearance using the SIGREEP model described in Reference 7 (approved in Reference 8) and for fuel assembly length clearance using the refined correlation discussed in References 1 and 9. All clearances were found to be adequate for Cycle 8. The clearance for fuel rod growth in the SCOUT and PROTOTYPE assemblies will be evaluated during the cycle 7 outage and modified if necessary. 4.1.4 CEA Design The replacement CEAs to be utilized for the changes described in Section 3.3 have essentially the same design as the original components (Reference FSAR) with the exception that all will include reconstitutable features which are similar to those used in a Calvert Cliffs Unit 2 demonstration CEA (Reference 10) . This reconstitutable design will also be used for the replacement of discharged CEAs. The full strength replacement CEA will use Ag-In-Cd pellets at the tip of all five control rods; the previous design used Ag-In-Cd pellets at the tip of just the four outer control rods. The very weak replacement CEA for the core center location will contain only Al 0 23 ""d Zi" "I Y P'11*** 1" lieu of Bg C and Ag-In-Cd pellets. 4.1.5 Removal of CEA Plugs Unit 1 is presently operating with CEA plugs installed in the locations originally occupied by Part Length Rods (PLRs). These CEA plugs will be removed for Cycle 8 to facilitate the installation of the Reactor Vessel Level Monitoring System and to expedite refueling outage operations. An assessment of the effects that removing these CEA plugs would produce has been completed. This assessment concluded that the removal of CEA plugs from all eight PLR positions can be effected safely for both Calvert Cliffs units. i 4.1.6 Metallurgical Requirements The metallurgical requirements of the fuel cladding and the fuel assembly structural members for the Batch X fuel are identical to those of the other fuel batches to be included in Cycle 8. Thus, the chemical or metallurgical performance of the Batch K fuel will remain unchanged from that the Unit 1 Cycle 7 fuel. 4.2 Hardware Modifications to Mitigate Guide Tube '4 ear I a All standard fuel assemblies which will be placed in CEA locations in Cycle 8 will have stainless steel sleeves installed in the guide tubes to prevent guide tube wear. A detailed discussion of the design of the sleeves in irradiated fuel ~ assemblies and their effect on reactor operation is contained in Reference 11 A modified short sleeve design will be used in Batch K fuel assemblies. This will allow for reconstitution of the 3atch X fuel assemblies without having to renove and, consequently, reinstall the guide tube sleeves. A discussion of the snort sleeve design is contained in Reference 12. . 4-2 '. l
~_ - ._ . . . _ _ - . __ .
M Cycle 8 will also . utilize one fuel assembly (Satch F, SCCUT) in a CEA location that was fabricated with modified guide tubes (see Reference 2) instead of sleeves to mitigate guide tube wear. This modified assembly has previously resided in a CEA position for two cycles. An examination for guide tube wear was conducted after one cycle of residence in a CEA position. The test results presented in Reference 13 showed no detectable wear. 4.3 Thermal Design The thermal performance of a composite fuel pin which envelopes the various + fuel assemblies present in Cycle 8 (fuel Batches F, G, H, J, and K and Datch E from Unit 2) has been evaluated using the FATES 3 version of the fuel evalcation model (References 14 and 15), as approved by the NRC (Reference 16). The analysis was performed with a history that modeled the power and burnup levels representative of the peak pin at each burnup interval, from beginning of cycle to end of cycle burnups. The burnup , range analyzed is in excess of that expected at end of Cycle 8. In j addition, the SCOUT and PROTOTYPE test pins were analyzed and found to be bounded in both temperature and pressure by the standard fuel batches in
. Cycle 8. Consequently, the test pins are not limiting with respect to thermal performance.
The augmentation factor is being removed from the Tech. Specs. and the values of several items used in the Incore Monitoring System, i.e., measurement-calculational uncertainty and axial fuel densification and I thermal expansion factor, are being lowered (See Section 9.0). These changes when coupled with an unchanging LOCA limit present the ' potential for core operation at a higher steady state local linear heat rate. This higher local power level which can result in more adverse fuel performance was included in the Cycle 8 analysis. ? 5 . ~I J 4-3 , n-- - a
i i
- j. 5.0 NUCLEAR DESIGN l
~ 5.1 Physics Characteristics
; 5.1.1 Fuel Management
! . The Cycle 8 fuel management employs a mixed central region as described in ! Section 3, Figure 3-1. The fresh Batch K fuel is ccmprised of two sets of
- assemblies, each having a unique enrichment in order to minimize. radial
! power peaking. There are 48 assemblies with an enrichment of 4.05 wt% U- ! 235 and 24 assemblies with an enrichment of 3.40 wt5 U-235. With this i loading, the Cycle 8 burnup capacity for full power operation is expected - to be between 13,700 MWD /T and 14,400 MWD /T, depending on the final Cycle 7 termination point. The Cycle 8 core characteristics have been examined for 1 Cycle 7 terminations between 12,900 and 13,900 MWD /T and limiting values
- established for the safety analyses. The leading pattern (see Section 3) is applicable to any Cycle 7 termination point between the stated extremes.
Physics characteristics including reactivity coefficients for Cycle 8 are l' listed in Table 5-1 along with the corresponding values from the reference , cycle. Please note that the values of parameters actually employed in safety analyses are different from those displayed in Table 5-1 and are ! typically chosen to conservatively bound predicted values with accommodation for appropriate uncertainties and allowances, i l Table 5-2 presents a summary of CEA shutdown worths and reactivity j allowances for the end of Cycle 8 zero power steam line break accident and
- a comparison to reference cycle data. The EOC zero power steam line i_ accident was selected since it is the most limiting zero power transion.t with respect to reactivity requirements and, thus, provides the basis for l . verifying the Technical Specification required shutdown margin.
. iable 5-3 shows the reactivity worths of the three CEA groups which are allowed in the core during critical / power conditions. These reactivity worths were calculated at full power conditions for Cycle 8 and the reference cycle. The configurations of CEA Groups 5 and 4 have been changed as described in Section 3; the configuration of Group 3 remains the ,1 same as in the reference cycle. The power dependent insertion limit (PDIL) curve is the same as for the reference cycle, l i 5.'1.2 Power Distributions Figures ' 5-1 through 5-3 illustrate the all rods out (ARO) planar ' radial powce distributions at BOC8, MOC8 and EOC8, respectively, 'that are characteristic o f. the high burnup end of the Cycle 7 shutdown window. These planar radial power peaks are characteristic of the major portion of the active core length between about 20 and 80 percent of the fuel height. The hith burnup end of the Cycle 7 shutdown window tends to increase . the power. peaking in this axial central region of the core for Cycle 8. The planar radial power distributions for the above region with CEA Grcup 5 fully inserted at beginning and end of Cycle 8 are shown in Figures 5 4 and 5-5, respectively, for the high burnup end of the Cycle 7 shutdown window. j 1 The radial power- distributions described in this section are calculated data . without uncertainties or other allowances. However, the single red power peaking vai6es do include the. increased . peaking that is
- characteristic of fuel rods adjoining the water holes in the fuel assembly 3 lattice. For both DNB and kw/ft safety and setpoint analyses in either N -
._.._.___._._.._-_._.J-L___.__.___,,__.__.._,
4 redded or unrodded configurations, the power peaking values actually used are higher-than those expected to occur at any time during Cycle 8. These conservative values, which are used in Section 7 of this document, establish the allowable limits for power peaking to be observed during operation. The range of allowable axial peaking is defined by the Limiting Conditions for Operation (LCOs) covering Axial Shape Index (ASI). Within these ASI
. limits, the necessary DNBR and kw/ft margins are maintained for a wide j range of possible axial shapes. The maximum three-dimensional or total peaking factor anticipated in Cycle 8 during normal base load, all rods out operation at full power is 1.92, not including uncertainty allowances.
[ 5.1.3 Safety Related Data 5.1 3 1 Ejected CEA Data The maximum reactivity worths and planar power peaks associated with an Ejected CEA Event are shown in Table 5-4 for Cycle 8 and the reference cycle. These values encompass the worst conditions anticipated during Cycle 8 for any expected Cycle 7 termination point. The values shown for Cycle 8 are the safety analysis values which are conservative with respect
- to the actual calculated values. The data for the full power condition
. remained unchanged relative to the reference cycle; however, the data for the zero power condition was revised due to the change in CEA configuration discussed in Section 3. 5.1 3 2 Dropped CEA Data i , The . Cycle 8 safety related data for this section are identical to the safety related data used in the reference cycle. . 5.1 3.3 Augmentation Factors Recently completed analyses (Reference 1) have demonstrated that the
- increased power peaking associated with the small interpellet gaps found in modern, i.e, pre-pressurized and non-densifying, fuel is insignificant compared to the uncertainties in the safety analyses and Tech.
Specs. Consequently, augmentation factors are being eliminated frem reload analyses (See Sections 4.3, 7.0 and 8.1) and the Tech. Specs. (See Section 9.0). 1 . 5.1.3.4 Fuel Temperature Coefficient Bias A negative bias of 15% is being added to the Fuel Temperature Coefficient (FTC) data used in the safety analyses to estalish consistency with the bias on Power Coefficient presented in the ROCS /DIT Topical (Reference 2). This negative bias is being used conservatively by selective application, i.e., FTC data which is bounding in the more negative direction is being adjusted to be even more negative while the FTC data which is bounding in
~ the less negative direction remains unadjusted. This application procedure when combined with the standard 2 15% uncertainty results in a negative adjustment of 30% for the coro negatively bounding data and. a positive adjustment of 15% for the less negatively bounding data.
5-2 ,,
5.2 Analytical Input to In-Cora Mansurem2nts In-core detector measurement constants to be used in evaluating the reload cycle power distributions will be calculated in the manner described in Reference 3, which is the same method used for the reference cycle. 5.3 Nuclear Design Methcdology Analyses have been performed in the same manner and with the same methodologies used for the reference cycle analyses. 5.4 Uncertainties in Measured Power Distributions The power distribution measurement uncertainties which are applied to Cycle 8 are the same as those applied to the reference cycle. 4 s
- 90 9
5-3
- - - - . . .-w-., . . , - . 4 - , , , , - - . - - - - , - . - ~ -w -c ,.--,-r,,-. - , , - , . . - - - - - ~ - . , - . . ,
TABLE 5-1 CALVERT CLIFFS UNIT 1 CICLE S NOMINAL PHYSICS CHARACTERISTICS Reference Cycle Units (Unit 1 Cycle 7) Cycle 8 Dissolved Boron Hot Full Power, All Rods Out Equilibrium Xenon Bcron Content for Criticality at BOC PPM 1090* 1200 Boron Worth 1 Hot Full Power BOC PPM /%Ao 105+ 107 Hot Full Power ECC PPM /%Ao 84 88 Reactivity Coefficients < 4-(CEAs Withdrawn) Moderator Temperature Coefficients, Hot Full power, Equilibrium Xenon s Beginning of Cycle 10-4ao/ F -0.2 -0.1 End cf Cycle 10-4a0 / F -2.2 -2.3 Doppler Coefficient Hot Zero Power BOC 10-5ao/ F -1.56 -1.56 I Hot Full Power BOC 10-5ac / F -1.28 -1.26 Hot Full Power EOC 10-540 / F -1.45 -1.44 1 Total Delayed Neutron Fraction, Seff BOC 0.00604 0.00604 ECC 0.00522 0.00516 4 - Neutron Generation Time, t' BOC- 10-0 sec 23.4 23.1 ECC 10-0 sec 29.8 29.2
+ A sligntly ir. correct value was reported in Reference 3 5-a
- TABLE 5-2 CALVERT CLIFFS UNIT 1 CYCLE 8 LIMITING VALUES OF REACTIVITY WORTHS AND ALLCWANCES FOR THE END-OF-CYCLE (EOC) HOT ZERO PCWER (HZP)
STEAM LINE RUPTURE ACCIDENT, %A0 Reference Cycle
- Cycle 8
- 1. Worth of all CEA's Inserted 9.1 93
- 2. Stuck CEA Allowance 2.6 2.6
- 3. Worth of all CEA's less Worth 6.5 6.7 of CEA Stuck Out**
4 Power Dependent Insertion 1.6 1.8 Limit CEA Bite at Zero Power
- 5. Calculated Scram Worth 4.9 4.9
- 6. Physics Uncertainty plus 0.6 0.6 Bias .
- 7. Net Available Scram Worth 43 4.3
- 8. Technical Specification 4.3 35 Shutdown Margin ,
9 Margin in Excess of Technical 0.0 0.8 Specification Shutdown Marsin
- Unit 1 Cycle 7.
** Stuck CEA is one which yields worst results for ECC HZP SLS, i.e., eorst combination of scrar worth and reactivity insertion with cooldown.
W
TABLE 5-3 CALVERT CLIFFS UNIT 1 CYCLE 8 REACTIVITY WORTH OF CEA REGULATING GROUPS AT HOT FULL POWER, %Ac Beginning of Cycle End of Cycle Regulating Reference
- Reference
- CEA's Cycle Cycle 8" Cycle Cycle 8##
Group 5 _ 0.53 0.28 0.64 0.36 Group 4 0.34 0.82 0.44 0.91 Group 3 0.99 0.94 1.07 1.02 4 Note Values shown assume sequential group insertion.
