ML20046B916

Cycle 5 Redesign Reload Safety Evaluation

ML20046B916
Person / Time
Site:	Millstone
Issue date:	07/31/1993
From:	Slagle W WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20046B910	List: B14563, Proposed Tech Specs Increasing Volume Requirements of Boric Acid Storage Sys of TS 3.1.2.6 in Order to Meet Requirements of Redesigned Core for Cycle 5 Operation ML20046B909 ML20046B916
References
NUDOCS 9308090016
Download: ML20046B916 (36)

Text

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MILLSTONE UNIT 3 L

CYCLE 5 REDESIGN RELOAD SAFETY EVALUATION

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July 1993 Edited by:

L W. H. Slagle i

Approved:

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E. H. Novendstern, Manage'r Fuel Licensing Integration 9308090016 930731 f'"#

Nuclear Manufacturing Divisions PDR ADOCK 05000423 p

PDR Lj htinghouse Electric Corporation W

Nuclear Manufacturing Divisions P. O. Bos 355 Pittsburgh, Penns3 tania 15230 FLI-93-026 1

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TABLE OF CONTENTS Section Title Page

1.0 INTRODUCTION

AND SU5151ARY...........................

I 1.1 In t rod uct io n......................................

I 1.2 General Description.................................

2 1.3 Conclusions and Assessment 2

2.0 REACTOR DESIGN 8

2.1 Mechanical Design..................................

8 2.2 Nucl ear Design.................................... 11 2.3 Thermal and IIydraulic Design.......................... 12 3.0 POWER CAPABILITY AND ACCIDENT EVALUATION

...........14 3.1 Power Lapability................................... 14 3.2 Accident Evaluation................................. 14 3.2.1 Kinetics Parameten............................ 16 3.2.2 Control Rod Wort hs............................ 16 3.2.3 Core Peaking Factors........................... 16 4.0 TECIINICAL SPECIFICATION CIIANGES 17 I

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5.0 REFERENCES

19 Appendix A : TECIINICAL SPECIFICATION MARKUPS........v........:.... - #A-1 i

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LIST OF TABLES Table Title Page 1

Fuel Assembly Design Parameters

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Afillstone Unit 3 - Cycle 5.................................. 22 2

Kinetics Characteristics N Loop Operation Afillstone Unit 3 - Cycle 5.................................. 23 3

End-of-Cycle Shutdown Requirements and Afargins N Loop Operation Afillstone Unit 3 - Cycle 5.................................. 24 I

LIST OF FIGURES 1

L Figure Title Page F

1 1

Core Loading Pattern Atillstone Unit 3 - Cycle 5.................................. 25 L

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Burnable Absorber and Source Rod Location Afillstone Unit 3 - Cycle 5.................................. 26 L.

3 K(Z) - Normalized F (Z) as a Function of Core IIeight n

for Four Loop Operation 1

i Afillstone Unit 3 - Cycle 5.................................. 27 L

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J MILLsroNE UNrr 3 - CYCLE 5 REDESIGN Jt1LY 1993 Y

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1.0 INTRODUCTION

AND

SUMMARY

1.1 Introduction This redesign Reload Safety Evaluation (RSE) presents an evaluation for Millstone Unit 3 Cycle 5, which demonstrates that the core reload will not adversely affect the safety of the plant for N loop operation.

This evaluation was accomplished utilizing the methodology described in WCAP-9273-NP-A,

" Westinghouse Reload Safety Evaluation Methodology"m for the incidents presented in the FSARm and Plant Safety Evaluation * (PSE) that are within Westinghouse's scope.

l The Millstone Unit 3 reactor operated in Cycle 4 with its first transition core of Westinghouse 17x17 4

VANTAGE 5H fuel assemblies. For Cycle 5 (expected startup October 1993) and subsequent cycles, it is planned to refuel the Millstone Unit 3 core with VANTAGE 5H fuel. The PSES provides justification for the transition from standard fuel to VANTAGE 5H fuel, thimble plug removal and the associated proposed changes

• to the Millstone Unit 3 Technical Specifications

• It also justifies the compatibility of standard with VANTAGE 5H fuel assemblies in a transition core and the operation with a full VANTAGE 5H core. This report is consistent with the PSES and other associated analyses **

which contain reanalyses applicable to the FSARm The Millstone Unit 3 Cycle 5 core has been redesigned as aresult of assembly re-caging and core shuffles

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that were specified due to grid-to-rod fretting findings at Salem and Beaver Valley..'The results of the redesign are:

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-- All once-burnt VANTAGE 5H fuel assemblies that were originallylocated on the baffle have been moved inboard.

Eight twice-burnt STANDARD fuel assemblies that were ' originally located inboard have been moved to baffle locations.

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Fourty-eight of the fresh Regi6n'7 VANTAGE SH fuel assemblies"were re-caged (refer to Section 2.1 for details of re-caging).

Ilased upon the above referenced methodology, only those incidents analyzed and reported in the FS AR*,

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PSE* and other associated analyses *" that are within Westinghouse's scope which could potentially be Rev. 07/14/93 12:10pm 1

l MILLSTONE UNrr 3 - CYCLE 5 REDESIGN JULY 1993 Y

affected by this redesign fuel reload have been reviewed for the Cycle 5 design described herein. The justihcation for the applicability of previous results is provided in Sections 3.1 and 3.2.

1.2 General Description The Millstone Unit 3 reactor core is comprised of 193 fuel assemblies arranged in the core loading l

pattern conHguration shown in Figure 1. The Cycle 5 core loading configuration features a low leakage pattern. During Cycle 4/5 refueling,36 fresh Region 7 VANTAGE SH assemblies (4.4 w/o U-235) and 48 fresh' re-caged Region 7 VANTAGE 5H asse'mblies (4 4'.wlo' U 235) will replace 9 Region 4A, 20 Region 4B, 32 Region 5A, and 23 Region 5B fuel assemblies. A summary of the Cycle 5 fuel inventory is given in Table 1.

Nominal core design parameters utilized for Cycle 5 are as follows:

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Four Loop (N)*

Core Power (MWt) 3411 System Pressure (psia) 2250 Core Inlet Temperature ( F) 557.0 l

RCS Thermal Design Flow (gpm) 378,400 Average Linear Power Density (kw/ft)"

5.434 1.3 Conclusions and Assessruent c

From the evaluation presented in this report, it is concluded that the Millctone Unit 3 Cycle 5 reload redesign does not result in the acceptable safety limits for any accident being exceeded and does not result in any unreviewed safety questions as defined in 10 CFR 50.59. The basis for this determination is r

delineated below. The reload design criteria are referenced throughout this document.

I i

Auumes 10% of steam generator tubes in each generator are plugged and corresponds to the worst plugging level of any steam generator.

Linear power demity based on average active fuel stack height of 143.7 in.

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MILLSTONE UNrr 3 - CYCLE 5 REDEslGN

.IULY 1993 1.

Will the r robability of an accident previousiv evaluated in the FS AR be increased?

