ML20134N129
| ML20134N129 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 08/30/1985 |
| From: | Poindexter C BALTIMORE GAS & ELECTRIC CO. |
| To: | Butcher E Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20134N133 | List: |
| References | |
| NUDOCS 8509050014 | |
| Download: ML20134N129 (47) | |
Text
B ALTI MORE GAS AND ELECTRIC CHARLES CENTER P.O. BOX 1475 BALTIMORE, MARYLAND 21203 CMmis H. PolNDEXTER vice Pats ocNT ENG4NCEmNG AND CONSTRUCT >ON August 30,1985 U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555 ATTENTION:
Mr. E. J. Butcher, Jr., Acting Chief Operating Reactors Branch #3 Division of Licensing
SUBJECT:
Calvert Cliffs Nuclear Power Plant Unit No. 2; Docket No. 50-318 Request for Amendment to Operating License DPR-69 Seventh Cycle License Application
REFERENCE:
A. E. Lundvall, Jr. to J. R. Miller letter dated February 22, 1985, Calvert Cliffs Nuclear Power Plant Unit 1; Docket No. 50-317, Amendment to Operating License DPR-53, Eighth Cycle License Application ENCLOSURE: Unit 2 Cycle 7 Reload License Submittal 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 a seventh cycle. The enclosed report presents a detailed description of the required Standard Technical Specifications (STS) with supporting safety analysis information to ensure conservative operations at a rated thermal power of 2700 MWth for Unit 2 Cycle 7.
Our present intention is to begin the Unit 2 refueling outage on October 18,1985, and to complete the outage and begin the approach to criticality on December 9, 1985, with return to power operations immediately thereafter.
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Mr. E. J. Butcher August 30,1985 Page 2 PROPOSED CIIANGES (BU&E FCR 85-3003)
This reload is enveloped by the analysis previously submitted for the current operating Unit 1 Cycle 8 (Reference). The proposed changes to the Unit 2 Technical Specifications will make the following Technical Specifications identical to the Unit 1 Technical Specifications: STS 3.1.1.1, decreases the required shutdown margin; STS 3.1.1.4, changes the allowed Moderator Temperature Coefficient (MTC); STS 4.2.1.4, removes the flux peaking augmentation factors. These changes are all enveloped by the analysis performed for the referenced cycle and the reconfiguration of Control Element Assemblics (CEAs). One Technical Specification change that will be different from Unit 1 is Technical Specification B.2.1.1 which changes the minimum DNBR limit from 1.23 to 1.21. This is consistent with the Departure from Nucleate Boiling Ratio (DNBR) limit for the CE-1 correlation which the NRC recently approved.
DETERMINATION OF SIGNIFICANT IIAZARD We have determined, based on the analytical information supplied in the enclosure, that this amendment request does not involve a significant hazards consideration.
The proposed changes are consistent with example iii of amendments considered not likely to involve significant hazards consideration, as shown in the Federal Register dated April 6,1983, page 14870. No fuel assemblies to be loaded into the Unit 2 Cycle 7 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 Unit 2 Cycle 7 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.
i
- 1. The probability or consequences of an accident evaluated is not significantly increased. The large break Loss of Coolant Accident was re-evaluated and its consequences were found to be slightly worse than previously reported, but still well below the license limit.
- 2. The reload application does not create the possibility of a new or different kind of accident from any previously evaluated.
- 3. The license reload does not involve a significant reduction in the margin for safety.
Small changes in shutdown margin, increase in moderator temperature coefficients, and decrease in DNBR ratio are compensated by the CEA reconfiguration and reference cycle analysis.
Mr. E. J. Butcher August 30,1985 Page 3 SAFETY COMMTITEE 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.
FEE DETERMINATION Pursuant to 10 CFR 170.21, we are including BG&E Check Number A348084 in the amount of $150.00 to the NRC to cover the application fee for this request.
Very tgily yours,,
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STATE OF MARYLAND:
- TO WIT CITY OF BALTIMORE:
Chris II. Poindexter, 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.
WITNESS my lland and Notarial Seal:
g2M Jo Ngary Public My Commission Expires:
7/
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Clip / JAM /Imt ec: D. A. Brune, Esquire G. F. Trowbridge, Esquire D. II. Jaffe, NRC T. Foley, NRC T. Magette, DNR l
i I
1
ATTACHMENT TO R-NE-CALVERT CLIFFS UNIT 2 CYCLE 7 RELOAN LICENSE SUBMITTAL I
Calvert Cliffs Unit 2 Cycle 7 Reload License Submittal Table of Contents Section 1.
Introduction and Sum mary............................................. 1-1 2.
Operating History of the Previous Cycle................................. 2-1 3.
General Descrip tion................................................... 3-1 4.
Fuel Syste m Design................................................... 4-1 5.
Nuc le ar Design....................................................... 5 -1 6.
Thermal-Hydraulic Design............................................. 6-1 7.
Transient Analysis.................................................... 7-1 8.
ECCS Performance Analysis
...........................................8-1 9.
Technical Specifications............................................... 9-1
- 10. S tar tup Testing.................................,..................... 10 -1 Re fe rences......... '............................................
11.
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1.0 INTRODUCTION
AND
SUMMARY
This report provides an evaluation of design and performance for the operation of Calvert Cliffs Unit 2 during its seventh fuel cycle, at full rated power of 2700 MWt.
All planned operating conditions remain the same as those for Cycle 6.
The core will consist of 148 presently operating Batch G and H assemblies, 60 fresh Ratch J assemblies, 8 Batch E assemblies previously discharged from Cycle 4 of Calvert Cliffs Unit 2, and 1 Batch 0 assembly previously discharged from Cycle 3 of Calvert Cliffs Unit 2.
In addition to the plant operating requirements which ordinarily create a need for flexibility in the termination point of the previous cycle, the malfunction of the Unit 1 electric generator resulted in the need for additional flexibility in the form of an extended Cycle 6 length.
Consequently, in place of the standard 1000 MWD /T end-of-cycle burnup range for the previous cycle, the Unit 2 Cycle 7 analyses have been performed for a Cycle 6 termination point between 12,200 MWD /T and 13,800 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 7 conditions would be enveloped, provided the Cycle 6 termination point falls within the expanded cycle burnup range discussed above. The analysis presented herein will accommodate a Cycle 7 length of up to 13,900 MWD /T.
The evaluations of the reload core characteristics have been conducted with respect to the Calvert Cliffs, Unit 1 Cycle 8 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.
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 previous cycle (Unit 2 Cycle 6, Reference 2), proposed modifica-tions to the plant Technical Specifications are provided and are justified by the analyses repncted herein.
- These proposed modifications are very similar to those approved (Reference 3) for the reference cycle.
All Cycle 7 analyses address fuel exposure explicitly.
