ML20149F369
| ML20149F369 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 02/12/1988 |
| From: | Tiernan J BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20149F376 | List: |
| References | |
| NUDOCS 8802170122 | |
| Download: ML20149F369 (66) | |
Text
'1 l
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BALTIMORE 1
GAS AND ELECTRIC CHARLES CENTER P. O.lLX.475 BALTIMORE. MARYLAND 21203 JOSEPH A.TIERNAN wer PacsiotNr NUCLEAR ENERGY February 12, 1988 U.S Nuclear Regulatory Commission Washington,DC 205"
ATTENTION:
Document Control Desk
SUBJECT:
Calvert Cliffs Nuclear Power Plant Unit Nos. 1 and 2; Docket Nos. 50-317 and 50-318 Request for Amendment i
Unit One Tenth Cycle License Application; I
Unit Two Axial Shape Index Region Enlarzement
REFERENCES:
(a) Letter from J.A. Tiernan (BG&E), to Document Control Desk t
(NRC), Docket No. 50-318, "Request for Amendment, Eighth Cycle License Application," dated February 6, 1987.
(b) Letter from R.A.
Capra (NRC), to J.A.
Tiernan (BG6E),
"Amendment No.
108 to Facility Operating License No.
DPR-69," dated May 4, 1987.
(c)
Letter from S.A.
McNeil (NRC), to J.A.
Tiernan (BC6E),
"Safety Evaluation of Topical Report CEN 348(B)-P, ' Extended Statistical Combination of Uncertainties' (TACS 64985 and 64086)," dated October 21, 1987.
Gentlemen:
The Baltimore Gas and Electric Company hereby requests an Auendment to its operating license for Calvert Cliffs Unit No. I to allow operation for a tenth cycle. The enclosure presents a detailed description of the required Standard Technical Specifications with supporting safety analysis information to ensure conservative operation at a rated thermal power of 2700 MWth for Unit 1 Cycle 10.
Our intention is to begin the Unit i refueling outage on April 8, 1988, and to complete the outage and begin the first Cycle 10 arproach to criticality on May 19, 1988, with a return to 100% power operation on May 26, 1988. The initial heatup to support the criticality will occur such.that Unit 1 Cycle 10 will enter MODE 4 on May 17, 1988. BG6E requests that approval for this amendment request be granted prior to that date (i.e., May 17, 1988) so as not to igpg the scheduled Unit 1 startup date, nr 8802170122 880212' PDR ADOCK 05000317 MCdC.k
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Document Control Desk February 12, 1988 Page 2 Unit 1 Cycle 10 is the first twenty-four month cycle reload core for Unit 1 at Calvert Cliffs Nuclear Power Plant (CCNPP). This is the second twenty-four month relosd core at CCNPP, with the first being the Unit 2 Cycle 8 reload core currently in operation. Like Unit 2 Cycle 8, Unit 1 Cycle 10 uses a low-leakage core design wherein previously used fuel assemblies - are placed around the periphery of the core with fresh assemblies located predominantly in the central core region. This configuration (fresh fuel enriched to 4.08 w/o) provides an expected cycle length of approximately 22,000 MWD /T.
The entire Unit 1 Cycle 10 reload transient safety analysis is enveloped by the analyses of the reference cycle, Unit 2 Cycle 8 (unless otherwise stated),
previously found acceptable by the NRC (as requested under Reference (a), and as approved under Reference (b)).
Eighteen design basis events were reviewed to determine the effect the Unit 1 Cycle 10 core design has on existing approved safety analysis. It has been determined that no design basis event, for Unit 1 Cycle 10, is outside of the results and conclusions of the safety analyses presented in the reference cycle.
As a result, no safety analyses for any of the design basis events are presented in this request for operating license amendinent.
With one exception, no new design methodologies were employed in designing the Unit 1 Cycle 10 reload core that were not used for previous reload core designs and hence previously approved by the NRC. The one e.:ception is the use of an improved method of catistically combining thermal-hydraulic uncertainties so as to remove unnecessary conservatism in th.
i. termination of the Departt.re from Nuclear.e Boiling Ratio (DNBR). This methm ogy is referred to as the Extended Statistical Combination of Uncertainties (ESCU),
which was approved by Reference (c).
All Technical Specification changes requested to support the reload core for Unit 1 Cycle 10 are presented in Change No. 1 below.
The Baltimore Gas & Electric Company also requests an amendment to its operating license for Calvert Cliffs Unit 2 to change three figures in the Unit 2 Technical Specifications. These figures broaden the range of core operability at lower power levels with respect to Axial Shape Offset limits. This reduces the risk of unnecessary reactor trips at lower power levels later in cycle life.
Change No. 2 lists those changes to the Technical Specifications requested for Unit 2.
The justification for these changas is included in the Attachment to this submittal.
CHANGE NO. 1 (BG&E FCR 88-3000)
The Technical Specification changes requested herein are to make the Calvert Cliffs Unit 1 Technical Specifications consistent with the analyses presented in the Attachment to this Unit 1 Cycle 10 Request for License Amendment. These Technical Specffication change requests are summarized as follows:
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e 6
Document Control Desk February 12, 1988 Page 3 1.
Modify Figure 2.2-2, page 2-12, as indicated in the enclosed Attachment to
)
accommodate the implementation of the 'ESCU methodology used in the thermal-hydraulic analysis of Unit 1 Cycle 10. This modification changes j
the coefficients in the PVAR equation on the figure.
)
2.
Replace Figure 2.2-3, page 2-13, with the revised Figure 2.2-3,.in the enclosed Attachment to accommodate the implementation of the ESCU methodology used in the reload design.
l 3.
Modify the text of.B 2.1.1, pages B 2-1 and B 2-3 as indicated in the enclosed Attachment. This text modification reflects the implementation of the ESCU methodology, including the replacement of the specific minimum i
DNBR value, with wording as approved by Reference 3 of Section 9 of the Attachment.
4.
Modify B 2.2.1, pages B 2-5 and B 2-6, to replace the minimum DNBR value of 1.21 with the phrase as indicated in the enclosed Attachment. The specific minimum DNBR value is being replaced, as part of the Juplementation of the j
ESCU methodology.
)
S.
Replace 3.1.1.1, 4.1.1.1.1 and 4.1.1.1.2, pages 3/4 1-1 and 3/4 12 (old) with the enclosed modified Technical Specifications which have been condensed to one page (3/4 1-1), leaving page 3/4 1-2 for use by the new Figure 3.1-lb.
This removes the applicability of these Technical Specifications to the Critical Modes.
J 6.
Insert c losed new Figure 3.1-lb after page 3/4 1-1. This becomes the new l
page 3/4 1-2. The shutdown margin is changed from a constant value to text which refers to a new Figure 3.1-b.
that provides a linearly increasing shutdown margin over cycle lifetime. This new figure also accommodates a l
revisea EOC shutdown margin of 5.0% delta k/k to supp3rt the Steam Line Rupture analysis.
7.
Modify 3.1.1.4, page 3/4 1-5, an indicated in the enclosed Attachment, which condenses subsections a and b, to a new subsection a that refers to a new Figure 3.1-la, reflecting a change in the allowable Moderator Temperature Coefficient (MTC) limit.
This change accommodates implementation of the 24-month cycle for Unit 1, eliminates startup delays, and facilitates a rapid power ascension program. Also, change subsection c to subsection b to account for section letter designator shifts.
8.
Insert enclosed new Figure 3.1-la, after page 3/4 1-5, to become new page 3/4 1-Sa. This figure presents the MTC
+.2x10'gimit curve changing the positive MTC limit above 70% power delta rho / F to a value which varieslinearlyfrom+.3x10'from delta rho / F at 100% power to +.7x10' delta rho / F at 70% power.
9 Replace Figure 3.1-2, page 3/4 1-27, with the enclosed modified Figure 3.1-2. The new figure reflects a change in the Transient Insertion Limit between 90% and 100% power, wherein the limit is increased from an allowed insertion limit which varies linearly from 35% for Bank 5 at 90% power to 25% at 100% power, to a constant value of 35% between 90% and 100% power.
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i Documsnt Control Desk February 12, 1988 Page 4 10.
Modify the text of B 3/4.1.1.1 and B 3/4,1,1,2, page B 3/4 1-1 (and adding page 3/4 1-1A) as indicated in the Attachment. 1his modification to the bases provides supporting justification for the proposed shutdown margin Technical Specification changes.
11.
Modify B 3/4.1.3, pages B 3/4 1-3 (and adding page B 3/4' l 3A) and B 3/4 1-5, as indicated in the Attachment. This bases modification is to clarify and. emphasize the function of the movable Control Element Assochl ic s i
(CEA's) in assuring adequate shutdown margin for the Critical Mode.s of l
operation.
12.
Replacts Figure 2.2-1 (Axial Power Distribution Trip Limiting Safety System Settings ( APD LSSS)) with the new modified Figure 2.2-1 as enclosed in the Attachment. This figure is associated with Technical Specification 2.2-1 (Table 2.2-1, Item 8). The new limits of this figure are wider in both the positive and negative ASI regions below 70% power.
13.
Replace Figure 3.2 2 (Linear Heat Rate Axial Flux Offset Control Limits) with the new modified Figure 3.2-2 as enclosed in the Attachment. This figure is associated with Technical Specification 3.2.1 (Linear Heat Rate Limiting Condition for Operation ( LHR LCO)). The new limits are wider in the negative ASI region below 50% power.
14.
Replace Figure 3.2-4 (DNB Axial Flux Offset Control Limits) with the new modified Figure 3.2 4 as enclosed in the Attachment. This figure is associated with Technical Specification 3.2.5 (DNB Limiting Condition for Operation (DNB LCO)). The new limits are wider in the negative ASI region below 50% power.
DETERMINATION OF SIGNIFICANT HAZARDS i
l We have determined, based on the analytical information supplied in the enclosed j
Attachment and Appendix, that this amendment does not involve a significant hazards consideration. Justification for this determination is presented below l
and in the enclosed Attachment and Appendix.
1 Four of the 96 fresh assemblies constituting the reload core are manufactured by Advanced Nuclear Fuels Corporation (ANF), and are included in the Unit 1 Cycle 10 core to serve as demonstration lead assemblies for arsessment of ANF fuel for use in core reloads at CCNPP. These four assemblies were included in all aspects of the reload design process including;
- nuclear, mechanical, thermal-hydraulic, transient and ECCS design. In all aspects, these assemblies j
are designed to be compatible with the fresh fuel (92 assemblies) provided by Combustion Engineering (C E) and previously used fuel remaining in the Cycle 10 core.