- Unit 1 Cycle 7.
- CEA configurations of Groups 5 and 4 have been modified for Cycle 8 as described in Chapter 3.
e
- O 5-6
- r.
TABLE 5-4 CALVEftT CLIFFS UNIT 1 CYCLE 8 CEA EJECTICN DATA Limiting Values Reference Cycle Unit 1 Cycle 8 Safety Analysis Value, Safety Analysis Value Maximum Radial Power Peak Full power with Bank 5 inserted; worst CEA ejected 3.6 3.6 Zero power with Banks 5+4+3 inserted; worst CEA ejected 9.4 9.5 Maximum Ejected - CEA Worth (%ac) Full power with Bank 5 inserted; worst CEA ejected 0.28 0.28 Zero power with Banks 5+4+3 inserted; worst CEA ejected 0.63 0.86 .
- Unit 1 Cycle 7
. Notes
- 1. Uncertainties and allowances are included in the above data.
~
- 2. The Cycle 8 safety. analysis values are conservative with respect to the actual Cycle 8 calculated values.
5-7 s .
l l 1 2 0.76 1.00 3 4 5 6 7 0.81 1.09 1.20 0.87 1.08 - X 8 9 10 11 12 13 0.91 1.20 1.00 0.96 0.79 0.82 1 14 15 16 17 18 19 20 ! 0.91 1.16 1.11 1.26 0.86 1.26 0.95 21 22 23 24 25 26 27 28 0.80 1.19 1.06 1.20 0.89 1.17 0.98 1.16 29 30 31 32 33 34 35 36 1.07 0.98 1.23 0.86 1.24 0.85 1.23 0.72 37 38 39 40 41 42 43 44 1.18 0.94 0.84 1.15 0.84 1.06 0.92 1.01 45 . 46 47 48 49 50 51 52 53 0.86 0.78 1.25 0.97 1.21 0.89 1.15 0.69
- 55 56 57 58 59 60 61 62 1.08 0.82 0.95 1.16 0.72 1.01 'O.69 0.46 NOTE: X = MAXIMUM 1. PIN PEAK = 1.54 CALVERT CLIFFS UNIT 1 CYCLE 8 Figure GAS & ELE T IC CO* ASSEMBLY RELATIVE POWER DENSITY AT BOC, 51 C lvert Cliffs EQUILIBRIUM XENON Nuclect Power Plent
+,
l 1 2 l
- 0.77 0.99 i
3 4 5 6 7 0.78 1.03 1.15 0.89 1.10 8 9 10 11 12 13 0.86 1.12 0.97 0.96 0.83 0.87 14 15 16 17 18 19 20 0.87 1.08 1.05 1.21 0.87 1.24 0.97 21 22 23 24 25 26 27 28 0.78 1.12 1.02 1.15 0.90 1.16 0.98 1.17 29 30 31 32 33 34 35 36 1.03 0.96 1.19 0.88 1*.24 0.88 1.24 0.78 37 38 39 40 41 42 43 44 1.16 0.96 0.87 1.16 0.88 1.10 0.96 1.07 0.78 46 47 48 49 50 51 52 53 l 0.89 0.82 1.25 0.99 1.24 0.94 1.21 0.76 , 54 , 1.00 55 56 57 58 59 60 61 62 1.10 0.87 0.97 1.17 0.78 1.07 'O.76 0.55 NOTE $ X = MAXIMUM 1. PIN PEAK = 1.45 - I i l l l BALWORE CALVERT CLIFFS UNIT 1 CYCLE 8 Figure GAS & ELECTRIC CO. ASSEMBLY RELATIVE POWER DENSITY AT 7 GWD/T Calvert Cliffs l EQUIL18RIUM XENON
- Nuclear Power Plant -
3-9
i i l l I 1 2 i l
, 0.80 1.00 !
3 4 5 6 7 0.79 1.02 ' 15
- i. 0.91 1.10 I 8 9 10 11 12 13 O.86 1.08 0.96 0.97 0.86 0.91 4
14 15 16 17 18 19 20 0.87 1.05 1.01 1.17 0.89 1.22 0.97 i 21 22 23 24 25 26 27 28 0.79 1.09 1.00 1.12 0.91 1.14 0.98 1.14 , 29 30 31 32 33 34 35 36 1.02 0.96 1.17 , 0.89 1.21 0.90 1.21 0.80 37 38 39 40 41 42 43 44 45 1.15 0.97 0.89 1.14 0.90 1.09 0.97 1.08 46 47 48 49 50 51 52 53 g 0.91 0.86 1.22 0.98 1.21 0.95 1.21 0.80 1.00 55 56 57 58 59 60 61 62 1.10 0.91 0.97 1.14 0.80 1.08 0.80 0.62 NOTE: X = MAXIMUM 1 PIN PEAK = 1.40 l l BALTIMORE CALVERT CLIFFS UNIT 1 CYCLE 8 Figure GAS & ELECT I CO. ASSEMBLY RELATIVE POWER DENSITY AT ECC, 53 Nuclear Power Plant EQUILIBRIUM XENON , l 5-10 t . - . . - . _ , . . .. - . - _. -. - . - . . . - - . - - - - - - - ~~-- - . - = - - - - - -
1 2 0.66 0.83 CEA BANK 5 3 4 5 6 7 LOCATIONS 0.83 1.08 1.11 0.74 0.84 8 9 10 11 12 13' 0.97 1.28 1.05 0.94
/ /
0.67 0.47 /
////
14 15 16 17 18 19 20 0.97 1.27 1.27 1.33 0.87 1.16 0.84 21 22 23 24 25 26 27 28 0.82 1.27 1.17 1.32 0.97 1.22 0.99 1.14 X 25, 30 31, 32 33 34 35 36 1.06 1.03 1.30 0.94 1.33 0.91 1.28 0.76 37 38 39 40 41 42 43 44 1.10 0.93 0.86 1.21 0.91 1.15 0.99 1.09 45 , 0.66 "40 47 48 49 50 51 52 53 54 0.73 0'.66 1.16 0.99 1.26 0.96 1.22 0.74 0.83 55 , 57 58 59 60 61 ,
/
0.84
/0 0.84 1.14 0.76 1.09 0.74 NOTE: X = MAXIMUM 1 PIN PEAK = 1.61 -
BALTIMORE Rgure CALVERT CLIFFS UNIT l CYCLE 8 GAS & ELECTRIC CO. ASSEMBLY RELATIVE POWER DENSITY WITH BANK 5 54' Calvert Cliffs INSERTED,HFP,BOC - Nucleer Power Plant
1 2 0.71 0'.85 l CEA BANK 5 l LOCATIONS 3 4 5 6 7 0.84 1.04 1.08 0.76 0.84 8 9 10 11 12 f 0.94 1.17 0.99 0.93 0.70 0.47
/
14 15 16 17 18 19 20 0.95 1.15 1.10 1.24 0.89 1.11 0.82 21 22 23 24 25 26 27 28 0.84 1.17 1.08 1.21 0.98 1.20 0.99 1.13 29 30 31 32 33 34 35 36 1.04 0.99 1.23 0.96 1.32 0.97 1.30 0.85 37 38 39 40 41 42 43 44 45 1.00 0.93 0.89 1.20 0.97 1.20 1.06 1.19 0.72 46 47 48 49 50 51 52 53 l g 0.76 0.70 1.12 0.99 1.30 1.05 1.34 0.88 x
- j. 0.85 55
,56/ / 57 58 59 60 61 0.84 .47/ 0.82 1.13 0.85 1.19 0.88 6
! NOTE: X = MAXIMUM 1. PIN PEAK = 1.50 l i l ! BALTIMORE CALVERT CLIFFS UNIT 1 CYCLE 8 Figure GAS & ELECTRIC CO. ASSEMBLY RELATIVE POWER DENSITY WITH BANK 5 Calvert Cliffs 5'~ INSERTED, H FP, EOC Nuclear Power Plant . i 5-12 __ - __- _, _
l 6.0 THERMAL HYDRAULIC DESIGN
,s 6.1 DNBR Analysis Steady state DNBR analyses of Cycle 8 at the rated power level of 2700 MWt have been performed using the T0hc computer code described in Reference 1, the CE-1 critical heat flux correlation described in Reference 2, and the simplified modeling methods described in Reference 3.
i y* A variant of TORC called CETOP, optimized for simplified modeling .E applications, was used in this cycle to develop the " design thermal margin model" described generically in Reference 3. Details of CETOP are discussed in Reference 4 CETOP was approved for use on Calvert Cliffs Units in Reference 5. CETOP is used only because it reduces computer costs l. significantly; no margin gain is expected or taken credit for. k Table L6-1 contains a list of pertinent thermal-hydraulic design parameters , applicable. to both safety analyses and the generation of reactor protective system setpoint information. The calculational factors (engineering heat ( flux factor, engineering factor on hot channel heat input, rod pitch and clad diameter factor) listed in Table 6-1 have been combined statistically I with other uncertainty factors at the 95/95 confidence / probability level (Reference 6) to define a design limit on CE-1 minimum DNBR when iterating on power as discussed in Reference 6 and approved by the NRC in Reference 5. The applicability of this minimum DNBR limit was verified for Cycle 8. 6.2 Effects of Fuel Bowing on DNBR Margin , , The effects of fuel rod bowing on DNS margin for Calvert Cliffs Unit 1 Cycle 8 have been evaluated using the methods described in Reference 7. These methods were approved by NRC in Reference 8. i Based upon these methods, a penalty of 0.3% DNBR is required to account for the adverse T-H effects of rod bow at an assembly average burnup of 30 GWD/T. An equivalent penalty of 0.4% in radial peak was applied in the , determination of ths Tech. Spec. limit on radial peak. A conservative l- (i.e., maximum over all operating ranges) conversion factor of -1.2% radial peak / % DNBR was used to determine the equivalent radial peak penalty. For those assemblies with an assembly average burnup in excess of 30 GWD/T, L the minimum best estimate margin available relative to more limiting-l peaking values present in other assemblies is greater than 55, exceeding the corresponding rod bow penalties based upon Refererice 7. Hence, sufficient available margin exists to offset rod bow penalties for assemblies with burnup greater than 30 GWD/T.
- 6-r .
L
TABLE 6-1 Cl.LVERT CLIFFS UNIT 1 THERMAL-HYDRAULIC PARAMETERS AT FULL PCWER** Reference + General Characteristics Unit Unit 1, Cycle 7 Cycle 8 Total Heat Output (core only) 2700 2700 MWg 10 BTU /hr 9215 9215 Fraction of Heat Generated .975 .975 In Fuel Rod Primary System Pressure psia 2250 2250 (Neminal) Inlet Temperature F 548 548 Total Reactor Coolant Flow gpg 381,600 381,600 (steady state) 10 lb/hr 143.8 143.8 Coolant Flow Through Coro 0 10 lb/hr 138.5 138.5 Hydraulic Diamettr ft 0.044 0.044 (nominal channel) Average Mass Velocity 100 lb/hr-ft2 2.59 2.59 Pressure Drop Across Core psi 11.1 11.1 (steady state flow irreversible AP over entire fuel assembly) Total Pressure Drop Across psi 34.7 34.7 - Vessel (based on steady state flow and nominal dimensions) Core Average Heat Flux BTU /hr-ft 2 183,000*** 182,300**** (Accounts for above fraction of heat generated in fuel
~
rod and axial densification factor) Total Heat Transfer Area ft 2 49,100*** 49,300****
.( Accounts for axial densification factor)
Film Coefficient at Average BTU /hr-ft2 - F 5930 5930 Conditions 6-2 .,,
TABLE 6-1 (continued) Reference + General Characteristics Unit Unit 1, Cycle 7 Cycle 8 Average Film Temperature F 31 31 Difference Average Linear Heat Rate of kw/ft 6.128** 6.09'*** Undensified Fuel Rod (accounts for above fraction of heat generated in fuel rod) Average Core Enthalpy Rise BTU /lb 66.5 66.5 Maximum Clad Surface F 657 657 Temperature Reference + Calculational Factors Unit 1 Cycle 7 Cycle 8 Engineering Heat Flux on Hot Channel 1.03* 1.03* Engineering Factor on Hot Channel 1.028 . 1.02* Heat Input Rod Pitch and Clad Diameter Factor 1.065* 1.065* = Fuel Densification Factor (axial) 1.01++ 1.01++ Notes
'These factors have been calbined statistically with other uncertainty
, factors at 95/95 confidence /rrobability level (Reference 6) to define a design limit on CE-1 minimum DNBR when iterating on power as discussed in Reference 6 and approved by the NRC in Reference 5. This limit was verified to be applicable to Cycle 8.
**Due to the statistical combination of uncertainties described in References 6, 9, and 10, the nominal inlet temperature and nominal primary sys*Jm pressure were used to calculate some of tbase parameters.