This RSE documents that the probability of an accident previously evaluated in the Millstone Unit 3 FSARS is not increased. The Cycle 5 reload core redesign meets all applicable design criteria and ensures that all pertinent licensing basis acceptance criteria are met. Though fuel and core design are not directly related to the probability of any previously evaluated accident, the demonstrated adherence to applicable standards and acceptance criteria precludes new challenges to components and systems that could increase the probability of any previously evaluated accident. Specifically, the mechanical changes as specified in Section 2.1 will not increase the probability of occurrence of an accident previously evaluated in the Millstone Unit 3 FSARS

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The clad integrity is maintained and the structural integrity of the fuel rods, fuel assemblies, and

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core is not affected. The ZIRLO clad fuel rod improves corrosion performance and dimensional stability. The other mechanical features, noted in Section 2.1, have no impact on fuel rod performance or dimensional stability as documented in the RSE nor will they cause the core to operate in excess of pertinent design basis operating limits. Therefore, the probability of occurrence of an accident previously evaluated in the FSARS has not increased.

2.

Will the consecuences of an accident previously evaluated in the FS AR be increased?

This RSE documents that the consequences of an accident previously evaluated in the Millstone Unit 3 FSAR* is not increased. The Cycle 5 reload core redesign is designed within the Technical Specification limits; therefore, it does not have any affect on the consequences of any accident. and does not affect any of the bases (assumptions, actions, etc) for the current analyses l

as described in the Millstone Unit 3 FSARS The Cycle 5 reload core redesign meets all applicable design criteria and ensures that all pertinent licensing basis acceptance criteria are met.

The demonstrated adherence to these standards and criteria precludes new challenges to components and systems that could: a) adversely affect the ability of existing components and systems to mitigate the consequences of any accident and/or; b) adversely affect the integrity of the fuel rod cladding as a fission product barrier. Furthermore, adherence to applicable standards L

and criteria ensures that these fission product barriers maintain design margin to safety.

Specifically, the mechanical changes as specified in Section 2.1 will not increase the consequences of an accident previously evaluated in the Millstone Unit 3 FSAR* The ZIRLO' clad material is similar in chemical composition and has similar physical and mechanical properties as that of Zircaloy-4. The other mechanical features, noted in Section 2.1, have no impact on chemical, physical or mechanical properties as documented in this RSE nor will they cause the core to operate in excess of pertinent design basis operating limits. Thus, clad integrity is maintained.

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Since the predictions presented in the FSARS are not sensitive to the mechanical changes specified in this report, the consequences of accidents previously evaluated in the Millstone Unit 3 Ret 07d4/93 12:lopm 3

MILLSTONE UNrr 3 - CYCLE 5 REDESIGN JULY 1993 FSAR* have not increased.

3.

May the possibility of an accident which is different from any already in the FSAR be created?

This RSE documents that the possibility of an accident which is different from any already in the Millstone Unit 3 FSAR* is not created. The Cycle 5 reload core redesign meets all applicable design criteria and ensures that all pertinent licensing basis acceptance criteria are met. The demonstrated adherence to these standards and criteria precludes new challenges to components and systems that could introduce a new type of accident. Specifically, the mechanical changes as specified in Section 2.1 will not create the possibility of an accident of a different type than any previously evaluated in the Millstone Unit 3 FSARm. The fuel assemblies containing the mechanical features noted in Section 2.1 will satisfy the same design bases""** as that used for fuel assemblies in the other fuel regions. All design and performance criteria will continue to be met and no new single failure mechanisms have been created as documented in this RSE nor will they cause the core to operate in excess of pertinent design basis operating limits. Therefore, the possibility of an accident of a different type than any previously evaluated in the FSARS has not been created.

I 4.

Will the probability of a malfunction of eauipment important to safety previousiv evaluated in the FSAR be increased?

[

This RSE documents that the probability of a malfunction of equipment important to safety I

previously evaluated in the Millstone Unit 3 FSARm is not increased. The Cycle 5 reload core redesign meets all applicable design criteria and ensures that all pertinent licensing basis acceptance criteria are met. Demonstrated adherence to applicable standards and acceptance criteria precludes new challenges to components and systems that could increase the probability of any previously evaluated malfunction of equipment important to safety. Specifically, the riechanical changes as specified in Section 2.1, in compliance with the methodology established in References 8,9 and 10, will not increase the probability of occurrence of a malfunction of L

equipment important to safety previously evaluated in the Millstone Unit 3 FSAR* No new performance requirements are being imposed on any system or component such that any design I

criteria will be exceeded as documented in this RSE nor will they cause the core to operate in excess of pertinent design basis operating limits. No new modes or new limiting single failures have been created with the mechanical features noted above. Therefore, the probability of L

occurrence of a malfunction of equipment important to safety previously evaluated in the FSARS has not increased.

Rev o7/14/93 12:lopm 4

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MILLSTONE UNrr 3 - CYCLE 5 REDES!GN JULY 1993 3

5.

Will the consecuences of a malfunction of eauipment important to safety previously evaluated in the FSAR be increased?

This RSE documents that the consequences of a malfunction of equipment important to safety previously evaluated in the Millstone Unit 3 FS AR"' are not increased. The Cycle 5 reload core redesign is designed within the Technical SpeciHcation limits; therefore, it does not have any affect on the consequences of any malfunction of equipment important to safety, and does not affect any er the bases (assumptions, actions, etc) for the current analyses as described in the Millstone Unit 3 FSARm The Cycle 5 reload core redesign meets all applicable design uiteria and ensures that all pertinent licensing basis acceptance criteria are met. The deuonstrated adherence to these standards and criteria precludes new challenges to component.c and systems that could: a) adversely affect the ability of existing components and systemc to mitigate the consequences of any accident and/or; b) adversely affect the integrity of the fuel rod cladding as a Hssion product barrier. Furthermore, adherence to applicable standards and criteria ensures Y

that these fission product barriers maintain design margin of safety. Specifically, the mechanical changes as specified in Section 2.1 will not increase the consequences of a malfunction of equipment important to safety previously evaluated in the Millstone Unit 3 FSARm The offsite l

dose predictions presented in the FSAR* are not sensitive to the fuel rod cladding material or other mechanical changes that do not alter the metallurgical composition of the core. The use

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of ZIRLO cladding material, or the other mechanical features mentioned in Section 2.1, do not change the performace requirements on any system or component such that any desigo. heria

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will be exceeded as do;umented in this RSE nor will they cause the core to operate m eAcess of L

pertinent design basis operating limits. No new modes or new limiting single failures have been created with any of the mechanical features mentioned above. Therefore, the consequences of L

a malfunction of equipment important to safety previously evaluated in the Millstone Unit 3 FSAR"' have not increased.

6.

May the possibility of a malfunction of ecuipment important to safety different from any already evaluated in the FSAR be created?

L This RSE documents that the possibility of a malfunction of equipment important to safety

[

different from any already evaluated in the Millstone Unit 3 FSARm is not created. The Cycle 5 reload core redesign meets all applicable design criteria and ensures that all pertinent licensing basis acceptance criteria are met. The demonstrated adherence to these standards and criteria e

precludes new challenges to components and systems that could introduce a new type of a malfunction of equipment important to safety. Specifically, the mechanical changes as specified

[

in Section 2.1 will not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the Millstone Unit 3 FSARA. All original I

L Rev. 07!!4/93 12. lopm 5

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l MILLSTONE UNrr 3 - CYCLE 5 REDESIGN JULY 1993 E

design and performance criteria continue to be met, and no new failure modes have been created for any system, component, or piece of equipment. No new single failure mechanisms have been

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introduced as documented in this RSE nor will they cause the core to operate in excess of pertinent design basis operating limits. Therefore, the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the FSARS has not been created.