The performance of Combustion Engineering 14x14 fuel at extended burnup is discussed in Reference 4 1-1
t 1
l 2.0 OPERATING HISTORY OF THE PREVIOUS CYCLE
~
Calvert Cliffs Unit 2 is presently operating in its sixth fuel cycle utilizing Batch H,
G, F,
E and 0 fuel assemblies (including eight Batch E assemblies from Unit 1).
Calvert Cliffs Unit 2 Cycle 5 began operation on June 29, 1984 and reached full. power on July 24, 1984 The Cycle 6 startup testing was reported to the NRC in Reference 5.
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 6 will terminate on or about October 18, 1985. The Cycle 6 termination point can vary between 12,200 MWD /T and 13,a00 MWD /T to accomodate the plant schedule and still be within the assumptions of the Cycle 7 analyres.
As of August 19, 1985, the Cycle 6 burnup had reached 11,329 MWD /T.
h 2-1
3.0 GENERAL DESCRIPTION The Cycle 7 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 7 is the removal of 69 assemblies (61 Unit 2 assembifes: 40 Batch F, 1 Ratch D, 20 Batch O/; 8 Unit 1 Batch E/ assembifes). These assemblies will be replaced by 40 fresh unshimmed Batch J assemblies (4.05 wt% U-235 enrichment), 20 fresh unshimmed Batch J*
assemblies (3.40 wt% U-235 enrichment), 8 Batch E assembifes (3.03 wt% U-235) discharged from Unit 2 Cycle 4, and 1 Batch 0 assembly (3.03 wt% U-235) discharged from Unit 2 Cycee 3.
Figure 3-1 shows the fuel management pattern to be employed in Cycle 7.
Figure 3-2 shows the locations of the poison pins within the lattice of twice-burned Batch G/ assemblies and the fuel rod locations in unshimmed assemblies.
This fuel management pattern will accommodate Cycle 6
termination burnups from 12,200 WD/T to 13,800 WD/T.
0 The Cycle 7 core loading pattern is 90 rotationally symmetric.
That is, if one quadrant of the core were rotated 90 into its neighboring 0
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 7 assembly burnup distribution for
~
a Cycle 6 termination burnup of 13,800 WO/T.
The initial enrichment of the fuel assembliet is also shown in Figure 3-3.
Figure 3 4 shows the end of Cycle 7' assembly burnup distribution.
The end of Cycle 7 core average exposure is approximately 29,700 WO/T and the average discharge exposure is approximately 41,400 WD/T.
The end of cycle burnups are based on Cycle 6 and Cycle 7 lengths of 13,800 WO/T and 13,900 WD/T, respectively.
3.1 Prototype CEA The prototype CEA is described in Reference 6.
Cycle 3 was the first cycle of irradiation for this CEA.
During the EOC-4 and EOC-5 outages this CEA was examined, as described in References 7 and 8, respectively.
This prototype CEA was utilized in the center core position during Cycles 3 through 6.
Que to the changes in CEA pattern which specify the use of a very weak CEA in the center location (see Subsection 3.2 below and the reference cycle submittal (Reference 1)), this CEA will be shifted to a new location.
3.2 CEA Patterns The comoosition of nine CEAs, the configurations of two CEA banks and, consequently, the overall CEA hank pattern are being changed for Cycle 7 Both these changes and the reasons for them are identical to those presented in support of the reference cycle (Reference 1).
3-1
TABLE.1-1 CALVERT CLIFFS UNIT 2 CYCLE 7 CORE LOADING Total Initial Number Total Initial Poison Poison of Poison Number Assembly Number of Enrichment Ratc urnup(
TgI Rods Per Loading and Non-Fuel of Fuel
- l. signstice Assemblies (wt1 U-235) 80C7 E007 Assenhly (wt% 0 C)
Rods Rods 4
i 1
l J
40 4.05 0
12,200 0
0 0
7040 i
J*II 20 3.40 0
17,400 0
0 0
3520 inI 4R 4.05 12,600 28,100 0
0 0
8448 It'III 24 3.40 16,900 31,100 0
0 0
4224 III G
48 4.00 27,400 40,900 0
0 0
8448 III j
G{2) 28 3.55 31,400 43,400 8
3.03 224 4704 E
R 3.03 25,600 37,900 0
0 0
1408.
D(3) 1 3.03 22,300 34,500 0
0 0
176 i.
TOTAL 217 15,800 29,700 224 37,968
) Cerried over from Cycle 6 tc Cycle 7 of Unit 2
) Twice burned Batch E fuel discharged from Unit 2 Cycle 4 l) Twice burned Ratch 0 fuel discharged from Unit 2 Cycle 3 j) Cycle 6burnupof13,800HWG/T q) Cycle 7 hurnup of 13,900 MWO/T l
j j
1 2
J J
3 4
5 6
7 J
J H
G/
H 8
9 10 11
'12 13 J
H G
H G
G 14 15 16 17 18 19 20 J
H*
H*
H E
J*
H*
21 22 23 24 25 26 27 28 J
H R*
G G
H G/
H*
29 30 31 32 33 34 35 36 J
G H
G J*
G/
J* -
G/
37 38 39 40 41 42 43 44 H
H E
H G/
G G
H*
45 46 47 48 49 50 51 52 53 i
54 G/,
G J*
G/
J*
G H
G J
55 56 57 58 59 60 61 62
^
H G
H*
H*
G/
H*
G 0
1 SALI!MORE GA5 & ELECTRIC CO.
CALVERT CLIFFS UNIT 2 CYCt.E 7 "I**
Calvert Cliffs CORE MAP Nuclear Power Plant 3-1
=
i UNSHIMMED ASSIMety l
ll G/ 8 POISON ROO ASSEMSLY X
X
_X X
-X lX 1
IX X
PUEL RCD LOCATION
>) PolSON ROO LOCATION i
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GAS & ELECTRIC CO.
