ML20238F247

Positive Moderator Temp Coefficient Licensing Rept for Millstone Unit 3

ML20238F247
Person / Time
Site:	Millstone
Issue date:	08/31/1987
From:	Rosenthal P, Sterdis R, Wengerd M WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20238E926	List: B12661, Proposed Tech Specs,Allowing Positive Moderator Temp Coefficient at Reactor Power Levels Less than 100% to Support Cycle 2 Reload ML20238E923 ML20238F235 ML20238F247
References
NUDOCS 8709160043
Download: ML20238F247 (179)

Text

{{#Wiki_filter:k 4, :n c 7 s B, r;.;G T'j_',, ' i t: ' 1, o a / o: .w;- - U - ' POSITIVE MODERATOR TEMPERATURE COEFFICIENT ~ LICENSING REPORT.FOR-MILLSTONE' UNIT 3 P. Rosenthal M. Wengerd R. Sterdis lm' August 1987 WE'TINGHOUSE ELECTRIC CORPORATION S P. O. Box 355' Pittsburgh, Pennsylvania 15230-9thhoC643h00 3 DR P 0657v;1D/081187 1

p i jf -~ -A L Fj L. ~ .c .t p [ 'p TABLE' 0F CONTENTS q v,-. ~t t Section-Pace List of Tables iii 5 1 ..i List.of Figures iv 1.0 Introduction 1-1 2.0 Transients Evaluated for a Positive Moder,ator Temperature 2-1 Coefficient, 2.1 RCCA Misa",ignment i 2-1 2.2 Startup' ct! an Inactive Reactor' Coolant Pump 2-1. [ 2.3 :Exc'essive geet Removal Due to Feedwater System Me.1 functions 2-2 2.4 Excess've Increase.in Secondary Steam Ficw 2-2 245 Main Steam 1(ge Depressurization/ Steam System Piping Failure 2-3 3.'S FeedwaterSyst5mhipeBreak 2-3 2.7 Inadvertent Oper.ntion of the ECCS at Power 2-4 2.8 Loss of Coolant Accidents 2-4 n 2.9 Steam Generator sube Rupture ^ 2-7 3.0 Transients Analyzed for a Positive Moderator Temperature 3-1 Coefficient f 3.1'UncontrolledRCCABanFWithdrawalfromaSuberItical 3-1 Condition i ' t-r 3.2 Uncontrolled Bank Withdrawal at Power 1-3 3.3 Loss of Forced Reactor Coolant Pump Flow ?-4 3.4 Reactor Coolar,', Pump Shaf t Seizure 3-6 ) ,3.5 Loss of External Load / Turbine Trip 3-8 3.6 RCOA Ejection 3-11 j (,- 'f f 3.7 Loss of Normal Feedwater,3 3-12 g ( ) 3.8. Inadvertent Opening of a pressurizer Safety or Relief Yelve 3 '.4 .f. L 1 l

e 0657v

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c -, - - - - - - - - - - - - - - - - -

m 9, I 4 >,.I TABLE OF CONTENTS (Continued) .a. Section . gge 1 g4 L'- . 4.0' RWST/ Accumulator Boron Concentration Increase Evaluation 4-1 4.1 Introduction 4-1 4.2-Non-LOCA Analyses' 4-1 r 4.3 LOCA' Analyses 4-6 4.4 LOCA Related Design Consideration 4-11 4.5 Summary and Conclusion 4-15 1 5.0 Control' System' Evaluation 5-1 1 6.0 Technical Specification' Changes 6-1 -.7.0 FSAR Changes' 7-1 8.0 References 8-1 9.0 Conclusions' 9-1 ' Appendix A - Technical Specification Markups { i l} t i i 4 ) l 0657v;1D/081187 ii J J u.

I h ' LIST OF TABLES j Table Title Pace

1.0-1 FSAR Accidents Evaluated for Positive Moderator 1-5 '-

. Temperature Coefficient Effects -a 3.1-1 Time sequence of Events for an Uncontrolled RCCA Bank 3-16 Withdrawal from a Suberitical Condition '3.2-1 Time Sequence of Events.for a RCCA Bank Withdrawal at Power-3-17 3,3-1 Time Sequence of Events' for Incidents which Result in a 3-19 Decrease in Reactor Coolant System Flow 3-20 3.4-1: . Summary of Results for a Locked Rotor Transient 3-21 3.5-1 . Time sequence of Events for a Turbine Trip 3.6-1 Time Sequence of Events for~an RCCA Ejection Accident 3-23 3.6-2 Parameters Used in the Analysis of the RCCA Ejection Accident 3-24 3-25. 3.7-1 Time Sequence.'of Events for a Loss of Feedwater 3-26 3.7-2 Time Sequence'of Events for a Loss of Feedwater Without Offsite Power Available 3.8-1 Time Sequence of Events for an inadvertent Opening 3-27 of a Pressurizer Safety or Relief Valve iii 0657v.1D/DB1187 G

i LIST OF FIGURES 4 Figure Title Page

1. 0-l '

