ML20210S989
| ML20210S989 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 02/06/1987 |
| From: | Tiernan J BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20210S994 | List: |
| References | |
| NUDOCS 8702180015 | |
| Download: ML20210S989 (64) | |
Text
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i BALTIMORE GAS AND ELECTRIC CHARLES CENTER P. O. BOX 1475
- BALTIMORE, MARYLAND 21203 o
JOSEPH A.TIERNAN VICE PntstDENT NucLtAn ENERGY
~.
t i February 6,1987 ,
d U. S. Nuclear Regulatory Commission
! Washington, DC 20555 ATTENTION: Document Control Desk l
[ SUB3ECT: Calvert Cliffs Nuclear Power Plant l Unit No. 21 Docket No. 50-318 Request for Amendment s 3
Eighth Cycle License Application l
REFERENCES:
(a) Letter from C.H. Polndexter (BG&E), to E.3. Butcher, Jr. (NRC), -
dated August 30,1985, Request for Amendment to Operating License DPR-69, Seventh Cycle License Application ,
Gentlemen:
The Baltimore Gas and Electric Company hereby requests an Amendment to its Operating License No. DPR-69 for Calvert Cliffs Unit No. 2 to allow operation for an
! eighth cycle. The enclosure presents a detailed description of the required Standard i
Technical Specifications with supporting safety analysis information to ensure conservative operation at a rated thermal power of 2700 MWth for Unit 2 Cycle 8.
Our present intention is to begin the Unit 2 refueling outage on March 14,1987, and to complete the outage and begin the approach to criticality on May 12, 1987, with return l to power operation immediately thereaf ter. -
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Unit 2 Cycle 8 is the first twenty-four month cycle for the Calvert Cliffs Nuclear Power ,
Plant. The design for Unit 2 Cycle 8 uses a low leakage core with previously utilized fuel ,,,
assemb!!es composing the core periphery. This configuration, plus new fuel enrichments w,f of 4.08 w/o, provide an expected cycle length of approximately 20,000 MWD /T. ,
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, s . Document Contr i Desk p' February 6,1987 ~ .a
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The majority of the Unit 2 Cycle 8 reload is enveloped by the analyses of the reference cycle ( Unit 2 Cycle 7, previously found acceptable by the NRC (Reference (a)).
Seventeen design basis events weib reviewed to determine the effect the Unit 2 Cycle 8 core design has on the safety analyses. Only two events, the Boron Dilution Event and the Control Element Assembly (CEA) Election Event, were not bounded by the reference cycle analyses, r . .
The Boron Dilution Event was anatWed for Unit 2 Cycle 8 to demonstrate that sufficient "J time is available for an operator't'o identify the cause and to terminate an approach to criticality for operating MODES 2, 3, and 4 considering the increased initial boron concentration of the 24-month cycle. MODES 1,5, and 6 are bounded by the reference cycle analyses. Analysis results proved sufficient time exists for mitigation of the consequences of the event.
'The'CEA Ejection Event was analyzed to establish generic values of CEA ejection worth and the post-ejected radial power peak for the zero power case. The most adverse cycle conditions were'used in the analysis and results yielded were acceptable in that the maximum total energy deposited in the hot spot of the fuel pin during the event is less than both the criterlon for clad damage and the fully molten centerline threshold. No l < fuel pin failures are calculated to occur. Less than one percent of the fuel reaches the j s . Inciplent centerline melt threshold.
PROPOSED CHANGES (BG&E FCR 87-3000)
The Technical Specification changes requested herein are to make the Calvert Cliffs i , Unit 2 Technical Specifications consistent with the analyses presented in the attachment
.s to this Unit 2 Cycle 8 license submittal. These Technical Specification change requests
! are summarized as follows:
l 1. Modify Figure 2.2-1, page 2-11, as indicated in the enclosed attachment to reduce
) the Acceptable Operation Region between 70% and 100% power. This modification is required to accommodate the implementation of the 24 month, low leakage cycle.
l 2. 0 200 F, from I For 3.5% TSdelta3/4.1.1.1, page k/k to 4.5% delta k/k.3/4 The 1-1, change shutdown marginthe shutdown revision a
s margin, T @p> or
, Line Rupture Analysis for the low-leakage core design. With the low leakage core, t
and due to the increased scram worth available, the Steam Line Rupture Analysis
- results were less limiting than those previously reported. Consequently, this analysis l 1s not included. l l l
- 3. For TS 3.1.1.4, 3/4 1-5, change the positive Moderator Temperature Coefficient (MTC) pagelimit to be bounded by the limit line of new Figure 3.1-1 A. The positive MTC limit above 70 % power is being raised to accommodate implementation of the 24-month cycle, eliminate start-up delays and to facilitate a rapid power ascension program.
, 4. F'or B 3/4.1.1.1 and B 3/4.1.1.2, page B 3/41-1, change the end of cycle shutdown
- margin, T requestedE >the same reasons as stated for TS 3/4.1.1.1.200 F, from 3.5% delta k I
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Document Central Desk February 6,1987 Page 3 DETERMINATION OF SIGNIFICANT HAZARDS We have determined, based on the analytical informatYon supplied in the enclosure, that this amendment does not involve a significant hazards consideration. No fuel assemblies to be loaded into the Cycle 8 core will be of new or different design than those used previously and found to be acceptable to the NRC. No proposed changes to the Technical Specifications for Cycle 8 involve acceptance criteria which are significantly different from those previously found acceptable to the NRC. The analytical methods used to determine conformance with the Technical Specifications and regulations are consistent with previous NRC approvals and involve no significant changes.
We conclude that the proposed reload license amendment does not involve a significant hazard consideration in that operation of the facility in accordance with the proposed y amendment would not:
- 1. involve a significant increase in the probability or consequences of an accident previously evaluated; The probability or consequences of an accident evaluated is not significantly increased. Seventeen design basis events were reviewed and three non-LOCA design basis events were reanalyzed. Only two design basis events were not bounded by the results of the previously accepted reference cycle. Their consequences were not significantly increased and all were within acceptance criteria. Both the large break loss of coolant accident and the small break loss of coolant accident were reanalyzed and their consequences were found to be bounded by the results of the reference cycle and somewhat improved.
OR, ii. create the possibility of a new or different type of accident from any accident previously evaluated; The proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.
l OR, iii. involve a significant reduction in a margin of niety.
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The proposed amendment does not involve a significant reduction in the margin j for safety. Changes in the shutdown margin and increase in Moderator ;
Temperature Coefficients are compensated by the CEA reconfiguration and ;
, the reanalysis performed in the enclosure.
SAFETY COMMITTEE REVIEW These proposed ~ changes to the Technical Specifications and our determination of significant hazards have been reviewed by our Plant Operations and Off-Site Safety Review Committees, and they have concluded that implementation of these changes will not result in an undue risk to the health and safety of the public.
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6 Document Contr@l Desk February 6,1987 Page 4 '
FEE DETERMINATION Pursuant to 10 CFR 170.21, we are including BG&E Check No. 1113029 in the amount of
$150.00 to the NRC to cover the application fee for this request.
Very truly yours, STATE OF MARYLAND:
- TO WIT:
CITY OF BALTIMORE :
Joseph A. Tiernan, being duly sworn states that he is Vice President of the Baltimore Gas and Electric Company, a corporation of the State of Maryland; that he provides the foregoing response for the purposes therein set forth; that the statements made are true :
and correct to the best of his knowledge, information, and belief; and that he was '
authorized to provide the response on behalf of said Corporation. l WITNESS my Hand and Notarial Seal: W h Notar Pub
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My Commission Expires: M /, /jf f#
3AT/DSE/Imt f f Wate ,
l Enclosures cc: D. A. Brune, Esquire
- 3. E. Silberg, Esquire A. C. Thadani, NRC S. A. McNeil, NRC T. E. Murley, NRC T. Foley/D. A. Trimble, NRC T. Magette, DNR l
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24-57a(84H2)/cp-1 ATTACHMENT TO B-87-006 CALVERT CLIFFS UNIT 2 CYCLE 8 LICENSE SUBMITTAL l
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24-57a(84H2)/cp-2 Calvert Cliffs Unit 2 Cycle 8 License Submittal i
Table of Contents Section j 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 J
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24-57a(86H2)/cp-3
1.0 INTRODUCTION
AND
SUMMARY
This report provides an evaluation of design and performance for the operation of Calvert Cliffs Unit 2 during its eighth fuel cycle, at full rated power of 2700 MWt. All -planned operating conditions remain the same as those for Cycle 7.- However, Cycle 8 will.be the first "24-month cycle" at Calvert Cliffs. It will also be the first Calvert Cliffs cycle to use a low-leakage fuel management. pattern.
The core will consist of 129 presently operating Batch J and H assemblies and 88 fresh Batch K assemblies.
Plant operating requirements have created a need for flexibility in-the Cycle 7 termination point. This need has been ' met by' using a -
Cycle 7 window ranging from 11,300 MWD /T to 13,300 MWD /T in the Cycle 8 analyses. However, subsequent to the completion of most of the Cycle 8 analyses, an agreement was reached between BG&E and the NRC, concerning a proposed Technical Specification change, which restricts the Cycle 8 MTC at HFP, equi'iibrium Xe conditions to negative values (Reference 1). To satisfy this requirement it is necessary that the Cycle 7 burnup be at least 12,000 MWD /T. This restriction is not constraining since it is presently certain that Cycle 7 will actually be shutdown well above 12,000 MWD /T. However, i this document will cite 11,300 MWD /T as the early Cycle 7 shutdown i point to assure consistency with the burnup range used in the analyses.