CALVERT CLIFFS UNIT 2 CYCLE 7 Calvert Cliffs ASSEMBLY FUEL AND OTHER MOD LOCATIONg 32 Nuclear hww 2'ent
. ~
l INITIAL ENRICHMENT BOC 7 BURNUP (MWD /T), EOC 6 = 13,800 MWD /T 1
J 2
J 4.05 4.05 0
0 3
J 4
J 5
H 6 G/
7 H
4.05 4.05 4.05 3.55 4.05 0
0 11,800 31,100 10,300 8
J 9 H
10 G
11 H
12 G
13 G
4.05 4.05 4.00 4.05 4.00 4.00 0
10,700 26,200 14,200 30,100 29,200 14 J
15 H' 16 H' 17 H
18
. E 19 J'
20 H'
4.05 3.40 3,40 4.05 3.03 3.40 3.40 0
16,300 16,600 13,100 25,500 0
17,500 21 J
22 H
23 H* 24 G
25 G
26 H
27 G/ 28 H*
4.05 4.05 3.40 4.00 4.00 4.05 3.55 3.40 0
10,700 16,600 26,300 27,400 15,500 32,100 16,950 29 J
30 G
31 H 32 G
33 J'
34 G/ 35 J' 36 G/
4.05 4.00 4.05 4.00 3.40 3.55 3.40 3.55 0
26,100 13,400 28,300 0
32,400 0
31,100 37 H
38 H
39 E
40 H 41 G/ 42 G 43 G 44 M'
4.05 4.05 3,03 4.05 3.55 4.00 4.00 3.40 45 J
11,300 14,200 25,600 15,600 32,400 26,800 26,800 17,450 4.05 0
46 G/
47 G
48 J' 49 G/ 50 J' 51 G 52 H
53 G
3.55 4.00 3.40 3.55 3.40 4.00 4.05 4.00 54 J
28,900 30,100 0
32,100 0
26,700 10,200 25,150 4.05 0
55 H
56 G
57 H'58 H* 59 G/ 60 H' 61 G 62 0
4.06 4.00 3.40 3.40 3.55 3.40 4.00 3.03 10,300 29,200 17,500 16,950 31,100 17,450 25,150 22,300 GA5 ELE T IC CO*
CALVERT CLIFFS UNIT 2 CYCLE 7 Figure ASSEMBLY AVERAGE BURNUP AT BOC
}'hC "'
3-3 AND INITIAL ENRICHMENT DISTRIBUTION pg
l 1
J 2
J 10,100 13,500 3
J 4
J 5
H 6 G/
7 H
11,200 14,300 25,300 42,200 25,800 8
J 9 H
10 G
11 H
12 G 13 G
I 11,900 26,400 40,300 29,900 42,700 41,700 14 J
15 H' 16 H* 17 H
18 E 19 J' 20 H*
11,900 29,600 31,300 29,700 37,900 17,600 31,950 21 J
22 H
23 H* 24 G
25 G 28 H 27 G/ 28 H'
11,200 26,400 31,200 40,100 41,400 31,000 44,400 30,750 29 J 30 G
31 H 32 G
33 J' 34 G/ 35 J' 36 G/
o 14,300 40,300 29,900 42,000 17,500 44,500 17,200 43,400 37 H
38 H
39 E 40 H
41 G/ 42 G
43 G
44 H*
25,400 29,900 37,900 31,000 44,500 39,800 40,500 31,600 g
j 10,200 46 G/ 47 G 48 J' 49 G/ 50 J'
51 G
52 H
53 G
54 J
40,300 42,700 17,600 44,400 17,300 40,600 26,900 39,250 13,500 56 H 56 G
57 H*58 H* 59 G/
60 H* 61 G 62 0
25,800 41,700 31,950 30,750 43,400 31,600 39,250 34,500 BALTIMORE GAS & ELECTRIC CO.
CALVERT CLIFFS UNIT 2 CYCLE 7 Ficure Calvert Cliffs ASSEMBLY AVERAGE BURNUP AT EOC (MWD /T) 34 Nucleer Power Plant
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4.0 FUEL SYSTEM DESIGN 4.1 Mechanical Design 4.1.1 Fuel Design The mechanical design for the Batch J reload fuel is identical to that of the Batch X fuel described in the reference cycle submittal (Calvert Cliffs Unit 1 Cycle 8, Reference 1). The mechanical designs of the Calvert Cliffs Unit 2 Batch H. G E and 0 fuel assemblies were described in References 2, 3, 4 and 5, respectively.
4.1.2 Clad Collapse A draft copy of an EPRI-sponsored report dealing with the phenomena of interpellet gap formation and clad collapse in modern PWR fuel rods was submitted to the NRC as Attachment 5 of Reference 6.
The final version of this report was subsequently issued as Reference 7 A synopsis of the report focusing on C-E manufactured modern fuel was also submitted to the NRC, at the same time as the draft report, as Attachment 4 of Reference 6.
Based upon the conclusion and recommendation of this synopsis that cycle specific clad collapse analyses are not necessary for modern C-E manufactured fuel, a cycle specific calculation was not prepared for the reference cycle.
NRC concurrence with this approach was presented in Reference 8.
Consistent with the above discussion, a cycle specific clad, collapse analysis was not prepared for Unit 2 Cycle 7.
Since clad collapsee has been effectively removed as an issue for modern C-E fuel provided the fuel design remains basically unchanged, discussion of clad collapse will be absent in subsequent ifcense submittals where the fuel design is virtually the same as that of the reference cycle.
4.1.3 Dimensional Changes All fuel assemblies in Cycle 7 with the exception of the eight Batch E fuel assemblies were reviewed for shoulder gap clearance using the SIGREEP model described in Reference 9 (approved in Reference 10).
The shoulder gap clearance evaluation for the eight Ratch E fuel assemolies was based upon their end of Cycle 7 fluences and the shoulder gap measurements taken on other fuel assemblies from the same batch of fuel.
All fuel assemblies were reviewed for fuel assembly length clearance using the refined correlation discussed in References 11 and 12. All shoulder gap and fuel assembly length clearances were found to be adequate for Cycle 7.
4.1.4 CEA Design The replacement CEAs to be utilized for the changes discussed in Section 3.3 have exactly the same design as those replacement CEAs installed in the reference cycle.
j
4.1.5 Removal of CEA Plugs Unit 2 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 7 for exactly the same reasons and with exactly the same justification as presented in support of the reference cycle.
4.1.6 Metallurgical Requirements The metallurgical requirements of the fuel cladding and the fuel assembly structural members for the Batch J fuel are identical to those of the reference cycle.
Thus, the chemical or metallurgical performance of the Batch J fuel will be the same as that of the Unit 1 Cycle 8 fuel.
4.2 Hardware Modifications to Mitigate Guide Tube Wear All standard fuel assemblies which will be placed in CEA locations in Cycle 7 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 13. A modified short sleeve design (Reference 14) which is identical to that used in the reference cycle will be used in Batch J fuel assemblies.
4.3 Thermal Design The thermal performance of a composite fuel pin which envelooes the various fuel assemblies present in Cycle 7 (fuel Ratches J, H, G E and 0) has been evaluated using the FATES 3 version of the fuel evaluation model (References 15 and 16), as approved by the NRC (Reference 17).
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 7.
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 thermal expansion factor, are being lowered (See Section 9.0).
These changes and their effects are identical to those considered in' the reference cycle.
4 5.0 NUCLEAR DESIGN 5.1 Physics Characteristics 5.1.1 Fuel Management The Cycle 7 fuel management employs a mixed central region as described in Section 3, Figure 3-1.
The fresh Batch J fuel is comprised of two sets of assemblies, each having a unique enrichment in order to minimize radial power peaking.
There are 40 assemblies with an enrichment of 4.05 wt". U-235 and 20 assemblies with an enrichment of 3.40 wt". U-235.
With this loading, the Cycle 7 burnup capacity for full power operation is expected to be between 12,600 MWD /T and 13,700 MWD /T, depending on the final Cycle 6 termination point. The Cycle 7 core characteristics have been examined for Cycle 6 terminations between 12,200 and 13,800 MWD /T and limiting values established for the safety analyses.