*** Based on a value of 400 shims and 5 non-fuel rods.
**** Based on a value of 256 shias and 5 non-fuel rods.
+ Reference cycle (Unit 1, Cycle 7) analysis is contained in Reference 11.
++This~value is conservative with respect to. existing calculations..
~
J 5-3 ,,
._._ . . , _ . ~ . _ _ _ _ . _ _ ._-
7.0 Transient Analysis This section presents the results of the Baltimore Gas. & Electric Calvert Cliffs Unit 1, Cycle 8 non-LOCA safety analysis. ; The Design Bases Events (DBEs) considered in the safety analysis are listed in Table 7-1. These events were categorized in the following groups:
- 1. Anticipated Operational Occurrences (A00s) for which the intervention of the Reactor Protection System (RPS) is necessary to prevent exceeding acceptable limits.
- 2. ACOs for which the intervention of the RPS trips and/or initial steady state thermal margin, maintained by Limiting Conditions for Operation (LCO), are necessary to prevent exceeding acceptable limits.
- 3. Postulated Accidents A re-evaluation of all DBEs was performed to determine the impact of the following changes.
- a. Fuel temperature coefficient (FTC) multiplier changed from +15% to -15f.,
_ 2
+30% (see Section 5.1.3.4).
- b. Opening pressure setpoint for SG safety valves assumed for the non-LOCA safety analysis (see Chapter 9)
Bank 1 Changes from 1000 psia to 1050 psia
- Bank 2 Changes from 1015 psia to 1050 psia Bank 3 Changes from 1030 psia to 1080 psia Bank 4 Changes from 1050 psia to 1080 psia
- c. Moderator Temperature Coefficient (MTC) Range Positive MTC range was changed from +0.5 to +0.7x15i Negative MTC range was changed from -2.5 to -2.7x10"aoao/ /FF '
- d. Shutdown margin was reduced from 4.3 to 3.5%ao. 7
- e. Proposed Tech. Specs. change to HPSI pump flow rate (see Chapter 9).
Table 7-2 summarizes the core parameters assumed in che Unit 1 Cycle 8 transient analysis and compares them to the values used in the Reference Cycle. Specific initial conditions for each event are tabulated in that event's section of the report. For some DBEs presented, certain initial core parameters were assumed to be more limiting than the actual calculateo Cycle'8 values (e.g., CEA worth at trip). This was done to bound future cy'.:les. For all DBEs that have results bounded by the Reference Cycle, the margin of safety has not degraded from that of the Reference Cycle. Those events wnose results were not boundec by the Reference Cycle, are presented herein.
'The value assumed is conservative with respect to the Technical Specification limit.
7-1 ,.
1 Fcr tho cytnts acceptance presented, criterion to beTable used 7-3 shows the reason for the reanalysis, the results obtained. in judging the results and a summary of the provided in the appropriate sections. Detailed presentations of the results of the 1 I s e e O t 1 k i j is 1
TABLE 7-1 CALVERT CLIFFS U?!IT 1, CYCLE 8 - DESIG?! BASIS EVEITS C0!!SIDERED I?! THE fl0!!-LOCA SAFETY A!!ALYSIS Results 7.1 Anticipated Operational Occurrences for which intervention of the RPS is necessary to prevent exceeding acceptable limits: 7.1.1 Boron Dilution Presented 7.1.2 Starjup of an Inactive Reactor Coolant Sounded by Reference Cycle Pump 7.1.3 Loss of Load Presented 7.1.4 Excess Load Presented 7.1.5 Loss of Feedwater Flow Bounded by Reference Cycle 7.1.6 Excess Heat Removal Due to Feedwater Bounded by Reference Cycle Malfunction 7.1.7 Reactor Coolant Systen Depressurization Bounded by Reference Cycle 7.1.8 Excessive Charging Event Bounded by Reference Cycle 7.2 Anticipated Operational Occurrences for which RPS trips-and/cr sufficient initial steady state thermal margin, maintained by the LCOs, are necessary to prevent exceeding the acceptable limits: 7.2.1 Sequential CEA Group Withdrawal 2 Bounded by Reference Cycle 7.2.2 Loss of Coolant Flow Bounded by Reference Cycle 7.2.3 Full Length CEA Drop Bounded by Reference Cycle 7.2.4 Transients Resulting from the Presented 3 Malfunction of Oge Steam Generator 7.2.5 Loss of AC Power Bounded by Reference Cycle 4 7 3 Postulated Accidents 7.3.1 CEA Ejection Presented 7.3.2 Steam Line Rupture Presented 7.3.3SteamGeneragerTubeRupture Bounded by Reference Cycle
.7.3.4 Seized Rotor Bounded by Reference Cycle I
Technical Specifications preclude this event during operation. , 2 Regaires High Power and Variable Hign Power Trip. 3 Requires Low Flow Trip.
" Requires trip on high differential steam gencastor pressure.
7-3 _,m . ._ _ , - - ~ , _ - - - _ _
TABLE 7-2 l CALVERT CLIFFS UNIT 1. CYCLE 8 ) CORE PARAMETERS INPUT TO SAFETY ANALYSES , FOR DNB AND CTM (CENTERLINE TO MELT) DESIGN LIMITS Reference Cycle Values ** Unit 1 Physics Parameters Units (Unit 1. Cycle 7) Cycle 8 Values 1 Radial Peaking Factors Fog DNB Margin Analyses (Fh) Unrodded Region Bank 5 Inserted 1.70* 1.70* 1.87+' 1.87*' FogPlanarRadialComponent (F ) of 3-D Peak (CYNLimitAnalyses) Unrodded Region 1.705 1.70* Bank 5 Inserted 1.875 1.87* Maximum Augmentation 1.055 1.0 Factor Moderator Temperature 10-4a0/ F -2.5 + +.5 -2.7 + +.7 Coefficient Shutdown Margin %ao -4.3 -3.5 Tilt Allowance % 3.0 3.0
#For DNBR and CTM calcu: ations, effects of uncertainties on these parameters were accounted for statistically. The procedures used in the Statistical Combination of Uncertainties (SCU) as they pertain to DNS and CTM limits are detailed in References 2, 3, and 4 These procedures have been approved by NRC for the Calvert Cliffs Units in Reference 5. , ** Reference 1 4 *The values assumed are conservative with respect to the Tecnnical Specifications limits.
7-4 s
TABLE 7-2 (continued) Reference Cycle Physics Parameters Values (Unit 1, Unit 1, Units Cycle 7) Cycle 8 Values Power Level MWt 2700' 2700' Maximum Steady State *F 548' 548' Temperature (Tin) Minimum Steady State psia 2200* RCS Pressure 2200* Reactor Coolant Flow 1061bm/hr 13C.5' 138.5' Negative Axial Shape .15' LCO Extreme Assumed Ip
.15'*
at Full Power (Ex-Cores) Maximum CEA Insertion 5 Insertion 25 25 at Full Power of Bank 5 Maximum Initial Linear KW/ft 16.0 16.0 Heat Rate for Transients Other than LOCA Steady State Linea.' KW/ft 22.0 22.0 Heat Rate for Fuel CTM Assumed in the Safety Analysis CEA Drop Time from sec 3.1 Removal of Power to 3.1 Holding coils to 90% Insertion Minimum DNBR (CE-1) 1.23' 1.23' 1 L r
'For DNBR and CTM calculations, effects of uncertainties on these parameters were accounted for statistically. The procedures used in the Statistical
' Combination detailed in References of Uncertainties 2, 3. (SCU) and 4 as they pertain to DMB and CTM limits are These procedures have been approved by NRC for the Calvert Cliffs Units in Reference 5.
*The values assumed Specifications limits.
are- conservative- with respect- to the Technical' 7-5 '
Table 7-3 DESIGN BASIS EVENT PRESENTED FOR UNIT 1 CYCLE 8 Reason for Acceptance Event Reanalysis Criterion Summary of Results
'(enanges relative to reference cycle)
Boron Dilution Decrease in Time to Results acceptable. Shutdown Margin Criticality no Further details in less than 15 Section 7.1.1. minutes for Modes 2, 3 and 4 Loss of Load Increase in Peak RCS pressure Peak RCS pressure Moderator less than 2750 calculated to Temperature psia be <2750 psia. Coefficient (MTC) Further details in and increase in Section 7.1.3. opening pressure setpoints of SG safety valves. Excess Load Decrease in DNBR and CTM Results acceptable. . negative MTC and SAFDL's not Further details in change in HPSI exceed ed. Section 7.1.4 pump flow. Transients Decrease in DNBR and CTM Results acceptable. Resulting from negative MTC and SAFDL's not Further details in the Malfunction increase in exceed ed . Section 7.2.4 of One Steam opening pressure - Generator setpoints of SG l safety valves. CEA Ejection Increase in post Total average Results show no pin ejected 3-D peak enthalpy < 200 experiences clad and ejected CEA cal /gm. damage or. incipient worth for hot Total centerline centerline melting, zero power case enthalpy <-310 Further details in l cal /gm. Section 7.3.1 Steam Line Rupture Changes in moderator Radiological Site boundary doses (Inside cooldown curve and dose is less are bounded by the Containment) available scram than 10CFR100.- outside containment worth at trip, and_ dose calculated for l change in HPSI Calvert Cliffs pump flow. Unit 1 Cycle 7. j
- Reference Cycle is specified for each. event in the subsequent sections.
7-6 .-
7.1.1 BORON DILUTION EVENT The Boron Dilution event is analyzed for Cycle 8 to demonstrate that sufficient time is available for an operator to identify the cause and to terminate an approach to criticality for suberitical Modes 2, 3, and 4 of operation. This event was reanalyzed on the basis of a reduction in shutdown margin for operational Modes 2, 3, and 4 as shown in Table 7.1.1-1. An inadvertent boron dilution adds positive reactivity, produces power and temperature increases, and during operation at power (for Mode 1 and 2) can cause an approach to both the DNBR and CTM limits. Since the TM/LP trip system monitors the transient behavior of core power level and core inlet temperature at power, the TM/LP trip will intervene, if necessary, to prevent the DNBR limit from being exceeded for power increase within the setting of the Variable High Power Level trip. For more rapid power excursions the Variable High Power Level trip initiates a reactor trip. The approach to the CTM limit is terminated by either the Local Power Density (ASI) trip, Variable High Power Level trip, or the DNBR related trip discussed above. The trip which is actuated depends on the rate of reactivity increase resulting from the dilution event. For a boron dilution initiated from hot zero power, critical, the power transient resulting from the slow reactivity insertion rate is terminated by the Variable High Power Level trip prior to approaching the limits. Table 7.1.1-1 compares the values of the key transient parameters assumed in each mode of operation for Cycle 8 and the reference cycle. The conservative input data chosen consists of high critical boron concentrations and low inverse boron worths. These choices produce the most adverse effects by reducing the calculated time to criticality. The time to criticality was determined by using the same expression as in the reference cycle (Unit 1, Cycle 5).. Table 7.1.1-2 compares the results of the analysis for Cycle 8 with those for the Reference Cycle. The key results are the minimum times required to lose prescribed negative reactivity in each operational mode. Modes 2, 3, and 4 - results are more limiting than the reference cycle due to a lower shutdown margin. As seen from Table 7.1.1-2, sufficient time exists for the operator to initiate appropriate action to mitigate the consequences of this event. f C
TABLE 7.1.1-1 KEY PARAMETERS ASSUMED IN THE BORON DILUTION ANALYSIS Unit 1 Unit 1 Parameter Cycle 58 Cycle 8 Critical Boron Concentration, PPM (All Rods Out, Zero Xenon) Startup (Mode 2) 1900 1900 Hot Standby (Mode 3) 1900 1900 Hot Shutdown (Mode 4) 1900 1900 Inverse Boron '4 orth, PPM /t ao Startup 65 65 Hot Standby 55 55 Hot Shutdown 55 55 Minimum Shutdown Margin Assumed,140 Startup -4.0 -3.5 Hot Standby -4.0 -3.5 Hot Shutdown -4.0 -3.5 ' Reference 7. 7-8 '
-_._ a
i TABLE 7.1.1-2 RESULTS OF THE BORON DILUTION EVENT Criterion for Minimum Time to Lose Time to Lose Prescribed Shutdown Prescribed Shutdown l Mode Margin (Min) Martin (Min) Unit 1 Unit 1 Cycle 5 Cycle 8 Startup 69.8 60 15 Hot Standby 59.6 50 15 Hot Shutdown 59.6 50 15 O e z O I m
7.1.3 LOSS OF LOAD EVENT The Loss of Load event is analyzed to demonstrate that the DNBR limit and the RCS pressure upset limit are not exceeded during Cycle 8. This event was reanalyzed due to an increase in the positive moderator temperature coefficient Technical Specifications limit and to an increase in the opening setpoint of l the main steam safety valves. The assumptions used to maximize RCS pressure during the transient are: a) The event is assumed to result from the sudden closure of the turbine stop valves without a simultaneous reactor trip. This assumption causes the greatest reduction in the rate of heat removal from the reactor coolant system, and thus results in the most rapid increase in primary pressure and the closest approach to the RCS pressure upset limit. ,
)
b) The steam dump and bypass system, the pressurizer spray system, and the power operated pressurizer relief valves are assumed not to be operable. This too maximizes the primary pressure reached during the transient. The Loss of Load event was initiated at the conditions shown in Table 7.1.3-1. The combination of parameters shown in Table 7.1.3-1 maximizes the calculated peak RCS pressure. The methods used to analyze this event are identical to those applied in the reference cycle. The initial core average axial power distribution for this analysis was assumed to be a bottom peaked shape. This distribution is assumed because it minimizes the negative reactivity inserted during the initial . portion of the scram following a reactor trip and maximizes the time required to mitigate the pressur A Moderator Temperature Coefficient (MTC) of
+.7X10~g and0 heat flux increases.