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 tennination point falls within the above discussed cycle burnup range used in the analyses. The analysis presented herein will accommodate a Cycle 8 length which varies from 20,000 to 21,500 MWD /T, depending upon the Cycle 7 shutdown burnup, including a coastdown in inlet temperature to 537 F and a coastdown in power to approximately 75%.
The evaluations of the reload core characteristics have been conducted with respect to the Calvert Cliffs Unit 2 Cycle 7 safety analysis described in Reference 2. Unit 2 Cycle 7 will hereafter be referred to as the " reference cycle" in this report, unless otherwise noted. This i.; the appropriate reference cycle because its design / safety basis is the one most recently reported to the NRC.
Specific core differences have been accounted for in the present analysis. In all cases, it has been concluded that either the reference 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 existing plant Technical Specifications are provided and are justified by the analyses discussed herein.
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~ 24-57a(86H2)/cp-4 The Cycle 8 analyses used the same methodology as the reference cycle in all areas except three.~ First, the large break LOCA analysis (Section 8.1)~ used the amended C-E large break evaluation model presented in Reference 3 and recently approved by the NRC in Reference 4.- Second, thermal performance data (Section 4.3). was generated using FATES 3B (Reference 5) which is a revised version of the FATES 3 fuel evaluation model (Reference 6). FATES 3B has received interim NRC approval (Reference 7). Third, fine mesh-pin-by-pin data was calculated by the MC Code.(Reference 8) in place of PDQ.
The performance of Combustion Engineering 14x14 fuel at extended burnup is discussed in Reference 9, which was approved in Reference
- 10. For Cycle 8 the batch average discharge will be considerably less than the 45,000 MWD /T criterion of that reference, but the burnup of approximately 0.2% of the fuel pins will be above the 52,000 MWD /T point discussed in Reference 9, if Cycle 7 and-8 are operated to their maximum burnups. However, since all Cycle 8 analyses address fuel exposure explicitly and the power levels of the few high burnup pins are low, the safety analyses documented herein are appropriate and valid for Cycle 8. Futhermore, the maximum . burnup (approximately 54,000 MWD /T) is well below the burnups which have been or are projected to be achieved in lead assembly demonstrations at Calvert Cliffs (appr'oximately 64,000 MWD /T).
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24-57a(86H2)/cp-5 2.0 OPERATING HISTORY OF THE PREVIOUS CYCLE Calvert Cliffs Unit 2 is presently operating in its seventh fuel cycle utilizing Batch J, H, G, E and D fuel assemblies. Calvert Cliffs Unit 2 Cycle 7 began operation on December 8, 1985 and reached full power on December 16, 1985. The Cycle 7 startup testing was reported to the NRC in Reference 1. 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 March 14, 1987. As of January 9, 1987, the Cycle 7 burnup had reached 11,642 MWD /T. If the -plant capacity remains high for the remainder of Cycle 7, it is probable that the high end of the burnup window (13,300 MWD /T), upon which the Cycle 8 analyses are based, will be exceeded. If any revisions to this document become necessary due to this potential extension of Cycle 7, they will be forwarded immediately.
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'24-57a(86H2)/cp-6 3.0 GENERAL DESCRIPTION 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 will be the switch from 18-month conventional fuel management to 24-month low-leakage fuel management. This change will entail the removal. of 88 irradiated assemblies (3 Batch H*, 48 Batch G, 28 Batch G/, 8 Batch E and 1 Batch D assemblies) and their replacement with 88 fresh assemblies at 4.08 wt% U-235 enrichment (16 unshimmed Batch K assemblies, 44 8-shimmed Batch K/ assemblies and 28 12-shimmed Batch K* assemblies).
Figure 3-1 shows the fuel management pattern to be employed in Cycle
- 8. Figure 3-2 shows the locations of the fuel and poison pins within the fresh K* and K/ shimed assemblies. This fuel management pattern will accommodate Cycle 7 termination burnups from-11,300 MWD /T to 13,300 MWD /T.
The Cycle 8 core loading pattern is 90' rotationally symmetric.
That . is, if one quadrant of the core were rotated 90' into 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 11,300 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 32,700 MWD /T and the average discharge exposure is approximately 41,000 MWD /T. The end of cycle burnups are based on a Cycle 7 length of 13,300 MWD /T and a Cycle 8 length of 20,000 MWD /T.
3.1 PROTOTYPE CEA The PROTOTYPE CEA is described in Reference 1. Cycle 3 was the first cycle of irradiation for this CEA. During the E0C-4 and E0C-5 outages this CEA was examined, as described in References 2 and 3, respectively. This PROTOTYPE CEA was utilized in the center core position from Cycles 3 through 6 and then shifted to a new position for Cycle 7.
3.2 CENTER CEA COMPOSITION The composition of the center CEA, which is part of the lead -bank (Bank 5), is being changed for Cycle 8. This modification is being made to support the change from 18-month conventional fuel management to 24-month low-leakage fuel management for Cycle 8.
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24-57a(86H2)/cp-7
, The specific change which will be made involves the replacement of the present very weak center CEA, which is composed of five Al 0 3 fingers, with a weak CEA, which will be composed of four outer Al 0 ;
fingers and a center Bg C finger. .This alteration will counter c$
the effect of the change to low-leakage fuel management which would have resulted in an increase in radial power peaking if the lead bank had not been modified. See Figure 3-5 for a description of the composition of the present and revised center CEA.
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24-57a(86H2)/cp-8 -
3 TABLE 3-1 4 CALVERT CLIFFS UNIT 2 CYCLE 8 CORE LOADING Total Initial Number Total Initial Poison Poison of Poison Number of Fuel Assembly Number of Enrichment Batchggynup(MWD { Rods Per loaffng and Non-Fuel Designation Assemblies (wt% U-235) B0C8 E0C8 Assembly (gms B / inch) Rods Rods K 16 4.08 0 17,200 0 0 0 2816 K* 28 4.08 0 24,600 12 .036 336 4592 i K/ 44 4.08 0 24,500 8 .036 352 7392
- J(1) 40 4.05 9,800 33,300 0 'O O 7040 J*(1) 20 3.40 14,100 36,900 0 0 0 3520 H(1) 48 4.05 24,800 41,200 0 0 0 8448 H*(1) 21 3.40 28,000 47,100 0 0 0 3696 Total '217 11,300 32,700 688 37,504 4
(1) Carried over from Cycle 7 to Cycle 8 of Unit 2. i . (2) Cycle 7 burnup of 11,300 MWD /T (3) Cycle 7 burnup of 13,300 MWD /T and Cycle 8 burnup of 20,000 MWD /T +
..-._._.__....___.____.._-__m__. . _ _ _ _ _ _ _ _ _ _- 1 y =
KEY X - sox NUMBER Y -sarca 1 2 H K 3 4 5 6 7 H K J K* J 8 9 10 11 12 13 i H K/ J K/ H J* 14 15 16 17 18 19 20 H K/ J K* H* K/ -J* 21 22 23 24 25 26 27 28 H K/ J J H* K/ J* K* 29 30 31 32 33 34 35 36 K J K* H* K* H K/ H* 37 38 39 40 41 42 43 44 J K/ H* K/ H J H J 45 46 47 48 49 50 51 52 53 K* H K/ J* K/ H K* J* 54 55 56 57 58 59 60 61 62 J J* J* K* H* J J* H* l l N L I CO. CALVERT CLIFFS UNIT 2 CYCLE 8 FIGURE CALVERT CLIFFS CORE MAP 3-1 NUCLEAR POWER PLANT 3-4
X X X X l X X ! X X l l i 8 - SHIM K/ ASSEMBLY : l X. X l 4 X X i i X X i l X X l X X X X a 12 - SHIM K* ASSEMBLY BALTIMORE CALVERT CLIFFS UNIT 2 CYCLE 8 FIGURE MS & ELECTRIC 00. FUEL AND SHIM LOCATIONS CALVERT CLIFFS IN FRESH SHIMMED ASSEMBLIES 3-2 NUCLEAR POWER PLANT 3-5
i i , l KEY ! i X -sATcH l XYXYX - INITIAL ENRICHf1ENT l XYXYXY - BOC BURNUP (MWD /T) I H 2 K 4.05 4.06 , 27,500 0 I j 3 H 4 K 5 J 6 K" 7 J ; 4.05 4.08 4.05 4.05 4.05 26,400 0 0,100 0 10,700 l 8 H 9 K/ 10 J 11 K/ 12 H 13 Ja i 4.05 4.05 4.05 4.06 4.05 3.40 26,400 0 9,600 0 23,500 14,200 i 14 H 15 K/ 16 J 17 K" 15 H" 19 K/ 20 J* 4.05 4.08 4.05 4.08 3.40 4.08 3.40 l 26,200 0 9,000 0 26,000 0 13,900 21 H 22 K/ 23 J 24 J 25 Ha 26 K/ 27 Ju 28 K" 4.05 4.08 4.05 4.05 3A0 4.06 3.40 4.06 26,500 0 9,000 11,500 26,300 0 14.300 0 29 K 30 J 31 K" 32 H* 33 K" 34 H 35 K/ 36 Ha 4.08 4.05 4.08 3A0 4.08 -4.05 4.06 3A0 O 9.600 0 27,500 0 23,100 0 26.600 37 J 38 K/ 39 Ha 40 K/ 41 H 42 J 43 H 44 J j 4.05 4.08 3A0 4.06 4.05 4.05 4.05 4.05 45 H 8,000 0 27,900 0 23,100 11.600 22,200 10,700 4.05 l 27,500 46 K" 47 H 48 K/ 49 Ja 50 K/ 51 H 52 K" 53 Ja 4.05 4.05 4.06 3AO 4.05 4.05 4.00 3AO 54 K 0 22,400 0 14.300 0 22,300 0 14,000 4.06 0 55 J 56 J' 57 J" 58 K" 59 Ha 60 J 61 Ja 62 H* 4.05 3A0 3A0 4.08 3A0 4.05 3A0 3A0 10.700 14,200 13,900 0 20,600 10,700 14,000 26,500 EOC 7= 11,300 MWD /T BALTIMME CALVERT CLIFFS UNIT 2 CYCLE 8 F10VRE GAS & ELECTRIC CO. ASSEMBLY AVERADE BURNUP AT BOC 3-3 CALVERT CLIFFS AND INITIAL ENRICHMENT DISTRIBUTION NUCLEAR POWER PLANT 3-6