The C-E fresh fuel assemblies for Unit 1 Cycle 10 are not different in design from _those used previously. The design is the same as was found to be acceptable to the NRC for use in the reference cycle. The analytical methods used to determine conformance with the Technical Specifications and regulations are consistent with previous NRC approvals and involve no significant changes.
I
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Documsnt Control Desk February 12, 1988 Page 5 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 amendment would not:
1.
involve a significant increase in the probability or consequences of an accident previously evaluated; The probability or consequences of all previously evaluated accidents are not increased by the proposed reload license amendment. To support the Unit 1 Cycle 10 reload core design, 18 design basis events were reviewed and one of those events was reanalyzed (the Steam Line Rupture Event). All 18 design basis events, including the Steam Line Rupture Event, were bounded by the results of the previously accepted reference cycle (Unit 2 Cycle 8).
An Emergency Core Cooling System (ECCS) performance analysis was performed for Unit 1 Cycle 10 to demonstrate compliance with 10 CFR 50.46, which presents the NRC Acceptance Criteria for ECCS's for light-water reactors. The analysis justifies an allowable Peak Linear Heat Generation Rate (PulGR) of 15.5 kw/ft. This PLHGR is equal to the existing limit for CCNPP, Units 1 and 2. The method of analysis and detailed results which support this value are presented in the enclosed Attachment to this request for license amendment.
Small Break Loss of Coolant Accident (SBLOCA) analyses confirm that the results previously reported to the NRC for Unit 1 Cycle 8 (SBLOCA reference cycle for Unit 1 Cycle 10) also bound the SBLOCA results for the Unit 1 Cycle 10 reload core design. This conclusion is supported in the enclosed Attachment to this request for license application.
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. The design of Unit 1 Cycle 10 closely follows that of the reference cycle, Unit 2 Cycle 8. The four ANF demonstration lead assemblies, included in the Unit 1 Cycle 10 core, do not impact the core design in any adverse manner. All nuclear, mechanical, thermal-hydraulic, transient and LOCA safety analyses performed for Cycle 10 core design, envelope the four ANF assemblies. The analyzed performance of those assemblies is determined to be very similar to that of the balance of the core. The Appendix to the Attachment to this submittal discusses the design of the ANF assemblies.
OR,
Document Control Desk February 12, 1988 Page 6 iii. involve a significant reduction in a margin of safety.
The proposed Technical Specification changes have an insignificant impact on the safety analyses for Unit 1 Cycle 10. With each proposed Technical Specification change, sufficient conservatism or margin of safety remains between the proposed limits of the changes and actual safety limits such as Specified Acceptable Fuel Design Limits (SAFDL's). Justification for this conclusion is provided in the enclosed Attachment to this submittal.
CHANGE NO. 2 (BG&E FCR 88-3001)
The following proposed changes are requested for the Unit 2 Technical Specifications and included in this submittal. These proposed changes for Unit 2 are identical to changes 12, 13, and 14, specified above under CHANGE NO. 1 for Unit 1 Cycle 10. The proposed changes are summarized below with a detailed explanation enclosed in the Attachment to this request for license amendment.
1.
Replace Figure 2.2-1 (Axial Power Distribution Trip Limiting Safety System Settings ( APD LSSS)) with the new modified Figure 2.2-1 as enclosed in the attachment. This figure is associated with Technical Specification 2.2-1 (Table 2.2-1, Item 8). The new limits of this figure are wider in both the positive and negative ASI regions below 70% power.
2.
Replace Figure 3.2-2 (Linear Heat Rate Axial Flux Offset Control Limits) with the new modified Figure 3.2-2 as enclosed in the Attachment. This figure is associated with Technical Specification 3.2.1 (Linear Heat Rate Limiting Condition for Operation { LHR LCO)). The new limits are wider in the negative ASI region below 50% power.
3.
Replace Figure 3.2 4 (DNB Axial Flux Offset Control Limits) with the new modified Figure 3.2-4 as enclosed in the Attachment. This figure is associated with Technical Specification 3.2.5 (DNB Limiting Condition for Operation (DNB LCO)). The new limits are wider in the negative ASI region below 50% power.
DETERMINATION OF SIGNIFICANT HAZARDS The proposed changes to the Unit 2 Technical Specifications are included in this Unit 1 Cycle 10 request for licenso amendment since they are identical to the ASI region changes requested for Unit 1 Cycle 10, and for the same reasons are equally applicable to Unit 2 Cycle 8.
The justifications presented for the Unit 1 Cycle 10, ASI region Technical Specifications (No.
12, 13, and 14 presented under CHA.GE NO. 1 above) are equally supportive of the requested N
changes for Unit 2 Cycle 8.
It is determined that these changes do not constitute a significant hazard consideration, in that they do not: i) involve a significant increase in the probability or consequences of an accident previously evaluated; 11) create the possibility of a new or different type of accident from any accident previously
Document Control Desk February 12, 1988 Page 7 evaluated; or 111) involve a significant reduction in a margin of safety, for the same reasons as stated for Unit 1 Cycle 10.
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.
FEE DETERMINATION Pursuant to 10 CFR 170.21, we are including BG&E Check No. 1196329 in the amount of $150.00 to the NRC to cover the application fee for this request.
Very truly yours, l Wi+2n STATE OF MARYLAND TO WIT :
COUNTY OF CALVERT I hereby certify that on t.he _/A day of 2 ar,2 ; o 19 1, before me, the subscriber, a Notary Public of the State of Maryland in and for County of Calvert, personally appeared Joseph A. Tiernan, being duly sworn, and 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 kt:owledge, information, and belief; and that he wac authorized to provide the response on behalf of said Corporation, al A. V;/ 4 WITNESS my Hand and Notarial Seal:
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Date JAT/DSE/imt cc:
D. A. Brune, Esquire J. E. Silberg, Esquire R. A. Capra, NRC S. A. McNeil, NRC l
W. T. Russell, NRC D. C. Trimble, NRC T. Magette, DNR t
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ATTACHMENT TO B-88-011 CALVERT CLIFFS UNIT 1 CYCLE 10 LICENSE SUBMITTAL l
24-29a(86II)/cp-2 Calvert Cliffs Unit 1 Cycle 10 License Submittal Table of Contents i
Section 1.
Introduction and Summary i
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 1
I 4
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1.0 INTRODUCTION
AND
SUMMARY
1 This report provides an evaluation of design and performance for the operation of Calvert Cliffs Unit 1 during its tenth fuel cycle, at full rated power of 2700 MWt.
All planned operating conditions remain the same as those for Cycle 9.
However, Cycle 10 will be the first "24-month cycle" for Unit 1. It will also be the first Unit 1 cycle to use a low-leakage fuel management pattern.
The core will consist of 121 presently operating Batch K and L assemblies and 96 fresh Batch M assemblies.
Of the 96 fresh Batch M assemblies, 92 will be manufactured b Combustion Engineering (C-E) and 4, by Advanced Nuclear Fuel (ANF)y The C-E assemblies will use BaC for neutronic shinning; the ANF assemblies will use Gd 0 and 3re to be included in the Cycle 10 core as part of the ekoht to qualify ANF fuel for 24-month cycle operation.
The analyses presented in the main body of this report support the neutronic modelling of the core, the safety evaluation of the core and the acceptable performance of the C-E fuel.
The discussion in the Appendix supports the performance of the ANF fuel.
Plant operating requirements have created a need for flexibility in the Cycle 9 termination point.
This need has been met by using an End-of-Cycle 9 window ranging from 9,800 MWD /T to 11,800 MWD /T in the Cycle 10 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 10 conditions would be enveloped, provided the Cycle 9 termination point falls within this end-of-cycle burnup range.
The analysis presented herein will accomodate a Cycle 10 length which varies from 21,400 to 22,700 MWD /T, depending upon the Cycle 9 ghutdown 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 8 safety analysis described in Reference 1.
Unit 2 Cycle 8 will hereafter be referred to as the "reference cycle" in this report, unless otherwise noted.
This is the appropriate reference cycle because its design / safety basis is the one most recently reported to the NRC and the basic system characteristics of the two reload cores are very similar Unit ? Cycir 8 also being a 24-month cycle which used l
low-leakage fuel management.
i i
1-1
)24-29a(8611)/cgh-4 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 1 Cycle 9,
Reference 2), proposed modifications to the existing plant Technical Specifications are provided and are justified by the analyses discussed herein.
Some of these proposed modifications are similar to those approved (Reference 3) for the reference cycle.
The Cycle 10 analyses used the same methodology as the reference cycle in all areas except one.
The setpoint analysis and relevant transient analyses used the Extended Statistical Combination of Uncertainties (ESCV) methodology described in Retence 4.
This methodology was recently approved by the NRC in e.eVerence 5 for application to the Calvert Cliffs reactors.
The perfonnance of Combustion Engineering 14x14 fuel at extended burnup is discussed in Reference 6, which was approvad in Reference 7.
For Cycle 10 the batch average discharge will be considerably less than the 45,000 MWD /T criterion of that reference, but the burnup of approximately 0.3% of the fuel pins will be above the 52,000 MWD /T point discussed in Reference 6, if Cycles 9 and 10 are operated to their maximum burnups.
However, since all Cycle 10 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 10.
Furthermore, the maximum pin burnup (approximately 54,100 MWD /T) is well below the burnups which have been or are projected to be achieved in lead assembly demonstrations at Calvert Cliffs (approximately 64,000 MWD /T).
)
I 1-2
24-29a(86II)/cgh-5 2.0 OPERATING HISTORY OF THE PREVIOUS CYCLE Calvert Cliffs Unit 1 is presently operating in its ninth fuel cycle utilizing Batch L, K, J. H, GX and E fuel assemblies (including 24 Batch E assernblies from Unit 2).
Calvert Cliffs Unit 1 Cycle 9 began operation on January 3, 1987 and reached approximate - full power conditions on January 15, 1987.
The Cycle 9 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 9 will tenninate on or about April 9, 1988. The Cycle 9 termination point can vary between 9,800 MWD /T and 11,800 PWD/T to accommodate the plant schedule and still be within the assumption of the Cycle 9 analyses. As of February 1, 1988, the Cycle 9 burnup had reached 9,103 MWO/T.
l 2-1
24-29a(8611)/cgh-6 3.0 GENERAL DESCRIPTION The Cycle 10 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 10 will be the' switch from 18-month, conventional fuel management to 24-month, low-leakage fuel management.