The loading pattern (see Section 3) is applicable to any Cycle 6 termination point between the stated extremes.
Physics characteristics including reactivity coefficients for Cycle 7 are listed in Table 5-1 along with the corresponding values from the reference cycle (Reference 1).
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.
Table 5-2 presents a summary of CEA shutdown worths and reactivity allowances for the end of Cycle 7 zero power steam line break accident and a comparison to reference cycle data.
The,EOC zero power steam line accident was selected since it is the most limiting zero power transient with respect to reactivity reouirements and, thus, provides the basis for verifying the Technical Specification required shutdown nargin.
Table 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 7 and the reference cycle.
The configurations of CEA Groups 5 and 4 are being changed as discussed in Section 3; the new configurations will be identical to those of the reference cycle.
The configuration of Group 3 wnich remains unchanged is also identical to that of the reference cycle.
The power dependent insertion limit (PDIL) curve is the same as that of the l
reference cycle.
5.1.2 Power Distributions Figures 5-1 through 5-3 illustrate the all rods out (AR0) planar radial power distributions at 80C7, MOC7 and
- E0C7, respectively, that are characteristic of the high hurnup end of the Cycle 6 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 heignt.
The high burnup end of tha Cycle 6 shutdown window tends to increase the radial power peaking in this central axial region of the core for Cycle 7.
The planar radial power distributions for the above region with CEA Grouc 5 l
fully inserted at beginning and end of Cycle 7 are shown in Figures 5 J and 5-5, respectively, for the high burnup end of the Cycle 6 shutdown window.
w
The radial power distributions descriceq in this section are calculated data without uncertainties or other allowances.
However, the single rod power peaking values do include the increased peaking that is characteristic of fuel rods adjoining the water holes in the fuel assembly lattice.
For both DNB 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 at any time during Cycle 7.
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 range of possible axial shapes.
The maximum three-dimensional or total peaking factor anticipated in Cycle 7 during normal base load, all rods out operation at full power is.1.91, not including uncertainty allowances.
5.1.3 Safety Related Data 5.1.3.1 Ejected CEA Data The Cycle 7 safety related data for this section are identical to the safety related data used in the reference cycle.
Subsequent to the submission of the reference cycle application (Reference 1), the maximum reactivity worth and planar power peak used in the analysis of the Ejected CEA Event at Hot Zero Power were revised.
Table 5 4 which contains these data is included herein solely to update these values to those used in the revised analysis.
Revision of these data did not affect the conclusions contained in Section 7.3.1 of Reference 1.
5.1.3.2 nroooed CEA Data The Cycle 7 safety related data for this section are identical to the safety related data used in the reference cycle.
5.1.3.3 Auomentation Factors A draft copy of an EPRI-sponsored report dealing with the phenomenon of interpellet gap formation in modern PWR fuel and a synopsis of this report focusing on only C-E manufactured modern fuel were submitted to the NRC as Attachments 5 and a, respectively, of Reference 2.
These documents presented the analyses which demonstrate that the increased power peaking associated with the small interpellet gaps found in modern, i.e, pre-pressurized and non-densi fying, fuel is insignificant compared to the uncertainties in the safety analyses and Tech.
Specs.
Consequently, augmentation factors were eliminated from the reference cycle analysis and the Unit 1 Tech. Specs.
NRC concurrence with this approach was presented in Reference 3.
Consistent with the above discussion, augmentation factors were eliminated from the Unit 2 Cycle 7 analysis and are being proposed for elimination in the Unit 2 Tech. Specs. (see Section 9).
Since gao formation and the resulting need for augmentation factors has been effectively removed as an issue for modern C-E fuel provided the fuel design remains basically unchanged, discussion of augmentation factors will be absent in subsecuent license submittals where the fuel design is virtually the same as that of the reference cycle.
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5.1.3.4 Fuel Temoerature Coefficient Bias The Cycle 7 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 In-core detector measurement constants to be used in evaluating the reload cycle power distributions will be calculated in the same manner as those for the reference cycle.
5.3 Nuclear Design Methodology Analyses have been pr:rformed in the same manner and with the same methodologies used for the reference cycle analyses.
5.4 Uncertainties in Measured Power Of stributions The power distribution measurement uncertainties to be applied to Cycle 7 are the same as those applied to the reference cycle.
o
i TABLE 5-1 CALVERT CLIFFS UNIT 2 CYCLE 7 NOMINAL PHYSICS CHARACTERISTICS Reference Cycle Unit 2 Units (Unit 1 Cycle R)
Cycle 7 Dissolved Boron Hot Full Power, All Rods Out Equilibrium Xenon Boron Content for Criticality at BOC PPM 1200 1090 Boron Worth Hot Full Power BOC PPM /taa 107 107 Hot Full Power EOC PPM /%3o A8 87
- Reactivity Coefficients (CEAs Withdrawn)
Moderator Temperature Coefficients, Hot Full power, Equilibrium Xenon Beginning of Cycle 10*#ao/ F
-0.1
-0.1 U
End of Cycle 10*#ao / F
-2.3
-2.3 0
l Dopoler Coefficient 10-5,joF
-1.56
-1.57 i
Hot Zero Power BOC 3
10-5,j F
-1.26
-1.25 3 o Hot Full Power BOC Hot Full Power EOC 10-533 f F
-1.44
-1.aa o
l Total Effective Delayed Neutron Fraction, Saff BOC 0.00604 0.00593 EOC 0.00516 0.00514 Neutron Gener1 tion Time,1*
ROC 10~6 sec 23.1 23.1 EOC 10-6 sec 23.2 2R.A 54
TABLE 5-2 CALVERT CLIFFS UNIT 2 CYCLE 7 LIMITING VALUES OF REACTIVITY WORTHS AND ALLOWANCES FOR THE ENO-OF-CYCLE (EOC) NOT ZERO POWER (HZP)
STEAM LINE RUPTURE ACCIDENT, %Ao Unit 2 Reference Cycle
- Cycle 7 I
1.
Worth of all CEA's Inserted 9.3 8.7 2.
Stuck CEA Allowance 2.5+
2.1 3.
Worth of all CEA's less Worth 6.8' 6.6 of CEA Stuck Out**
4 power Dependent Insertion 1.8 1.8 Limit CEA Bite at Zero Power 5.
Calculated Scram Worth 5.0+
4.8 6.
Physics Uncertainty plus 0.7*
0.6 Bias 7
Net Available Scram Worth 4.3 4.2 8.
Technical Specification 3.5 3.5 Shutdown Margin 9.
Margin in Excess of Technical 0.8 0.7 Specification Shutdown Margin
- Unit 1 Cycle 8.