despite ao /F was conservatively assumed in this analysis at full power a 10~Jechnical Specification x 10~'ao /"F. An MTC of
+0.7 x do /oF is allowed by thelimit of +0.2 Specification Technical to exist at or below 70% power. This conservative assumption of full power with the maximum positive MTC was used to bound the event. This MTC in conjunction with the -
increasing coolant temperatures, enchances the rate of change of heat flux and ' the pressure at the time of reactor trip. A Fuel Temperature Coefficient (FTC) corresponding to beginning of cycle conditions was used in the analysis. This FTC causes the least amount of negative reactivity feedback to mitigate the transient increases in both the core heat flux and the pressure. The multiplier on the FTC used in the analyses is shown in Table 7.1 3-1. The lower limit on initial RCS pressure less uncertainties is used to maximize the t ate of change of pressure, and thus peak pressure, following t' rip. The Loss of Load event, initiated from the conditions given in Table 7.1.3-1, results in a high pressurizer pressure trip signal at 5.8 seconds. At ,8.8 acconds, the primary pressure reaches its maximum value whien is less than 2750 psia. The increase in secondary pressure is limited by the opening of the main steam safety valves, which open at 6.1' seconds. The minimum main steam safety valve opening setpoint is increased from 1000 psia to 1050 psia. This setpoint change results in a maximum secondary pressure of 1095 psia at 12.1 seconds after the initiation of the event. ( l 7-10 'i
The event was also analyzed to demonstrate that the acceptable DNBR limit is not violated. The minimum transient DNBR calculated for the event is greater than the DHBR SAFDL. Table 7.1.3-2 presents the sequence of events for this event. Figures 7.1.3-1 to 7.1.3-5 show the transient behavior of core power, core heat flux, RCS coolant temperatures, the RCS pressure, and the steam generator pressure. The results of this analysis demonstrate that the Loss of Load event will not result in a DNBR that violates the DNLR SAFDL and that the peak RCS pressures will not exceed the upset pressure limit of 2750 psia. D 7-11 '.-
TABLE 7.1.3-1 ' KEY PARAMETERS ASSUMED IN THE LOSS OF LOAD A'NkLYSIS TO MAXIMIZE CALCULATED RCS PEAK PRESSURE Unit 2' Unit 1 Parameter Units Cycle 5 Cycle 8 Initial Core Power Level MWt 2754 2754 Initial Core Inlet Coolant Temperature F 550 550 Core Coolant Flow X100 lbm/hr 133.9 133.9 Initial Reactor Coolant System Pressure psia 2154 2154 Initial Steam Generator Pressure psia 864 843 Moderator Temperature Coefficient X10-"ao / F +.5 +.7 Doppler Coefficient Multiplier -
.85 .85
.CEA Worth at Trip 54o -4.7 -4.7 Minimum Main Steam Safety Valve psia 1000 1050 Opening Setpoint
' Reference 9.
e l I l l l 7-12 '
TABLE 7.9.3-2 SEQUENCE OF EVENTS FOR THE LOSS OF LOAD' EVENT TO MAXIMIZE CALCULATED RCS PEAK PRESSURE TIME (SEC) SETPOINT , EVENT or VALUE ! 0.0 Loss of Secondary Load - 5.8 High Pressurizer Pressure Trip Signal 2435 psia
- Generated 6.1 Steam Generator Safety Valves open 1050 psia 6.4 Pressurizer Safety Valves open 2500 psia 7.2 CEA's begin to drop into core -
8.8 Maximum RCS Pressure <2750 psia 12.1 Maximum Steam Generator Pressure 1095 psia 12.4 Pressurizer Safety Valves are fully closed 2400 psia D L e e I
- ~
7-13 o
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7.1.4 EXCESS LOAD EVENT. The Excess Load Event was reanalyzed to demonstrate that the SAFDL'S are not violated during Cycle 8. The reanalyses was necessary to include the effects of.a more negative MTC, a lower CEA worth available at trip and changes in the HPSI flow characteristics. The High Power Level and Thermal Margin / Low Pressure (TM/LP) trips provide primary protection to prevent exceeding the DNBR SAFDL during this event. Additional protection is provided by other trip signals. including high rate of change of power, low steam generator water level, and low steam generator pressure. The approach to CTM SAFDL is terminated by either the Local Power Density trip, Variable High Power Level trip or the DNB related trip discussed above. In this analysis, credit is taken only for the action of the High Power trip in the determination of the minimum transient DNBR. The most limiting load increase events at WP (Hot Full Power) and HIP (Hot Zero Power) conditions, for approach to the SAFDL'S, are due to the complete [ opening of the steam dump and bypass valves and the complete opening of the c turbine control valves, respectively. The Excess Load Event at HFP was initiated at the conditions given in Table 7.1.4-1. An MTC of -2.7 X 10- do / F was assumed in this analysis. This MTC, in conjunction with the decreasing coolant inlet temperature, enhances the rate of increase of heat flux at the time of reactor trip. An FTC corresponding to beginning-of-cycle conditions with an multiplier of 0.85 was used in the analysis since this FTC causes the least amount of negative reactivity change for mitigating the transient increase in core heat flux. The L ~ minimum CEA worth assumed to be available for shutdown at the time of reactor trip for full power operation is 4.35 60. The analysis conservatively assumed I that the worth of boron injected from the safety injection tank is -1.00%ao per
- 105 PPM. The pressurizer pressure control system was assumed to be inoperable
! because this minimizes the RCS pressure during the event and therefore reduces j the calculated DNBR. l The HFP Excess Load Event results in a High Power trip at 7.1 seconds. The minimum DNBR calculated for the event at the conditions specified in Table 7.1.4-1 is greater than the design limit of 1.23. For the Excess Load Event initiated from HFP conditions, SIAS is generated at 50.3 seconds at which time the RCP's are manually tripped by the operator. The l coastoown of the pumps decreases the rate _of decay heat removal and therefore !. keeps the RCS coolant temperatures and pressure at higher values. l l Auxiliary feedwater flow is delivered to both steam generators at 187:5 I seconds. The feedwater flow causes . additional cooldown of the RCS. -The decreasing temperatures in combination with a negative MTC inserts positive reactivity which enables the core to approach criticality. The negative
- reactivity inserted due to the CEAs and boron injected via the Hign Pressure Safety Injection (HPSI) pumps however is sufficient to maintain the core suberitical up to 600 seconds when .the operator is assumed' to - terminate the auxiliary feedwater flow to both steam generators.
Table 7.1.4-2 presents the sequence of events for an Excess -Load Event initiated at HFP concitions. Figures 7.1.4 1 to 7.1.4-6 snow the NSSS transient response for power, heat flux, RCS temperatures, RCS pressure, steam . - i generator pressures and reactivities. - [- _ _ ._ _ _ - - _ -A
The Excess Load Event at HZP was initiated at the conditions given in Table 7.1.4-3. The MTC value assumed in the analysis is the same as for the full power case for the reasons previously given. However, the FTC corresponding to beginning-of-cycle conditions with an multiplier of 0.85 was used in the zero power cases since this FTC cause the least amount of negative reactivity change for mitigating the transient increase in core heat flux. The minimum CEA shutdown worth available is conservatively assumed to be -3.5%t.o. The 'results of the analysis show that a variable high power trip occurs at 13.0 seconds. The minimum DNBR calculated during the event is greater than 1.23 and the CTM limit is not exceeded. The sequence of events for the HZP case is presented in Table 7.1.4-4 Figures 7.1.4-7 to 7.1.4-12 show the NSSS transient response for core power, core heat flux, RCS temperature, RCS pressure, steam generator pressures and reactivities. Note that the core remains suberitical at all times after trip for an Excess Load Event initiated from HZP conditions. For both the full and zero power Excess Load Events the DNBR and CTM SAFDL's are not violated. 1 se 8 7-20 ' - -
TABLE 7.1.4-1 KEY PARAMETERS ASSUMED FOR HOT FULL POWER EXCESS LOAD EVENT ANALYSIS Unit 2' Unit 1 Parameter Units Cycle 5 Cycle 8 Init'ial Core Power Level + MWt ?700 2700 Core Inlet Temperature
- O F 548 548 Reactor Coolant System Pressure + psia 2200 2200 Core Mass Flow Rate + X100 1bm/hr 138.4 138.4 Moderator Temperature Coefficient X10 as/0F -2.5 -2.7 CEA Worth Available at Trip % ao -4.3 -4.3 l
[ Doppler Coefficient Multiplier -
.85 .85 Inverse Boron Worth PPM /%Ao 105 105 Auxiliary Feedwater Flow Rate lbm/see 175 175 High Power Level Trip Setpoint % of Full Power 110 107++
t-Low S. G. Water Level Trip Setpoint ft 30.9 30.9
' Reference 9. ,
*For DNBR calculations, effects of uncertainties on these parameters were combined statistically.
++ Temperature decalibration'is being included explicitly.
e e
- 7-21
TABLE 7.1.4-2 SEQUENCE OF EVENTS FOR THE EXCESS LOAD. EVENT AT HOT FULL POWER TO CALCULATE MINIMUM DNBR
- Time (sec) Event Setcoint or Value 0.0 Complete Opening of Steam Dump and Bypass Valves at Full Power _
7.1 High Power Trip Signal Generated 107% of full power 7.5 Trip Breakers Open 8.0 CEA's Begin to Drop Into Core 8.4 Maximum Power 114% of full power 8.9 Minimum DNBR (CE-1) >1.23 9.2 Low Steam Generator Level Trip 30.9 ft. , Setpoint Reached 28.0 Rampdewn of main Feedwater Flow 8% of full power , Completed main feedwater flow 49.4 Pressurizer Empties
~
l 50 3 Safety Injection Actuation Signal 1578 psia Initiated; Manual Trip of RCP's 50.4 RCS Pump Coastdown Begins - 76.2 Main Steam Isolation Signal 548. psia 170.1 Isolation of Main Feedwater Flow ~ to Both Steam Generators 187.5 Auxiliary Feedwater Flow Delivered 175 lbm/see to to both Steam Generators each, steam generator 484.2 Pressurizer Begins to Refill a 600.0 Operator Terminates Auxiliary -2.00% ao
-Feedwater Flow to Both Steam Generators; Total Reactivity e
9
- e *
- - . - , . . , , - _ . - , m __-, . . - - - , - , , - - - , - --.-.c
I TABLE 7.1.4-3 KEY PARAMETERS ASSUMED FOR HOT ZERO POWER EXCESS LOAD EVENT ANALYSIS Unit l' Unit 1 Parameter Units Cycle 6 Cycle 8 Initial Core Power Level
- MWt 1 1 Core Inlet Temperature
- F 532 532 Reactor Coolant System Pressure
- psia 2225 2200 Core mass Flow Rate
- X100 lbm/hr 141.3 141.3 Moderator Temperature X10-"do/ F -2.5 -2.7 Coefficient CEA Worth Available at Trip % do -4.0 -3.5 Loppler Coefficient Multiplier -
.85 .85 Inverse Boron Worth PPM /%ao 100 105 Variable High Power Trip % of full 40 ', 40 Setpoint power
~
' Reference 8.
*For DNBR calculations, effects of uncertainties on these parameters were combined statistically. .
s s 7-23 _ _ _ _
i l' l TABLE 7.1.4 3 SEQUENCE OF EVENTS FOR EXCESS LOAD EVE?iT AT HOT ZERO POWER CONDITIO!IS TO CALCULATE MAXIMUM LHR Time (see) Event Setcoint or Value 0.0 Turbine Admission Valve Opens 120% Steam Flow at Full Power 13.0 variable High Power Trip Signal 40% of full power Generated 13.4 Trip Breakers open - 13.5 Core Power Reaches Maximum; 75.1 % of full power 13.9' CEA's begin to drop into Reactor - Core J f s
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7.2.4 A00's RESULTING FROM THE MALFUNCTION OF ONE STEAM GENERATOR The transient resulting from the malfunction of one steam generator is analyzed for cycle 8 to determine the initial margin that must ' be maintained by the LCO's such that in conjunction with the RPS (asymnetric steam generator protective trip ASGPT), the DNBR and fuel centerline melt design limits are not exceeded. Reanalysis of this event was required due to changes in the Moderator Temperature Coefficient fMTC) and the Main Steam Safety Valve Opening Setpoints. The methods used to andyze these events are consistent with those used in the reference cycle (Unit 1 Cycle 6). The four events which affect a single generator are identified below: 1 Loss of Load to One Steam Generator
- 2. Excess Load to One Steam Generator
- 3. Loss of Feedwater to One Steam Generator 4 Excess Feedwater to One Steam Generator Of the four events described above, it has been determined that the Loss of Load to One Steam Generator (LL/1SG) transient is the limiting asymmetric event. Hence, only the results of this transient are reported.