This change will entail the removal of 96 irradiated assemblies (72 Unit 1 assemblies: 3 Batch K*,
48 Batch J,16 Batch J*
1 Batch H and 4 Batch GX; 24 Unit 2 Batch E assemblies).
These assemblies will be replaced by 96 fresh assemblies:
16 unshimed Batch M assemblies and 76 12-shimed (B C) Batch M* assemblies at 4.08 wt% U-235 4
enrichment; 4 12-shimed (Gd 0 ) Batch MX demonstration - assemblies (see Section 3.1) with an as,setnbly average enrichment of 3.85 wt%
2 U-235.
Figure 3-1 shows the fuel management pattern to be employed in Cycle 10.
Figure 3-2 shows the locations of the fuel and poison pins within the fresh M*
and MX shimed assemblies.
This fuel management pattern will accomodate Cycle 9 tennination burnups from 9,800 MWD /T to 11,800 MWD /T.
The Cycle 10 core loading pattern is '90' rotationally symetric.
That is, if one quadrant. of the core were rotated 90* into its neighboring quadrant, each assembly would be aligned with a similar i
assembly. This similarity includes batch type,. number of fuel rods, initial enrichment and burnup.
Figure 3-3 shows the beginning of Cycle 10 assembly burnup distribution for a Cycle 9 termination burnup of 9,800 MWD /T.
The initial enrichment of the fuel assemblies is also shown in Figure 2-3.
Figure 3-4 shows the end of Cycle 10 assembly burnup distribution.
The end of Cycle 10 core average exposure is approximately 32,800 MWD /T and the average discharge exposure is approximately 40,800 MWD /T. The end of cycle burnuos are based on a Cycle 9 length of 11,800 MWD /T and a Cycle 10 length of 21,400 MWD /T.
3.1 DEMONSTRATION ASSEMBLIES i
All previously loaded demonstration assemblies and components, consisting of the 4 PROTOTYPE assemblies, a segmented test rod from the SCOUT assembly installed in a PROTOTYPE assembly at the beginning of Cycle 9, and the Batch H high burnup assembly, will be discharged at the end of Cycle 9.
Four new demonstration assemblies, designated as Batch MX, will be loaded into Cycle 10.
These assemblies are being manufactured by Advanced Nuclear Fuel (ANF) and will contain Gadolinium as the burnable poison material.
Each assembly will consist of 164 4.08 wt% U-235 enriched fuel pins and 12 fuel bearing Gadolinium poison pins.
The poison pins will contain natural Uranium and 10 wt%
3-1
24-29a(86f1)/cgh-7 Gd 0. The fuel and poison pin arrangement for these demonstration ds e blies is shown in Figure 3-2.. A further discussion concerning these assemblies is contained in the Appendix.
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 10. This modification is being made to support. the change from 18-month, conventional fuel management to 24-month, low-leakage fuel management for Cycle 10.
The specific change and the technical reason for it are identical to those of the reference cycle (Reference 1).
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3-2 l
24-29a(86II)/cgh 8 TABLE 3-1 CALVERT CLIFFS UNIT 1 CYCLE 10 CORE LOADING Initial Non-Fuel Total Number Assembly Bearing Initial of Non-Fuel Total Average B,C Poison 8 C Poison Bearing 8 C Number 3
4 Batch.Byyyup(MWD /(() Assembly(gms B / inch)
Rods Rods Assembly Number of Enrichment Rdds Per Ldagfng Poison of Fuel Designation Assemblies (wt% U-235)
BOC10 E0C10 M
16 4.08 0
17,800 0
0 0
2816 M*
76 4.08 0
26,100 12
.036 912 12,464 MX(3) 4 3.85 0
24.800 0
0 0
704 L(4) 40 4.05 8,700 33,700 0
0 0
7040 o,
2, L*(4) 12 3.40 12,500 37,300 0
0 0
2112 K(4) 48 4.05 23,000 40,600 0
0 0
8448 K*(4) 21 3.40 26,000 46,000 0
0 0
3696 Total 217 10,100 32,800 912 37,280 (1) Cycle 9 burnup of 9,800 MWD /T (2) Cycle 9 burnup of 11,800 MWD /T and Cycle 10 burnup of 21,400 MWD /T (3) MX fuel, which is being manufactured by ANF, contains 12 Gd 02 3 (10 wt%) fuel bearing (nat. u) poison pins and 164 fuel pins (4.08 wt% u-235) per assembly.
(4) Carried over from Cycle 9 to Cycle 10 of Unit 1.
t KEY-X
- BOX NUMBER Y
BATCH 1
2 K
M 3
4 5
6 7
K M
L M*
L 8
9 10 11 12 13 K
M*
L M*
K L*
14 15 16 17 18 19 20 K
M*
L M*
K M*
K 21 22 23 24 25 26 27 28 K
M*
L M*
K*
M*
L*
M*
29 30 31 32 33 34 35 36 M
L M*
K*
M*
L M*
K 1
37 38 39 40 41 42 43 44 l
L M*
L K*
'K*
M*
45 46 47 48 49 50 51 52 53 M*
L*
M*
K*
E L
54 55 56 57 58 59 60 61 62 L
L*
K M*
K M*
L K*
BALTIMORE GAS & ELECTRIC CO.
CALVERT CLIFFS UNIT 1 CYCLE 10 FICURE CALVERT CLIFFS CORE MAP 3-1 NUCLEAR POWER PLNIT 3-4
_ -. _ -. _, _ _.. _ _.~
-X X:
X.
X:
X X
l I
X X
~
X X:
X' X.
12-Shim M* Assembly (8 C) 4 or 12 Shim MX Assembly (Gd 0 /UO )
23 2
Fuel Red Location Poison Rod Location (B C or M 4
23 BALTIMCRE CAS & ELECTRIC CO.
CAL'!ERT CLIFFS UNIT 1 CYCLE 10 FIGURE CAL'!ERT CLIFFS FUEL A2;D SdIM LOCATIONS 3-2 SUCLEAR PC'.*ER PLC;T IN FRESH SHIMMED ASSEMBLIES 3-5
KEY X
- BATCM 1
K 2
M YYY
- INITIAL ENRICHMENT 4.05 4.08 ZZZZZ
- BURNUP (MWD /T) 25,200 0
3 K
4 M
5 L
6 M*
7 L
4.05 4.08 4.05 4.08 4.05 24,400 0
7,300 0
9,700 8
K 9
M* 10 L 11 M* 12 K 13 L*
4.05 4.08 4.05 4.08 4.05 3.40 23,300 0
7,900
.0 21,600 12,600 14 K 15 M* 16 L 17 M* 18 K 19 M* 20 K
4.05 4.08 4.05 4.08 4.05 4.08 4.05 23,200 0
8,600 0
23,900 0
19,800 21 K 22 M* 23 L 24 M* 25 K* 26 M* 27 L* 28 M*
4.05 4.08 4.05 4.08 3.40 4.08 3.40 4.08 24,700 0
8,500 0
26,000 0
12,500 0
29 M 30 L 31 M* 32 K* 33 M* 34 L 35 M* 36 K
4.08 4.05 4.08 3.40 4.08 4.05 4.08 4.05 O
7,800 0
26,000 0
10,200 0
19,800 37 L 38 M* 39 K 40 M* 41 L 42 K* 43 K* 44 M*
4.05 4.08 4.05 4.08 4.05 3.40 3.40 4.08 45 K
7,300 0
23,100 0
10,100 26,100 25,000 0
4.05 25,200 46 M* 47 K 48 M* 49 L* 50 M* 51 K* 52 MX 53 L
4.08 4.05 4.08 3.40 4.08 3.40 3.85 4.05 54 M
0 21,600 0
12,400 0
26,700 0
9,700 4.08 0
55 L 56 L* 57 K 58 M* 59 K 60 M* 61 L 62 K*
4.05 3.40 4.05 4.08 4.05 4.08 4.05 3,40 9,700 12,600 19,800 0
19,800 0
9,700 26,200 i
i EOC 9 - 9,800 MVD/T i
BALTIMORE GAS & ELECTRIC Co.
CALVERT CLIFFS UNIT 1 CYCLE 10 FIGLTE CALVERT CLIFFS ASSEMBLY AVERAGE BURNUP AT BOC 3-3 NUCLEAR POWER PIRIT AND INITIAL ENRICHMENT DISTRIBUTION 3-6
KEY X
- BATCH 1
K 2
M ZZZZZ
- BURNUP (MWD /T) 36,300 17,200 3
K 4
M 5
L 6
M*
7 L
35,500 18,300 28,600 23,700 34,000 8
K 9
M* 10 L 11 M*
12 K
13 L*
35,400 21,400 33,700 26,900 44,900 36,000 14 K 15 M*
16 L 17 M* 18 K 19 M* 20 K
35,400 21,700 34,700 27,700 47,600 27,600 44,000 21 K 22 M* 23 L 24 M*
25 K* 26 M* 27 L*
28 M*
l 35,800 21,500 34,700 27,400 47,000 27,900 38,000 28,400 l
29 M 30 L 31 M*
32 K*
33 M* 34 L 35 M*
36 K
18,300 33,700 27,800 47,100 27,20'O 36,600 27,300 44,100 37 L 38 M*
39 K 40 M* 41 L 42 K* 43 K* 44 M*
45 K 28,600 27,000 47,000 27,900 36,500 45,200 44,600 26,300 36,400 46 M* 47 K 48 M* 49 L* 50 M* 51 K*
52 tiX 53 L
54 M 23,600 44,800 27,600 37,900 27,200 46,000 24,800 35,600 17,200 55 L 56 L*
57 K 58 M*
59 K 60 Me 61 L 62 K*
1 34,000 36,000 44,000l28,400 44,100 26,300 35,600 46,700 EOC 9 - 11,800 MWD /T ECC 10 - 21,400 MWD /T BALTIMORE CAS & ELECTRIC CO.
CALVERT CLIFFS UNIT 1 CYCLE 10 FIGURE CALVERT CLIFFS ASSEMBLY AVERAGE BURNUP AT EOC 34 NUCLEAR POWER PLANT 3-7
^
24-29aN6I1)/cgh-9 4.0 FUEL SYSTEM DESIGN 4.1 MECHANICAL DESIGN 4.1.1 Fuel Design The mechanical design for the Batch M, C-E reload fuel is identical to that of the Batch K fuel described in the reference cycle submittal (Calvert Cliffs Unit 2 Cycle 8,
Reference 1).
The mechanical designs of the Batch L and K fuel assemblies were described in Reference 2.
4.1.2 Dimensional Changes All C-E fuel assemblies in Cycle 10 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 10.
4.1.3 CEA Design The replacement CEA to be utilized for the change discussed in Section 3.2 will have the same reconstitutable feature as the replacement CEA installed in the reference cycle for the same purpose.