7.

Will the marcin of safety as defined in the BASES to any technical specifications be reduced?

This RSE document aat the margin of safety as defined in the Bases to any Millstone Unit 3 Technical SpecificuunsF is not reduced. The Cycle 5 reload core redesign meets all applicable design criteria and ensures that all pertinent licensing basis acceptance criteria are met. It has been determined that the Millstone Unit 3 VANTAGE 5H reload design and safety analysis limits remain applicable, and that these limits are supported by the applicable Millstone Unit 3 l

Technical Specifications

• for Cycle 5 (refer to Section 4.0). Specifically, the mechanical changes as specified in Section 2.1 will not reduce the margin of safety as defined in the basis for any Technical Specification

• The use of these fuel assemblies will take into consideration normal core operating conditions within the allowable technical specification values. For each cycle reload core, these fuel assemblies will be speci6cally evaluated using approved reload design methods

• and fuel rod design models and methods ("'

This will include considerations of the core physics analysis peaking factors and core average linear heat rate effects. Therefore, the margin of safety as defined in the Bases to the Millstone Unit 3 Technical Specifications

• has not Imen reduced.

l Based upon the preceding information and the following:

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1) the End-Of-Cycle 4 burnup is between 19,500 and 21,080 MWD /MTU:

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2) the Cycle 5 burnup will not exceed the predicted End-Of-Full Power Capability (with control rods fully withdrawn and approximately 0 to 10 ppm of residual boron at the Cycle 5 rated power condition of 3411 MWt) plus up to 1,500 MWD /MTU of power coastdown operation;

3) there is adherence to plant operating limitations given in the Technical Specifications

• and the Core Operating Limits Report (COLR);

Rev. 07/22/93 12:16pm 6

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MILLSTONE UNrr 3 - CYCLE 5 REDESIGN JULY 1993 3

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4) NRC approval of the Technical Specification changes in Reference 7 is received prior to Cycle 5 core loading; and

5) TheTechnical Specification changes specified in Section4.0 and Appendix A are implemented

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for Cyc!c 5 prior to entry into Mode 2.*

there are no unreviewed safety questions or Technical Specifications changes required, other than the changes as specified in items 4 and 5 above, as a result of the Millstone Unit 3 Cycle 5 core design.

Therefore, the Cycle 5 reload redesign is licensable under 10 CFR 50.59 and requires no prior NRC approval.

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F To mund an emergency Technical specification chang,e, adminstrative contmis could he implenwnted until the Technical Specificaion change is approved by the NRC This option would be at Northeast Utilitien discretion, by Rev. 07/22/93 2:37pm 7

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MILLsrONE UNrr 3 - CYCLE 5 REDESIGN JULY 1993 Y

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2.0 REACTOR DESIGN 2.1 Mechanical Design The mechanical design of the Region 7 VANTAGE 5H fuel assemblies is the same as the Regions 6A and 6B fuel assemblies except for the following: (1) modifications to remove the cusp and keyhole in the top nozzle assembly, (2) shorter guide thimble tube fabricated with ZIRLO*, (3) shorter instrumentation tube fabricated with ZIRLO", (4) mid-grids fabricated with ZIRLO", (5) IFM grids fabricated with ZIRLO', (6) shorter fuel rod clad with ZlRLO", (7) shorter plenum spring, (8) annular axial blanket pellets, (9) IFBA coated fuel pellets with increased B-10 loading, (10) chrome plated enhanced performance RCCAs, and (11) 48 re; caged Region 7. fuel ~ assemblies. A description of each of these fuel design improvements follows.

1 (1)

The top nozzle assembly has been modified to remove the keyhole and cusp to prevent top nozzle l

spring hang-up on the nozzle cusp during operation. The modified design uses a through slot to accommodate the spring tang. This generic change has been made to all top nozzle designs so that the parts and manufacturing processes may be standardized. The removal of the cusp and keyhole design in the top nozzle does not compromise the performance of any safety-related system nor result in any adverse effect on any analysis, since this change does not affect the normal plant operating parameters, the safeguards systems actuation, or the assumptions and input I

parameters used in these analyses.

l (2,3,6) The guide thimble tubes, instrumentation tubes and fuel rod cladding have all been fabricated with ZlRLO" Each respective length has been reduced, resulting in a slightly shorter overall fuel assembly height. These changes accommodate extended burnups beyond the 60,000 MWD /MTU licensing limit. The extended burnups would result in additional growth of the assembly. Fuel y

rod irradiation experience indicates that ZIRLO fuel rods grow approximately half as much as those of standard Zircaloy-4 Based on this growth, the Region 7 ZIRLO" fuel assemblies should grow no more than previous Zircaloy-4 based fuel assembly designs. ZIRLO" also e

L provides increaced corrosion resistance over that of standard Zircaloy-4. The ase of guide thimble tubes, instrumentation tubes and fuel rod cladding fabricated with ZI(LO' does not k

compromise the performance of any safety-related system nor result in any advarse effect on any f

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MILLSTONE UNrr 3 - CYCLE 5 REDEslGN JULY 1993 3

analysis, since this change does not affect the normal plant operating parameters, the safeguards systems actuation, or the assumptions and input parameters used in these analyses""

(4,5)

The mid-grid and IFM grid assemblies used in Region 7 are similar to those used in the previous region, except that each was fabricated with ZIRLO' The use of ZIRLO' provides increased corrosion resistance as well as increased dimensional stability of the grid assemblies at high

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burnups. The use of mid-grids and IFM grids fabricated with ZIRLO does not compromise the performance of any safety-related system nor result in any adverse effect on any analysis, since this change does not affect the normal plant operating parameters, the safeguards systems actuation, or the assumptions and input parameters used in these analyses"*

L (7)

The plenum spring used in each Region 7 fuel rod is similar to that used in the previous region, except that the spring free length has been decreased to accommodate the shorter ZIRLO' clad

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fuel rods. This revised spring provides a holddown force comparable to the previous spring.

I The use of the shorter plenum spring does not compromise the performance of any safety-related system nor result in any adverse effect on any analysis, since this change does not affect the r

normal plant operating parameters, the safeguards systems actuation, or the assumptions and input l

4 parameters used in these analyses'*

(8)

The annular axial blanket pellets used in all Region 7 fuel rods are identical to those of the previous region, except for the addition of an annulus through the center of each pellet and the l

L removal of the dish on each end of the pellets. Annular axial blankets pellets reduce neutron p

leakage, improve fuel utilization and provide additional void volume to accommodate increase L

fission gas release due to extended burnups beyond the 60,000 MWD /MTU licensing limit. The use of annular axial blankets pellets does not compromise the performance of any safety-related system nor result in any adverse effect on any analysis, since this change does not affect the normal plant operating parameters, the safeguards systems actuation, or the assumptions and input

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g parameters used in these analyses"

The IFBA coated fuel pelles, utilized in Region 7, are identical to those used in the previous (9) region, except that the pellets have an increased B-10 loading in the diboride coating. IFBAs provide power peaking and moderator temperature coefficient control. The use ofincreased B-10 loading in the diboride coating of the IFBA coated fuel pellets does not compromise the L

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MILL 5 TONE UNIT 3 - cycle 5 REDESIGN JULY 1993 3

performance of any safety-related system nor result in any adverse effect on any analysis, since this change does not affect the normal plant operating parameters, the safeguards systems actuation, or the assumptions and input parameters used in these analyses"*

(10)

Enhanced Performance Rod Cluster Control Assemblies (EP-RCCAs) will be utilized in Cycle 5.