KEY XY - 6ATcH XYXYX - BURNUP 1 H 2 K 38,100 16,700- . 3 H 4 K 5 J 6 K" 7 J 37,000 17,700 28,500 22.700 34,000 8 H 9 K/ 10 J 11 K/ 12 H 13 J' 37,900 2I.000 34.200 25.700 45,300 36,t 00 14 H 15 K/ 16 J 17 Ka 18 Ha 19 K/ 20 J' 37,700 21,300 33,900 25,700 47,500 26,000 37.200 21 H 22 K/ 23 J 24 J 25 H" 26 K/ 27 J" 28 K" 37,000 21,000 33.900 35,700 47.000 25.400 37,000 20,300 29 K 30 J 31 K" 32 H* 33 K* 34 H 35 K/ 36 H" 17.700 34.200 25.700 46.300 24.500 45.300 25.800 47,900 37 J 38 K/ 39 H* 40 K/ 41 H 42 J 43 H 44 J
- H 28,500 25.800 47,500 25,500 45,400 35.800 44,000 34.600 j 38,100 46 K" 47 H 48 K/ 49 J" % K/ 51 H 52 K" 53 J8 l I 1 54 22.800 44.400 37,800 K 26.100 25.900 44.100 24,200 35.300 a
16,700 55 J 56 J' 57 J" 58 K" 59 Ha 60 J 61 J" 62 H* 34,000 36,100 37,200 26,300 47,900 34,600 35.300 43,900 EOC7=13,300 MWD /T EOC8:20,000 MWD /T I BALTIMORE GAS & ELECTRIC CO. CALVERT CLIFFS UNIT 2 CYCLE 8 F10VRE ASSEMBLY AVERA0E BURNUP AT EOC (MWD /T) 3-4 N RP PWT 3-7
1 l CYCLE 7 CYCLE 8 i O -^'2 o 3 l h -B 4 0 1 BALTit10RE GAS & ELECTRIC CO. CALVERT CLIFFS UNIT 2 FIGURE : CALVERT CLIFFS CENTER CEA FOR CYCLES 7 AND 8 3-5 NUCLEAR POWER PLANT 3-8 l
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4.0 FUEL SYSTEM DESIGN 4.1 MECHANICAL DESIGN 4.1.1 Fuel Design The mechanical design for the Batch K reload fuel is identical to that of the Batch J fuel described in the reference cycle submittal (Calvert Cliffs Unit 2 Cycle 7, Reference 1) except some Batch K fuel assemblies contain burnable poison rods. The design of these poison rods is essentially the same as the poison rods that operated in the reference cycle in the Batch G/ assemblies. The only changes in the design are the poison loadings (gms B-10/in) and the use of hollow spacers rather than solid spacers. The mechanical design of the Batch H fuel assemblies was described in Reference 2. 4.1.2 Dimensional Changes All fuel assemblies in Cycle 8 were reviewed for shoulder gap clearance using the SIGREEP model described in Reference 3 (approved in Reference 4) and for fuel assembly length clearance using the . refined correlation discussed in References 5 and 6. All clearances were found to be adequate for Cycle 8. 4.1.3 CEA Design The replacement CEA to be utilized for the change discussed in Section 3.2 will have virtually the same reconstitutable feature as those replacement CEAs installed in the reference cycle except all five fingers of the replacement CEA for Cycle 8 are reconstitutable. 4.1.4 Metallurgical Requirements The metallurgical requirements of the fuel cladding and the fuel assembly structural members for the Batch K fuel are identical to those of the reference cycle. Thus, the chemical or metallurgical performance of the Batch K fuel will be the same as that of the Unit 2 Cycle 7 fuel. i The metallurgical requirements for fuel cladding and assembly structural members have not changed for many years and are not anticipated to change in the future. Consequently, this sub-section , will be deleted from future license submittals unless there is a change in the metallurgical requirements.
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i 4.2 HARDWARE MODIFICATIONS TO MITIGATE GUIDE TUBE WEAR All fuel assemblies 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 used in the Batch H fuel assemblies and their effect on reactor operation is contained in 1 4-1
24-57a(86H2)/cp-10 Reference 7. A modified short sleeve design (Reference 8) was used in Batch J of the reference cycle. This new short sleeve design will also be used in the Batch K fuel assemblies. The short sleeve design discussed in Reference 8 is considered to be the permanent modification to the basic fuel assembly design for the purpose of mitigating guide tube wear. This design will be used for all future Calvert Cliffs reloads as further modifications are not anticipated. Consequently, this sub-section will -be deleted from future license submittals unless there is a further change. 4.3 THERMAL DESIGN The thermal performance of a composite, standard fuel pin which envelopes the various fuel assemblies present in Cycle 8 (Batches K. J and H) has been evaluated using the FATES 3B version of the fuel evaluation model (References 9, 10, and 11). FATES 3B has received interim NRC approval (Reference 12). 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 was in excess of that expected at end of Cycle 8. 4-2
24-57a(86H2)/cp-11 5.0 NUCLEAR DESIGN 5.1 PHYSICS CHARACTERISTICS 5.1.1 Fuel Management The Cycle 8 fuel management employs a low-lea kage pattern as-described in Section 3, Figure 3-1. The fresh Batch K fuel is comprised of three sets of assemblies, all having the same enrichment of 4.08 'wt% U-235 but each containing a unique number of shims in order to minimize radial power peaking. There are '16 unshimmed assemblies, 44 assemblies with 8 shims and 28 assemblies with 12 shims. With this loading, the Cycle 8 burnup capacity for full power operation is expected to be between 19,100 MWD /T and 20,400 MWD /T, depending on the final Cycle 7 termination point. The Cycle 8 core characteristics have been examined for Cycle 7 terminations between 11,300 and 13,300 MWD /T and limiting values established for the safety analyses. The loading pattern (see ; Section 3) is applicable to any Cycle 7 termination point between i the stated extremes. Physics characteristics including reactivity coefficients for Cycle 8 the are listed in cycle reference Table(Reference 5-1 along )with the corresponding
- 1. Please note that thevalues valuesfrom 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.
Table 5-2 presents a summary of CEA shutdown worths and reactivity allowances for the end of Cycle 8 zero power steam line break accident and a comparison to reference cycle data. The E0C zero power steam line break accident was selected since it is the most limiting zero power transient with respect to reactivity requirements and, thus, provides the basis for verifying the Technical Specification required shutdown margin. Table 5-3 shows the reactivity worths of the three CEA groups which l are allowed in the core during critical / power conditions. These ! reactivity worths were calculated at full power conditions for Cycle j 8 and the reference cycle. The composition of the center CEA, which is part of Bank 5, is being changed as described in Section 3.2. The power dependent insertion limit (PDIL) curve is the same as that l of the reference cycle. 5.1.2 Power Distributions Figures 5-1 through 5-3 illustrate the all rods out (AR0) integrated radial power distributions at B0C8, M0C8 and E0C8, respectively, that are characteristic of the high burnup end of the Cycle 7 shutdown window. The high burnup end of the~ Cycle 7 shutdown window tends to increase the integrated 1-pin radial power peaking. The 5-1 l
~
. 24-57a(86H2)/cp-12 a
integrated radial power distributions with CEA Group 5 fully ! inserted at beginning and end of Cycle 8 are shown in Figures 5-4 i and 5-5, respectively, for the high burnup end of - the Cycle 7 shutdown wirdow. 1 The radial power ' distributions described in this- section are calculated data without uncertainties or other allowances. However, , - the single rod power peaking values do include'the increased peaking. i that is characteristic of fuel rods adjoining the water holes. in the - l , fuel assembly lattice. For both DN3 and kw/ft safety and setpoint analyses in either rodded or. unrodded configurations, the _ power . peaking values actually used are higher than those expected to occur i at any time during Cycle 8. . These conservative values, which are i used in Section 7 of this document, establish the_ allowable limits. ! for power peaking to be observed during operation. '! i The range of allowable axial peaking is defined by the Limitin 1 Conditions for Operation (LCOs) covering Axial Shape Index (ASI)g- . 1 Within these. ASI limits, the necessary DNBR and kw/ft margins are i , maintained for a wide 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 l 1 1.90, not including uncertainty allowances, a 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 are the safety analysis values. The data for the full power condition remain unchanged relative to the reference cycle. However, the data for the zero power condition s were revised in order to perfonn a generic analysis. The data for both power conditions are conservative (highly so for the zero power condition) with respect to actual calculated values. 5.1.3.2 Dropped CEA Data The Cycle 8 safety related data for this section are identical to the safety related data used in the reference cycle. 5.2 ANALYTICAL INPUT TO IN-CORE MEASUREMENTS I In-core detector measurement constants to be used in' evaluating the 2 reload cycle power distributions will be calculated by the coarse mesh code ROCS and the fine mesh code MC, as described in References 2 and 3. The use of ROCS and MC .in place of PDQ, which had been
; used previously, is consistent with the change in design procedure j described in Section 5.3.