The event is initiated by the inadvertent closure of a single main steam isolation valve. Upon the loss of load to the steam generator, its pressure and temperature increase to the opening pressure of the secondary safety valves. This analysis assumed the new opening setpoint pressure of 1050 psia for the Main Steam Safety Valves. The intact steam generator " picks up" the lost load, which causes its ' temperature and pressure to decrease, thus causing the core average inlet temperature to decrease and enhancing the asymmetry in the reactor inlet temperature. In the presence of a negative MTC this causes an
- increase in core power and radial peaking on the " cold-side" of the core. The most most negative MTC value of -2.7X10~4ao / F is used in the analysis.
The LL/1SG event results in the greatest asymmetry in core inlet temperature distribution and the most limiting DNBR for the transients resulting from the malfunction of one steam generator. The LL/1SG was initiated at the initial conditions given in Table 7.2.4-1. A reactor trip is generated by the Asymmetric Steam Generator Trip at 3.0 secones based on high differential pressure between the -steam generators. A new differential pressure trip value of 186.0 psis was used to include maximum measurement uncertainty. - t Table 7.2.4-2 presents the sequence' of events for the Loss of Load to One Steam Generator. The transient behavior of key NSSS parameters are presented in Figures 7.2.4-1 to 7.2.4-3. event is greater than the DNBRThe limitminimum of.1.23. transient DNBR calculated for this 1 G 4 7-37
An allowable initial linear heat generation rate of 18.5 KW/ft could exist as an initial condition without exceeding the acceptable fuel to centerline melt of 22 KW/ft during this transient. The actual initial linear heat generation rate will be less since the Linear Heat Rate LCO is based on the more limiting linear heat rate for LOCA (e.g., 15.5 Dl/ft). The event initiated from the extremes of the LCO in conjunction with the ASGPT protective trip will not lead to DNBR or a centerline fuel temperature which exceed the DNBR and centeline to melt design limits. C e
- 4 e
7-33
TABLE 7.2.4-1 KEY PARAMETERS ASSUMED IN THE ANALYSIS OF LOSS OF LOAD TO ONE STEAM GENERATOR Unit l' Unit 1 Units Cycle 6 Cycle 8 j Initial Core Power + MWt 2700 2700 Initial Core Inlet Temperature
- F 548 548 Initial Reactor Coolant psia 2225 2200 System Pressure
- Moderator Temperature Coefficient X10-4ao/ F -2.5 -2.7
- Doppler Coefficient Multiplier -
.85 .85 ASGPT Setpoint psid 175- 186 ' Reference 8. *For DNBR calculations, effects of uncertainties on these parameters were combined statistically.
S e *
, . - - - - ~ . , ,,-...,g -,
---O,., , - , , - - , -.- - . , , -
1 TABLE 7.2.4-2 SEQUENCE OF EVENTS FOR LOSS OF LOAD . TO ONE STEAM GENERATOR TIME (sec) .E7ENT SETPOINT or VALUE 0.0 Spurious closure of a single main - stesa isolation valve 0.0 Steam Flow from unaffected steam operator - increases to maintain turbine power 3.0 ASGPT' setpoint reached (differential pressure) 186 psid
- l 3.9 Trip breakers open -
i 4.4 CEA's begin to insert 4.4 Atmospheric Dump and Byp ss valves open - 6.0 Minimum DN3R occurs >1.23 6.6 Main Steam Safety Valves open on 1050 psia isolated steam generator 10.2 ,, Maximum steam generator pressure 1074 psia
'ASGPT-Asymmetric Steam Generator Protection Tri p D
e _ _ , , , , , _ . . w-~-' '"' * * ' ' ~ '
( '
. H0 i i i i 100 -
) -
$E z 80 -
e R ' E
" 60 -
e2 E E (. [== a 40 20 - - 0 O 10 20 30 40 50 TIME, SEC0(IDS P
, cAs (ALTIP.CPCELEcT.=icca. LOSS. OF LOAD /1 STEAM GE?iERATOR EVEi!T FIGL calver: cittf: CORE PONE? VS TII'.E
([. nuclear power plant 7.2.d 1 7-41 ' b -
( .. 120 i i i i 100 -
. =: -
8 u_ 80 - o th d , S
' 60 -
!:2 u
( < 40 - M S 20 - 0 0 10 20 30 40 50 TIME, SECONDo
!ai,sa$[$$$!ca.
LOSS OF LOAD /1 STEAM GENERATOR EVENT FIC calver: ci f res
.~ CORE AVERAGE HEAT FLUX VS TIME 7,2 L_Muclear Pcuer Plant I 7-42 -
C ~ 700 t- i i l u_ 650 -
.o _
g T OUT E
& I si 600 -
T y AVG 5 l 5 M 550 - M M 3 5 500 - Tin _ e l D x 450 - ' ' ' I i 400 0 10 20 30 40 50 TIIiE, SECONDS - d ! BALT!l<0RE GAS & ELECTRIC CO. LOSS OF LOAD /1 STEA?! GENEPATOR EVE!r k-nu!i}NcSfNant i REACTOR C00LAi!T SYSTEii TEliPERATURES VS TI?E 7.2,- 7-43 .
I C .__ 2400 i i i i 2000 - 5 . C1. . S
= 1600 -
W E
- CL.
5 y 1200 - en
!E 5
(. b 800 - a l D - 5
=
400 - l l 0 l 0 10 20 30 40
. 50
. TIME'l SECONDS L
i s* l AALTI.MOR:
~ -
cAssstscTatico. l ( calver: c1tfis Nuclear Pcuer Plant l LOSS OF LOAD /l STEAM GENERATOR ?!E.31T REACTOR C00LAtlT~ SYSTEM PRESSURE VS' TIME FIGB 7'2'4) l 7-44 '. L i
C 1200 l i I I AFFECTED 1000 -
~
5 E
,, 800 -
W Bi E c y 600 - ' NAFFECTED - R EE u m (,.
- 400 -
e , le 200 _ i i 0 ' ' ' ' O 10 20 30 40 1 ' _ 50 TIME, SEC0flDS . l BALTIMORE ( , as a EtscTatc ca.. LOSS'0F LOAD /1 STEAM GE.'lERATOR EVENT FIac calver: c11frs
?!uclear ?cwer Plant STEAM GENERATOR PRESSURE VS TIME 7*2*4-7-45 '
7.3.1 CEA EJECTION EVENT The zero power case of the CEA Ejection event is analyzed for Cycle 8 to determine the fraction of fuel pins that exceed criteria for clad damage. Reanalysis is required due to increases i- the CEA ejected worth and the post-ejected radial power peak for the zero power case. The analytical method employed in the analysis of this event is the NRC , approved Combustion Engineering CEA Ejection method which is described in CENPD-100-A, (Reference 6). The key parameters used in this event are listed in Table 7.3.1-1. These key parameters are selected to add conservatism to the procedure outlined in Figure 2.1 of Reference 6, which is then used to determine the average and centerline enthalpies in the hettest spot of the h'd rod. The calculated enthalpy values are compared to threshold enthalpy values to determine the amount of fuel exceeding these thresholds. The threshold enthalpy values are: Clad Damage Threshold: Total Average Entnalpy = 200 cal /gm Incipient Centerline Melting Threshold: . Total Centerline Enthalpy = 250 cal /gm Fully Molten Centerline Threshold: Total Centerline Enthalpy = 310 cal /gm To bound the most adverse conditions during the cycle, the most limiting of either the Beginning of Cycle (BOC) or End of Cycle (ECC) parameter values were used in the analysis. A BCC Doppler defect was used since it produces the 1 east amount of negative reactivity feedback tojitigate the transient. A BOC moderator temperature coefficient of +0.7X10 ao /F was used because a positive MTC results in positive reactivity feedback and thus increases coolant temperatures. Although t is also an increase from the previous cycle values of +0.5X10~ge ao/MTC F, value it has only a second order effect on the ' results of the analysis. EOC delayed neutron fractions and neutron kinetics were used in the analysis to produce the highest power rise during the event. The zero power CEA ejection event was analyzed assuming the core is initially operating at 1 MWt. At zero power, a 7ariable Overpower trip is conservatively assumed to initiate at 405 (30% + 10% uncertainty) of 2754 MWt and terminate the event. The zero power case was analyzed, assuming the value of 0.05 seconds for - the total ejection time, which is consistent with the Reference Cycle.- The power transient produced by a CEA ejection initiated for the zero power case is shown in Figure 7.3.1-1. The results of the zero power CEA ejection cm analyzed (Table 7.3.1-2) show that the maximum total energy deposited durtrui the event in the pin is less than both the criterion for clad damage (i.e., 200 cal /gn) and the incipient centerline melt threshold of 250 cal /gm. Consequently, no fuel pin failures are calculated to occur. I- .
TABLC 7.3.1-1 i ^ KET PARAMETERS ASSUMED IN THE CEA EJECTION ANALYSIS ZERO POWER CASE 6 Unit 2* Unit 1 Parameter Units Cycle 5 Cycle 8 4 Core Power Level MWt 1. 1 Ejected CEA Worth ~ t .63 .86 Post-Ejected Radial Power Peak - 9.4 9.5 Axial Power Peak - 1.60 1.64 CEA Bank Worth at Trip Sao -1.5 - 1.5 Doppler Coefficient Multiplier -
. .35 .35 Moderator Temperature Coefficient X10-"ao/ F .5 .7 Delayed Neutron Fraction -
.0044 .0044
- Reference 9.
t l 7-47
- w . , ..e~ w- r 4 - , - +- -, ,--,-m, , m: *,,, .
TABLE 7.3.1-2 CEA EJECTION EVENT RESULTS I Unit 2 Unit 1 Cycle 5 Cycle 8 Zero Power Total Average Enthalpy of 145. <200 Hottest Pellet (cal /gm) Total Centerline Enthalpy of 199. <250 Hottest Fuel Pellet (cal /gm) Fraction of Rods that Suffer Clad 0 0 < Damage (Average Enthalpy 2; 200 cal /gm) Fraction of Fuel Having at least Incipient 0 0 Centerline Melting (Centerline Enthalpy
> 250 cal /gm)
Fraction of Fuel Having a Fully Molten 0 0 Centerline Condition (Centerline Enth,alpy 2; 310 cal /gm) e G 7-43
10.'O' i _ i i _, ZERO POWER
~ ~
. 'CEA EJECTED WORTH E 0.86%so '
8s . CN Ws l', O --
= ..
g .. g .. u .. e: .. . W _ E .. W
- 8 -
0.1 :
. 6 -
~
l03' ' ' ' ' 0 l' 2 3 4 ,5 TIME, SECONDS gGAS & Co. CEA EJECTI0il EVEili FIsuas calvert citfrs . fluclear Pcwer Plant CORE POWER VS TIME, 7 ;3 ' ~1 -1 l
- .1 _ _ _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ - - - _ _ _ - -
7.3.2 Steam Lit.e. Ruoture Analysis The steam line rupture (SLB) event has been reanalyzed for Calvert Cliffs Unit 1 Cycle 8. The purpose of this reanalysis is to extend the conservatively enveloping analysis performed in suppo-t of Unit 1 Cycle 7 to represent Unit 1 Cycle 8. There have been no changes in plant parameters or core characteristics which would impact the pre-trip SLB events; therefore, the results of the Unit 1 Cycle 7 analysis remain valid for this class of breaks, Changes did, however, occur in core characteristics which necessitated the reanalysis of the post trip steam line break. Previous analysis has shown that the consequences of an inside containment post-trip SLB are more adverse than those of the outside containment breaks due to the action of the flow restrictors. The post-i trip inside containment breaks were initiated from both Hot Full Power (HFP) and Hot Zero Power (HZP) and each was performed with and without a Loss of AC (LOAC) power on turbine trip. The acceptance criteria for this postulated accident is that the site boundary doses will be within the l 10CFR100 guidelines and that coolable geometry will be maintained. Analysis Assumotions and Initial Conditions for SLB Inside Containment The SLB event was initiated from the conditions listed in Table 7.3.2-1 The moderator temperature coefficient (MTC) of reactivity assumed in the analysis corresponds to end of cycle, since this MTC results in the , greatest positive reactivity change during the RCS cooldown caused by the steam line rupture. Since the reactivity change associated with moderator 4 feedback varies significantly over the moderator density covered in the analysis, a curve of reactivity insertion versus density rather than a single value of MTC, is assumed in the analysis. The moderator coolcown curve assumed 'in the analysis is given in Figure 7.3.2-1. This moderator l cooldown curve was conservatively calculated assuming that on reactor trip l- the control element assembly which yields the most severe combination of scram worth and reactivity insertion is stuck in the fully withdrawn position. The reactivity change asceiated with the fuel temperature decrease was also based on an end of cycle Doppler coefficient because this fuel temperature coefficient (FTC), in conjunction with the decreasing fuel i temperatures, causes the greatest positive reactivity insertion during the i steam line rupture event. The Doppler multiplier on the FTC assumed in the analysis is given in Table 7.3.2-1 The a fraction assumed was the maximum i absolute value including uncertainties for end of cycle conditions. This too is conservative since it maximizes suberitical multiplication and, thus, enhances the potential for Return-To-Power (R-T-P). The analysis also assumed a conservatively low value of boron reactivity worth of -1.0".a0 per i 85 PPM for safety injection flow from the High Pressure Safety Injection l . pumps. ' l The minimum CEA worth assumed to be available for shutdown at the time of reactor trip is 6.335ao at the maximum allowed power level and 3 . 5 ",ao at zero power. This available scram worth was calculated for tne stuck rod wnich produced the moderator cooldown curve in Figure 7.3.2-1 l T During a return-to-power, negative' reactivity crecit was assuneo in the l analysis.- This negative reactivity crecit is cue to the local neatup of the inlet fluid in the not channel, whien occurs near the location of the ! stuck CEA. This credit is based on three-dimensional coupled neutronic-7-50 '. _ _ . _ _ _ _ _ , - ~ . _ . _ __ _ . ..__ _ ,__ _ __ . . _ _ _ , . _ _ . _ _ _ _ _ _ _ _ _ .