- These are revised values; revision of these values was made soon after the submission of Reference 1 and did not affect the conclusion demonstrated in i
Table 5-2 of Reference 1.
s-s
e TABLE 5-3 CALVERT CLIFFS UNIT 2 CYCLE 7 REACTIVITY WORTH OF CEA REGULATING GROUPS AT HOT FULL POWER, %au Beginnina of Cycle End of Cycle Regulating Reference
- Unit 2 Reference
- Unit 2 CEA's Cycle Cycle 7 Cycle Cycle 7 Group 5 0.9.8 0.28 0.36 0.39 Group A 0.82 0.83 0.41 0.88 Group 3 0.94 0.91 1.02 1.02 Note Values shown assume sequential group insertion.
- Unit 1 Cycle a o
I i
i 5-6 t
i TABLE 5-4 CALVERT CLIFFS UNIT 2 CYCLE 7 CEA EJECTION DATA Limiting Values Reference Cycle Unit 2 Cycle 7 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 Ranks 5+4+3 inserted; worst CEA ejected R.9+
R.9 Maximum Ejected CEA Worth (% ao)
Full power with Bank 5 inserted; o
worst CEA ejected 0.28 0.2R o
Zero power with Banks 5+4+3 inserted; worst CEA ejected 0.77+
0.77
- Unit 1 Cycle 8 (Reference 1)
- These are revised values; revision of these values was made soon after the submission of Reference 1 and did not affect the conclusions contained in Section 7.3.1 of Reference 1.
Notes 1.
Uncertainties and allowances are included in the above data.
2.
The safety analysis values are conservative with respect to the actual calculated values.
l l
l 5-7
1 l
J l
1 2
0.72 1.00 4
3 4
5 6
7 0.84 1.00 0.99 0.78 1.14 I
X 8
9 10 11 12 13 0.89 1.22 1.03 1.16 0.87 0.85 l
14 15 16 17 18 19 20 0.89 0.98 1.07 1.24 0.85 1.30 1.03 21 22 23 24 25 26 27 28 0.83 1.22 1.07 0.97 0.98 1.12 0.87 0.97 29 30 31 32 33 34 35 36 1.07 1.03 1.22 0.98 1.28 0.84 1.24 0.86 37 38 39 40 41 42 43 44 0.99 1.15 0.84 1.11 0.84 0.88 0.94 0.98 46 0.71 46 47 48 49 50 51 52 53 0.79 0.86 1.29 0.86 1.26 0.96 1.18 0.94 j
56 56 57 58 59 60 61 62 i
1.14 0.85 1.03 0.97 0.86 0.98 0.94 0.78 i
i NOTE: X = MAXIMUM 1.P PEAK = 1.52 BALTIMORE CALVERT CLIFFS UNIT 2 CYCLE 7 Figure GAS & ELECTRIC CO.
ASSEMBLY RELATIVE POWER DENSITY AT BOC l
Calvert Cllifs EQUILlBRIUM XENON 51 Nuclear Power Plant O
__-,--_.m.
.._,,____.,_-.-,...__.__-.,._.-_,_.~~,--,,,-c
.,m.-
r 1
2 0.74 1.00 3
4 5
6 7
0.81 1.04 0.99 0.81 1.15 8
9 10 11 12 13
~
0.87 1.15 1.00 1.15 0.89 0.89 14 15 16 17 18 19 20 0.87 0.95 1.04 1.20 0.86 1.28 1.04 21 22 23 24 25 26 27 28 0.81 1.15 1.04 0.96 0.97 1.13 0.88 0.99 29 30 31 32 33 34 35 36 1.04 1.00 1.19 0.95 1.25 0.86 1.25 0.88 37 38 39 40 41 42 43 44 0.99 1.15 0.86 1.12 0.86 0.92 0.96 1.01 45 0.75 46 47 48 de 50 51 52 53 0.83 0.89 1.29 0.89 1.26 0.97 1.20 0.98 y
X l*#
56 56 57 58 59 60 61 62 1.15 0.89 1.04 0.99 0.88 1.01 0.98, 0.84 NOTE: X = MAXIMUM 1 PIN PEAK = 1.45 BALTIMORE CALVERT CLIFFS UNIT 2 CYCLE 7 Rgure GAS & ELECTRIC CO.
ASSEMBLY RELATIVE PCWER DENSITY AT 7 GWD/T Calvert Cliffs EQUILIBRIUM XENON 52 Nuclear Power Plant
e e
l I
2 0.80 1.03 X
3 4
5 6
7 4
0.82 1.04 1.01 0.86 1.17 I
8 9
10 11 12 13 0.88 1.11 0.99 1.13 0.92 0.91 l
~
14 15 16 17 18' 19 20 0.88 0.94 1.01 1.16 0.88 1.25 1.02 21 22 23 24 25 26 27 28 0.82 1.11 1.01 0.95 0.96 1.11 0.90 0.99 s
29 30 31 32 33 34 35 36 1.04 0.99 1.16 0.94 1.22 0.88 1.22 0.89 37 38 39 40 41 42 43 44 1.01 1.13 0.88 1.11 0.88 0.92 0.96 1.00 46 0.80 46 47 48 49 50 51 52 53 54 0.87 0.92 1.24 0.90 1.22 0.97 1.18 0.97 1.01 86 56 57 58 59 60 61 62 1.17 0.91 1.02 0.99 0.89 1.00 0.97 0.86 NOTE: X = MAXIMUM 1 PIN PEAK = 1.44 GAS & El.E T IC CO*
CALVERT CLIFFS UNIT 2 CYCLE 7 Figure Calvert Cliffs ASSEMBLY RELATIVE POWER DENSITY AT EOC 53 Nuclar Pm Plant EQUILIBRIUM XENON
I N-
[ CEA BANK 5 0.66 0.86
/'
LOCATION
/
3 4
5 6
7 0.89 1.12 0.96 0.67 0.90 lll 8
9 10 11 12
,13 j//
i 0.96 1.30 1.07 1.13 0.73
/ 0.48
/ /
6 14 15 16 17 18 19 20 l
0.96 1.06 1.15 1.30 0.85 1.19 0.88 l
21,
22 23 24 25 26 27 28 u
0.09 1.30 1.15 1.04 1.05 1.16 0.86 0.96
)
i' 29 30 31 32 33 34 35 36 1.1n 1.07 1.29 1.02 1.36 0.89 1.30 0.90 L
X M
M M
M 41 42 4
4 y
0.95 1.12 0.85 1.15 0.89 0.95 1.01 1.06 i
0.2,
46 47 48 49 50 51 52 53 0.68 0.73 1.19 0.86 1.32 1.04 1.29 1.03 56 57 58 59 60 61 62'
/-
0.00 0.48 0.88 0.96 0.30 1.06 1.03 0.80
/
/
NOTS: X = MAXIMUM 1.P!N PEAK = 1.59
^
N L
GAS ELE T IC CO*
CALVERT CLIFFS UNIT 2 CYCLE 7 Figure ASSEMBLY RELATIVE POWER DEN 3lTY WITH BANK 5 54 Nuc ow Plant S RTED, HP, BOC
//
1 2
CEA BANK 5
//
LOCATION 0.71 0.87 3
4 5
6 7
0.88 1.06 0.95 0.72 0.89 8
9 10 11 12 13 0.96 1.20 1.03 1.09 0.74
/ 0.48
//
14 15 16 17 18 19 20 0.96 1.04 1.10 1.23 0.88 1.13 0.85 21 22 23 24 25 26 27 28 0.88 1.20 1.10 1.03 1.03 1.15 0.90 0.98 29 30 31 32 33 34 35 36 1.06 1.03 1.23 1.02 1.32 0.94 1.30 0.95 37 38 33 40 41 42 43 44 0.96 1.09 0.88 1.15 0.94 1.01 1.05 1.10 45 46 47 48 49 50 51 52 53 l
0.73 0.75 1.13 0.90 1.31 1.06 1.30 1.07
{
X l
55
,56 57 58 59 60 61
,62' O.89
/ 0.48 0.85 0.98 0.95 1.10 1.07
/ 0.86 I
NOTE: X = MAXIMUM 1-PIN PEAK = 1.47 i
GAS ELE T !C CO*
CALVERT CLIFFS UNIT 2 CYCLE 7 Figure Calvert Cliffs ASSEMBLY RELATIVE POWER DENSITY WITH BANK 5 55 Nuclear Power Plant RTED, HFP, EOC t
\\
t 6.0 THERMAL HYDRAULIC DESIGN 6.1 DNBR Analysis Steady state DNBR analyses of Cycle 7 at the rated power level of 2700 MWt have been performed using the TORC 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.