4.2 THERMAL DESIGN The thermal performance of a composite, standard fuel pin which envelopes the various C-E fuel assemblies which will be present in Cycle 10 (Batches M, L and K) has been evaluated using the FATES 3B version of the fuel evaluation model (References 7,
8 and 9).
FATES 3R has received NRC approval (Reference 10) for application to the Calvert Cliffs reactors. The analysis was performed with a history that nodeled 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 10.
)
4-1 1
~-,
. e
. 29a(8611)/cp-10 5.0 NUCLEAR DESIGN 5.1 PHYSICS CHARACTERISTICS 5.1.1 Fuel Management j
The Cycle 10 fuel management employs a low-leakage pattern as-described in Section 3, Figure 3-1.
The fresh Batch M ~ fuel is comprised of three sets of assemblies, all using non-poison fuel pins - of just one enrichment (4.08 wt% U-235), with each set containing either a unique number of shims per assembly or in the case of the Gadolinium demonstration assemblies, a different neutrotrhc poison.
The unique number of shims per assembly was chosen to minimize radial power peaking and to control BOC MTCs.
Specifically, Batch M consists of 16 unshirrined assemblies, 76 assemblies with 12 B C shims per assembly and 4 assemblies with 12 Gd 0 shims per assdnbly.
With this loading, the Cycle 10 burnup caha$ity for full power operation is expected to be between 20,600 MWD /T and 21,800 MWD /T, dependirig on the final Cycle 9 termination point.
The Cycle 10 core characteristics have been examined for Cycle 9 terminations between 0,000 and 11,800 MWD /T and limiting values established for the safety analyses.
The loading pattern (see Section,3) is applicable to any Cycle 9 termination point between the stated extremes.
Physics characteristics including reactivity coefficients for Cycle 10 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 accomodation for appropriate uncertainties and allowances.
A table has traditionally been included in Section 5 which demonstrated that adequate shutdown margin would be available at the end-of-cycle, hot zero powers, critical condition to meet the shutdown requirement of the most limiting event at that condition, j
4 1
1.e., Steam Line Rupture.
Such a table has been emitted from this submittal due to the requested elimination of the applicability of.
the shutdown margin Technical Specification to the critical modes, i
1.e., Medes 1 and 2 (Xeff >1.0).
The details concerning this change-to the Technical Specifications, the reasons for it and the l
supporting discussions are contained in Section 9.0.
Table 5-2 shows the reactivity worths of the three CEA groups which are allowed in the core during critical conditions.
These reactivity worths were calculated at full power conditions for Cycle 10 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, such that it will be identical to that of the reference cycle.
N 5-1
j 24-29a(8611)/cp-11 l
The power dependent insertion limit (PDIL) curve is being chan ed slightly relative to the PDIL of the reference cycle. Specifical y, the allowable insertion of the lead bank (Bank 5) at 100% power is being increased from 25% to 35%.
This revised PDIL, which is shown in Figure 5-1, has been used in the generation of all Cycle 10 safety and setpoint data.
5.1.2 Power Distributions i
Figure 5-2 through 5-4 illustrate the all rods out (ARO) integrated radial power distributions at BOC10, M0C10 and E0C10, respectively, that are characteristic of the high burnup end of the Cycle 9 shutdown window.
The high burnup end of the Cycle 9 shutdown window tends to increase the integrated 1-pin radial power peaking. The integrated radial power distributions with CEA Group 5 fully inserted at beginning and end of Cycle 10 are shown in Figures 5-5 and 5-6, respectively, for the high burnup end of the Cycle 9 shutdown window.
The performance of the four ANF assemblies is discussed in the l
Appendix.
The locations of the ANF assemblies were selected to i
yield maximum 1-pin peaks in the ANF assemblies which are predicted to be at least 10% below the maximum 1-pin peak in the core for standard operating conditions (see Figures 5-2 through 5-6).
l 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 l
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 10.
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 j
threr.-dimensional or total peaking factor anticipated in Cycle 10 during normal base load, all rods out operation at full power is l
1.85, not including uncertainty allowances.
5.1.3 Safety Related Data (Ejected CEA and Drop CEA Data)
The Cycle 10 safety related data for this section are identica'. to l
the safety related data used in the reference cycle.
l 5-2
24-29a(86f1)/cp-12 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 performed in the same manner and with the same methodologies used for the reference cycle analyses.
The neutronic modeling of the ANF Gadolinium demonstration assemblies was performed by C-E in accordance with the methods described in C-E's Gadolinium topical report (Reference 2).
5.4 UNCERTAINTIES IN MEASURED POWER DISTRIBUTIONS The power distribution measurement uncertainties to be applied to Cycle 10 are the same as those applied to the reference cycle.
5-3
l 24-29a(8611)/cgh-13 TABLE 5-1 CALVERT CLIFFS UNIT 1 CYCLE 10 NOMINAL PHYSICS CHARACTERISTICS Reference Cycle Unit 1+
Units (Unit 2 Cycle 8)
Cycle 10 Dissolved Baron Hot Full Power, All Rods Out, Equilibrium Xenon Baron Content for Criticality at B0C PPM 1490 1419 Inverse Boron Worth Hot Full Power, EOC PPM /%ap 121 122 Hot Full Power, EOC PPM /%Ao 87 85 Moderator Temperature Coefficient Hot Full power, Equilibrium Xenon, CEAs Withdrawn Beginning of Cycle 10-4ap/*F 0.0*
-0.2 j
End of Cycle 10-4ap/*F
-2.3
-2.3 Doopler Coefficient Hot Zero Power, SOC 10' op/* F
-2.06**
-2.01 Hot Full Power, BOC 10' op/*F
-1.53**
-1.49 Hot Full Power, E0C 10' ap/*F
-1.81**
-1.80 i
TotalEffectiveDe}ayed Neutron Fraction, eff BOC 0.00629 0.00638 EOC 0.00516 0.00513 Neutron Generation Time, t*
BOC 10-6 sec 20.5 20.4 E0C 10-6 sec 28.6 29,3 BOC10 data were calculated using the early Cycle 9 shutdown burnup of
+
9,800 MWD /T; EOC10 data were calculated using appropriate end of Cycle 10 conditions.
To make the Unit 2 Cycle 8 data consistent with the methods being used to generate the comparable Unit 1 Cycle 10 data, this Moderator Temporature Coefficient is being revised relative to the value shown in Reference 1.
The new value is the result of the application of a revised bias, i
To make the Unit 2 Cycle 8 data consistent with the methods being used to generate the comparable Unit 1 Cycle 10 data, these Doppler Coefficients
)
are being revised relative to the values shown in Reference 1.
Thesc new values are the result of the use of three dimensional methods and the application of an appropriate bias.
5-4
24-29a(8611)/cgh-15 TABLE 5-2 CALVERT CLIFFS UNIT 1 CYCLE 10 REACTIVITY WORTH OF CEA REGULATING GROUPS AT HOT FULL POWER, %oo 1
Beginning of Cycle End of Cycle Regulating Reference
- Unit 1 Reference
- Unit 1 CEA's Cycle Cycle 10 Cycle Cycle 10 Group 5 0.36 0.44 0.40 0.40 Group 4 0.79 0.61 0.80 0.88 Group 3 0.86 0.84 1.03 1.01 Note Values shown assume sequential group insertion; values are biased.
- Unit 2 Cycle 8 i
1 5-5
I l
l i
1.00
- -1.00,Gp 5 @ 35%
I I
0.90
"-0.90,Gp 5 @ 35%
1 x
0.80
- 0.75,Gp 5 @ 50%
0.70 70.Gp 5 @ 60%
f* g,Gp 5 @ 854 j
- g56,Gp4@50%
DngTerml ShrtTerm 0.40 Steady State Steady State Insertion l Insertion 0.30 Limit Limit Grp 5 @ 25%
Grp 4 @ 20%
0.20 f
0.20,Gp 3 @ 604 0.10 0.00 0.0,Cp 3 @ 60%
i Allowable BASSS 4C-- Gn 5 @ 55%
Operating Region
- "E " _
t i
i i
i i
i i
i 0
20 40 60 80 100 0
20 40 60 136.0 108.8 81.6 54.4 27.2 0
136.0 108.8 81.6 54.4 4.=
4 I
i i
i i
i O
20 40 60 80 100 j
136.0 108.8 81.6 54.4 27.2 0
a 4 CEA INSERTION INCHES CEA WITHDRAVN I
CEA GROUP INSERTION LIMITS VS. FRACTION OF ALLO'JA3LE THERMAL PO'4ER FOR EXISTING RCP COMBINATION BALTIMORE GAS & ELECTRIC CO.
CALVERT CLIFFS UNIT 1 CYCLE 10 FIGURE CALVERT CLIFFS PDIL FOR REGUIATING CROUPS 51 j
NUCLEAR POWER PIANT 5-6 i
1 I
1 cb 1
K 2
M 0.43 0.90 I
?
3 K
4 M
5 L
6 M*
7 L
0.39 0.94 0.99 1.16 1.18 l
i 8
K 9
M* 10 L 11 M* 12-K 13 L*
0.40 0.96 1.21 1.28 1.03 1.07 14 K
15 M*
16 L 17 M* 18 K 19 M* 20 - K 0,.40 0.94 1.15
-1.26' 1.00 1.28 1.09 1
21 K 22 M* 23 L 04 M* 25 K* 26 M* 27 L* 28 M*
0.39 0.96 1.15 1.21 0.83 1.24 1.08 1.28 1
29 M 30 L 31 M*
32 K*
33 M*
34 L 35 M* 36 K
i 0.94 1.22 1.27 0.83 1.19 1.14 1.20 1.01 i
37 C 38 M* 39 K 40 M* 41 L 42 K* 43 K* 44 h 1.00-1.29 1.02 1.25 1.14 0.75 0.76 1.1.
43 K
+
i j
46 M* 47 K 48 M* 49 L* 50 M*
51 K* 52 MX 53 L
1.16 1.04 1.28 1.08 1.19 0.73 1.03 1.10 54 M
4
++
O.90 55 L 56 L*
57 K 58 M*
59 K 60 M* 61 L 62 A*
1.18 1.07 1.09 1.28 1.01 1.12 1.10 0.84 1
j I
, NOTE : & = MAXIMUM 1. PIN PEAK - 1.54
++ - MAXIMUM 1. PIN PEAK IN ANF Gd ASSEMBLY - 1.32 BALTIMORE GAS & ELECTRIC CO.