I thermal-hydraulic calculations performed with the HERMITE/ TORC code (References 10 and 11). The magnitude of the credit'is similar to that used in the Calvert Cliffs Unit 1 Cycle 7 steam line break event (Reference 1). 3-D power distribution peaks (Fq) were also determined by the HERMITE/ TORC methodology. The limiting SLB case in this analysis was found to be the Hot Full Power, non-LOAC case. The Fq used for this case corresponded to operation at the extremes of the HFP ASI LCO limits. The actual value of Fq i's a function of both fission power and ccre flow. The analysis only credited the Icw steam generator pressure trip. An analysis trip setpoint of 600.0 psia was assumed in the analysis. This represents the Technical Specification setpoint of 685.0 psia and an uncertainty of 85.0 psia. The analysis also assumed that a Main Steam Isolation Signal (MSIS) is generated when secondary pressure reaches 600.0 psia. This represents the Technical Specification setpoint of 685.0 psia and an uncertainty of 85.0 psia. A Main Steam Isolation Valve (MSIV) closure time of 6.9 seconds (includes valve closure time and signal processing delay time) was conservatively assumed in the analysis. The analysis conservatively assumed that following reactor trip, the main feedwater flow is ramped down to 8% of full power feedwater flow in 20 seconds and that the main feedwater isolation valves are completely closed in 80 seconds after a low steam generator pressure or. a main steam isolation signal. These assumptions are consistent with Technical Specification limits. The analysis assumption; regarding the auxiliary feedwater actuation setpoint, the associated time delays, and the AFW flow through each leg are given below. They were conservatively chosen to initiate AFW flow sooner and deliver the maximum AFW flow to the ruptured steam generator, which maximizes the primary cooldown and enhances the potential R-T-P. The auxiliary feedwater Technical Specification actuation setpoint is 45% of steam generator level wide range indication with an uncertainty of 118%. Auxiliary feedwater (AFW) was conservatively assumed to initiate at time of reaccor trip, which in all cases resulted in AFW init)Lation at a level far above the Technical Specification actuation setpoint plus uncertainties. This was done to ensure the analysis results would remain bounding in the event of any future revision of the Technical Specification or uncertainty. Time delays associated with the AFW pumps were conserva-tively set to zero resulting in instantaneous flow, even;in loss of AC cases. This is consistent with the enveloping nature of the , analysis. All flow from the AFW pumps is conservatively directed to the damaged steam
+
generator until automatic isolation of that steam generator. AFW pump flow is assumed to be at a runout value of 1300 spm. The analysis also included isolation of the ruptured steam generator when the steam generator differential pressure reached the analysis setpoint of 365.0 psid. This represents a Technical Specification setpoint of 135.0 psid and an uncertainty of 230.0 psid. In addition, a 20.0 second time delay was assumed in the analysis to close the AFW isolation (i.e., block) valves. These assumptions are conservative since it delays the isolation of AFW to the ruptured steam generator. 7-51 (
+ %>
l
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A screty injsetion cetustion analysis sstpsint of 1578.0 psia, which is conservative compared. to current Technical Specifications and existing uncertainties, was assumed in the analysis. The analysis conservatively assumed that on a Safety Injection Actuation Signal (SIAS), only one High Pressure Safety Injection (HPSI) pump starts. In addition, a maximum time delay of 30 seconds for HPSI pumps to accelerate to full speed was assumed in the analysis. In case of LOAC power, additional time delays were included in the analysis. It included 10.0 seconds for the diesel generators to start and reach speed following the LOAC and 5.0 seconds for the HPSI pump to be loaded on line regardless of which sequencer (i.e.. shutdown or LOCA) was initiated. l The post-trip minimum DNBRs were calculated using the MacBeth correlation 1 (Reference .12) with the Lee non-uniform mixing correlation factor l (Reference 13). l Results for SLB Inside Containment l l l It was found that the SLB events with Loss of AC (LOAC) power on turbine trip, initiated from either Hot Full Power or Hot Zero Power conditions, do ! not approach the SAFDL on DNBR as closely as did the LOAC cases in the I Unit 1 Cycle 7 analysis; consequently, the Unit 1 Cycle 7 results remain valid. This is due to improved modeling of the coolant flow in the hot t channel under natural circulation conditions based upon the results of the HERMITE/ TORC methodology. This improved flow is used together with the MacBeth correlation in the calculation of the minimum DNBR value. The SL3s without a LOAC do not approach the SAFDL on DNBR. They do, however, present a challenge to the Linear Heat Generation Rate (LHGR) e SAFDL. This occurs because the higher core flow reduces the negative
- reactivity inserted by the local heatup of the moderator in the hot channel. This results in a maximum post trip LHGR. Therefore, the results of the largest inside containment SLB without LOAC on turbine trip are presented herein.
l l' The sequence of events for the 6.305 ft 2 SLS without LOAC on turbine trip initiated from HFP conditions is given in Table 7.3.2-2. The break si::e of 2 6.305 ft was determined in Reference 1 to - be the most' severe. The ! reactivity insertion as a function of time is presented in Figure 7.3.2-1 l The NSSS responses during -the transient are given in Figures 7.3.2-2 l through 7.3.2-7. The results of the analysis show that the HFP SLS causes the secondary
- l. pressure to rapidly decrease until a reactor trip on low steam generator
! pressure is initiated at 2.5 seconds. The CEAs drop into the- core at 3.9 j seconds and terminate the power and heat flux increases. Auxiliary l feedwater is initiated at runout flow to the damaged side steam generatar at time of trip. . At 21.8 seconds one HPSI pump is loaded on line and at 51.8 seconds the HPSI pump reaches full speed. The Steam Generator Isolation Analysis Setpoint is reached at 2.5 .secones. At 3.4 seconds, the MSIVs begin to close and are completely closed at 9.4 seconds. The blowdown from the intact steam generator is terminated . at this. time. l S 7-32
An AFW isolation signal based on steam generator differential pressure is initiated at 8.5 seconds. At 27.9 seconds, the AFW block valves associated with the steam generator with the lowest pressure (i.e., ruptured steam generator) are completely closed. The continued blowdown from the ruptured steam generator causes the core reactivity to approach criticality. The ruptured steam generator blows dry at 67.6 seconds, which terminates the cooldown of the RCS. A peak reactivity of .42%Ao at 70.0 seconds is obtained. A peak R-T-P of 9.29". consisting of 5.36% fission power and 3 935 decay power, is procuced at 66.3 seconds. Less than one percent of the fuel exceeds the centerline melt limit; the DNBR limit is not approached for this event. Conclusions Site boundary doses are bounded by the outside containment doses calculated for Calvert Cliffs Unit 1 Cycle 7 which remain valid. The DNBR results of the steam line break, inside containment of the Unit 1 Cycle 7 analysis continue to envelope the Unit 1 Cycle 8 results. The LHGR results of the steam line break, inside containment predict that less than one percent of the fuel would exceed the centerline melt limit, thus, ensuring a coolable geometry. Therefore, the results of the outside and inside containment SLS events are acceptable for Calvert Cliffs Unit 1 Cycle 8. b e 8 e
- 7-33 - -
TABLE 7.3.2-1 KEY PARAMETERS ASSUMED IN THE INSIDE CONTAINMENT STEAM LINE BREAK EVENT INITIATED FROM HFP Parameter Units Value Initial Core Power MWt 2754.0 Initial Core Inlet *F 550.0 Temperature Initial RCS Pressure psia 2300.0 Initial Steam Generator psia 860.0 Pressure Low Steam Generator psia 600.0 Pressure Trip Setpoint Steam Generator psid 365.0 Differential Pressure Setpoint safety Injection psia 1578.0 Actuation Signal Minimum CEA Worth 5a0 ^
-6.33 Available at Trip s Doppler Multiplier -
1.30 Moderator Cooldewn Curve %ao vs. See Figure density 7.3.2-1 , Inverse Boron Worth PPM /%ao 35.0 l Effective MTC x10~"ac/ F -2.7 3 Fraction (including .0060 uncertainty) l 9 e d k
TABLE 7.3.2-2 2 HFP 6.305 FT BREAK WITHOUT LOAC, INSIDE CONTAINMENT Time (sec) Event Seteoint or Value 0.0 Steau Line Break Occurs 6.305 ft 2 2.5 Low Steam Generator Pressure 600.0 psia Analysis Trip Setpoint is Reached;
" Steam Generator Isolation Analysis Setpoint is Reached; 3.4 Trip Breakers Open; Main Feed Rampdown Begins; MSIVs Begin to Close 3.9 CEAs Enter Core ,
8.5 Steam Generator Differential 365.0 psid Pressure Setpoint Reached 9.4 Main Steam Isolation Valves Fully Closed 21.4 Pressurizer Empties 21.8 Safety Injection Setpoint 1578 psia is Reached 27.9 AFW Block Valves Closed Providing Auxiliary Feedwater to Intact S.G. only. 51.8 Safety Injection Pumps up to l-Full Speed i 66.3 Peak Power 9.29% of 2700 MWt 67.6 Affected Steam Generator Blows , Dry, RCS Cooldown Stopes 70.0 Peak Reactivity .42%ao l 83.4- Main Feedwater Isolation Valves --- Completely Closed 115.0 Safety Injection Baron Begins Appearing in the Core in
- Inch' easing Amounts Ensuring Shutdown I
i- i l .- 7-35
l l l l 10 _. i g .1 _~
~ ~
8 _ _ 7 q : :
<3 6
x 3 HOT FULL POWER -::
~
5 - -- E s u ._- _ 6
=
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. x -
S 3 L 2 x - W : : g 2 _ 1 L l-0 . 2 .
-1 1 2:
~
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2 40 45 50 55 G0 MODERATOR DENSITY, LBM/FT3 BALTIMORE GAS & ELECTRIC CO. STEAll LINE BREAK EVEllT FIG 0RE calvert citffs fluclear Power Plant MODERATOR REACTIVITY VS MODERATOR DENSITY 7.3.2-1 ' 7-54 _ __J
4 1 140 i i i i i i i HOT FULL POWER g 100 - 8 s [ o 80 M f 60 - - E c_ f 40 -
- 20 - -
m 0 0 20 40 60 80 100 120 140 160 TIME, SECONDS l BALTIMORE l l GAS & ELECTRIC CO. STEAM LINE BREAK EVENT INSIDE CONTAINMEf1TFIGURE l calvert cliffs fluclear Power Plant CORE POWER VS TIME j 7.3.2-2, s I 7-57
120 i i i i i i i g 100 / HOT FULL POWER x: 8 - N 80 - ls w A 60 - 5 u h 40 -
!s!
, 20 - w 0 '
- O 20 40 60 80 100 120 140 160 TIME, SECONDS l
l l f I l BALTIMORE GAS & ELECTRIC CO. STEAM LINE BREAK EVENT INSIDE CONTAINMENT FIGURE calvert c1tffs Nuclear Power Plant CORE HEAT FLUX VS TIME - 7.3.2 3 L em
i i b 650 i i i i i i i d T HOT FULL POWER
$ 600 OUT -
3 TAVG W 550 - 5 bi 500 - e 5
$ 450 -
g Tr; i
- C5
$ 400 i i i , ,
0 20 40 60 80 100 120 140 160 TIME, SECONDS BALTIMORE GAS & ELECTRIC co. STEAM LINE BREAK EVENT INSIDE CONTAINMENT FIGURE calvert cliffs Nuclear Power Plant REACTOR COOLANT SYSTEM TEMPEPATURES VS TIME 7.3.2-4 _ _ _ _ _ - - _ - - - _ _ _ _l-3 9
1 t 2500 4 5 , , , , , , , 20 k2000 -
. HOT FULL POWER -
g u I 1500 - - W s2 v>
+ 1000 - -
5 8 g 500 - t3 US
= 0 ' i I ' ' ' I O 20 40 60 80 100 120. 140 1EO-TIME, SECONDS BALTIMORE GAS & ELECTRIC CO. STEAM LINE BREAK EVENT INSIDE CONTAINMENT FIGURE calvert cliffs REACTOR COOLANT SYSTEM PRESSURE VS TIME -
fluclear Power Plant 7.3.2_5 7-60 .