A variant of TORC called CETOP, optimized for simplified modeling applications, was used in this cycle to develop the " design thermal margin model" described generically in Reference 3.
Details of CETOP are I
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 significantly; no margin gain is expected or taken credit for.
Table 6-1 contains a list of pertinent thermal-hydraulic design parameters applicable to both safety analyses and the generation of reactor protective system setooint 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 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 S.
The statistically derived DNRR limit is being reduced from the value of 1.23 to a value of 1.21.
The 1.23 value was approved in Referenca. 5.
The reduction results from NRC approval in the Safety Evaluation Report (SER) of Reference 7 of a reduced CE-1 DNBR limit for CE's 14x14 fuel.
At the l
' time the Statistical Combination of Unceru inties (SCU) analysi,s was l
approved for the Calvert Cliffs units, NRC review of Reference 8 which dealt with the applicability of the CE-1 CHF correlation to rods with I'
nonuniform Axial Power Distributions (APD) was incomolete.
An interim CE-1 DNBR limit of 1.19 was thus used in the original SCU analysis.
In the SER (Reference 7) on CE's nonuniform APD topical report, the NRC reduced the CE-1 DNBR limit from 1.19 to 1.15 for 14x14 fuel.
The SCU DNBR limit is being correspondingly reduced from 1.23 to 1.21.
The 1.21 SCU DNBR limit includes the following penalties imposed by the NRC in their review of the SCU analysis (Reference 5):
Critical Heat Flux (CHF) correlation cross validation penalty (5%
increase in standard deviation of CHF correlation uncertainty distribution),
T-H code uncertainty penalty (5%, equal to two standard deviations).
The 1.21 SCU ONBR limit also includes a 0.006 DNBR rod bow penalty which accounts for the adverse effects of rod bowing on CHF for la 14 fuel with x
burnup not exceeding 45 GWD/T.
The axial fuel densification factor is being reduced from 1.01 to 1.002 to make it consistent with existing calculations.
The value of this parameter is being changed in the Unit 2 Tech. Specs via this submittal (see Section 9). This change was made in other safety analyses for the reference cycle and an identical change was made in the Unit 1 Tech. Specs. via the reference cycle submittal (Reference 9).
6-1
6.2 Effects of Fuel Rowing on DNBR Margin The effects of fuel-rod bowing on DNB margin for Calvert Cliffs Unit 2 Cycle 7 have been evaluated using the methods described in Reference 10.
These methods were approved by NRC in Reference 11.
Based upon these methods, a penalty of 0.006 DNBR units is required to account for the advsrse T-H effects of red bcw at an assembly average burnup of 45 GWD/T.
This penalty was included in the derivation of the newly developed DNBR limit discussed above.
For Unit 2 Cycle 7 no assembly is projected, based upon the burnup ranges discussed in previous sections, to have an average burnup greater than 45 GWD/T (see Figure 3-4).
S O
e O
e A
4
e TABLE 6-1 CALVERT CLIFFS UNIT 1 THERMAL-HYDRAULIC PARAMETERS AT FULL POWER **
Reference +
Unit 2, General Characteristics Unit Unit 1, Cycle 8 Cycle 7 Total Heat Output (core only)
MWg 2700 2700 10 BTU /hr 9215 9215 Fraction of Heat Generated
.975
.975 In Fuel Rod Primary System Pressure psia 2250 2250 (Nominal)
Inlet Temperature OF 548 548 Total Reactor Coolant Flow gpm 381,600 381,600 (steady state) 10 lb/hr 143.8 143.8 D
6 Coolant Flow Through Core 10 lb/hr 138.5 138.5 Hydraulic Diameter ft 0.044 0.044 (nominal channel)
Average Mass Velocity 10" lb/hr-ft2 2.59 2.59 Pressure Orop Across Core psi 11.1 1k.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) 2 Core Average Heat Flux BTU /hr-ft 182,300***
180,700****
(Accounts for above fraction of heat generated in fuel rod and axial densification factor) 2 Total Heat Transfer Area ft 49,300***
49,700****
(Accounts for axial densification factor) 2 Film Coefficient at Average BTU /hr-ft,og gg3g 593g Conditions 6-3
TABLE 6-1 (continued)
Reference +
Unit 2, General Characteristics Unit Unit 1, Cycle 8 Cycle 7 O
Average Film Temperature F
31 31 Difference Average Linear Heat Rate of kw/ft 6.09***
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 OF 657 657 Temperature Reference' Calculational Factors Unit 1, Cycle 8 Unit 2, Cycle 7 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) 1.01
- 1.002
- Notes
- These factors have been combined statistically with other uncertainty 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 verified to be applicable to Cycle 7.
- 0ue to the statistical combination of uncertainties described in References 6,
12, and 13, the nominal inlet temoerature and nominal primary system pressure were used to calculate some of these parameters.
- Based on a value of 256 shims and 5 non-fuel rods.
- Based on a value of 224 shims.
- Reference cycle (Unit 1, Cycle 8) analysis is contained in Reference 3.
"This value was conservative with respect to existing calculations.
"This value is consistent with existing calculations.
6-4
I 7.0 TRANSIENT ANALYSIS The Design Basis Events (08Es) considered in the Unit 2 Cycle 7 safety analyses are listed in Table 7-1.
Core parameters input to the safety analyses for evaluating approaches to DNB and centerline temperature to melt fuel design Ifmits are presented in Table 7-2.