These have a thin chrome electroplate applied to the absorber rodlet cladding to provide increased resistance to cladding wear. In addition, the absorber diameter is reduced slightly at the lower extremity of the rodlets in order to accommodate absorber swelling and minimize cladding interaction. The use of the EP-RCCAs does not compromise the performance of any safety-related system nor result in any adverse effect on any analysis, since this change does not affect the normal plant operating parameters, the safeguards systems actuation, or the assumptions and input parameters used in these analyses.

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(11)

Fuel assembly re-caging: 48 of the' original 84 Region 7 fuel assemblies were re-caged. The re-caging process invoh*ed removing the top nozzle assemblp, bottom nozzle assembly _and fuel rods from each assembly.' The fuel rods were then "re-caged" 'in new skeletons having ZIRLO" guide thimble and instrumentation tubing and Zircaloy-4 mid-grids and IFMs; The existing top and bottom nozzle' assemblies were re-used on the re caged fuel assemblies; ~ Due to schedular constraints. Zircaloy-4 mid-grids and IFMs were used for the re-caging in lieu of ZIRLO*.

During the assembly of the re-cage skeletons, the axial orientation of the Zircaloy-4 mid and IFM grids was modified such that the odd. numbered Zircaloy-4 grids with referetce to the bottom

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(bottom inconel grid is grid #1, second grid (Zircaloy-4) from~ the bottom is #2, etc.) are rotated 90 degrees in the clockwise direction. For the Region 7 re-caged fuel assemblies,' this ~means that grid #3 and each of the three IFM grids (#5, #7, and #9) are rotated as described'above.

The purpose for the rotation is to minimize the 'susceptibilitpof the fuel assembly.to flow induced

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vibration. Since there are no physical changes to _the grids or their axial positions, this change

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will have no impact on pressure dropsf DNB performance, or other thermohydraulic criteria.

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All fuel assembly design criteria remain valid with no impact to the fuel.

Thimble plug usage or removal is justified for the Cycle 5 core as documented in the PSE*

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MILu; TONE UNrr 3 - CYCLE 5 REDESIGN JULY 1993 3

Table 1 presents a comparison of pertinent design parameters of the various fuel regions for Cycle 5.

Fuel rod design evaluations for Millstone Unit 3 Cycle 5 fuel were performed using NRC approved

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models"O, and the NRC approved high burnup design methods", to demonstrate that all of the NRC Standard Review Plan fuel rod design bases are satisfied.

Westinghouse has had considerable experience with Zircaloy clad fuel and VANTAGE 5 fuel assemblies.

This experience is summarized in WCAP-8183, " Operational Experience with Westinghouse Cores,"".

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2.2 Nuclear Design

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The nuclear design models for the Cycle 5 transition core are based on the PHOENIX-P*' and Advanced Nodal Code (ANC)"* computer codes. PHOENIX-P is a two-dimensional transport theory based code

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which calculates lattice physics constants. ANC is an advanced nodal analysis theory code capable of two-dimensional and three-dimensional calculations. These supplement the standard " Westinghouse Reload Safety Evaluation Methodology"*

The Cycle 5 core loading is designed to meet the ECCS limit of F x P 5; 2.60 x K(Z)* for four loop e

operation.

Relaxed Axial Offset Control (RAOC) methodology"* will be employed in Cycle 5 to enhance operational

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flexibility. RAOC makes use of available margin by expanding the allowable.il band, particularly at reduced power. The analysis for Cycle 5 indicates that no changes to the safety parameters are required for RAOC operation.

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Table 2 provides a comparison of the Cycle 5 kinetics characteristics with the current limits based on the results of accident analyses reported in the PSES and other associated analyses" submitted to the NRC.

it can be seen from the Table 2 parameters that all of the Cycle 5 values fall within the bounds of the r

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reference analysis limits. These parameters are evaluated in Section 3.0. Table 3 provides the control rod worths and requirements at the most limiting condition during the cycle. The available shutdown margin exceeds the minimum required.

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MILLSTONE UNrr 3 - CYCLE 5 REDESIGN JULY 1993 Y

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For Cycle 5 all remaining Hafnium RCCAs will be replaced by Ag-In-Cd EP-RCCAs. This change has been evaluated to allow the exchange of the Ag-In-Cd EP-RCCAs for any number of Hf RCCAs. This L,

is possible since both RCCA designs have similar neutronic characteristics. The largest change in total rod worth during the cycle due to the change in RCCAs is less than 100 pcm*. Core peaking factors change by less than 1 %. As a result, the core performance characteristics of the Ag-In-Cd EP-RCCAs remain essentially the same. The available shutdown margin will exceed the minimum required shutdown

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margin and all other Technical Specification limits related to nuclear design will be met for the RCCA configuration described above.

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The loading pattern for Cycle 5 is shown in Figure 1. The Millstone Unit 3 Cycle 5 burnable absorber I

loading will consist of 4608 fresh IFBA located in the Region 7 fuel assemblies (Figure 2). Secondary L

sources are located in each of two Region 6B fuel assemblies (location H-03 and H-13) (Figure 2).

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L 2.3 Thermal and Ilydraulic Design The DNB core limits and safety analysis used for Cycle 5 are based on conditions given in Sections 1.2 and 3.0. Fuel temperatures were calculated using the new fuel thermal model""

Tests and analyses have confirmed that the VANTAGE 5H fuel assemb'ies are hydraulically compatible with the standard fuel assemblies" The NRC approved Revised Thermal Design Procedure (RTDP)"0 F

and the WRB-1 correlation"* are used in the analyses of standard fuel assemblies for normal operation L

and most DNB related accidents. As reported in Reference 17, RTDP and the WRB-2 correlation are used for the VANTAGE 5H fuel. The improved THINC IV modeling scheme'2"' is used. For those

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FS AR* accidents cases where a statistical approach is not applied, a fixed value method is used where plant uncertainties are directly accounted for in the calculations. Should reactor coolant conditions also s

fall outside the range of applicability for the WRB-1 or the WRB-2 correlation, the W-3 DNBR correlation is used, as indicated in the PSE"' for specific accidents.

The following table shows that relationships between the DNBR correlation limit, design limit DNBR, and the safety analysis limit DNBR values used for the Cycle 5 design (when using RTDP, the WRB-1 g m = 10'.1p R ev. 07/15 N3 114Nn

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l MILLSTCME UNrr 3 - CYCLE 5 REDESIGN JULY 1993 Y

and WRB-2 correlations).