. - . _ _ _ _ - _ . . _ _ _ _ . _ _ ~ . _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ , _ _ _ _ _ . _ _ _ _ _ _
900 i i i , i i i 800 i - HOT FULL POWER . 700
-\ .
20 g 600 E ! !?> W c. 500 1 8 g 400 k - 300 - G " W v> 200 - 100 -
- 0 ' ' ' ' ' ' '
0 20 40 60 80 100 120 140 160
~
TIME, SECONDS GAS & E C CO. STEAM'LIllE BREAX EVENT INSIDE CONTAINMEllT FIGURE caivert c1tfrs STEA!! GENERATOR PRESSURE VS TI!1E Nuclear Power Plant 7.3.2-6 l 7-61 - t
6 , , , , , ,
, MODERATOR HOT FULL POWER 4 -
4 I 2 - DOPPLER d 0 BORON 1 v7 TOTAL - ~
;-- /# N E -2 -
T 3-D - t; C5 cc _4 -
-6 -
SCRAM .
-8 0 20 40 60 80 100 120 140 160 TIME! SECONDS ,
t I' GAS & E C CO. STEAM LINE BREAK EVENT INSIDE CONTAINMEllTFIGURE calvert c11ffs REACTIVITIES VS TIfE Nuclear Power Plant 7.3.2_7 7-62
j 8.0 ECCS ANALYSIS 8.1 Large Break Loss-of-Coolant Accident i : 8.1.1 Introduction and Summary l An ECCS performance analysis was performed for Calvert Cliffs Unit 1 Cycle 8 to descnstrate con;pliance with 10CFR50.46 which presents the NRC Acceptance Criteria for Emergency Core Cooling Systems for Light Water Reactors
- '(Reference 1). The analysis justifies an allowable Peak Linear Heat l Generation Rate (PLHGR) of 15.5 kw/ft. This PLHGR is equal to the existing i limit for Calvert Cliffs Unit 1. The method of analysis and detailed results which support this value are presented in the following sections.
8.1.2 Method of Analysis The NRC approved C-E large break evaluation model(2) was used to
- re-evaluate ECCS performance for the limiting large break LOCA.
blowdown hydraulic calculations employed in the cycle 6 evaluation (ge the reference cycle, apply to Cycle 8 since there have been no significant changes to RCS hardware characteristics. Refill /reflood hydraulic
- calculations were performed for Cycle 8 to account for the lowering of the minimum containment pressure from 14.7 psia to 13.7 psia and for the
- reduction in the minimum HPSI fl These computations were performed l using the NRC approved CCMPERC-IIgw).code. The hot rod clad temperature i and oxidation calculations were performed to account fer removal of the hot j rod augmentation factors, the revised refill /reflood hydraulics.
the fuel rod conditions specific to Cycle 8 g The NRC approved STRIKIN-II 1 code was used for this purpose. Burnup dependent calculations were performed te determine the limiting { conditionsfortheECCgerformanceanalysis. The nuclear and fuel thermal performance (FATES 3) data used as input to the ECCS analysis t considered high burnup effects specific to the Cycle 8 reload. Two STRIKIN- ! II temperature calculations were performed: One at a -low red average l burnup which yields the maximum initial fuel stored energy and one at a i high burnup which yields the highest initial rod pressure. The low burnup l maximum fuel stored energy case was demonstrated to be limiting. I
- The temperature calculations for both cases were performed for the 1.0
- D The break spectrum analysis performed for Unit 1 Cycle
! 2gPD'determined break. that the 1.0 DES /PD is the limiting break since it yields I the highest residual fuel stored energy at the end of the blowdown period and, therefore, yields the highest peak cladding temperature during the , reflood period. [ The NRC approved PARCHC8) code was utilized in the cycle 8 evaluation to ! produce the steam cooling heat transfer coefficients during the late i reflood period when reflood rates are below one inch per second. Previous reload cycle analyses used a more simplified but unnecessarily conservative i constant minimum steam heat transfor value. As a result of this analysis improvement, the peak clad temperatures reported herein are significantly reduced.
- DES /PD - Double Ended Slot at Pump Discharge i
i .. i < S-l
8.1.3 R sults Table 8.1-1 summarizes the results calculated for the two rod average burnup cases selected. A summary of the fuel parameter input values is shown in Table 8.1-2. For comparison purposes, the corresponding values of the reference cycle analysis (Unit 1 Cycle 6) are also presented in Tables 8.1-1 and 8.1-2. A list of the significant parameters displayed graphically for the limiting case (Figures 8.1-1 through 8.1-9) is presented in Table 8.1-3. The decrease in intial containment pressure in the refill /reflood hydraulic analysis resulted in a lower transient minimum containment pressure as shown on Figure 8.1-7. Consequently, slightly lower reflood rates (Figure 8.1-8) and slightly lower heat transfer coefficients (Figure 8.1-5) were calculated in comparison to the reference analysis. The results of the evaluation confirm that 15.5 kw/ft is an acceptable value for the PLHGR in Cycle 8. As shown in Table 8.1-1, the peak clad temperature, maximum local oxidation and core wide clad oxidation values of 1896 0F, 5.25 and <.515, respectively, are well below the 10CFR50.46 acceptance criteria limits of 2200 F, 17% and 1%, respectively. As shown in Table 8.1-1, the burnup with the maximum initial stored energy in the fuel (943 MWD /MTU) resulted in the highest peak clad temperature. The transient results for the limiting case are presented in Figures 8.1-1 to 8.1-9. The fuel cladding is predicted to rupture, as well as achieve its peak temperature, during the reflood period (at 37 seconds and 240 seconds, respectively). The high burnup (50,000 MWD /MTU) case resulted in a ' peak clad temperature of 1887*F, 9F lower than that for the maximum initial fuel stored energy case. This case assumed that the rod operates at a PLHGR of 15.5 kw/ft. This assumption is very conservative since a rod at such a high burnup would actually be at a power level significantly below 15.5 kw/ft. At 15.5 kw/ft the pin pressure was sufficient to cause an earlier hot red rupture time. However, the lower initial stored energy at that burnup causes a lower peak clad temperature than the case at 943 MWD /MTU, as shown in Table 8.1-1. 8.1.4 conclusions As discussed above, conformance to the ECCS criteria is summarized by the analysis results presented in Table 8.1-1 The most limiting case results in a peak clad temperature of 1896 0 F,. which is well below the acceptance limit of 2200 0F. The maximum local and core wide values for zirconium oxidation percentages, as shown in Table 8.1-1, remain well below the acceptance limit values of 17% and 11, respectively. Therefore, operation of Unit 1 Cycle 8 at a PLHGR of 15.5 kw/ft and a power level of 2754 MWT (102% of 2700 MWT) results in compliance with the 10CFR50.46 acceptance , cetteria. 8-2 9
Table 8.1-1
SUMMARY
OF ECCS PERFORMANCE RESULTS FOR CALVERT CLIFFS 1 CYCLE 8 FOR THE LIMITING BREAK SIZE (1.0 DES /PD) Limiting Case High Burnup Case (Maximum Initial Fuel (Maximum Initial Parameter Stored Energy) Rod Pressure) Unit 1 Unit 1 Unit 1 Unit 1 Cycle-6 Cycle-8 Cycle-6 Cycle-8 Rod Average Burnup, MWD /MTU 3000 943 34,000 50,000 Peak Clad Temperature 2038 1896 2024 1887 (PCT),'F Time of PCT, seconds 249 240 248 237 Time of clad rupture, seconds 31.9 36.5 10.6 10.3 Peak local oxidation, % 8.5 5.20 8.4 5.31 Core wide oxidation, % 4 0.51 <0.51 < 0.51 4: 0.51 6
=
8-3
Table 8.1-2 CALVERT CLIFFS 1 CYCLE 8 FUEL PARAMETERS Unit 1 Unit 1 Cycle-6 Cycle-8 Quantity Value Value Reactor Power Level (102% of Nominal) (Mwt) 2754 2754 Average Linear Heat Rate (102% of Nominal) (kw/ft) 6.44 6.37 Hot Channel Peak Linear Heat Generation Rate (kw/ft) 15.5 15.5 Hot Assembly Peak Linear Heat Generation Rate (kw/ft) 13.14 12.52
- Gap Conductance at PLHGR (Btu /hr-ft2 *F) 2025 2447
- Fuel Centerline Temperature at PLHGR (*F) 3634 3586
- Fuel Average Temperature at PLHGR (*F) 2213 2160
- Hot Rod Gas Pressure (Psia) 1251 1216
- Hot Rod Burnup (MWD /MTU) 3000 943 Hot Rod Augmentation Factor (Maximum) 1.04 1.00
- Initial fuel rod parameters, in STRIKIN-II, which yield the limiting ECCS performance results (maximum peak cladding temperature).
O P 8-4 , , 4
Table 8.1-3 CALVERT CLIFFS 1 CYCLE 8 ANALYSIS PLOTS FOR LIMITING CASE Variable Figure Number Peak Clad Temperature 8.1-1 Hot Spot Gap Conductance 8.1-2 Peak Local Clad Oxidation 8.1-3 Temperature of Fuel Centerline, Fuel Average, 8.1-4 Clad and Coolant at Hottest Node Hot Spot Heat Transfer Coefficient 8.1-5 ' Hot Rod Internal Gas Pressure 8.1-6 Containment Pressure 8.1-7 Mass Added to Core During Reflood 8.1-8 Water Level in Downcomer 8.1-9 t 0 3-5 - I
FIGURE 8.1-1 CALVERT CLIFFS I C CLE 8 1.0 x D0l;BLE ELIDED SLOT BREAK Ill PufiP'LISCliAfGE LEG PEAK CLAD TEMPERATURE 2200 2CCC O 1800 '^ '. . N
/ -
/ N -
16CC j cl 'I , ~ l \ & j\ l\ l 1 ' x N~ L~'~~
~.
g 14CC, !i l , , 5 ! Il I PEAK CLAD TEPERATURE f4EE
;I "N y l --- RUPTURE PODE l
w M d
- 12CC
, !lI l
' l l
1CCC i 8CC i l 6CC 100 2CC 3CC 4CC ECC SCC 7r TIME. SECCtlCS 8-6
FIGURE 8.1-2 CALVERT CLIFFS I C CLE 8 1.0 x DOUBLE EllDED SLOT BREAK Ifi PUMP DISCHARGE LEG liOT SPOT CAP C0llDUCTAllCE ISCC 14CC 120Cl i i f o' I I, ' c-{ 1 C C C , Y m v i i 8CC S E O o S SCC a u t, Q 5 \ l
- 4. C C ',y 1 -
I t 2C0 C I 100 2CC 3CC 400 5C0 600 7f
"^
7.iME, SECCNCS
FIGURE 8l-3 CALVETTCL[fFSIC'YCLE8 1.0 x DOUBLE ENDED SLOT BREAK IN PUhP DISCliARGE LE PEAK LOCAL CLAD OXIDATIO'N 16 14 i 12 iC , , PEAK CLAD TE'FERATURE PCEE " --- RUPTURE t0DE
- s S 8 e
e 5 % 6 d 4 .
/,' -
/
/
/ .
2 /
/ /
/
~
A' i i ,l
!CC 2CC 3CC ACC 500 SCC 7C TIME. SECCNCS -
8-8
~~
FIGURE 8.1-4 CALEEfRTCLIFFSICYCLE8 1.0 x DOUBLE ENDED SLOT BREAK IN PL'fiP DISCllARGE LEG TEtiPERATURE OF FUEL CENTERLIllE, FUEL AVERACE, CLAD AND C00LAi1T OF HOTTEST [10DE 35CC 3CCC
\ -
25CC i . E' ' g 2CCC ; F ' AVERAGE FUEL' i y (; ,
~ ^
' FUEL CEllTERLIllE Y >j'!ky l ' l
~
@1500 CLAD l u ICCC ,
o l 5CC i C00LAf1T
\ /\
4 100 2CC 3CC 4CC 500 SCC 70 TIM 5. S5CCNCS 3-9
~
- FIGURE 8'l-5 CALVERT CLIFFS I CYCLE 8 1.0 x DOUBLE EiiDED SLOT BREAK Ill PUNP DISCl!ARGE LEE HOT SPOT HEAT TRAllSFER COEFFICIEliT 150 14C c'
ol 12C C i 4 i
~ .x :
if l m .
, ICC i 15 2
5d I y v SC I ts iii E. SC , 4 u . 4C 20 f
\,_
e , 100 2CC 3CC 400 5CC SCC 7 C' TIME. SECCNCS '~.