As indicated in Table 7-1, no reanalysis was performed for any of the DBEs since key transient input para.neters all lie within the bounds (conservative with respect to) of the reference cycle values (Unit 1, Cycle 8 Reference 1). The results and conclusion's quoted in the reference cycle analysis for all DBEs are valid 4or Unit 2, Cycle 7.
t e
7-1
TABLE 7-1 CALVERT CLIFFS UNIT 2, CYCLE 7 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 Roron Oilution Not Reanalyzed 7.1.2 StargupofanInactiveReactorCoolant 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 7.1.5 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:
7.2.1 Sequential CEA Group Withdrawa1 2 Not Reanalyzed 3
7.2.2 Loss of Coolant Flow Not Reanalyzed 7.2.3 Full Length CEA Orop Not Reanalyze.d 7.2.4 Transients Resulting from the Not Reanalyzed 3
MalfunctionofOgeSteamGenerator 7.2.5 Loss of AC Power Not Reanalyzed 7.3 Postulated Accidents 7.3.1 CEA Ejection Not Reanalyzed 7.3.2 Steam Line Rupture Not Reanalyzed 7.3.3 SteamGeneragorTubeRupture Not Reanalyzed 7.3.4 Seized Rotor, Not Reanalyzed 1Technical Specifications preclude this event curing operation.
2Requires High Power and Variable High Power Trip.
3Requires Lcw Flow Trip.
- Requires trip on high differential steam generator pressure.
7-2
TABLE 7-2 CALVERT CLIFFS UNIT 2, CYCLE 7 CORE PARAMETERS INPUT TO SAFETY ANALYSES-FOR ONB AND CTM (CENTERLINE TO MELT) DESIGN LIMITS Reference Cycle **
Physics Parameters Units (Unit 1, Cycle 8)
Unit 2, Cycle 7 Radial Peaking Factors For ONB Margin Analyses T
(Fr)
Unrodded Region 1.70*'*
1.70***
Bank 5 Inserted 1.87**+
1.g7**+
For 01anar Radial Component (F
) of 3-D Peak (C M Limit Analyses)
Unrodded Region 1.70*
1.70*
Bank 5 Inserted 1.87*
1.87*
Maximum Augmentation
- 1.0 1.0 Factor Moderator Temperature 10-#ao/ F
-2.7 - +.7
-2. 7 - +. 7 0
Coefficient Shutdown Margin (Value
%ao
-3.5
-3.5 Assumed in Limiting EOC Zero Power SLB)
Tilt Allowance 3.0 3.0
- 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 ONB 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
+The values assumed are conservative with respect to the Technical Specification limits.
++Since the need. for augmentation factors has been effectively removed as an issue (see Section 5.1.3.3), this item will be eliminated in future license submittals.
7-3
TABLE 7-2 (continued)
Reference Cycle Safety Parameters Units (Unit 1, Cycle 8)
Unit 2, Cycle 7 Power Level MWt 2700*
2700*
O Maximum Steady State F
548*
548*
Temperature 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**+
E LC0 Extreme Assumed at Full Power (Ex-Cores)
Maximum CEA Insertion
". Insertion 25 25 at Full Power of Rank 5 Maximum Initial Linear KW/ft 16.0 16.0 Heat Rate for Transient 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) 1.23*
1.21*
- 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 ON8 and CTM limits are detailed in References 2, 3 ar.d 4 These procedures have been approved by NRC for the Calvert Cliffs Units in Reference 5.
+The values assumed are conservative with respect to the Technical Specification limits.
7-J
8.0 ECCS ANALYSIS 8.1 Large Break Loss-of-Coolant Accident 8'.1.1 Introduction and Summary An ECCS performance analysis was performed for Calvert Cliffs Unit 2 Cycle 7 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 7 consisted of an evaluation of the differences between Cycle 7 and the reference cycle, Unit 1 Cycle 8.
Acceptable ECCS performance was demonstrated for the reference cycle in Reference 2 and approved by the NRC in Reference 3.
The NRC approved C-E large break evaluation model (Reference 4) was used to re-evaluate ECCS performance for the limiting large break LOCA.
The blowdown hydraulic calculations, refill /reflood hydraulics calculations and steam cooling heat transfer coefficients of the reference cycle apply to Unit 2,
Cycle 7 since there have been no significant changes to RCS, hardware characteristics, Therefore, only the fuel rod conditions specific to Cycle 7 require a re-evaluation of the ECCS hot rod clad temperature and oxidation calculations. The NRC approved STRIKIN-II (Reference 5) code was used for this purpose.
Hot rod augmentation factors were removed, as was done for the reference cycle.
Burnup dependent calculations were performed to determine the limiting conditions for the ECCS performance analysis.
The nuclear and fuel thermal performance (FATES 3 (Reference 6)) data used as input to the ECCS analysis considered high burnup effects specific to the Cycle 7 reload. Two STRIKIN-II temperature calculations were performed:
one at a low rod average burnup which yields the maximum initial fuel stored energy and one at a high burnup which yields the highest initial rod pressure.
The low burnup maximum fuel stored energy case was demonstrated to be limiting.
The temperature calculations for both cases were performed for the 1.0 Double Ended Slot at Pump Discharge (OES/PD) break.
The break spectrum analysis performed for Unit 1 Cycle 2 (Reference 7) determined that the 1.0 OES/PO is the ifmiting break since it yields the highest residual fuel stored energy at the end of the blowdown period and, therefore, yields the highest peak cladding temoerature during the reflood period.
8.1.3 Results 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 comoarison purposes, the corresponding values of the reference cycle analysis (Unit 1 Cycle 81 are also presented in Tables 8.1-1 and 8.1-2.
A list of the significant parameters disolayed graphically for the limiting case (Figures 8.1-1 through 8.1-6) is presented in Table 8.1-3.
8-1
- 4
~ _ _ _.
..--.__u_
The results of the evaluation confirm that 15.5 kw/ft is an acceptable value for the PLHGR in Cycle 7.
As shown in Table 8.1-1, the peak clad temperature, maximum local oxidation and core wide clad oxidation values 0
of 14a5 F, 5.4% and <.51%,
respectively, are well below the 10CFR50.46 acceptance criteria limits of 220n F, 17% and 1 *., respectively.
These results have been confirmed for up to 100 plugged tubes per steam generator.
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 of 1945 F.
The transient results for the limiting case are presented in Figures 8.1-1 to 8.1-6 The fuel cladding is predicted to rupture, as well as achieve its peak temperature, during the reflood period (at 33 seconds and 261 seconds, respectively).
The high burnuo (50,100 MWO/MTU) case resulted in a peak clad temperature 0
0 of 1929 F, 16 F 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.
For the high burnup case, at 15.5 kw/ft the pin pressure was sufficient to cause an earlier hot rod rupture time.
However, the lower initial stored energy at that burnup causes a lower peak clad temoerature 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 sumarized by the analysis results presented in Table 8.1-1.
The most limiting case results 0
. in a peak clad temperature of 1945 F, which is well below the acceptance U
limit of 22dO F.