17x17 Standard VANTAGESH DNB Correlation WRB-1 WRB-2 Correlation Limit 1.17 1.17 Design Limit (Typ)*

1.25 1.26 (Thm)"

1.24 1.24 Safety Limit (Typ)*

1.27 (N Loop) 1.69 (N Loop)

(Thm)"

1.26 (N Loop) 1.65 (N Loop)

The margin between the design and safety analysis limit DNBRs is more than sufficient to cover the rod how penalties (1.3% or less) applied b both fuel types and the transition core penalty applied to the VANTAGE SH fuel assemblies (12.5 % or less). The transition core penalty is a function of the number of VANTAGE 5H fuel assemblies in the core c.nd is determined using the methodology of Reference 21 which has been approved by the NRC for application to VANTAGE 5 transition cores.

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[

[

[

[

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TYP = Typcal Cell.

THM = Thimble Cell.

L e',ev. 07/15/93 II:49am 13 l.

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MittsToss UNrr 3 - cycle 5 RrocsloN JULY 1993 3

3.0 POWER CAPABILITY AND ACCIDENT EVALUATION r

L 3.1 Power Capability F

h The plant power capability for three and four loop operation for Cycle 5 has been evaluated considering

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the consequences of those incidents examined in the FSARS It is concluded that the core reload will p

b not adversely affect the ability to safely operate at 100 percent of Rated Thermal Power (RTP) for four Icop operation during Cycle 5.

For overpower transients, the fuel centerline temperature limit of l

f 4700 Y can be accommodated with margin in the Cycle 5 core using the methodology described in Reference 1. NRC appreved models "' were used for fuel temperature evaluations. The LOCA limit o

can be met by maintaining F at or below 2.60 for four loop operation at 100% power according to the o

normalized F envelope (shown in Figure 3 and depicted in the COLR). This limit is satisfied by the q

pos er control maneuvers allowed by the Technical Specifications *, which assure that the Final Acceptance Cciteria (FAC) limits are met for a spectrum of Small and Large Break LOCAs.

3.2 Accident Evaluation Tne ett Ms of the reload, including the mechanical design changes described in Section 2.1, on the design aasis and pastulated incidents analyzed in the FSAR* and within Westinghouse's scope were examined.

In all cases, it was found that the effects were accommodated within the conservatism of the initial e

[

assumptions used in the -

sly applicable safety analyses. The changes do not adversely impact the non-LOCA and LOCA sa.cy yses Thus, the conclusions of the FSARS are valid.

~

r L

lt should be noted that for Cycle 5 operation, Westinghouse has neither reanalyzed nor evaluated the event, Startup of an Inactive Reactor Coolant Pump at an incorrect Temperature (FSAR Section 15.4.4).

The event, as currently considered in the FSAR*, has been administratively precluded and is therefore

[

no longer within the Westinghouse scope.

L IFBA fuel was evaluated to determine the potential impact on the FSAR* Large Break and Small Break LOCA analyses. This fuel design Ns been used in other PWR fuel operations since 1987. NRC The non-LOCA and LOCA

• s.' ws are performed in ammiance with the methelogy in WCAP-12610.

I L

R ev. 07/14/93 12:10pm

]4 s

s MILT 5 TONE UNrr 3 - cycle 5 RrDESIGN JULY 1993 approval of the design and Safety Evaluation Report for IFBA fuel with natural boron is included in WCAP-10444-P-A, Addendum 2"3 Reference 12 documents a safety evaluation for the application of IFB A fuel with enriched boron which will be used in the Millstone Unit 3 Cycle 5 core. The evaluation concluded that the enriched boron did not adversely affect core safety considerations. The IFB A fuel rods have been shown to be less limiting than the FSAR* analyses of record for non-IFBA fuel. While the results of Reference 12 have since been called into questien, a plant specific evaluation for Millstone

(

Unit 3 concluded that the analyses are bounding. The limiting fuel type assumption for FSARA LOCA analyses continues to apply and the IFBA fuel rod is bounded by the FSARS analyses of record. Thus operation of Cycle 5 with IFBA fuel meets with the requirements of 10 CFR 50.46.

Large Break LOCA analyses have been traditionally performed using a symmetric, chopped cosine axial pov'er shape.

Recent calculations have shown that there was a potential for top-skewed power i

distributions to result in peak cladding temperatures (PCT) greater than those calculated with a chopped cosine axial power distribution. Westinghouse has developed a process, which was applied to the reload for Millstone Unit 3 Cycle 5, that reasonably ensures that the cosine remains the limiting power distribution, by defining npropriate power distribution surveillance data if required. This process, called the Power Shape Sensitivity Model (PSSM), is described in topical report (WCAP-12909-P) and further clarified in ET-NRC-91-3633, both of which are currently under NRC review.

I In May 1991, Westinghouse transmitted to the NRC this report titled, " Westinghouse ECCS Evaluation q

Model: Revised Large Break LOCA Power L)istribution Methodology" which describes the process that c

l Westinghouse is now utilizing to more accurately account for the effect of power distribution in the core L

reload design. In January 1991, the implementation of this approach was discussed with the NRC. In y

i L

a May 1991 meeting with the NRC, Westinghouse again told the NRC that it planned on implementing the PSSM process shortly after the topical report was submitted. Westinghouse indicated in the b

transmittal letter of the topical (NS-NRC-91-3578), its intent to implement the PSSM process for future reload design applications. As of this time, the NRC staff has not formally approved the use of PSSM mett.odology but is aware of its utilitization. PSSM is currently used in many other operating puts.

Westinghouse has utilized this process in the. Reload Safety Analysis for Millstone Unit 3, Cycle 5.

A core reload can typically affect accident analysis input parameters in the following areas: core kinetic characteristics, control rod worths, and core peaking factors. Cycle 5 parameters in each of these three areas were examined as discussed below to ascertain whether revisions to the accident analyses l

l Rev 07U4/93 12:10pm 15 I

l l

MILLSTONE UNrr 3 - CYCLE 5 REDESIGN JULY 1993 3

assumptions were required.

3.2.1 Kinetics Parameters A comparison of Cycle 5 kinetics parameters with the current limits is presented in Table 2. All kinetics values remain within the bounds of the analysis limits.

3.2.2 Control Rod Worths Changes in control rod worths may affect differential rod worths, shutdown margin, ejected rod worths, and trip reactivity. Table 2 shows that the maximum differential rod worth of two RCCA control banks moving together in their highest worth region for Cycle 5 meets the current limit. Table 3 shows that the Cycle 5 shutdown margin requirements are satisfied.

3.2.3 Core Peaking Factors Peaking factors for the dropped RCCA incidents we.e evaluated based on the NRC approved dropped rod methodology *' Results for N loop operation show that the DNB design basis is met for ali dropped rod events initiated from full power. Peaking factors following control rod ejection are within the bounds of the analysis limits documented in the PSEA The peaking factors for the misalige rod and the reanalyzed steaniline break have been evaluated and the minimum DNBR safety analysb acceptance criteria are satisfied"'

s I

w a

s Rev. 07/14/93 12:10pm

]6 s

[

J MrusroNE UNrr 3 - CYCLE 5 REDESIGN

.IULY 1993 Y

b 4.0 TECIINICAL SPECIFICATION CIIANGES A review of the Millstone Unit 3, Cycle 5 Reload Safety Ev:Juation fRSE) has been performed relative to the impact of thi.c ME on the Millstone Unit 3 Technical Specifications"' and the Core Operating Limits Report (COLR) which will be updated for Cycle 5. This review was performed on the Millstone Unit 3 Technical Specification

• inclusive of Amendment 75 which was effective upon commencement of the design process.