FIGURE 8,1-6 I CALVERTCLIFFSUii[TICYCLE8 1.0 x DOUELE ENDED SLOT BREAK Ifi PUMP DISCliARGE LEE fl0T R0D INTERNAL GAS PRESSURE 1400 , , , ,
= . PSIA 1200 Ihi m L _
1000 - _ RUPTURE AT 36.5 SEC c_ 800 - - J E a y C00 - - c. 400 - - 200 - - 0 O 20 40 E0 80 100 TII;E, SEC0 IDS s-tt .,
FIGURE 8','l-7 CALVERT CLIFFS UNIT I CYCLE 8 1.0 x D0bBLE EllDED SLOT BREAK IN PUMP DISCHARGE LEG CONTAINMEllT PRESSURE E0 , , , , 50 - 40 - 5 E d 30 -- 5 0
? -
20' - 10 - 0 0 80 160 240 320 400 TIME AFTER CONTACT, SEC
.. O
~
FIGURE 8.1-8 CALVERT CLIFFS I CYCLE 8-1.0 x DOUBLE ENDED SLOT BREAK II; PUFiP DISCliAREE LEG FASS ADDED TO CORE DURIliG REFLOOD i 120000 TIME, SEC REFLC00 RATE 100000 0.0 - 10.0 2.520 IN/SEC - 10.0 - 91.1 1.242 IN/SEC 91.1 - 400.0 0.766 IN/SEC u-5 8c000 < 8 fM 60000 M 5 40000 [ 20000 ! c- 8 8 8 8 8* 9 e a e a 8 S el M " TIME RFTER CONTRCT, SEC e --
~
FIGURE 8,'l-9 CALVERT CLIFFS I CYCLE 8' 1.0 x DOUBLE ENDED SLOT BREAK Ill PUMP DISCHARGE LEG WATER LEVEL Ill DOWl1 COMER e 18 000 y, ,__ _,_::::=.
, =
1o.000 1 12 000 i. t y 9 000 g - a 6.000 [ 3 000 e 8 k k k 5 I 9 5 8 ? E 5 C - OJ M v
. TIME RFTER CONTRCT, SEC W
i i 8.2 Small Break Loss-of-Coolant Accident 8.2.1 Introduction and Summary The ECCS performance evaluation for the small break loss-of-coolant I accident (LCCA) for Calvert Cliffs Unit 1 Cycle 8, presented her ' demonstrates conformance with the acceptance criteria of 10CFR50.46g,. 4 The evaluation demonstrates acceptable small break LOCA ECCS perfor:re.See at a peak linear heat generation rate (PLHGR) of 15.5 kw/ft and a reactor power level of 2754 Mwt (102% of 2700 Mwt) with an assumed small reduction in high pressure safety injection flow capacity. Injection flow from one ! charging pump was assumed in this analysis. A revised axial power distribution which is intended to envelope present and future cycles was
- used. It is expected that the analysis presented herein will apply directly to other cycles of both Calvert Cliffs units.
8.2.2 Melacd of Analysis Reference 9 presented the Calvert Cliffs small break LOCA ECCS performance at a reactor power level of 2611 Mwt (102% of 2560 Mwt) and a PLHGR of 15.8 kw/ft. In reference 9, the 0.1 ft break in the reactor coolant pump discharge leg was identified as the worst small break si::e. 10, the'0.1 ft 2 break was reanalyzed at a reactor power level In Reference of 2754 Mwt (102% of 2700 Mwt) and a PLGHR of 16.0 kw/ft, hereafter referred to as the reference analysis. This lctter analysis resulted in a peak clad temperature and a peak local clad oxidation percentage of 1940 0F and 7.965, reagectively. The evaluation presented herein is a re-analysis of the 0.1 ft break at a reactor power level of 2754 Mwt with an assumed ' reduction in high pressure safety injection flow capacity. The calculation was performed using Combustion Engineering's NRC approved Small Break Evaluation Model as described in References 9 and 11. Evaluation af small, break transients involves the use of the following com codes. Blowdown hydraulics are calculated using the CEFLASH-4ASgr code. Fuel rod temper and ci calculated using the STRIKIN-IIgres and PARCHg oxidation codes. percentages are Details of the interfacing of these codes are discussed in Reference 11. The significant core and system parameters which changed frem the reference analysis are listed in Table 8.2-1. The peak linear heat rate assumed was 15.5 kw/ft which is the current maximum allowed by the Specification. Technical In addition to the assumed high pressure safety injection flow, the flow from one charging pump is credited. The charging pump delivers flow to two cold legs. For breaks in a reactor coolant pump discharge leg, it is assumed that all injection flow delivered to the broken leg spills out the break. After accounting for instrument error and applying a conservatively determined flow split, credit is taken for a minimum of 13 gpm delivereo to the intact cold leg. To permit the assumed reduction in high pressure safety injection flow, the allowable axial shape index is being reduced in the Tech. Specs. from .15 to .10 ASIU (See Section 9.0). Increased low pressurizer pressure setpoints for reactor trip and safety injection actuation relative to the reference analysis were credited. This latter change is based on the present Technical Specification values which were increased after the reference analysis subaittal. (The values shown in the table account for measurement uncertianty under accident conditions). The Main Steam Safety Valve Setpoint Technical Specification change discussed in Table 9-1 has been incorporated in this analysis. ,
Tha tttcl ficw from ena high prcssura safety inj ction pump assum;d in this analysis is given in Table 8.2-2. The reduction in high pressure safety injection pump flow will allow greater flexibility .in the surveillance testing of these pumps. 8.2.3 Results The results of the ECCS performance analysis for the 0.1 ft 2 break are summarized in Table 8.2-3. The peak clad temperature is 1877 F with a peak local clad oxidation percentage of 4.91% and a core-wide oxidation percentage less than 0.632%. The results are comparable or slightly lower than those of the reference analysis. The important transient parameters which have been plotted as a function of time (Figure 8.2-1 through 8.2-8) are listed in Table 8.2-4 8.2.4 Conclusion A reanalysis of the ECCS performance at an assumed reduction in high pressure safety injection pump flow capacity for the worst small break LOCA has been performed for Calvert Cliffs Unit 1 Cycle 8 at a reactor power level of 2754 Mwt and a PLHGR of 15.5 kw/ft. The analysis demonstrated a peak clad temperature of 1877 F and a peak local clad oxidation percentage of 4.91%, thgby demonstrating appreciable margin relative to the Acceptance Criteria for the worst small break LOCA. Therefore, it can be concluded that operation of Calvert Cliffs Units 1 Cycle 8 at the reactor power level of 2754 Mwt is acceptable. It is expected that this conclusion will apply directly to other cycles of both Calvert Cliffs units. D o S-16
I Table 8.2-1 System Parameters and Initial Conditions Calvert Cliffs l Value Parameters Present Analysis Reference Analysis i l Peak Linear Heat Rate 15.5 kw/ft 16.0 kw/ft Charg ing Pump Minimum Flow 13 gpm delivered none to intact leg Axial Shape Index (Most .16 ASIU .21 ASIU Negative Value Including Uncertainty) Low Pressurizer Pressure Setpoints Reactor Trip 1741 psia 1728 psia Safety Injection Actuation 1591 psia 1578 psia 6 8-17
Table'8.2-2 Calvert Cliffs Units 1 and 2 HPSI Pump Flow for Small Break Analysis 3 RCS Present Analysis Reference Analysis Pressure, osig Flow, GPM Flow, GPM 1276.0 0.0 0.0 1266.3 0.0 35.0 1250.0 69.50 104.5 1200.0 145.50 180.5 i 1150.0 188.25 223.25 1100.0 225.25 261.25 1050.0 254.75 289.75 1000.0 278.50 313.50 ! 900.0 321.25 356.25 800.0 368.75 403.75 700.0 402.0 437.0 [ , 600.0 435.25 470.25
~ 500.0 473.25 508.25 l 300.0 535.0 570.0 0.0 - 606.25 641.25 l
e
,b l
;c. . .
L a-is
1 Table 8.2-3 Times of Interest and Fuel Rod i Performance Summary for 0.1 ft Break Time for HPSI Pump On 56 sec. , Time for LPSI Pump and SI Tanks On a Time Hot Spot Peak Clad Temperature Occurs 1650 sec. Maximum Clad Surface Temperature 1877*F Elevation of Hot Spot (from Bottem of Core) 10.8 ft Peak Local Clad Zirconium Oxidation 4.92% Core Wide Zirconium Oxidation 40.632% l l a ~ Calculation terminated before LPSI pumo or SI tank actuation.
'~
8-19 -
\; , Table 8.2-4 Variables Plotted as a Function of Time Calvert Cliffs Units 1&2 0.1 ft2/PD Variable Ficure No.
Normalized Total Core Pcwer 8.2-1 Inner Vessel Pressure 8.2-2 Break Flow Rate 8.2-3 Inner Vessel Inlet Flow Rate 8.2-4 Inner Vessel Two-Phase Mixture Volume 8.2-5 Heat Transfer Coefficient at Hot Spot 8.2-6 Coolant Temperature at Hot Spot 8.2-7 Clad Surface Temperature at Hot Spot 8.2-8 9
' ~
S-20
FIGURE 8.2-1 0.1FTjALVERTCLIFFSLNITSIANDII COLD LEG BREAK AT PULP DISCHARGE NORMALIZED TOTAL CORE POWER , (SEALL BREAK AllALYSIS) . 1.2
/
1.0 5 Bi c_ 0.8 i ~ U E3 a
;5 0.0 E2 -
S Z? a HE 0.4 l 8
==
0.2 k l 0.0 0 10 20 30 40 50 TIME, SEC em 6 . , , . . . , - - . . - , . - - - - - ' - - '
FIGURE 8.2-2 0.1FT{ALVERTCLIFFSUNITSIANDII COLD LEG BREAK AT PLflP DISCHARGE INNER VESSEL PRESSURE (SMALL BREAK ANALYSIS) 2400 2000 5
!C N 1600 u
c_ i l 0 1200 W 5 - i 5 800 N e* 0 0 500 1000 1500 2000 2500 TIfiE, SEC 3-22 ..- . .
FIGURE 8.2-3 0.1FT{ALVERTCLIFFSUNITSIANDII COLD LEG BREAK AT PUMP DISCliARCE BREAK FLOW RATE (SMALL BREAK ANALYSIS) 2400 2000 S 1600 R 5 h c:: 1200 5 Y d s 800 i 400 l - 0 O 500 1000 1500 2000 2500 l TIME, SEC l l 8-23
FIGURE 8.2-4 . jALVERTCLIFFSUNITSIANDII 0.1 FT COLD LEG BREAK AT PUMP DISCHARGE INNER VESSEL INLET FLOW RATE (SMALL BREAK ANALYSIS) 35000 28000 21000 8
~
s2 a . ! .J 14000 E 5 d 7000 w 0 l l l -7000 l 0 500 1000 1500 2000 2500 l
- TIME, SEC 3-24
FIGURE 8.2-5 jALVERTCLIFFSUNITSIANDII 0.1 FT COLD LEG EREAK AT PUMP DISCHARGE INI1ER VESSEL TWO-PHASE MIXTURE VOLUME (SMALL BREAK ANALYSIS) 6000 i 5000 m 4000
$ 3000 E
. E .
w
-E
=
c- 2000 h~ ' TOP OF CCRE w - 1000 , BOTICti 0F CORE 0 0 500 1000 1500 2000 2500 TIME, SEC s-25
- _ _ . . _ ~ -_. _ . _ , _ _ . _ . _ _ _ .._..__ _ _ . . _ _ . _ - . _ _ _ _
FIGURE 8.2-6 0.1 FT{ALVERT COLD LEG BREAK ATCLIFFS UNITS I AND II PUMP DISCHARGE HEAT TRANSFER COEFFICIENT AT HOT SPOT (SMALL BREAK ANALYSIS) 100,000 . w l'- 10,000 . b b Eii ~ EE p 1,000 _ l = : u - h l 5 100 :- Os : l -
\
= 10 i E 5
1 1 l 0 500 1000 1500 2000 2500 TIME, SEC L __ _ 3
- 2 6 __ _ .. _ _ _
Freuas 8.2-7
- 0.1FT{ALVERTCLIFFSUNITSIAi4DII COLD LEG BREAK AT PUMP DISCHARGE COOLANT TEMPERATURE AT HOT SPOT i (SMALL BREAK AllALYSIS) 1200 1000 p 800 id
# \
. b 600 .
=
yv 400 200 O O 500 1000 1500 2000 2500 TIME, SEC 8-27 -
FIGURE 8.2-8 gALVERTCLIFFSUNITSIANDII 0.1 FT COLD LEG BREAK AT PUMP DISCHARGE CLAD SURFACE TEMPERATURE AT HOT SPOT (SMALL BREAK ANALYSIS) 2200 s 1900 C
, 1600 s
E i g 1300 .
$ f s " /
1000 m \ 700 400 0 500 1000 1500 2000 2500 TIME, SEC
~
3-28}}