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 1%, respectively.
Therefore, coeration of Unit 2 Cycle 7 at a PLHGR of 15.5 kw/ft and a power level of 2754 MWT (102% of 2700 MWT) is in compliance with the 10CFR50.26 acceptance criteria.
Reference 8 notified the NRC of a potential non-conservatism in one element of the C-E large break LOCA evaluation model (Reference 4).
As stated in Reference 8, the effect of this non-conservatism on peak clad temperature (PCT) is expected to be well within the 255 margin to the U
0 2200 F PCT limit which the current analysis demonstrates.
Consequently, it has been concluded that this potential non-conservatism will not alter the previously discussed determination that the operation of Unit 2 Cycle 7 would be in compliance with the 10CFR50.46 acceptance criteria.
3-2
TABLE 8.1-1 Summary of ECCS Performance Results for Calvert Cliffs 2, Cycle 7 for the Limiting Break Size (1.0 DES /PD) as Compared to Reference Cycle (Unit 1, Cycle 8)
Limiting Case High Burnup Case (Maximum Initial Fuel (Maximum Initial Rod Parameters Stored Energy)
Pressure)
Unit 1 Unit 2 Unit 1 Unit 2 Cycle 8 Cycle 7 Cycle 8 Cycle 7 Rod Average Burnup 943 943 50,000 50,100 MWD /MTU Peak Clad Temperature IR36+
1945 1822+
1929 (PCT),OF Time of PCT, Seconds 254+
261 254+
263 Time of Clad Rupture, 35.0+
33.0 10.5+
10.7 Seconds Peak Clad Oxidation, %
3.70+
5.40 3.75*
5.40 Core Wide Oxidation, %
<.51
<.51 4.55
<.51
+These are revised values; revision of these values was made soon after the submission of Reference 2 and did not affect the conclusions contained in Section 8.1 of Reference 2.
o 8-3
TABLE 8.1-2 Calvert Cliffs Unit 2 Cycle 7 Fuel Parameters as Compared to Unit 1 Cycle 8 Unit 1 Unit 2 Cycle 8 Cycle 7 Fuel Parameters Values Values Reactor Power Level (102% of Nominal) MWT 275a 2754 Average Linear Heat Rate (102% of Nominal) kw/ft 6.37 6.37 Hot Channel Peak Linear Heat Generation 15.5 15.5 Rate kw/ft Hot Assembly Peak Linear Heat Generation 12.52 13.42 Rate kw/ft 2
- Gap Conductance at PLHGR (B/hr-ft UF) 2057*
1937
- Fuel Centerline Temperature at PLHGR (UF) 3655' 3649
- Fuel Average Temperature at PLHGR ( F) 2221+
2228
~
- Hot Rod Gas Pressure (psia) 1186*
1188
- Hot Rod Burnup (MWD /MTU) 943 943 Mot Rod Augmentation Factor (Maximum) 1.00 1.00
~
- Are initial fuel rod parameters, in STRIKIN-II, which yield maximum PCT.
+These are revised values; revision of these values was made soon after the submission of Reference 2 and did not affect the conclusions contained in Section 8.1 of Reference 2.
1 i
8-4
TABLE 8.1-3 Calvert Cliffs Unit 2 Cycle 7 Reload Analysis plots for Limiting Case Variable Figure Number Peak Clad Temperature 8.1-1 Hot Spot Gap Condutance 8.1-2 Peak Local Clad Oxidation 8.1-3 Temperature of Fuel Centerline, 8.1 4 Fuel Average, Clad and Coolant at Hottest Node Hot Spot Heat Transfer Coefficient 8.1-5
- Hot Rod Internal Gas Pressure 8.1-6 O
e G
,s wr
-,-w gr
-g
FIcuaE 8.1-1 CALVERT CLIFFS 2 CYCLE 7 1.0 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG PEAK CLAD TEMPERATURE 2200 2000
~
/
1800 s
's
/
g l
i
/
~ ~
1600 f
,, ~ ~ ~, '
p f\\
iE' 1400 l
',j PEAK CLAD MERAME tCCE b
RUPTURE NCCE t
1200 l,'
18 F
1000 800 l
600 100 200 300 400 500 S00 70C TIME SECONOS 8-6
FIGURE 8.1-2 CALVERT CLIFFS 2 CYCLE 7 1.0 x DOUBLE ENDED SLOT BREAK IN PUNP DISCHARGE LEG HOT SPOT GAP CONDUCTANCE 1600 1400 1200 1000 y
I 800 f2ii
<5
.=
$ 600 i
Oa5
[ 4CG l 5
200 100 200 300 400 500 SCO 700 TIME.
SECONOS
FIsuas 8,'l-3
~
CALVERT CLIFFS 2 CYCLE 7 1,0 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG PEAK LOCAL CLAD 0.XIDATION 16 4
i 14 12 10 w
s 8
~
E 5
PEAK CLAD TENE'RATLRE NCCE g
---aumas sooE d
6 4
/
i i
,i f
2 Y
100 200 300 400 500 S00 7Cn
~
- TIME, SECONOS 8-8
FIsuas 8.1 if CALVERT CLIFFS 2 CYCLE 7 1.0 x DOLBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG TEMPERATURE OF FUEL CENTERLINE, AVERAGE FUEL, CLAD AND COOLANT OF HOTTEST NODE 3500 3000 2500 RJE. CENTERLINE
/
/
. AVERAGE FUEL 2000 k
"o f
m1500 f
1000 500 s
ccCLANT
\\
/
100 200 300 400 500 600 7C 7IME.
SECONOS 8-9
Frauas 8. 1-5 CALVERT CLIFFS 2 CYCLE 7 1.0 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE HOT SPOT HEAT TRANSFER COEFFICIENT 160 140 C
120 4
-s iE
=
_ g 100
~
=. -
b
$ 80 5
E=
60 5
=
40 20 k
100 200 300 400 500 600 "00 TIME.
SECONOS
FIGURE 8.1-6 CALVERT CLIFFS UNIT 2 CYCLE 7 1.0 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG HOT ROD INTERNAL GAS PRESSURE 1409 P
= 1188 PSIA j INITIAL 1200./
~
~
RUPTURE AT 33.0 SEC 5
800
!!2 J5 h
600 400 200 0
0 20 40 60 80 100 TIME, SECONDS 8-11
R.2 Small Rreak Loss-of-Coolant Accident Analyses have confirmed that the reported small break loss-of-coolant accident results (Reference 2) of the reference cycle - Unit 1 Cycle 8 -
bound Calvert Cliffs Unit 2 Cycle 7.
These results have been approved by the NRC in Reference 3.
Therefore, acceptable small break LOCA ECCS performance is demonstrated at a peak linear heat rate of 15.5 kw/ft and a reactor power level of 2754 Wt (102% of 2700 Wt ).
This acceptable performance has been confirmed with up to 100 plugged tubes per steam generator.
S 8-12