As a result of this review, no fuel related Technical Specification changes for Millstone' Unit 3 are required for Cycle 5 operation, except thy given in Reference 7* and those specified below and in

(

Appendix A.

{

An evaluation of the adequacy of the boration volume in Technical Specifications 3.1.2.5 and 3.1.2.6 has been performed. The Mode 1 - 4 Technical Specification requires that enough boric acid be contained

{

in the Boric Acid Storage Tank and RWST to borate the plant to cold shutdown (Mode 5). His volume was found to be 21,802 gallons for the BAT, which is above the current Technical Specification limit of 21,020 gallons. Therefore, the current Technical Speci&ation must be revised. This calculation was also performed with the RWST as the boration source, and no changes to the RWST volume are required.

To define :he mode at which the current borated volume would be acceptable, evaluations were performed with shutdown initiating from Modes 2 and 3. The evaluation with shutdown'beginning in M6de 2 generated a minimum boration volume of 21,367-galloas which is greater than the current Technical Specification requirement. He required volume to borate from Mode 3 to Mode 4 was calculated to be c

18,068 gallons. This volume is below the current minimum Technical Specification borated volume and thus validates the current Technical Specification required volume for entrance into Mode-3.

The minimum required volume of 21,802 is necessary for entrance into' Mode 1 and 2.

L The Millstone Unit 3 Technical Specifications

• ensure that the pirnt operates in a manner that provides acceptable levels of protection for the health and safety of the public. The Technical Specifications

• are based upon assumptions made in the safety and accident analyses, including those relating to the core w

This RSE supports the use of ZIRLO' clad".ng as documented in *teference 7. NRC approval of Reference 7 is expected

.n mid-September.

Ret 07/2m93 3 05pm 17

L MILLSTONE UNrr 3 - CYCLE 5 REDESIGN JULY 1993 3

r L

redesign. This ensures adequate margin to the regulated acceptance criteria for the accident analyses.

Since it has been concluded that the core design parameters and assumptions utilized in the accident analyses remain appropriate, the conclusions in the Millstone Unit 3 FSAR* are valid. Therefore, the regulated margin of safety as defined in the Bases of the Technical SpecificationP is not affected by Cycle 5 operations.

E

[

[

L L

F E

Rev. 07,22/93 12 42prn 18

1 L

Ml MILL 5T NE UNrr 3 - CYCLE 5 REDESIGN JULY 1993 Y

f

5.0 REFERENCES

L 1.

Davidson.

S.

L., et al.,

" Westinghouse Reload Safety Evaluation Methodology,"

WCAP-9273-NP-A, July 1985.

2.

Final Safety Analysis Report Millstone Generating Station, Unit 3, USNRC Docket No. 50423, as amended through October 1992.

[

3.

Letter from D. W. Perone (Westinghouse) to J. A. Camp (NEU), " Plant Safety Evaluation for Millstone Generating Station Unit 3, VANTAGE 5H Fuel Upgrade,"

90NE*-G-0075, August 29,1990.

4.

Letter from D. L. Fuller (Westinghouse) to J. A. Camp (NEU), " Transmittal of Technical Specif; cations," NEU-90-596, August 29,1990.

5.

Technical Specifications Millstone Generating Station, Unit 3, USNRC Docket No. 50423 as amended through Amendment 75.

6.

Letter from D. L. Fuller (Westinghouse) to J. A. Camp (NEU), " Northeast Utilities Service Company, Millstone Unit 3, Boron Dilution Analysis," NEU-90-627, November 5,1990.

7.

Letter from J. F. Opeka (NNECO) to Document Control Desk (NRC), " Millstone Nuclear Power Station Unit No. 3, Proposed Revision to Technical Specifications Design Features - Fuel Assemblies," Docket No. 50-423, B14378, March 15,1993.

8.

Davidson. S. L. (Ed.), et al., " Reference Core Report VANTAGE 5 Fuel Assembly,"

WCAP-10444-P-A (Proprietary), September 1985.

9.

Davidson, S. L. and Nuhfer, D. L. (Eds.), " VANTAGE + Fuel Assembly Reference Core Report," WCAP-12610, June 1990.

g-Rev. 07/22/93 12:42pm

]9

(

MILLSTONE UNrr 3 - CYCLE 5 REDESIGN JULY 1993 Y

10.

Davidson, S. L. (Ed.), et al., " Extended Burnup Evaluation of Westinghouse Fuel,"

WCAP-10125-P-A, December 1985.

I1.

Weiner, R. A., et al., " Improved Fuel Performance Models for Westinghouse Fuel Rod Design

[

and Safety Evaluations," WCAP-10851-P-A, August 1988.

12.

Letter from W. J. Johnson (Westinghouse) to M. W. Hodges (NRC), " Application of Enriched Baron in the Westinghouse Integral Fuel Burnable Absorber Design," NS-NRC-89-3454, e

[

September 6.1989.

[

13.

Slagle, W. H., " Operational Experience with Westinghouse Cores (through December 31, L

1991)," WCAP-8183, Revision 20, February 1993.

14.

Nguyen, T. Q., et al., " Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores " WCAP-11597-A, June 1988.

15.

Liu, Y.

S.,

et al.,

" ANC:

Westinghouse Advanced Nodal Computer Code,"

WCAP-10966-NP-A, September 1986.

r l

L 16.

Miller, R. W., et al., " Relaxation of Constant Axial Offset Control, F Surveillance Technical q

Specification," WCAP-10216-P-A, June 1983.

l L

17.

Davidson, S.

L.

(Ed.), " VANTAGE 5 Fuel Assembly Reference Core Report,"

WCAP-10444-P-A, September 1985; Addendum 2-A, " VANTAGE 5H Fuel Assembly,"

April 19h8.

18.

Friedland, A. J. and Ray, S., " Revised Thermal Design Procedure," WCAP-l l397-P-A, April 1989.

(

19.

Motley, F. E., et al., "New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles With Mixing Vane Grids," WCAP-8762-P-A, July 1984.

~

a Rev 07/22/93 12 42pm 20 E

L MILLSTONE UNrr 3 - cycle 5 REDESIGN ll!LY 1993 3

20.

Friedland, A. J. and Ray, S.,

" Improved THINC IV Modeling for PWR Design,"

WCAP-12330-P, September 1989.

21.

Schueren, P, and McAtee, K. R., " Extension of Methodology for Calculating Transition Core

(

DNBR Penalties," WCAP-ll837-P-A, January 1990.

(

22.

Leech, W. J., et al., " Revised PAD Thermal Safety Model," WCAP-8720 Addendum 2 (Proprietary), October 1982.

23.

Letter from C. O. Thomas (NRC) to E. P. Rahe, Jr. (Westinghouse), " Acceptance for Referencing of Licensing Topical Report WCAP-8720, Addendum 2, ' Revised PAD Code Thermal Safety Model*," December 9,1983.

24.