5-2
.----, -~--- . - -,---
24-57a(86H2)/cp-13
.5.3 NUCLEAR DESIGN METHODOLOGY-Analyses have been performed with the coarse mesh code ROCS and the fine mesh code MC (Reference 3). ROCS was used in the Cycle 8 analyses in the same manner.that it was used in the . reference' cycle analyses. MC replaces PDQ as the calculator of pin-by-pin data and was _ used in a very similar manner. to the' way PDQ was used in the reference cycle analyses. .5.4 UNCERTAINTIES IN MEASURED POWER DISTRIBUTIONS The power distribution measurement- uncertainties to be applied to Cycle 8 are the same as those applied to the reference cycle.
i 1 I 5-3
-. . = . . . . - _ _ _ _ _ _ _ _ - _ _ _ _
24-57a(86H2)/cp-14 TABLE 5-1 CALVERT CLIFFS UNIT 2 CYCLE 8 NOMINAL PHYSICS CHARACTERISTICS Reference Cycle Unit 2+ Units (Unit 2 Cycle 7) Cycle 8 Dissolved Boron Hot Full Power, All Rods Out, Equilibrium Xenon Boron Content for y Criticality at BOC PPM 1245 1490 Baron Worth Hot Full Power BOC PPM /%Ao 107 121 Hot Full Power E0C PPM /%Ap 87 87 Moderator Temperature Coefficient Hot Full power, Equilibrium Xenon, CEAs Withdrawn Beginning of Cycle 10-4ap/'F -0.1 *
+0.05
- End of Cycle 10-4ap/*F -2.3 -2.3 Doppler Coefficient Hot Zero Power BOC 10 ap/*F -1.57 -1.65 Hot Full Power B0C 10 ap/ F -1.25 -1.31 Hot Full Power E0C 10 ap/*F -1.44 -1.58 TotalEffectiveDejayed Neutron Fraction, eff BOC 0.00593 0.00629 E0C 0.00514 0.00516 Neutron Generation Time, t*
BOC 10- sec 23.1 20.5 E0C 10- sec 28.4 28.6
+BOC8 data were calculated using the early Cycle 7 shutdown burnup of 11,300 MWD /T; E0C8 data were calculated using the late Cycle 8 shutdown burnup of 13,300 MWD /T and a Cycle 8 length of 20,000 MWD /T.
"A value of 1090 ppm was reported in Reference 1; this value corresponded to a late Cycle 6 shutdown burnup of 13,800 MWD /T. The revised value of 1245 ppm corresponds to an early shutdown burnup of 12,200 MWD /T and is !
consistent with the other Unit 2 Cycle 7 data and the Unit 2 Cycle 8 data.
- This value corresponds to a Cycle 7 shutdown burnup of 11,300 MWD /T. At a l Cycle 7 shutdown burnup of 12,000 MWD /T, the B0C8 HFP MTC would be slightly ;
negative. It is presently certain that Cycle 7 will shut down well above : 12,000 MWD /T (see Section 1.0). I l l 5-4
24-57a(86H2)/cp TABLE 5-2 CALVERT CLIFFS UNIT 2 CYCLE 8 LIMITING VALUES OF REACTIVITY WORTHS AND ALLOWANCES FOR THE END-0F-CYCLE (E0C) HOT ZERO POWEP, (HZP) STEAM LINE RUPTURE ACCIDENT, %Ao Unit 2 Reference Cycle
- Cycle 8
- 1. Worth of all CEA's Inserted 8.7 9.4 l
- 2. Stuck CEA Allowance 2.1 1.7
- 3. Worth of all CEA's less Worth 6.6 7.7 I of CEA Stuck Out** ,
- 4. Power Dependent Insertion 1.8 2.0 Limit CEA Bite at Zero Power
- 5. -Calculated Scram Worth 4.8 5.7
- 6. Physics Uncertainty plus 0.6 0.7 Bias
- 7. Net Available Scram Worth 4.2 5.0 l
- 8. Technical Specification 3.5 4.5 Shutdown Margin
- 9. Margin in Excess of Technical 0.7 0.5 Specification Shutdown Margin -
- Unit 2 Cycle 7.
** Stuck CEA is one which yields worst results for E0C HZP SLB, i.e., worst l combination of scram worth and reactivity insertion with cooldown.
l 5-5
24-57a(86H2)/cp-16 TABLE 5-3 CALVERT CLIFFS UNIT 2 CYCLE 8 REACTIVITY WORTH OF CEA REGULATING GROUPS AT HOT FULL POWER, %Ao Beainning of Cycle End of Cycle Regulating Reference
- Unit 2 Reference
- Unit 2 CEA's Cycle Cycle 8 Cycle Cycle 8 Group 5 0.27+ 0.36++ 0.37+ 0.40**
Group 4 0.79+ 0.79 0.84+ 0.80 Group 3 0.86+ 0.86 0.97+ 1.03 l Note Values shown assume sequential group insertion. i
- Unit 2 Cycle 7
+ Unbiased values were reported in Reference 1. Revised biased values are reported herein.
++The composition of the center CEA, which is part of Bank 5, is being modified for Cycle 8 as described in Section 3.2.
l 4 1 l 5-6
__- . _. - . . - . - - =. 24-57a(86H2)/cp-17 TABLE 5-4 CALVERT CLIFFS UNIT 2 CYCLE 8 CEA EJECTION DATA
- 7 Limitino Values Reference Cycle Unit 2 Cycle 8 ;
Safety Analysis Value* Safety Analysis Value - 1 Maximum Radial Power Peak-Full power with Bank 5 inserted; worst CEA
- ejected 3.6 3.6
, Zero power with t Banks 5+4+3'
- inserted; worst CEA ejected 8.9 10.5 i
1 Maximum Ejected CEA Worth (%Ap) Full power with i Bank 5 inserted;
- worst CEA ejected 0.28 0.?8 Zero power with i Banks 5+4+3 inserted; worst CEA ejected 0.77 1.00 t
- Unit 2 Cycle 7 (Reference 1) l 1
Note l Uncertainties and allowances are included in the above data. , i 2
- 5-7 !
- = _ . . - . . __. .- .. - -. .. - ._ _ _ _ .
1 2 0.38 0.83 3 4 5 6 7 0.39 0.93 0.96 1.09 1.08 ) 8 9 10 11 12 13 0.43 1.05 121 1.29 0.97 0.99 14 15 16 17 18 19 20 0.43 1.07 124 126 0.83 129 1.06 l 21 22 23 24 25 26 27 28 0.40 1.06 1.24 1.16 0.79 123 1.05 127 29 30 31 32 33 34 35 36 0.93 122 1.26 0.80 1.16 1.00 129 0.83 37 38 39 40 41 42 43 44 45 0.97 1.31 0.85 125 1.01 1.16 1.01 1.13 l 1 0.38 i 46 47 48 49 50 51 52 53
.I 1.06 1.30 1.01 1.18 0.94 54 ;
- O.83 55 56 57 58 59 60 61 62 1.0d 0.99 1.06 127 0.83 1.13 0.94 0.73 NOTE: X= MAXIMUM 1-PIN =1.56 i
BALTIMORE GAS & ELECTRIC CO CALVERT CLIFFS UNIT 2 CYCLE 8 FIGURE
! CALVERT CLIFFS ASSEMBLY RELATIVE POWER DENSITY AT BOC 5-1 NUCLEAR POWER PLANT 5-8 i
1 2 0.42 0.83 3 4 5 6 7 OA2 0.88 0.94 1.14 1.05' 8 9 10 11 12 13 0A6 1.05 1.13 129 0.97 0.97 I 14 15 16 17 18 19 20 OA6 1.06 1.15 1.29 0.88 1.31 1.05 l 21 22 23 24 25 26 27 28 OA2 1.05 1.15 1.10 0.84 128 1.06 1.33 29 30 31 32 33 34 35 36 0.88 1.13 129 0.84 124 1.00 129 0.86 l l 37 38 39 40 41 42 43 44 45 0.94 129 0.88 128 1.00 1.10 0.98 1.06 OA2 46 47 48 49 50 51 52 53 54 1.14 0.99 1.32 1.06 1.30 0.98 1.21 0.94 55 56 57 58 59 60 61 62 1.05 0.97 1.05 1.33 0.86 1.08 0.94 0.77 X i NOTE: X = MAXIMUM 1-PIN =1.53 l BALTIMORE GAS & ELECTRIC CO. CALVERT CLIFFS UNIT 2 CYCLE 8 FIGURE ! CALVERT CLIFFS ASSEMBLY RELATIVE POWER DENSITY AT I IK MWD /T 5-2 NUCLEAR POWER PU NT 5-9
-e,- g.- - - - -..-p g e-. e m m e- P ----
"aw*
. ,. y . .