CALVERT CLIFFS UNIT 1 CYCLE 10 FIGURE 4
j CALVERT CLIFFS ASSERBLY RELATIVE POWER DESNSITY 52
+
NUCLEAR POWER PLANT AT BOC, EQUILIBRIUM XENON T-
\\
\\
t l
5-7 3
d 1.
1 a.
li.,.?)
.h..
1 K
2 M
0.41 0.78 3
.K 4
M 5
L 6
M*
7 L
0.40 0.84 0.90 1.08 1.01 y
8 K[9 M* '10 L 11 M*
12 K 13 L*
0.45 0 99 1.12 1.26 0.96 0.96 t
14 K 13 M* 16 L 17
- (* 18 K 19 M+ 20 K
0.45 1.01 1.14 1.31 1.00 1.30 1.04 21 K 22 M* 23 L 24 M*
25 K* 26 M* 27 L* 28 M*
0.40 0.99 1.14 1.29 0.89 1.32 1.08 1.35 29 M ~0 L 31 M*
32 K* 33 M*
34 L 35 M*
36 K
O.84 1.12-1.31 0.89 1.29 1.15 1.30 1.05 37 L 38 M* 39 K 40 M* 41 L 42 K* 43 K* 44 M*
f~
0.90 l'.23 1.01 1.33 1.15 0.80 0.83 1.24 43 g
+
0.41 46 M*,.47','K 48 P* 49 L*
50 M* 51 K* $2 MX 53 L
1.08 0.5 1.30 1.08 1.29 0.81 1.11 1.11 54 M
++
u '*7 8
/
55 L
SF L*
57 K 58 M*
59 K 60 M* 61 L 62 K*
4.
l 1.0 0.96 1.04 1.35 1.05 1.24 1.11 0,86
- ', s.
i
." /
NOTE: + = FAXIMUM 1-PIN PEAK - 1.54
++ = MAXIMUd 1 PIN PEAK IN ANF Gd ASSEMBLY - 1.32 j
.i
/b
_[ ' 'k'62.S & ELECTRIC CO.
CALVERT CLIFFS UNIT 1 CYCLE 10 FIGURE BALTIMORE
/
CALVERT CLIFFS ASSE, GLY RELATIVE POWER DESNSITY 53 NUCLEAR POWER PLANT AT 10 M'4D/T, EQUILIBRIUM XENON
,, t r~
d is
- *ve 5-8 s '
s, 1
.[
I' V
1 K
2 M
0.46 0.82 3
K 4
M 5
L 6
M*
7 L
0.46 0.85 0.90 1.15 1.00 8
K 9
M* 10 L 11 M*
12 K 13 L*
0.54 1.11 1.09 1.26 0.94 0.90 14 K 15 M*
16 L 17 M*
18 K
19 M* 20 K
0.54 1.14 1.14 1.31 0.97 1.26 0.97 21 K 22 M* 23 L 24 M* 25 K* 26 M* 27 L* 28 M*
0.46 1.10 1.14 1.32 0.90 1.29 1.01 1.28 29 M 30 L 31 M*
32 K* 33 M*
34 L 35 M*
36 K
0.85 1.09 1.31 0.90 1.27 1.06 1.26 1.00 37 L
38 M* 39 K 40 M* 41 L 42 K* 43 K* 44 M*
0.90 1.26 0.98 1.29 1.06 0.78 0.83 1.26 45 K
45 M* 47 K 48 M* 49 L* 50 M* 51 K* 52 MX 53 L
!.15 0.94 1.26 1.01 1.26 0.82 1.14 1.06 54 M
55 L 56 L*
57 K 58 H+ 59 K 60 M* 61 L 62 K*
1.00 0.90 0.97 1.28 1.00 1.26 1.06 0.83 l
NOTE: + - MAXIMUM 1-PIN PEAK - 1,48
++ - !dAXIMUM 1 PIN PEAK IN ANF Cd ASSEMBLY - 1.28 EALTIMORE CAS & ELECTRIC CO, CALVERT CLIFFS UNIT 1 CYCLE 10 FIGURE CALVERT CLIFFS ASSEMBLY RELATIVE POER DESNSITY 5-4 NUCLEAR P0'JER PL'diT AT EOC, EQUILIBRIt'M XENON 5-9
5 CEA BANK 5 1
g, g
LOCATION 0.37 0.76 3
K 4
M S
L 6
M*
7 L
0.42 0.96 0.94 0.97 0.90 8
K 9
M* 10 L 11 M*
12 K
13 L*
0.44 1.04 1.27 1.25 0.85 0.52U 14 K 15 M* 16 L 17 M*
18 K 19 M* 20 K
0.44 1.04 1.26 1.34 1.01 1.19 0.93 21 K 22 M* 23 L 24 M* 25 K* 26 M* 27 L* 28 M*
0.42 1.04 1.27 1.33 0.90 1.32 1.10 1.29 29 M 30 L 31 M*
32 K* 33 M* 34 L 35 M*
36 K
0.97 1.28 1.35 0.90 1.31 1.24 1.29 1.08
+
37 L 38 M*
39 K 40 M* 41 L 42 K* 43 K* 44 M*
0.95 1.26 1.03 1.32 1.24 0.83 0.84 1.22 45 K
0.38 46 M* 47 K 48 M* 49 L* $0 M* 51 K* 52 MX 53 L
0.97 0.85 1.19 1.10 1.28 0.81 1.12 1.18 54 H
++
0.76 55 L 56 L* 57 K 58 M* 59 K 60 M* 61 L 62 K*
0.90 0.52 0,93 1.29 1.08 1.22 1.18 0.78, NOTE: + - MAXIMLH 1 PIN PEAK - 1.64 H ~ MAXIMLH 1 PIN PEAK IN Gd ANF ASSEMBLY - 1.42 BALTIMORE CAS 6 ELECTRIC CO.
CALVERT CLIFFS UNIT 1 CYCLE 10 FIGURE CALVERT CLIFFS ASSEMBLY RELATIVE POWER DESNSITY 5-5 NUCLEAR POWER PLANT WITH BANK $ INSERTED, HFP, BOC 5-10 l
CEA BANK 5 1
K 2
M LOCATION 0.41 0.71 3
K 4
M 5
L 6
M*
7 L
0.49 0.87 0.37 0.99 0.78 8
K 9
M*
10 L 11 M*
12 K 13 L*
0.59 1.19 1.13 1.24 0.79 0.43A 14 K 15 M*
16 L 17 M* 18 K 19 M* 20 K
0.59 1.25 1.23 1.3v 0.98 1.18 0.84 21 K 22 M* 23 L 24 M* 25 K* 26 M*
27 L* 28 M*
0.49 1.19 1.23 1,42 0.96 1.35 1.03 1.28
+
29 M
30 L 31 M*
32 K* 33 M*
34 L 35 M*
36 K
0.87 1.13 1,39 0.96 1.37 1.14 1.34 1.06 37 L 33 M* 39 K 40 M* 41 L 42 K* 43 K* 44 M*
0.87 1.24 0.99 1.35 1.14 0.84 0.89 1.34 45 K
0.41 46 M* 47 K 43 M* 49 L* 50 M* 51 K* 52 MX 53 L
0.99 0.79 1:18 1.03 1.33 0.87 1.20 1.10 54 M
++
0.70 55 L 56 L*
57 K
58 M*
59 K 60 M* 61 L 62 K*
0.78 0.43 0.84 1.28 1.06 1.34 1.10 0.71 A
A
!'OTE : + - MAXIMUM 1-PIN PEAK = 1,61
++ = MAXIMUM 1 PIN PEAK IN ANF Cd ASSEMBLY - 1.36 BALTIMORE CAS & ELECTRIC CO, CALVERT CLIFFS UNIT 1 CYCLE 10 FIGURE CALVERT CLIFFS ASSEMBLY RELATIVE PO'a'ER DESNSITY 5-6 NUCLEAR POER PIRIT k'ITH BANK 5 INSEP.TED, HFP, EOC 5-11
24-29a(8611)/cgh-16 6.0 THERMAL HYDRAULIC DESIGN 6.1 DNBR ANALYSIS Steady state DNBR analyses of Cycle 10 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 using the Extended Statistical Combination of Uncertainties (ESCU) methods, described in Reference 6 and approved in Reference 7, to derive an overall uncertainty allowance which, when used with the CE-1 CHF correlation design limit of 1.15 for 14x14 fuel, provides a 95/95 probability / confidence level of assurance against DNB occurring during steady state operation or anticipated operational occurrences. The statistically derived ESCU uncertainty allowance 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.
6.2 EFFECTS OF FUEL BOWING ON DNBR MARGIN The effects of fuel rod bowing on DNB margin for Calvert Cliffs Unit 1 Cycle 10 have been evaluated using the NRC approved methods described in Reference 8.
Based upon these methods, a DNBR penalty of 0.006 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 presently included in the ESCU uncertainty allowance, as discussed above.
For assemblies which will attain burnups creater than 45 t
GWD/T, the power peaking will be significantly lower than that of other assemblies, providing ample margin to offset increases in the l
rod bcw penalties for these high burnup assemblies.
]
6-1
i 24-29a(86f1)/cgh-17 TABLE 6-1 CALVERT CLIFFS UNIT 1 CYCLE 10 THERMAL-HYDRAl'LIC PARAMETERS AT FULL POWER
- Reference **
Unit 1 General Characteristics Unit Unit ? Cycle 8 Cycle 10 Total Heat Output (core only)
MWg 2700 2700 10 BTV/hr 9215 9215 Fraction of Heat Generated
.975
.975 In Fuel Rod Primary System Pressure psia 2250 2250 l
(Nominal) i Inlet Temperature
'F 548 548 4
Total Reactor Coolant Flow gpg 381,600 381,600 (steady state) 10 lb/hr 143.8 143.8 6
Coolant Flow Through Core 10 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 I
(steady state ficw irreversible P cver entire fuel assembly)
Total Pressure Drop Across psi 34.7 34.7 Vessel (based on steady state flow and nominal dimensions) i 2
Core Average Heat Flux 8TU/hr-ft 182,900***
184,000****
(Accounts for above fraction of heat generated in fuel rod and axial densification factor) 2 Total Heat Transfer Area ft 49,100***
48,800****
(Accounts for axial densification factor) 2 Film Coefficient at Average BTV/hr-ft
- F 5930 5930 l
Conditions 1
1 6-2 j
24-29a(86II)/cgh-18 TABLE 6-1 (continued)
Reference **
Unit 1 General Characteristics Unit Unit 2 Cycle 8 Cycle 10 Average Film Temperature
'F 31 31 Difference Average Linear Heat Rate of kw/ft 6.16***
6.20****
Underisified Fuel Rod (accounts for above fraction of heat generated in fuel rod)
Average Core Enthalpy Rise BTU /lb 66.5 66.5 Maximum Clad Surface
'F 657 657 Temperature 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 Notes Due to the Extended Statistical Combination of Uncertainties methodology described in Reference 6, the nominal inlet temperature and nominal primary system pressure were used to calculate some of these parameters.