Morita, T., et al., " Dropped Rod Methodology for Negative Flux Rate Trip Plants,"

WCAP-10298-A, March 1983.

l

[

l

% 07/22/93 12.42pm 2]

)

l

_____________.________________________________________J

!L 4

~

[

j MiustoNE UNrr 3 - cycle 5 REDESIGN JULY 1993 Y

r L

Table 1 Fuel Assembly Design Parameters Millstone Unit 3 - Cycle 5 r

L Region 5B 6A 6B 7

7' Enrichment (w/o U-235)"

4.506"*

4.191*"*

4.496*"*

4.401"*"

4.401~~~

[

Geometric Density" 95.07 95.66 95.47 95.18 95.18

(% theoretical)

Number of Assemblies 21 32 56 36 48 Approximate Burnup at 35,500 25,800 21,900 0

0

~

Beginning of Cycle 5 (M WD/MTU)""**

The c numetnblies twve been re-caged an dc.cribed in Section 2.1'of this RSE.

F

~

All values are un-built.

Contains Natural UO; (0.728 w/o U-235) in six inch axial blankets at each end of fuel pellet stack.

l

]

"" Contains Natural UO; (0.717 w/o U-235) in six inch axial blankets at each end of fuel pellet stack.

• ~~

Contaunr. Natural UO, (0 7$$ w/o U-235) m six inch annalar axial blankets at a.ach end of fuel pellet stack.

~~~* Based on an EOC4 bumup of 20.330 MWD /MTU.

f

(

Rev. 07/22/93 12.42pm 22 m._

- - - - - - - - - - - - - - ^

4 I

MILLSTONE UNrr 3 - CYCLE 5 REDESIGN JULY 1993 Y8, u-r L

Table 2 Kinetics Characteristics N Loop Operation y

L Millstone Unit 3 - Cycle 5 r

L Reference Cycle 5 Changes to Analysis LimitN" Current Limits Most Positive Moderator

+ 5 <; 70% RTP f

Temperature Coefficient

+ 5 to O Ramp from L

(pcm/ F)*

70% to 100% RTP Most Negative Unrodded

-56.5 Moderator Temperature Coefficient (pcm/ F)*

Doppler Only Temperature

-3.2 to -0.91 Coefficient (pcm! F)*

Least Negative Doppler - Only

-9.55 to -6.05 Power Coefficient, Zero to HFP (pcm/ F)*

Least Negative Doppler - Only

-900 (4 loop)

Power Defect at BOC (pcm)*

-690 (3 loop)

F Most Negative Doppler - Only

-19.4 to -12.6 L

Power Coefficient. Zero to HFP (pcm/ F)*

Delayed Neutron Fraction 0.40 to 0.70 An.(%)

Maximum Differential Rod Worth 146.67 of Two Banks Moving Together at HZP (pcm/in)*

r pcm = 10 do.

Rev. 07/22/93 12:42pm 23

a l

[

l MrLLsToxE Unrr 3 - CYCLE 5 REDESIGN JULY 1993 3

l Table 3 i

End-of-Cycle Shutdown Requirements and Margins N Loop Operation Millstone Unit 3 - Cycle 5 4 Loop (N)

Cvele 5 Control Rod Worth (Woo)

All Rods Inserted Less Most 6.70

)

[

Reactive Stuck Rod

(

(1) Less 10%

6.03

(

Control Rod Reauirements (% An)

Reactivity Defects (Combined Doppler, 3.37 Moderator Temperature, Void, Redistribution, and Rod Repositioning Allowance)

Rod Insertion Allowance 0.40 u

7 (2) Total Requirements 3.77 L

Shutdown Marcin f(l)42)1 (% Ap) 2.26 7

L Rr_ quired Shutdown Margin (% Ac) 1.30 Rey, 07/22/93 12:42pm 24

MILLSTONE UNrr 3 - CYCLE 5 REDESIGN JULY 1993 3

f L

Figure 1 Core Loading Pattern Millstone Unit 3 - Cycle 5 F

P N

M L.

K.

J.

H.

G, P,

E, D

C B

A SB 7

7 5B 7

7 5B E59 G03 GOG E39 GIO G16 E7 0

$B 7

7 7

SB 6A 63 7

7 7

SB 2

E72 G22 G31 G34 F60 F21 F59 G36 G42 G24 E56 5B 7

6B 6B 7

6A 6B 6A 7

6B 6B 7

5B E62 Ol' F51 F'S G4' Fii F47 F05 G59 F7 8 F67 G02 E"1 7

6B 6B 7

6A 7

6B 7

6A 7

6B 6B 7

G25 F62 F38 C62 F17 072 F60 G73 F30 G46 F43 F63 G26 SE 7

6B 7

6A 7

6B 7

6B 7

6A 7

6B 7

5B

~

E6C G43 FSB GS1 Fil G52 FS2 G67 F71 G68 F01 G64 F76 G30 E55 7

7 7

6A 7

6B t' B 6A 6B 6B 7

6A 7

7 7

~

G22 G44

';6 5 F24 G58 F77 F45 F14 F39 F87 G61 F13 G45 G32 C04 7

6B 6A 7

6B 6B 6A 7

6A 6B 6B 7

6A 6B 7

-7 013 F'2 FIS G71 F70 F41 F03 G57 F12 F44 F69 G54 F09 F64 G14 5B 6A 6B 6B 7

6A 7

5B 7

6A 7

6B 6B 6A SB 90

- 8 E43 F08 F46 FB1 G78 F20 G76 E40 G70 F19 G84 F74 F37 F22 E35 7

6B 6A 7

6B 6B 6A 7

6A 6B 6B 7

6A 6B 7

~ '

Gil F56 F27 G69 F68 F48 F29 G66 F28 F34 F50 G75 F32 r58 G12 7

7 7

6A 7

6B 6B 6A 6B 6B 7

6A 7

7 7

~

G07 G39 G53 F25 G80 F82 F40 F10 F35 F86 GS1 F04 G60 G40 G09 5B 7

6B 7

6A 7

6B 7

6B 7

6A 7

6B 7

SB

~

E65 G3' F83 G74 F07 G63 F53 G77 F54 G79 F23 G83 F84 G38 E69 7

6B 6B 7

6A 7

6B 7

6A 7

68 6B 7

G27 F49 F36 G48 F02 G82 F79 G49 F18 G55 F33 F66 C21 5B 7

6B 6B 7

6A 6B 6A 7

6B 6B 7

SB 13 E69 G01 F65 T'3 G50 F26 F42 F06 G56 F85 F61 GIS E76 5B 7

7 7

6B 6A 6B 7

7 7

5B 14

~

E58 G28 G33 G35 F57 F16 F55 G29 G41 G23 E73 5B 7

7 5B 7

7 5B 15 E66 G06 GOS E34 G19 C18 E64 00 LEJ E!m R

R REGIG!J IDE!JTIFIER y

ID ID FUEL ASSEMBLY IOE!JTIFICATIG!J R ev. 07/21'93 12-42pm 25 L

MILLSTONE UNrr 3 - CYCLE 5 REDESIGN JULY 1993

[

Figure 2 Burnable Absorber and Source Rod Location Millstone Unit 3 - Cycle 5 F

F M

M L

M.

J.

H.

C,.

F, E,

D C

B A

e DCPD DCFD RCCA DCFD CCFD DCPD DCPD RCCA DCFC RCCA RCCA DCPD RCCA DCFD DC PD RCCA DCPD DCFD

~

46I 64I 64I 64I 64I 4SI DCFD CCFD DCFD DCFD DCFD RCCA RCCA 4SSA RCCA RCCA DCFD DCFD DCPD 801 90I R CA DCPD DCPD DCFD DCPD ROCA DCPD RCCA DCPD RCCA DCFD RCCA DCFD 4

48I 80I BOI GOI 80I 48I D PD DCFD DCPD DCPD DCPD DCPD DCPD DCFD RCCA RCCA DCPD DCPD RCCA RCCA DCFD

- 5 64I BOI 90I 80I BOI BOI 64I FCCA DCFD DCPD DCPD DCFD RCCA DCFC DCFD RCCA DCPD RCCA DCPD RCCA DCPD DCPD

- 6 64I 80I SCI 80I 80I 64I DCFD DCFD DCPD DCFD DCPD RCCA DCPD DCPD DCPD DCPD DCPD DCPD RCCA DCFD RCCA

- i SOI 60I 80I DCPD DCPD DCPD DCPD 900 DCFD FCCA DCPD ROCA RCCA RCCA RCCA RCCA DCFD RCCA DCPD

- S BOI BOI 80I BOI DCFD DCPD DCPD ROCA DCFD RCCA DCFD DCPD DCFD DCPD DCPD DCPD ROCA DCPD DCFD

- 9 80I BOI BOI RCCA DCPD DCPD DCPD DCFD RCCA J

D PD DCPD RCCA DCPD RCCA DCPD RCCA DCPD DCPD

- 10

{

64I BOI SOI 80I SCI b4I OCFD DCFD DCPD DCPD DCPD DCPD DCPD DCFD RCCA RCCA DCPD DCPD RCCA RCCA DCPD