(. .,~. l
's A
-.+
,~
s 1 2 - OA7 0.88 - i l 3 4 5 6 7 % 0.46 0.59 0.% 123 [1.d7 I
! A 8 9 10 11 12 13 ;'
O.51 1.11 1.12 1.30 0.97 0.95 - 14 15 16 17 16 19 20 0.51 1.12 1.14 1.33 0.89 126 1.00 . 1 X - 1 l -: e ', j 21 22 23 24 25 26 27 26 s J , OA6 1.11 1.14 1.07 0.85 127 1.01 1.31 1 29 30 31 32 33 34 35 36 . 0.85 1.12 1.33 0.66 126 0.96 124 0.53 37 38 39 0.% 40 41 42 43 44 /) 4 129 0.69 1.27 0.96 1.02 0.92 ^ l .00 .m " l OA7 ' 46 47 48 49 50 51 52 53 ' 54 122 0.96 126 1.01 124 0.92 120 0.9d " 0.se 55 56 57 58 59 60 61 62 1.07 0.95 1.00 1.31 0.63 1.00 0.90 0.76 "'* NOTE: X = MAXIMUM I-PIN =1.51 , BALTIMORE ' GAS & ELECTRIC CO' CALVERT C_lFFS UNIT 2 CYCLE 8 FIGURE CALVERT CLIFFS ASSEMBLY RELATIVE POWER DENSITY AT EOC 5-T' NUCLEAR POWER PLANT
- . ^^<
5-10 J-
< 1 CEA BANK 5 LOCATION s
1 2 0.33 0.70 1 3 4 5 6 7 OA2 0.95 0.92 0.92 0.83 f 8 9 10 11 12 13 OA7 1.13 125 126 0.50 OAS
~
14 15 16 17 15 19 20
. 0A7 1.17 1.34 1.33 0.84 120 0.91 ,
' _ 21 22 23 24 25 26 27 26 l 0A2 1.13 1.34 125 0.05 1.30 1.07 127 l 1
s 29 30 31 32 33 34 35 36 ) 2 0.95 126 1.33 0.86 126 1.08 1.38 0.90 X j 37 38 39 40 41 42 43 44 4 0.93 127 0.65 1.31 1.09 1.27 1.10 124 0.33 46 47 48 49 50 51 52 53 54 0.93 0.83 121 1.06 1.@ 1.11 IM 1.00 d.70 55 56 57 58 59 60 61 62 0.83 OA6 0.91 127 0.90 124 1.00 0.69 NOTE: X = MAXit1Uri 1-PIN = l.66 BALTIMORE CALVERT CLIFFS UNIT 2 CYCLE 8 F10URE GAS & ELECTRIC CO. ASSEMBLY RELATIVE POWER DENSITY WITH BANK 5 CALVERT CLIFFS 5-4 INSERTED, HFP,BOC NUCLEAR POWER PLANT l I 5-11 2 __ _ _
\
CEA SANK 5 ' LOCATION 1 2 OA2 0.75 t 3 4 5 6 7 0A9 0.89 0.90 1.04 0.82 8 9 10 11 12 13 0.57 1.18 1.14 1.24 0.80 OA4 14 15 16 17 18 19 20 0.57 1.21 1.21 1.39 0.91 1.19 0.87 X 21 22 23 24 25 26 27 28 OA9 1.18 1.21 1.15 0.93 1.34 1.04 1.32 29 30 31 32 33 34 35 36 0.89 1.14 1.39 0.94 I.38 1.05 1.33 0.91 37 38 39 40 41 42 43 44 45 0.90 1.24 0.94 1.34 1.05 1.13 1.03 1.12 OA2 46 47 48 49 50 51 52 53 1.04 0.81 1.19 1.04 1.33 1.03 1.32 0.98 54 55 56 57 58 59 60 61 62 0.82 OA4 0.87 1.32 0.91 1.12 0.98 0.69 NOTE: X = MAXIMUM l-PIN = 1.60 BALTIMORE CALVERT CLIFFS UNIT 2 CYCLE 8 FIGURE GAS & ELECTRIC CO. ASSEMBLY RELATIVE POWER DENSITY WITH BANK 5 5-CALVERT CLIFFS INSERTED, HFP,EOC NUCLEAR POWER PLANT 5-12
'24-57a(86H2)/cp-18 l
l 6.0 THERMAL HYDRAULIC DESIGN 6.1 DNBR ANALYSIS Steady state DNBR analyses of Cycle 8 at the rated power level of-2700 MWt have been performed using the TORC ' and CETOP computer codes, the CE-1 critical heat flux correlation and simplified modeling methods, as described in References 1 through 4 and approved in Reference 5. Table 6-1 contains a list- of pertinent thermal-hydraulic design parameters applicable to both the safety analyses and the generation of reactor protective system setpoint information. The calcula-tional 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 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 statistically derived DNBR limit was reduced from 1.23 to 1.21 for the reference cycle analyses (Unit 2 Cycle 7) and for the Unit 2 Technical Specifications in Reference 7. The 1.21 SCU DNBR limit includes a 0.006 DNBR rod bow penalty which accounts for the adverse effects of rod bowing - on CHF for 14x14 fuel with burnup not exceeding 45 GWD/T. The Unit 2 Cycle 8 thermal hydraulic analyses use the reduced SCU DNBR limit of 1.21. , 6.2 EFFECTS OF FUEL B0 WING ON DNBR MARGIN The effects of fuel rod bowing on DNB margin for Calvert Cliffs Unit 2 Cycle 8 have been evaluated using the ' NRC approved methods described in Reference 8. Based upon these methods, a penalty of 0.006 DNBR units is required to account for the adverse T-H effects of rod bow at an assembly average burnup of 45 GWD/T. This penalty is included in the presently used DNBR limit of 1.21, as discussed-above. For Unit 2 Cycle 8 thirty-two assemblies are projected to have average burrups greater than 45 GWD/T (see Figure 3-4). An examination of the power .versus burnup histories of these high burnup assemblies indicates that power peaking at burnups above 45 GWD/T is significantly lower than that of other assemblies, providing ample margin to offset increased rod bow penalties for assemblies with burnups greater than 45 GWD/T. 6-1 l
57a(86H2)/cp-19 TABLE 6-1 CALVERT CLIFFS UNIT 2 THERMAL-HYDPAULIC PARAMETERS AT FULL POWER
- 4 Reference ** Unit 2 General Characteristics Unit Unit 2 Cycle 7 Cycle 8 Total Feat Output (core only) MWg 2700 2700 10 BTU /hr 9215 9215 Fraction of Heat Generated .975 .975 In Fuel Rod Primary ystem Pressure psia 2250 2250 (Nominal 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 Core 106 lb/hr 138.5 138.5 Hydraulic Diameter ft 0.044 0.044 (nominal channel) 6 2 Average Mass Velocity 10 lb/hr-ft 2.59 2.59 Pressure Drop Across Core psi 11.1 11.1 (steady state flow irreversible P 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 BTV/hr-ft 180,700*** 182,900**** (Accounts for above fraction of heat generated in fuel rod and axial densification factor)
?
Total Heat Transfer Area ft 49,700*** 49,100**** (Accounts for axial densification factor) 2 Film Coefficient at Average BTU /hr-ft - F 5930 5930 Conditions 6-2
_ ._ ._ . . . .~ . - _ _ _ 24-57a(86H2)/cp-20 TABLE 6-1
.(continued)
Reference ** Unit 2 General Characteristics Unit Unit 2 Cycle 7 Cycle 8 Average Film Temperature
*F 31 -31 Difference-Average Linear Heat Rate of kw/ft 6.09***- 6.16****
Undensified Fuel Rod. (accounts;forabove
-fraction of heat generated in fuel rod)
~
j Average Core Enthalpy Rise- BTU /lb 66.5 66.5 i Maximum Clad Surface- F 657 657 Temperature Calculational Factors Engineering Heat Flux on Hot Channel 1.03+ 1.03+ ! Engineering Factor on Hot Channel 1.02+ 1.02+ Heat Input Rod Pitch and Clad Diameter Factor 1.065+ 1.065+ Fuel Densification Factor (axial) 1.002 1.002 l Notes Due to the statistical combination of uncertainties described in References 6, 9, and 10, the nominal inlet temperature and nominal primary system pressure were used to calculate some of these parameters.
** Reference cycle analysis (Unit 2 Cycle 7) is contained in Reference 7.
*** Based on a value of 224 shims.
**** Based on a value of 688 shims.
+ These factors have been combined statistically with other uncertainty i factors at 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. This limit was j verified to be applicable to Cycle 8.
6-3 l
24-57a(86H2)/cp-21 _ 7.0 TRANSIENT ANALYSIS , The Design Basis Events (DBEs) considered _in 'the Unit = 2 ~ Cycle 8
-safety analyses -are listed in Table 7-1. Core parameters' input-to ,
i the safety analyses for evaluating approaches to DNB and centerline-temperature to melt. fuel design limits are presented'in Table 7-2.- As indicated in Table 1, the .only reanalyses included herein ~ are - for the Boron Dilution and CEA Ejection events. No other reportable reanalyses were performed for any-other DBE's since all results for-I these DBE's lie within 'the bounds of (or are conservative with : i respect to) the ' reference cycle values (Unit' 2 ~ Cycle 7 -Reference . 1). i- The major characteristics of the ' Unit 2 Cycle 8 transient. analyses-
; are the following:
1
- a. With the exception of data -for-the Boron Dilution, Full Length l CEA Drop, CEA Ejection and Steam Line Rupture events, all key
- input parameters to the transient- analyses lie within the-bounds of those of the reference cycle.