- Reference cycle analysis (Unit 2 Cycle 8) is contair.cd in Reference 9.
- Based on a value of 688 shims.
- Based on a value of 912 B C shims.
l 4
+ Thesa factors have been combined statistically with other uncertainty fact ars at the 95/95 confidence / probability level (Reference 6) to derive an ovrrall uncertainty allowance, as discussed in Reference 6 and approved by the NRC in Reference 7.
This allowance was verified to be applicable to Unit 1 Cycle 10.
l 6-3 l
24-29a(86I1)/cp-19 7.0 TRANSIENT ANALYSIS The Design Basis Events (DBEs) considered in the Unit 1 Cycle 10 safety analyses are listed in Table 7-1.
Core parameters input to the safety analyses for evaluating approaches to DNB and cer,terline 1
temperature to melt fuel design limits are presented in Table 7-2.
l As indicated in Table 7-1, no reportable reanalysis was performed for any DBE, since all results for these DBEs lie within the bounds of (or are conservative with respect to) the reference cycle values (Unit 2 Cycle 8. Reference 1).
As noted in Table 7-2, the shutdown margin Technical Specification is being changed from a singular value to a value functionalized versus time in cycle.
As part of this change, the BOC value is i
being lowered to - 3.5%Ao and the BOC value is being raised to -
5.0%Ao.
This change does not require any reportable reanalyses for the following reasons:
1)
All previously reported OBE analyses except that for the Steam Line Rupture event assumed a shutdown margin of only - 3.5%ao, as discussed in Reference 2.
The singular reference cycle value of -4.Stao was based upon the requirement of the Unit 2 Cycle 8 Steam Line Rupture event.
2)
The results of the Steam Line Rupture event reanalyzed for Unit 1 Cycle 10 were less limiting than those previously reported.
The shutdown margin at EOC is being raised to be consistent with this new analysis.
Evaluations were performed on DBEs that have input into the DNB LCO
~
setpoints and Technical Specification Limits.
These DBEs were re-evaluated using the Extended Statistical Combination of Uncertainties (ESCU) methodology and the corresponding ESCO SAFDL described in Reference 3 and approved in Reference 4.
None of these DBEs' transient results changed from those previously reported; only the DNBR required margins were evaluated using the ESCU methods. As part of this margin evaluation, the change in short-term transient insertion limit at 100% power to permit Bank 5 to be inserted 35%
instead of 25% was reviewed.
The results were determined to be bounded by those of the reference cycle.
Also, as a part of the ESCU analysis the Excess Load event was recategorized from Category 7.1 (see Table 7-1) to Category 7.2.
Previously, (the Excess Load event was an Anticipated Operational Occurrence A00) for which intervention of the RPS prevented i
acceptable limits from being exceeded; whereas, now it is an A00 for which RPS trips and/or sufficient initial steady state thermal i
margin, maintained by the LCOs, prevent acceptable limits from being l
exceeded.
Parametric analyses were performed to calculate the
)
l initial steady state margin required. This required margin was less than the initial steady state margin required for Sequential CEA Group Withdrawal and Loss of Coolant Flow which are the limiting A00s in Category 7.2 with respect to initial steady state margin
]
requirements, j
7-1
24-29a(8611)/cp-20 i
i
+
i TABLE 7-1 CALVERT CLIFFS UNIT 1 CYCLE 10 DESIGN BASIS EVENTS CONSIDERED IN THE NON-LOCA SAFETY ANALYSIS Analysis Status 7.1 Anticipated Operational Occurrences for l
which intervention of the RPS is necessary to prevent exceeding acceptable limits:
7.1.1 Boron Dilution Not Reanalyzed 7.1.2 Star}upofanInactiveReactorCoolant Not Reanalyzed Pump 7.1.3 Loss of Lead Not Reanalyzed 7.1.4 Loss of Feedwater Flow Not Reanalyzed 7.1.5 Excess Heat Removal due to Feedwater Not Reanalyzed Malfunction 7.1.6 Reactor Coolant System Depressurization Not Reanalyzed 7.1.7 Excessive Charging ~ Event-Not Reanalyzed 7.2 Anticipated Operational Occurrences for which RPS trips and/or sufficient initial l
steady state thermal trargin, maintained by I
the LCOs, are necessary to prevent exceeding the acceptable limits:
i 2
7.2.1 Sequential CEA Group 3 Withdrawal Not Reanalyzed 7.2.2 Loss of Coolant Flow Not Reanalyzed 7.2.3 Full Length CEA Drop Not Reanalyzed 7.2.4 Transients Resulting from the Not Reanalyzed 4
Malfunction of One Steam Generator 7.2.5 LossofAC[cwer*
Not Peanalyzed 7.2.6 Excess load Not Reanalyzed l
7.3 Postulated Accidents 7.3.1 CEA Ejection NotReanalgzed 7.3.2 Steam Line Rupture Reanalyzed i
7.3.3 SteamGenerajorTubeRupture Not Reanalyzed 7.3.4 Seized Rotor Not Reanalyzed 7.3.5 Feed Line Break Not Reanalyzed ITechnical Specifications preclude this event during operation.
2Requires High Power and Variable High Power Trip.
3Requires Low Flow Trip.
- Requires trip on high differential steam generator pressure.
5Results for Unit 1 Cycle 10 are less limiting than those previously reported.
7-2
24-29a(86f1)/cp-21 l
TABLE 7-2 CALVERT CLIFFS UNIT 1 CYCLE 10 CORE PARANETERS INPUT TO SAFETY ANALYSES FOR DNB AND CTM (CENTERLINE TO PELT) DESIGN LIMITS Reference Cycle
- Physics Parameters Units (Unit 2 Cycle 8)
Unit 1 Cycle 10 Radial Peaking Factors For ONB Margin Analyses T
(Fr)
Unrodded Region 1.70**'+
1.70**'+
Bank 5 Inserted 1.904***+
1.904**'+
ForylanarPadialComponent 3
(F
)of3-0 Peak (CiNLimitAnalyses)
Unrodded Region 1.70**
1.70**
Bank 5 Inserted 1.904**
1.904**
Moderator Temperature 10~4ao/*F
-2.7 + +.7
-2.7 + + 7 Coefficient I
%ao
-4.5
-3.5 9 BOC++
to 4
-5.0 0 E0C Tilt Allowance 3.0 3.0 l
Pcwer Level MWt 2700**
2700**
Maximum Steady State
- F 548**
548**
Inlet Temperature Minimum Steady State psia 2200**
2200**
RCS Pressure 0
Reactor Coolant Flow 10 lbm/hr 138.5**
138.5**
2 Negative Axial Shape I
.15**'+
.15**'+
LC0 Extreme Assumed P
at Full Power (Ex-Cores) 2 j
4 7-3
24-29a(8611)/cp-22 I
J TABLE 7-2 I
(continued)
'l i
Reference Cycle
- Safety Parameters Units (Unit 2 Cycle 8)
Unit 1 Cycle 10 l
Maximum CEA Insertion
% Insertion 25 35 1
at Full Power of Bank 5 Maximum Initial Linear KW/ft 16.0 16.0 Heat Rate for Transients Other than LOCA Steady State Linear KW/ft 22.0 22.0 Heat Rate for Fuel CTM Assumed in the Safety l
Analysis l
CEA Drop Time from sec 3.1 3.1 j
Removal of Power to Holding Coils to 90%
Insertion Minimum DNBR (CE-1) 1.21**
1.15**
4 l
Notes Reference 1 For DNBR and CTM calculations, effects of uncertainties on these l
parameters were accounted for statistically. The procedures used in the Statistical Combination of Uncertainties (References 5, 6 and 7) and Extended Statistical Combination of Uncertainties (Reference 3) programs have been applied to DNB and CTM limits. These procedures have been approved by NRC for the Calvert Cliffs Unitsin Reference 8 and 4, re,pectively.
+
l The values assumed are conservative with respect to the Technical Specification limits.
4
++ A variable shutdown margin, ranging from -3.5%ao at BOC to -5.0%ao at E0C, was assumed in the Unit 1 Cycle 10 analyses, due to the requested change in the shutdown margin Technical Specification (see Section 9.0).
l i
)
7-4
~
24-29a(86II)/cgh-27 8.0 ECCS ANALYSIS 8.1 LARGE BREAK LOSS-OF-COOLANT ACCIDENT 8.1.1 Introduction and Sumary An ECCS performance analysis was performed for Calvert Cliffs Unit 1 Cycle 10 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 1 Cycle 10 was performed using 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 8) was presented in Reference 10.
The method of analysis for Cycle 10 is identical to the reference cycle large break LOCA ECCS performance analysis.
Blowdown hydraulics, refill /reflood hydraulics and hot rod temperature calculations were performed with the fuel parameters which bound Cycle 10 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 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 which bound Cycle 10 and are expected to bound future cycles.
The fuel performance data were calculated by FATES 3B (Reference 13) which is a revised I
version of the FATES 3 fuel evaluation model (Reference 14). FATES 3B has received NRC approval (Reference 15).
Core wide clad oxidation calculations were also performed in this analysis.
Burnup dependent hot red calculations were perfonted with STRIKIN-II to determine the initial fuel conditions which result in the highest I
peak clad temperature (PCT).
The Unit 1 Cycle 10 analysis considered up to 500 plugged tubes per steam generator and a 40 second safety injection pump response time.
It also covered an expanded range of safety injection tank (SIT) operating conditions (i.e., minimum SIT pf) essure of 195 psia, and SIT liquid volume between 1090 and 1190 ft A break spectrum analysis was performed which determined that the i
0.6 Double Ended Guillotine at Pump Discharge (DEG/PD) break yields the highest peak clad temperature.
This break size was also the limiting break size for the reference cycle.
A sumary of the fuel parameter input values for Cycle 10 and the reference cycle is shown in Table 8.1-1.
8-1
24-29a(86I1)/cgh-27a1
)
8.1.3 Results The burnup dependent hot rod calculations demonstrated that the burnup with the highest initial fuel stored energy results in the highest PCT.
This occurred at a low hot rod burnup of 1000 MWD /T.