- 11 64I SGI BOI BCI BOI BOI 64I RCCA DCPD DCPD DCPD DCPD ROCA DCFD RCCA DCPD RCCA DCPD RCCA DCPD 10 4EI BOI 80I 801 BCI 4EI DCPD DCFD DCFD DCFD DCPD RCCA RCCA 4SSA RCCA RCCA DCPD DCPD DCPD 13 80I 80I RCCA DCPD RCCA RCCA DCFD RCCA DCPD DCFD RCCA DCPD DCPD 14 1

1 DCPD DCPD DCPD DCFD RCCA DCFD DCPD 15 j

LEC.END 00 TYFE TYPE : CCMPCNEITC TYPE esel esel h"MER OF FRESH IFF3A RCDS mt :- m sEs-vrr; e

LA..

.M A !. s.t FLvM N is JE F.? TA

-?;WTRDL OF J rJ*NWN ROCA

• /JA tHB E 8 ff 900LE?5 CN JECON!ARY SOURCE A33EMBLY R ev. 07/22S3 12 A2pm 26

Q.

9'.

]

. MrusrONE UNrr 3 - CYCLE 5 REDEslGN JULY 1993 -

3 L.

Figure 3 i

K(Z) - Normalized F (Z) as a Function of Core Height q

L-For Four Loop Operation Millstone Unit 3 - Cycle 5 5

L' r;

1.2 1

L J,

1.1 -

1 :

C

[:

0.9 -

vj o.s -

<a'k

(

O.7 -

1

' O.6 -

EN.I ELEVATig o.5 -

g I0 e.orr N

O.4 -

3.0 6.0fi

[,

E e

0.925

12. ort z

O.3 -

h 0.2 -

0.1 -

o.

0 2

4 6

8 10 12 Corn E3evotion f

[

Rev. 07/22/93 12:42pm

'27

t

[

l MILLSTONE UNrr 3 - CYCLE 5 REDESIGN JULY 1993 f

1 i

f I

L Appendix A TECIINICAL SPECIFICATION MARKUPS b

r

[

I L

[

[

[

[

[

[

[

x.am,> i m,.

x.,

s March 11, 1991 REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCFS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum the following barated water source (s) shall be OPERABLE as required by Specification 3.1.2.2:

a.

A Boric Acid Storage System with:

2.t s@

b 1)

A minimum barated water usable volume of-ikhNhile gallons, 2)

A boron concentration between 6300 and 7175 ppm, and 3)

A minimum solution temperature of 67'F.

b.

The refueling water storage tank (RWST) with:

1)

A minimum contained borated water volume of 1,166,000

gallons, 2)

A baron concentration b'etween 2700 and 2900 ppm, 3)

A minimum solution temperature of 40'F, and 4)

A maximum solution temperature of 50*F.

ApptICABitITY: H0 DES 1, 2, 3, and 4.

{

ACTION:

(

a.

With the Boric Acid Storage System inoperable, restore the system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STAND 8Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOWN MARGIN equivalent to at least the limits as shown in Figure 3.1-4 at 200*F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next

{

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

[

i MILLSTONE - 1) NIT 3 3/4 1-18 Amendment No. 60 ocer

.\\

e March 11, 1991

[

EfACTIVITY CONTROL SYSTD4S

_B A_S ES BORATION SYSTEMS (Continued)

[

HARGIN from expected operating conditions of equivalent to that reoui red __bv Figure 3.1-5 after xenon dec_ay and cocidown to 200*VThe maxiinum expecte (boration capability requirement occurs at un.1 rom full power equilibrium e

[

Lxenon condi_tions_ and _ requires a usable volume of 21,020 callons of 6300 pp.

barated water f rom the Doric acid ~ storage tanks or 1,166,000 gallons of 2700 ppm borated water from the refueling water storage tank (RWST).

M l

ninimum RWST volume of 1,166,000 gallons is specified to be consistent with t

[

ECCS requirement.

'\\

With the RCS temperature below 200*F, one Boron Injection System is f

acceptable without single failure consideration on the basis of the stable L

reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Baron Injection System becomes inoperable.

The limitation fbr a maximum of one centrifugal charging pump to be OPER-ABLE and the Surveillance Requirement to verify all charging pumps except the j

[-

zass addition pressure transient can be relieved by the operation of a single required OPERABLE pump to be inoperable below 350*F provides assurance that a j

l PORV.

I I

The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 1.3% Ak/k after xenon decay and cooldown from 200*F to 140eF. This condition requires either a usable volume of 4100 gallons of r

6300 ppm borated water from the boric acid storage tanks or 250,000 gallons of 2700 ppm borated water from the RWST. The unusable volume in each boric acid storage tank is 1300 gallons.

l The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

L The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.0 and 7.5 for the solution recirculated uithin containment after a LOCA.

This pH band minimizes the evolution of c

i iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

[

The minimum RWST solution temperature for MODES 5 and 6 is based on analysis assumptions in addition to freeze protection considerations.

The ainimum/ maximum RWST solution tegeratures for MODES 1, 2, 3 and 4 are based on ana1ysis assumptions.

l The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6.

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MILLSTONE - UNIT 3 3 3/4 1-3 Amendment No. J2,60 0027

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INSERT A 1

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The maximum boration capability (minimum boration volume) requirement is established to conservatively bound expected operating conditions throughout core operating life. The initial RCS boron

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concentration is based on a minimum expected hot full power or hot zero power condition (peak xenon).

The final RCS boron concentration assumes that the most reactive control rod is not inserted into the

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core. This set of conditions requires a minimum usable volume of 21,802 gallons of 6300 ppm

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