- j. b. An evaluation of data generated by FATES 3B (Reference 2), wh_ich j is a - revised version of the FATES 3 fuel evaluation- model
, (Reference.3), showed no significant impact. - FATES 3B has received interim NRC approval (Reference 4).
I
- c. A reanalysis of the Boron Dilution event was performed to
! accommodate changes in input-data due to the implementation' of 24-month cycles.
I d. An evaluation of the Full: Length CEA Drop event ~ showed that a j revised Doppler curve (changed to accommodate the implementation ' i of 24-month cycles) does not have a significant--impact on the 4 analysis.
- ' e. A reanalysis of the CEA Ejection event was performed. to establish generic values for the CEA ejection worth and the -
, post-ejected radial power peak'_for the' Hot Zero Power.
- condition,
- f. A reanalysis of the Steam Line Rupture event to accommodate the implementation of 24-month . cycles yielded results that were less limiting than those previously reported. i l
l i
- 7-1
- q
24-57a(86H2)/cp-22 TABLE 7-1 CALVERT CLIFFS UNIT 2 CYCLE 8 DESIGN BASIS EVENTS CONSIDERED IN THE NON-LOCA SAFETY ANALYSIS Analysis Status 7.1 Anticipated Operational Occurrences for which intervention of the RPS is necessary to prevent exceeding acceptable limits: 7.1.1 Boron Dilution Reanalyzed 7.1.2 Star}up of an Inactive Reactor Coolant' Not Reanalyzed Pump 7.1.3 Loss of Load Not Reanalyzed 7.1.4 Excess Load Not Reanalyzed 7.1.5 Loss of Feedwater Flow Not Reanalyzed 7.1.6 Excess Heat Removal due to Feedwater Not Reanalyzed Malfunction 7.1.7 Reactor Coolant System Depressurization Not Reanalyzed 7.1.8 Excessive Charging Event Not Reanalyzed. 7.2 Anticipated Operational Occurrences for which RPS trips and/or sufficient initial steady state thermal margin, maintained by the LCOs, are necessary to prevent exceeding the acceptable limits: 2 7.2.1 Sequential CEA Group 3 Withdrawal Not Reanalyzed 7.2.2 Loss of Coolant Flow NotReanagyzed 7.2.3 Full Length CEA Drop Evaluated 7.2.4 Transients Resulting from the 4 Not Reanalyzed I. Malfunction of Oge Steam Generator i 7.2.5 Loss of AC Power Not Reanalyzed J 7.3 Postulated Accidents 1 1 7.3.1 CEA Ejection Reanalyzed 7.3.2 Steam Line Rupture Reanalyzed 6 7.3.3 SteamGenerajorTubeRupture Not Reanalyzed 7.3.4 Seized Rotor Not Reanalyzed I Technical Specifications preclude this event during operation. 2 Requires High Power and Variable High Power Trip. 3 Requires Low Flow Trip, j 4 Requires trip on high differential steam generator pressure. 5 Conclusions previously reported are valid for Unit 2 Cycle 8. 0 Results for Unit 2 Cycle 8 are less limiting than those previously reported. 7-2 __ , -I
24-57a(86H2)/cp-23 TABLE-7-2 CALVERT CLIFFS UNIT 2 CYCLE 8 CORE PARAMETERS INPUT TO SAFETY ANALYSES FOR DNB AND CTM (CENTERLINE TO MELT) DESIGN LIMITS Reference Cycle
- Physics Parameters- Units (Unit 2 Cycle 7) Unit 2 Cycle 8 Radial Peaking Factors ForTDNB Margin Analyses (Fr)
Unrodded Region 1.70***+ 1.70***+ Bank 5 Inserted 1.87**'+ 1.87**'+ ForylanarRadialComponent , (F ) of 3-D Peak (CM Limit Analyses) Unrodded Region 1.70** 1.70** f Bank 5 Inserted 1.87** 1.87** Moderator Temperature 10-4ap/* F -2. 7 + +. 7 -2. 7 + +. 7 Coefficient Shutdown Margin (Value %Ao -3.5 -4.5 Assumed in Limitin E0C Zero Power SLB Tilt Allowance % 3.0 3.0 Power Level MWt 2700** 2700** Maximum Steady State F 548** 548** l Inlet Temperature l Minimum Steady State psia 2200** 2200** RCS Pressure 6 Reactor Coolant Flow 10 1bm/hr 138.5** 138.5** Negative Axial Shape I .15**'+ .15***+ LC0 Extreme Assumed P at Full Power (Ex-Cores) i
- 7-3 4
24-57a(86H2)/cp-24 TABLE 7-2 (continued) Reference Cycle
- Safety Parameters Units (Unit 2 Cycle 7) Unit 2 Cycle 8 Maximum CEA Insertion % Insertion 25 25 at Full Power of Bank 5 Maximum Initial Linear KW/ft 16.0 16.0 Heat Rate fcr Transients Other than LOCA Steady State Linear KW/ft 22.0 22.0 Heat Rate for Fuel CTM Assumed in the Safety Analysis CEA Drop Time from sec 3.1 3.1 Removal of Power to Holding Coils to 90%
Insertion Minimum DNBR (CE-1) 1.21** 1.21** Notes
- Reference 1
** For DNBR and CTM calculations, effects of uncertainties on these parameters were accounted for statistically. The procedures used in the Statistical Combination of Uncertainties (SCU) as they pertain to DNB and CTM limits are detailed in References 5, 6 and 7. These procedures have ,
been approved by NRC for the Calvert Cliffs Units in Reference 8. l l
+
The values assumed are conservative with respect to the Technical Specification limits. 7-4 1
24-57a(86H2)/cp-25 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 subcritical Modes 2, 3, and 4 of operation. This event was reanalyzed to account for an increase in the critical boron concentrations for operational Modes 2, 3, and 4 as shown in Table 7.1.1-1. Table 7.1.1-1 compares the values of the key transient parameters assumed in each mode of operation reanalyzed for Cycle 8 and the corresponding reference cycle value. The conservative input data chosen consists of high critical boron concentrations and low inverse boron worths. These choices produce the most aoverse effects by reducing the calculated time to criticality. The time to criticality was determined by using the same expression as used in the reference cycle (Unit 1 Cycle 8, Reference 9). 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 those of the reference cycle due to a higher critical boron concentration. 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. l l l 7-5
24-57a(86H2)/cp-26' TABLE 7.1.1-1 KEY PARAMETERS ASSUMED IN THE BORON DILUTION ANALYSIS Unit 1 , Unit 2 Parameter Cycle 8 Cycle 8 Critical Baron Concentration, ppm (All Rods Out, Zero Xenon) Startup(Mode 2) 1900 2100 Hot Standby (Mode 3) 1900 2100 Hot Shutdown (Mode 4) 1900 2100-Inverse Boron Worth, ppm /%ao Startup 65 65 Hot Standby 55 55
- Hot Shutdown 55 55 Minimum Shutdown Margin Assumed, %ao Startup -3.5 -3.5+
i Hot Standby -3.5 --3.5+ Hot Standby -3.5 -3.5+
- Reference 9.
- The value assumed is conservative with respect to the revised Technical Specification limit (see Section 9.0).
i
- 7-6
24-57a(86H2)/cp-27 TABLE 7.1.1-2 RESULTS OF THE BORON DILUTION EVENT 1 Criterion for Minimum Time to Lose Time to Lose Prescribed Shutdown Prescribed Shutdown Mode Margin-(min) Margin (min) Unit 1 Unit 2 Cycle 8- Cycle 8 Startup 60 55 15 Hot Standby 50 45 15 Hot Shutdown 50 45 15 f I 1 i l 7-7 j i
. ~ .. . - .
! 24-57a(8'6H2)/cp-28 A 7.3.1 CEA Ejection Event ] i The zero power case of the CEA ejection event was reanelyzed for
, Unit 2 Cycle 8 to determine the ' maximum erergy deposited in the j fuel. The reanalysis was performed to establish generic values for L the CEA ejection worth and the post-ejected radial power peak for
- the zero power case.
4 i The analytical method employed in the analysis of this event is the NRC approved Combustion Engineering CEA Ejection method which is i described in CENPD-190;A (Reference 10). ! The key parameters used in this' event are listed in Table 7.3.1-1. These key parameters are selected to determine the maximum average j and centerline enthalpfes in the hottest spot of the hot rod. The. t calculated enthalpy values are compared to threshold enthalpy values t ! to determine the amount of fuel exceeding these- thresholds. The ! threshold enthalpy values are: j Clad Damage Threshold: j Total Average Enthalpy = 200 cal /gm Incipient Centerline Melting Threshold: - l Total Centerline Enthalpy = 250 cal /gm l' 1
- Fully Molten Centerline Threshold
- 2 j Total Centerline Enthalpy = 310 cal /gm
; To bound the most adverse conditions during the cycle, the most '
i limiting of either the Beginning of Cycle (BOC) or End of Cycle (EOC) parameter values were used in the' analysis.. A BOC Doppler defect was used since it produces the least amount of negative
! reactivity feedback to mitigate the transient. A BOC moderator
! temperature coefficient of +0.7X10-4 Ap/*F was used because a i positive MTC results in positive reactivity feedback and thus
- increases coolant temperatures. EOC delayed neutron fractions and l
neutron kinetics were used in the analysis to produce.the highest
- power rise during the event.