Table 8.1-2 presents the results of this limiting case.
For comparison purposes, the corresponding values of the reference cycle analysis (Unit 2 Cycle 8) are also presented in Tables 8.1-2.
The peak clad temperature, maximum local clad oxidation and core wide clad oxidation values of 1983'F, 4.1% and <.51%, respectively, are well below the corresponding 10CFR50.46 acceptance criteria limits of 2200*F, 17% and 1%.
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.
The high burnup fuel conditions result in a PCT of 1971*F, 12*F below the limiting case presented above.
The peak local oxidation of 6.3% is well within the acceptance limit of 17%.
A review of the effects of initial operating conditions on these results was performed.
It was determined that over the ranges of initial operating conditions, as specified in the Technical Specifications, operation of the plant at a PLHGR of 15.5 kw/ft is an acceptable limit for Cycle 10 operation.
8.1.4 Conclusions As discussed above, conformance to the ECCS criteria is sumarized by the analysis results presented in Table 8.1-2.
The most limiting case results in a peak clad temperature of 1983*F, which is well below the acceptance limit of 2200*F.
The maximum local and core wide values for zirconium oxidation percentages remain well below the acceptance limit values of 17% and 1%, respectively. Therefore,
)
operation of Unit 1 Cycle 10 at a PLHGR of 15.5 kw/f t and a power 1
level of 2754 MWT (102% of 2700 MWT) is in compliance with the 10CFR50.46 acceptance criteria.
\\
8.2 SMALL BREAK LOSS-OF-COOLANT ACCIDENT Analyses have confirmed that the reported small break loss-of-coolant accident results of the reference cycle (Unit 1 Cycle 8. Reference 16) bound Unit 1 Cycle 10.
These results have been approved by the NRC in Reference 17.
Therefore, acceptable small break LOCA ECCS performance is demonstrated at a peak linear i
heat rate of 15.5 kw/f t and a reactor power level of 2754 MWt (102%
of 2700 MWt).
This acceptable performance has been confimed with up to 100 plugged tubes per steam generator.
The expanded range of SIT operating conditions and increased safety injection pump response time considered in Section 8.1 were not evaluated for the small break LOCA evaluation.
)
l 8-2
24-29a(8611)/c9 -29 h
TABLE 8.1-1 Calvert Cliffs Unit 1 Cycle 10 Fuel Parameters as Compared to Unit 2. Cycle 8 Unit 1-Unit 2 Cycle 10 Cycle 8 Fuel Parameters Values Values Reactor Power Level (102% of Nominal), MWT 2754 2754 Average Linear Heat Rate (102% of Nominal), kw/ft 6.45 6.45 Hot Channel Peak Linear Heat Generation 15.5 15.5 Rate, kw/ft i
Hot Assembly Peak Linear Heat Generation 13.68 13.34 Rate, kw/ft 2
- Gap Conductance at PLHGR, B/hr-ft
'F 2105 2267
- Fuel Centerline Temperature at PLHGP, 'F 3633 3740
- Fuel Average Temperature at PLHGR, 'F 2207 2246
- Hot Rod Gas Pressure, psia 1209 1189
- Hot Rod Burnup, MWD /MTV 1000 943
- Are initial fuel rod parameters, in STRIKIN-II, which yield maximum PCT.
l 8-3
n 24-29a(8611)/cgh-30 TABLE 8.1-2 Summary of ECCS Performance Results for Calvert Cliffs Unit 1 Cycle 10 for the Limiting Break Size (0.6 DEG/PD) as Ccmpared to the Reference Cycle (Unit 2 Cycle 8)
Limiting Case (Maximum Initial Fuel Parameters Stored Eneroy)
Unit 1 Unit 2 Cycle 10 Cycle 8 Rod Average Burnup, 1000 943 MWD /MTV Peak Clad Temperature 1983 1903 (PCT),'F Time of PCT, Seconds 246.5 234 Time of Clad Rupture, 25.7 25.6 Seconds Peak Clad Oxidation, %
4.14 3.30 Core Wide Oxidation,'%
<.51
<.51 8-4
24-29a(8611)/cgh-31 TABLE 8.1-3 Calvert Cliffs Unit 1 Cycle 10 Variables Plotted as a Function of Time for the Limitina large Break Variable Figure Nuttber 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 Curing Reflood 8.1-8 Feak Clad Temperature 8.1-9 Hot Spot Gap Conductance 8.1-10 Peak local Clad 0xidation 8.1-11 Tenperature of Fuel Centerline, 8.2.-12 Fuel Average, Clad and Coolant at Hottest Node j
l Hot Spot Heat Transfer Coefficient 8.1-13 Hot Rod Internal Gas Pressure 8.1-14 i
8-5
Figure 8.1-1 CALVERT CLIFFS UNIT 1 CYCLE 10 0.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG CORE POWER L2-Li - i l
i i
l-I 0.9 -
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Figure 3.1 -2 CALVERT CLIFFS UNIT 1 CYCLE 10
@.6 x DOUBLE ENDED QUILLOTINE BREAK IN PUMP DISCHARGE LEO PRESSURE IN CENTER HOT ASSEMBLY NODE 2.4 2.2 -
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Figure 3.1-3 CALVERT CLIFFS U NIT 1 CYCLE 10 0.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG LEAK FLOW E0 11 0 -
00 -
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o 80-W 4
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Figure 8.1-4 CALVERT CLIFFS U NIT 1 CYCLE 10 0.6 x DOUBl..E ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG HOT ASSEMBLY PLOW, BELOW HOT SPOT l
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Fig u re 8.1 - 5 CALVERT CLIFFS UNIT 1 CYCLE 10 0.6 x DOUBLE ENDED OUILLOTINE BREAK IN PUMP DISCHARGE LEO HOT ASSEMBLY FLOW, ABOVE HOT SPOT 30 l
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i Figure 3.1 - 0 CALVERT CLIFFS UNIT 1 CYCLE 10 i
C.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG HOT ASSEMBLY QUALITY i
NODE 13, BELOW HOTTEST REGION
--- NODE 14, AT HOTTEST REGION
- - NODE 15, ABOVE HOTTEST REGION l
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Fig ure 8.1-7 CALVERT CLIFFS U NIT 1 CYCLE 10 0.6 x DOUBLE ENDED GUILLOTINE BREAK IN l8 UMP DISCH ARGE LEG CONTAINMENT PRESSURE 60 1
50 s i
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Figure s.1 - s CALVERT CLIFFS UNIT 1 CYCLE 10 0.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG MASS ADDED TO CORE DURING REFLOOD 12 0 D0,
TIME, SEC REFLOOD RATE 10 0 -
O.00-10.0 2.461 IN/SEC 10.0-76.0 1.165 IN/SEC 7e.
-4
.789 IN/SEC
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Fig u re s.1 - 9 CALVERT CLIFFS U NIT 1 CYCLE 10 0.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEO PEAK CLAD TEMPERATURE 2.4 2.2 -
l 2-IBM b0 ~ l;l[
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8-14
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Figure s.1-10 CALVERT CLIFFS UNIT 1 CYCLE 10 0.6 x DOUBLE ENDED QUILLOTINE BREAK IN PUMP DISCHARGE LEQ HOT SPOT GAP CONDUCTANCE l
[8 L7-LS-LS-L I4-F L3-h,.
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J Pigure 3.1-11 CALVERT CLIFFS UNIT 1 CYCLE 10
@.S x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG PEAK LOCAL CLAD OXID ATIO N TT -
E-l s,
g_
O-e -l Y
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0 0j 7-o 0 1 5 Jl l
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200 400 M.SEC 8-16 1
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g,-
Figure 0.1-12 CALVERT CLIFFS JJHIT 1 CYCLE 10 0.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCH ARGE LEG TEMPERATURE OF FUEL CENTERLINE, FUEL AVERAGE, CLAD AND COOLANT AT HOTTEST NODE 3-l i
x 23 -
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FUEL AVERAGE 22-CENTERUNE FUEL 6
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8-17
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'4 Figure s.1-13 CALVETT CLIFFS UNIT 1 CYCLE 10 7.8 d DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG HOT SPOT HEAT TRANSFER COEFFICIENT B0 l70-70 0 -
Y 15 0 -
FI 146 --] i e
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200 400 INE SEC 8-18
4 Figure 8.1-14 CALVERT CLIFFS UNIT 1 CYCLE 10 O.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG i
HOT ROD INTERNAL GAS PRESSURE l
[2 \\
PINITI AL= 120 8.2 PSIA ll -
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20 40 60 80 10 0 TIME, SEC 8-19
24-29a(8611)/cgh-32 l
9.0 TECHNICAL SPECIFICATIONS Two sets of requests for Technical Specification changes are presented in this section.
The first set involves only changes to the. Unit 1 Technical Specifications. A summary of these changes is presented in Table 9-1 in the form of:
- 1) an action statement for each change, 2) the reason for each change and 3) a reference to the supporting analyses which demonstrate acceptable safety analyses results for each change.
Following Table 9-1 either the existing Technical Specification page with the intended modifica tion, an already modified page or a new figure is provided for each Technical Specification for which a change is being requested.
The second set of requested changes involves changes to both the Unit 1 and Unit 2 Technical Specifications.
A summary of these changes is presented in Table 9-2 in the form of a description of each change, the reasons for it and documentation which supports each change.
Following Table 9-2 an already modi fied page is provided for each Technical Specification for which a change is requested.
Four of the five sets
- of Technical Specifications changes being requested involve changes not implemented previously. The fifth set has already been implemented in the Unit 2 Technical Specifications.
Specifically, revising of the applicability and form of the SHUTDOWN MARGIN specifications, increasing the Transient Insertion Limit between 90 and 100% power and widening of the Axial Shape Index limits at lower power levels are all new changes.
Implementation of the Extended Statistical Combination of Uncertainties methodology is also a new change, although the nethodology has already been approved.
The change in the positive MTC limit between 70 and 100%
power was previously proposed for Unit 2 Cycle 8 in Reference 1 and approved in Reference 2.
l I
)
- Text plus figure, Specification plus bases, Related Specifications, etc.
A 9-1
24-29a(86II)/cgh-33 TABLE 9-1 CALVERT CLIFFS UNIT 1 CYCLE 10 TECHNICAL SPECIFICATION CHANGES Tech. Spec.
No. and Page Action Explanation Support /Use Figure 2.2-2 Modify the coefficients in The curve fit for the Thermal The setpoint analysis per-page 2-12 the PVAR equation listed Margin / Low Pressure Trip formed for Unit 1 Cycle 10 on Figure 2.2-2, as in-setpoint is being revised to which utilized the ESCO dicated.
accommodate the implementation methodology supports this of the Extended Statistical change.