9 The zero power CEA ejection event was analyzed assuming the core is
- initially operating at 1 MWt. At -zero power, a Variable Overpower i i trip is conservatively assumed to initiate at 40% (30% + 10% ;
i uncertainty) of 2700 MWt and terminate the event. 1 + The zero power case was analyzed, assuming the value of 0.05 seconds i for the total ejection time, which is consistent with the reference cycle (Unit 1 Cycle 8 Reference 9). I , The power transient produced by a CEA ejection initiated for the ' zero power case is shown in Figure 7.3.1-1. i I l 7-8 ,
. _ _ _ - - . -_ _ ._......~.. _ ,_,____-, _ ,__,_.. .. _._. i
24-57a(86H2)/cp-29 The.results of the zero power CEA ejection case analyzed (See' Table. 7.3.1-2) show that the maximum total energy- deposited in the hot spot of the pin during the event is less than both the criterion for clad damage (i.e., 200 cal /gm) and the fully molten canterline threshold of 310 cal /gm. Consequently, no fuel pin failures are calculated to occur. Also, an acceptably small fraction (less than 1%) of the fuel reaches the incipient centerline melt. threshold of 250 cal /gm. 1 3 7-9
r
'24-57a(86H2)/cp-30 TABLE 7.3.1-1 i
KEY PARAMETERS ASSUMED IN THE CEA EJECTION ANALYSIS ZERO POWER CASE Unit 1 Unit 2 Parameter Units Cycle 8 Cycle 8 Core-Power Level MWt 1. 1. Ejected CEA Worth %Ap .77 1. Post-Ejected Radial Power Peak - 8.9 10.5 Axial Power Peak - 1.64 1.64 CEA Bank. Worth at Trip %ap -1.5 -1.5 Doppler Multiplier ,
.85 .85 Moderator Temperature Coefficient x10-4Ap/*F .7 .7 Delayed Neutron Fraction -
.0044 .0044 Reference 9 4
l 1 j 7-10 l
24-57a(86H2)/cp-31 TABLE 7.3.1-2 CEA EJECTION EVENT RESULTS Reference Cycle Current Cycle Unit 1 Cycle 8 Unit ? Cycle 8 Zero Power Total Average Enthalpy of Hottest <200' <200 Pellet (cal /gm) Total Centerline Enthalpy of <250 276 Hottest Fuel Pellet (cal /gm) Fraction of Rods that Suffer Clad 0 0 Damage (Average Enthalpy 1 200 cal /gm) Fraction of Fuel Having at least 0 <.01 Incipient Centerline Melting (Centerline Enthalpy 1 ?50 cal /gm) Fraction of Fuel having a Fully 0 0 Molten Centerline Condition (Centerline Enthalpy 1 310 cal /gm) 1 7-11 1
80,0
. i i i i _
N ZERO POWER CASE CEA EJECTED WORTH = 1%Ap - 10,0 - i - N . c k m u_ O ., E T-U u. 1.0 .-.- _ a: _ s " l 2 '- y '- 8 .- 0.1 -_
~
\
0,3 I I I i 0 1 2 3 14 5 TIME [ SECONDS CO. CEA EJECTION EVENT FIGURE GAS & Calvert Cliffs CORE POWER VS TIME 7.3.1-1 Nucl0ar Power Plant 7-12 _ __ _
24-57a(86H2)/cp-32 8.0 ECCS ANALYSIS 8.1 LARGE BREAK LOSS-0F-COOLANT ACCIDENT 8.1.1 Introduction and Summary An ECCS performance analysis was performed for Calvert Cliffs Unit 2 Cycle 8 to demonstrate compliance 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 Generation Rate (PLHGR) of 15.5 kw/ft. This PLHGR is equal to the existing limit for Calvert Cliffs Units 1 and 2. The method of analysis and detailed results which support this'value are presented in the following sections. 8.1.2 Method of Analysis The ECCS performance analysis for Calvert Cliffs Unit 2 Cycle 8 was performed us.ing the 1985 Evaluation Model which is described in References 2 through 8 and was approved by the NRC in Reference 9. The reference cycle analysis (Unit 2 Cycle 7, Reference 10) utilized the previously approved Large Break LOCA Evaluation Model. Except for the n'odel differences, the method of analysis for Cycle 8 is identical to the reference cycle large break LOCA ECCS performance analysis. I l Blowdown hydraulics, refill /reflood hydraulics and hot rod ! temperature calculations were performed with the fuel parameters which bound Cycle 8 at a reactor power level of 2754 MWt. The blowdown hydraulics calculations were performed with the CEFLASH-4A code (Reference 5) while the refill /reflood hydraulics calculations were performed with the COMPERC-II code (Reference 6). The hot rod- ; 4 clad temperature and clad oxidation calculations were performed with the STRIKIN-II and PARCH codes (References 11 and 12, respectively). These calculations utilized fuel performance data calculated by l FATES 3B (Reference 13) which is a revised version of the FATES 3 fuel evaluation model (Reference 14). FATES 3B has received interim NRC approval (Reference 15). Core wi6s clad oxidation calculations were also performed in this analysis. Burnup dependent hot rod calculations were~ performed with STRIKIN-II to determine the initial fuel conditions which result in the highest peak clad temperature (PCT). This study demonstrated that the burnup with the highest initial fuel stored energy results in the highest PCT. This occurred at a hot rod burnup of 943 MWD /T. A break spectrum analysis was performed which determined that the 0.6 Double Ended Guillotine at Pump Discharge (DEG/FD) break yields the highest peak clad temperature. A summary of the fuel parameter input values for Cycle 8 and the reference cycle is shown in Table
- 8.1-1.
4 8-1
2 24-57a(86H2)/cp-33 8.1.3 Results " Table 8.1-2 presents the analysis results for the 0.6 DEG/PD break i which produces - the highest -peak clad temperature. For ' comparison -
- purposes, the corresponding values of the reference cycle analysis
! (Unit 2 Cycle 7) are also~ presented in Tables 8.1-2. The resultstof. L the evaluation confirm that 15.5 kw/ft isan acceptable value- for F - the PLHGR LC0Lin Cycle 8. The -peak clad temperature, maximum local ' 1 clad oxidation and core wide clad- oxidation values of 1903 F, 3.3% ) and <.51%, respectively, are well - below the 10CFR50.46 . acceptance
- criteria limits of 2200"F, 17%:and 1%,-respectively. These results i have been confirmed for:up to 100 plugged tubes per steam generator.
- A list of the significant parameters displayed graphically for the
- limiting case (Figures 8.1-1 through 8.1-14)~ is presented in Table 8.1-3.
i A review of the effects of initial operating - condi.tions- or these results was performed. It was determined that over the ranges of initial operating conditions, as specified in the Technical i Specifications, operation of the plant at a PLHGR of 15.5 kw/ft is
- an acceptable limit for Cycle 8 operation.
i l 8.1.4 Conclusions i i As discussed above, conformance to the ECCS criteria is summarized - i by the analysis results presented in Table 8.1-2. The most limiting case results in a paak clad temperature of 1903*F, which is well' 4 below the acceptance limit of 2200'F. The maximum ' local and core wide values for zirconium oxidation percentages, as shown in Table ! 8.1-2, remain well below the acceptance limit values of 17% and 1%,
- respectively. Therefore, operation-of Unit 2 Cycle 8 at a PLHGR of.
15.5 kw/ft and a power level of -2754 PWT (102% of 2700 MWT) is in !. compliance with the 10CFR50.46 acceptance criteria.. 8.2 SMALL BREAK LOSS-0F-COOLANT ACCIDENT f
; Analyses have confirmed that the reported small break loss-of-coolant accidert results of the reference cycle- (Unit 1 Cycle 8, Reference 16) bound Unit 2 Cycle 8. These results have been approved by the NRC in Reference 17. Therefore, acceptable - , ! small break LOCA ECLS performance is demonstrated at a peak linear =
j heat-rate of 15.5 kw/f' and a reactor power level of 2754 MWt (102% i of 2700 MWt). This acceptable performance has been confirmed with up to 100 plugged tubes per steam generator. l } t l 1 8-2
. , - - - . , _ _ _ _ - _ , , _ , - -_._-,__,_,..-._~,,-_..,~,,.m -
~,_ ._. _ , _ , , _ , - - _ _ . , . _ _
24-57a(86H2)/cp-34 TABLE 8.1-1 Calvert Cliffs Unit 2 Cycle 8 Fuel Parameters as Compared to Unit 2 Cycle 7 Unit 2 Unit 2 Cycle 7 Cycle 8 Fuel Parameters Values Values Reactor Power Level (102% of Nominal), MWT 2754 2754 Average Linear Heat Rate (102% of Nominal), kw/ft 6.37 6.45 Hot Channel Peak Linear Heat Generation 15.5 15.5 Rate, kw/ft Hot Assembly Peak Linear Heat Generation 13.42 13.34 Rate, kw/ft
- Gap Conductance at PLHGR, B/hr-ft2 *F 1937 2267
- Fuel Centerline Temperature at PLHGR, *F 3649 3740
- Fuel Average Temperature at PLHGR, "F 2228 2246 J
- Hot Rod Gas Pressure, psia 1188 1189
- Hot Rod Burnup, MWD /MTU 943 943 i
*Are initial fuel rod parameters, in STRIKIN-II, which yield maximum PCT.
.{ 8-3 (
24-57a(86H2)/cp-35
. TABLE 8.1-2 Sumary of ECCS Performance Results for-i Calvert Cliffs Unit 2 Cycle 8 for the Limiting Break Size
~(0.6 DEG/PD)'as Compared to the Reference Cycle (Unit 2 Cycle 7)'
l Limiting Case t (Maximum Initial Fuel ! Parameters Stored Energy) i Unit 2 Unit'2
-Cycle 7 Cycle 8 Rod Average Burnup, 943 ~943 MWD /MTU
- Peak Clad Temperature 1945 1903
- (PCT),'F
! Time of PCT, Seconds 261 234 i
i Time of Clad Rupture, 33.0 25.6 i Seconds 4 Peak Clad Oxidation, % 5.40 3.30 c Core Wide Oxidation, % <.51 <.51 , I j i i. 1
! l
- 1 i l i
1 1 I 8A I
~ 24-57a(86H2)/cp-36 TABLE 8.1-3 Calvert Cliffs Unit 2 Cycle 8 Variables Plotted as a Function of Time for the Limiting large Break Variable Figure Number Core Power 8.1-1 Pressure in Center Hot Assembly Node 8.1-2 Leak Flow 8.1-3 Hot Assembly Flow (below hot spot) 8.1-4 Hot Assembly Flow (above hot spot) 8.1-5 Hot Assembly Quality 8.1-6 Containment Pressure 8.1-7 Mass Added to Core During Reflood 8.1-8 Peak Clad Temperature 8.1-9 Hot Spot Gap Conductance 8.1-10 1, Peak Local Clad Oxidation 8.1-11
- Temperature of Fuel Centerline, 8.1-12
; Fuel Average, Clad and Coolant at Hottest Node Hot Spot Heat Transfer Coefficient 8.1-13 Hot Rod Internal Gas Pressure 8.1-14 i
8-5
Figure 8.1-1 ,e CALVERT CUFFS UNIT 2 CYCLE 8 0.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG CORE POWER 1.2001 1.0000 l 8000
.soco 2
5 1 \ a l I
$ .400C R
2000
\
k 000% a a a a a 8 8 8 8 8 8 l 9 9 9 9 9 9
- ~ m , o TIME IN SEC 8-6
I Figure 8.1-2
+
CALVERT CUFFS UNIT t CYCLE 8 0.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG PRESSURE IN CENTER H0T ASSEMBLY N0DE 2400 0 2000 0 l 1600.0 3 i w 's
- 0-
\
1200 0
\
La o w w 800.0 uJ l s' s e N c_ 's i 400.0 x 0 0 a a a 8 8 8 8 8 8
) N 5 N 5 TIME IN SEC 8-7
; Figure 3.1-3 i CALVERT CUFFS UNIT 2 CYCLE 8 0.6 x DOUBLE ENDED OUILLOTINE BREAK IN PUMP DISCHARGE LEG LEAK FLOW 120000 i
100000 l l l l l u 80000 y Pump Side N --- RV Side cm l J 60000 ! u.2 l & 1 1 l QC
\
2 40000. o '\ J \ l l
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! N - I. l
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20000
%'N N l
l (lx M.'s % l l N I C O O O C3 O O O O
- O O O O O O o . O. .
- o. .
O o C o A - - N N l T!ME IN SEC 8-8
Figure 8.1-4 CALVERT CUFF 8 UNIT 2 CYCLE 8 0.6 x DWBLE ENDED QUILLOTINE BilEAK IN PUMP DISCHARGE LEO NOT ASSEMBLY FLOW, BELOW HOT SPOT l 30.000 20.000 , k u 10.000 w 1 v) C .k
\
.000 ,
w N / t p 5 J 2 C
-10.000! I l
l _.J j' l l
, , l I
-20.000! '
i
-30.000 o o e o l o o o o o a I 8 8 9 9 9 i S f
- 0 TIME IN SEC 8-9
Figers 8.1-5 ' , CALVERT CUFF 8 UNIT 2 CYCLE 8 0.8 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG _ HOT ASSEMBLY FLOW, ABOVE HOT SPOT 30 000, , 20.000 f a 10.000 w ,I vs N ' AQ I co ! 000 ' w a
\*
F
)
a:: r -10.000 I o i J u.
-20.000 1
-30 000 e
8 o 8 9 8 8. 8 8 9 9 9 a S 2 R g TIME IN SEC 8-10
7 , Figure 8.1-6 CALVERT CLIFFS UNIT 2 CYCLE 8 _T- -r" 0.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG - HOT ASSEMBLY QUALITY - Node 13, Below Hottest Region
---Node 14, At Hottest Region I -
--Node 15, Above Hottest Region _
1.0000 , qg1 f j I /g -
.f -
I
/l
'I 8000 ,
l
\
I r/ (
-)
!I f \/r / -
6000 lj I,
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' j / '
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}/,%
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lll I1./ ! I il il l l" i ; II l .: 2000 j , l ll ' 1 j l l , 1 0000 c a a a j o o e o o o ; o e o o o o l
- c. o . . . i O m o e D ~ ~ N N TIME IN SEC 8-11
+, i-
[e " Figure 8.1-7 - CALVERT CUFFS UNIT 2 CYCLE 8
^
0.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG CONTAINMENT PRESSURE
.~
d 60 000 e.. , 50 000 40 000 --! \ E m E 30.0C0t. 5 e E 20.000 l 4
'i0.000 i
.C00'! a a a a l
O O Q C C o o . . . o - c o o o o e e N e r - N m v TIME IN SEC A
^
8-12
+
Figure 0,1-8 CALVERT CUFFS UNIT 2 CYCLE 8 0.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEO MASS ADDED TO CORE DURING REFLOOD 120000-100C00 tur, we unoon un 0.00-12.0 1.540 IN/8EC 12.0-80.0 1.10014/SEC 88.0-400.0 0.790 IN/SEC 80000 E a 60000 # S S
- a g 40C00 E
20000 l l d i a 8 9 a 8 ? 8 8 x -
~ m ,
TIME AFTER CONTACT. SEC l l 8-13 I
Figure 8,1-9 CALVERT CLIFFS UNIT 2 CYCLE 8 0.6 x DOUBLE ENDED QUILL 0 TINE BREAK IN PUMP DISCHARGE LEG PEAK CLAD TEMPERATURE 2200 g 2000 s
^~
1800 I 1500 s N
\ N,
' u- ,f ! l a
*1400 -
g[ ! : i g i I ' 2 i l t , i E 1200' !,lt ' ' t , o ! 1 5 u I i i i 1 ' l l t l l 1000 , i . I : , 4 800 - 1 600 i i l 100 200 : _: 0 400 S00 600 J' l 7IME. ECON 05 8-14
Figure 4.1-10 CALVEllT CUFF 8 UNIT 2 CYCLE 4 0.8 x DOUBLE ENDED QUILL 0 TINE BitEAK IN PUMP DISCHAllGE LIG HOT SPOT GAP CONDUCTANCE 1600 1400 1200 Q 1000 Q 4 E s [ 800 tf 5 8 3 600 8 a 400 [ ._ l 200 l 1 100 200 300 400 500 600 70 TIME, 3 ECONO 3 8-15 .
Figure 8.1-11 CALVERT CLIFFS UNIT 2 CYCLE 8 0.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG PEAK LOCAL CLAD OXIDAfl0N ISi 14 12 10 I i i i 1 e , i l g i ! I x f' ! l i e j l i 4 ! i m _ l g ; , o ' ! ! 5 u
; I !
l l l A i :
! j i +
.- j
/ ! l 2 /
\ ;
l i 100 2L- .00 40C 5CO SCO _. : i TIME. 5E'GND5 8-16
Figure 8.1-11 CALVERT CLIFFS UNIT 2 CYCLE 8 0.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG TEMPERATURE OF FUEL CENTERLINE, FUEL AVERAGE, CLAD AND COOLANT AT HOTTEST N0DE 3500 3000 I l l l 2500 i
! I FUEL AVERAGE
\ '
l /! I 20Co ij j '. 'I X c'"N i
, il ' i a ! eus t g 1:-,_
1- h 5 - t . a r!( : i , W i j,) ! li : 1000I ' I l I \ l I 500 !
\ cooun
\ /
i 100 20C 5^0 400 500 hCO Of TIME. iECONC' 8-17 I
Figure 8.1-13 CALVERT CLIFFS UNIT 2 CYCLE 8 0.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG HOT SPOT HEAT TRANSFER COEFFICIENT ISO 140 l 120 & i gl ' k , a a 100 l j i i s i E i- l m . N w ., n 1 i i i U cu , E l m j ' ;' l
- 1 0 -
l o l j E ' f ! t l w l'
. 4 i
n ; >' i ,' ,' l . !
-s = '
l M . i i j !
- j ; ,
l t
> i j l i !
15 l ; i x i 40 l' I t l t ' 1 l i l 20 l - l
,. / 'A \
/ i l' I
'/ I I i 1
', -l 100 200 300 400 500 6C0 ~C-T[ME, 3EcgNg3 8-18 1
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