Combination of Uncertainties (ESCU) methodology (see below).
Figure 2.2-3 Replace Figure 2.2-3 with See change for Figure 2.2-2.
See change for Figure 2.2-2.
y3 A>
page 2-13 enclosed modified Figure 2.2-3.
B 2.1.1 Modify the text, as indica-The text is being modified, The ESCU methodology of pages B 2-1 and ted.
including the replacement of a Reference 3 was approved in B 2-3 specific minimum DNBR value with Reference 4.
The use of a the approved phrase from phrase in place of a specific Reference 3, to reflect the minimum DNBR value was recom-implementation of the ESCU mended in Reference 3 and methodology. The ESCU approved in Reference 4.
methodology was used in the Unit 1 Cycle 10 analyses to demonstrate additional thermal margin to the DNB SAFDL. This additional margin was used to justify the change in the HFP PDIL (see below) and to offset margin losses in other areas due to the implementation of 24-month cycles for Unit 1.
7 i
24-29a(86I1)/cgh-34 TABLE 9-1 (Cont'd)
CALVERT CLIFFS UNIT 1 CYCLE 10 TECHNICAL SPECIFICATION CHANGES 4
Tech. Spec.
No. and Page Action Explanation Support /Use B 2.2.1 Replace the minimum DNBR The specific minimum DNBR value See change for Spec.
pages B 2-5 and value of 1.21 with the is being replaced, as part of B 2.1.1.
B 2-6 indicated phrase.
the implementation of the ESCU methodology, with the approved phrase from Reference 3.
1.a) The elimination 3.1.1.1,4.1.1.1.1 Replace Specs. 3.1.1.1,
- 1) The applicabiilty of these of the applicability of and 4.1.1.1.?
4.1.1.1.1 and 4.1.1.1.2 specs to the critical modes is these specs to the pages 3/4 1-1 and with enclosed modified being removed and the critical modes does not 3/4 1-2 (old) specs. which have been surveillance requirements which affect the safety cal-condensed to one page relate to critical conditions culations in any manner.
(3/4 1-1), leaving page are being eliminated.
Adequate SHUTDOWN MARGIN 3/4 1-2 for use by the Presently, these specs which are for the critical modes new figure 3.1-lb.
in the Boration Control section has always been assured (3/4.1.1) specify that the via calculations, using SHUTDOWN MARGIN listed therein the maximum pre-trip CEA is applicable to the critical insertion allowed by the modes. However, adequate Transient Insertion SHUTDOWN MARGIN for the critical Limit of the PDIL (Spec.
modes can only be assured via 3.1.3.6) to minimize the the specs in the Movable Control calculated value.
Assemblies section (3/4.1.3),
bora"inn control being 1.b) Table 9-la contains irrelevant in these modes to the a detailed description maintenance of SHUTDOWN MARGIN.
of the role of individual Consequently, to remove this specifications in the incor.31stency the applicchility Movable Control of these specs to the critical Assemblies section modes has been removed along (3/4.1.3) in assuring with the appropriate adequate SHUTDOWN MARGIN surveillance requirements.
for the critical modes.24-29a(8611)/cgh-35 TABLE 9-1 (Cont'd)
CALVERT CLIFFS UNIT 1 CYCLE 10 TECHNICAL SPECIFICATION CHANGES Tech. Spec.
No. and Page Action Explanation Support /Use
- 2) The SHUTDOWN MARGIN is being
- 2) See Change No. 1 for new changed from a constant value to Figure 3.1-lb.
text which simply refers to a new Figure 3.1-lb.
See Change No. I for new Figure 3.1-lb.
Figure 3.1-1b Insert enclosed new Figure
- 1) The SHUTDOWN MARGIN is being
- 1) The variation of the (new) 3.1-lb after page 3/4 1-1.
functionalized versus time in specified SHUTDOWN page 3/4 1-2 cycle.
MARGIN with time in u)I.
(new)
Presently, the specified cycle has been SHUTDOWN MARGIN is a singular incorporated in all value, independent of time in appropriate safety cycle, boron concentration, etc.
analyses (see Section 7.0).
However, the required SHUTDOWN MARGIN varies throughout the cycle as a function of fuel depletion, boron concentration and moderator temperature. To more realistically reflect this variation the SHUTDOWN MARGIN has been functionalized versus time in cycle via the addition of a new figure.
- 2) The shutdown margin is being
- 2) The results of the increased at EOC from 3.5% Ak/k revised Steam Line to 5.0% Ak/k to support a Pupture analysis were revised Steam Line Rupture less limiting than those analysis.
previously reported.
Consequently, this analysis was not reported in section 7.0.24-29a(8611)/cgh-36 TABLE 9-1 (Cont'd)
CALVERT CLIFFS UNIT 1 CYCLE 10 1ECHNICAL SPECIFICATION CHANGES Tech. Spec.
No. and Page Action Explanation Support /Use 3.1.1.4 Modify Spec. 3.1.1.4, as See change for new Figure 3.1-la.
See change for new Figure page 3/4 1-5 indicated: Condense 3.1-la.
Subsections "a" and "b" to a new Subsection "a" which simply refers to a new Figure 3.1-la, and change Subsection "c" to Subsection "b" without modifying the contents of this.
subsection.
?
u,24-29a(86II)/cgh-37 TABLE 9-1 (Cont'd)
CALVERT CLIFFS UNIT 1 CYCLE 10 TECilNICAL SPECIFICATION CHANGES Tech. Spec.
No. and Page Action Explanation Support /Use Figure 3.1-la Insert enclosed new Figure The positive MTC limit above 70%
This change is identical (new) 3.1-la af ter page 3/4 1-5.
power ig being raised from to the change requested varies linearly from +.3x10-gh
+.2x10 ap/*F to a value whi in Reference 1 and page 3/4 1-Sa approved in Reference 2.
(new)
Ap/ F af 100% power to
+.7x10 Ap/*F at 70% power.
The Feedline Break This change is being made to (FLB) analysis accommodate the implementation which supports this of 24-month cycles, to eliminate change in MTC limit was startup delays and to facilitate discussed in Reference a rapid power ascension program.
1.
This FLB analysis has been determined to be applicable to Unit 1 Cycle 10.
4, The rapid power ascension program which will be used for Unit 1 Cycle 10 takes credit for this change in MTC limit. This program is identical to the program described in Section 10.0 of Reference 1.
I
24-29a(8611)/cgh-38 TABLE 9-1 (Cont'd)
CALVERT CLIFFS UNIT 1 CYCLE 10 TECHNICAL SPECIFICATION CHANGES Tech. Spec.
No. and Page Action Explanation Support /Use Figure 3.1-2 Replace Figure 3.1-2 with The Transient Insertion Limit This change has been factored page 3/4 1-27 enclosed modified Figure between 90% and 100% power is into all of the Unit 1 Cycle 10 3.1-2.
being increased from an allowed physics, safety and setpoint insertion limit which varies analyses (see Figure 5-1 and linearly from 35% for Bank 5 Table 7-2).
at 90% power to 25% at 100%
power, to a constant value of 35%. This increase is being made to enhance the ability to control axial oscillations near 1)
E0C.
u B 3/4.1.1.1 and Modify the text, as See Change No. I for specs.
See Change No. I for B 3/4.1.1.2 indicated.
3.1.1.1, 4.1.1.1.1. and Specs. 3.1.1.1, page B 3/4 1-1 4.1.1.1.2; see Change No. I for 4.1.1.1.1. and new Figure 3.1-lb.
4.1.1.1.7; see Change No. 1 for new Figure 3.1-1b.24-29a(86II)/cgh-39 o-TABLE 9-1 (Cont'd)
CALVERT CLIFFS UNIT 1 CYCLE 10 TECHNICAL SPECIFICATION CHANGES Tech. Spec.
No. and Page Action Explanation Support /Use B 3/4.7.3 Modify the text, as The text is being modified to a) The addition of these pages B 3/4 1-3 indicated.
clarify and emphasize the clarifying and amplifing and B 3/4 1-5 function of the Movable Control statements does not Assemblies section (3/4.1.3) affect the safety in assuring adequate SHUTDOWN analyses in any manner, MARGIN for the critical modes.
as discussed above for This change is being made in Specs. 3.1.1.1, etc.
conjunction with the removal of the applicability of Specs.
b) Table 9-la contains a 3.1.1.1, 4.1.1.1.1 and 4.1.1.1.2 detailed description of to the critical modes (see the role of individual above).
specifications in the u>
os Movable Control Assemblies section (3/4.1.3) in assuring adequate SHUTDOWN MARGIN for the critical modes.
I
24-29a(8611)/cp-40 TABLE 9-la SHUTDOWN MARGIN TECHNICAL SPECIFICATIONS FOR CRITICAL MODES Adequate SHUTDOWN MARGINS at critical conditions (Modes 1 and 2 (Keff 21.0) are maintained throughout the cycle for all power levels by the following specifications:
1)
Spec. 3.1.3.1 requires that all full length CEAs be OPERABI.E, i.e.,
as it pertains to SHUTDOWN MARGIN requirements, all CEAs are not either immovable as a result of excessive friction or mechanical interference, or known to be untrippable; ACTION Statement 'a' provides a remedy, if this requirement is violated; Spec. 4.1.3.1.2 specifies testing to ensure that this operable condition is maintained.
2)
Spec. 3.1.3.5 requires that all shutdown CEAs be withdrawn. to the proper position; the ACTION Statement provides a remedy, if the CEAs are not sufficiently withdrawn; Spec. 4.1.3.5 specifies testing to ensure that the shutdown CEAs are positioned properly.
3)
Spec. 3.1.3.6 requires that all regulating CEA groups be withdrawn to the proper position, i.e., as it pertains to SHUTDOWN MARGIN, all regulating CEAs are within the Tran:iient Insertion Limits; ACTION Statement 'a' provides a remedy, if this requirement is violated; Spec. 4.1.3.6 specifies testing to ensure that the regulating CEAs are withdrawn within the Transient Insertion Limits.
4)
Spec. 3.1.3.1 requires that all CEAs in a group be properly aligned, i.e., as it pertains to SHUTDOWN MARGIN, all CEAs in a regulating group shall be within 7.5 inches of all other CEAs in their group; ACTION Statements 'b' through 'h' provide remedies, if the CEAs are not properly aligned; Specs. 4.1.3.1.1 and 4.1.3.1.3 specify testing to ensure that the proper alignment is maintained.
9-9 i
.-.