Moderator Temperature Coefficient Versus Rated Thermal Power 1-4 i 3.1-1 Neutron Flux Transient for an Uncontrolled Rod Withdrawal from 3-28 a Suberitical Condition 3.1-2 Thermal Flux Transient for an Uncontrolled Rod Withdrawal 3-29 from a Suberitical Condition 3.1-3 Fuel and Clad Temperature Transients for an Uncontrolled Rod 3-30 Withdrawal from a Suberitical Condition 3.2-1 Nuclear Power, Core Heat Flux, and Core Average 3-31 Temperature for a RCCA Bank Withdrawal at Full Power withMinimumReactivityFeedback(70PCM/SECRate) 3.2-2 Pressurizer' Pressure, Water Volume,'and DNBR for a RCCA 3-32 Bank Withdrawal at Full Power with Minimum Reactivity feedback (70 PCM/SEC Rate) 3.2-3 Core Average Temperature, Heat Flux, and Nuclear Power 3-33 for a RCCA Bank Withdrawal at Full Power with Minimum Reactivity feedback (3 PCM/SEC Rate) 3.2-4 Pressurizer Pressure, Water Volume, and DNBR for a 3-34 RCCA Bank Withdrawal at Full Power with Minimum Reactivity Feedback (3 PCM/SEC Rate) 3.2-5 Nuclear Power, Heat Flux and Core Average Temperature 3-35 for a RCCA Bank Withdrawal at Full Power with Maximum Reactivity feedback (70 PCM/SEC Rate) f iV 0657v.1D/081187

g. LIST OF FIGURES (Continued) Title Pace Figure' '3.2-6 Pressurizer. Pressure, Water Volume,.and DNBR for a 3-36 RCCA Bank Withdrawal at Full-Power with Maximum ReactivityFeedback(70PCM/SECRate) 3.2-7 Nuclear Power, Heat Flux and Core Average Temperature 3-37 for a RCCA Bank Withdrawal at Full Power with Maximum Reactivity Feedback-(3.PCM/SEC Rate) 3.2-8' Pressurizer Pressure, Water Volume and DNBR for a RCCA 3-38 Bank Withdrawal at Full Power with Maximum Reactivity Feedback (3 PCM/SEC Rate) 3.2-9 Nuclear Power, Core Heat Flux, and-Core Average 3-39 Temperature for a' RCCA Bank Withdrawal at full Power with Minimum Reactivity Feedback (70 PCM/SEC Rate). (N-1LoopOperation) 3.2-10 Pressurizer Pressure, Water Volume, and DNBR for a RCCA 3-40 Bank Withdrawal at Full Power with Minimum Reactivity feedback (70 PCM/SEC Rate) (N-1 Loop Operation) 3.2-11 Core Average Temperature, Heat Flux, and Nuclear Power 3-41 for a RCCA Bank Withdrawal at Full Power with Minimum Reactivity feedback (3 PCM/SEC Rate) (N-1 Loop Operation) 3.2-12 Pressurizer Pressure, Water Volume, and DNBR for a 3-42 RCCA Bank Withdrawal at Full Power with Minimum Reactivity Feedback (3 PCM/SEC Rate) (N-1 Loop Operation) 06s7v;1o/081187 V L

p 6 i I LIST OF FIGU'RES (Continued). FigureJ Title Page-l 3.2-13~ ! Nuclear' Power, Heat. Flux and Core Average. Temperature 3 i for a RCCA Bank, Withdrawal at Full Power with Maximum Reacthityfeedback(70PCM/SECRate) -(N-1 Loop Operation) . 3'. 2-14. Pressurizer Pressure, Water. Volume, and DNBR,for a '3-44 RCCA Bank Withdrawal at Full Power with Maximum LReactivity feedback (70 PCM/SEC. Rate) '(N-1 Loop Operation) 3.2-15 . Nuclear Power,: Heat Flux and Core Average Temperature 3-45. for a RCCA Bank Withdrawal at Full Power with Maximum Reactivity Feedback (3 PCM/SEC Rate) j (N-1 Loop 0peration) 3.2-16 Pressurizer Pressure, Water Volume and DNSR for a RCCA 3-46 Bank Withdrawal at Full Power with Maximum Reactivity ' Feedback (3 PCM/SEC Rate) (N-1 Loop Operation)' 3.2-17 Minimum DNBR vs. Reactivity Insertion Rate for a RCCA 3-47 Bank Withdrawal at Full-Power 3.2-18 Minimum DNBR vs. Reactivity Insertion Rate for a RCCA 3-48 Bank Withdrawal at 60% Power 3.2 Minimum DNBR vs. Reactivity Insertion Rate for a RCCA 3-49 Bank Withdrawal at 10% Power / [ 3.2-20 Minimum DNBR vs. Reactivity Insertion Rate for a RCCA 3-50 Bank Withdrawal From 75% Power (N-1 Loop Operation) 0657v:1D/081187 vi

fq y !? y LIST OF FIGURES (Continued) Figure' Title Page 3.2-21 Minimum DNBR vs.. Reactivity Insertion Rate for a RCCA 3-51 Bank Withdrawal From 10% Power (N-1 Loop Operation) 3.3-1. Flow Transients for Four Loops in Operation, One Pump 3-52 Coasting Down 3.3-2 . Nuclear Power and Pressurizer Pressurizer Transients for Four 53 Loops in Operation, One Pump Coasting Down 3.3 Average and Hot Channel Heat Flux. Transients for Four Loops 3-54 in Operation, One Pump Coasting Down 3.3-4 DNBR versus Time for Four Loops in Operation, One Pump 3-55 Coasting'Down 3.3-5 Core Flow Coastdown for Four Loops in Operation, Four Pumps 3-56 Coasting Down 3.3-6 Nuclear Power and Pressurizer Pressure Transients for Four 3-57. Loops in Operation, Four Pumps Coasting Jown 3.3-7 Average and Hot Channel Heat Flux Trar.aients for Four Loops 3-58 in Operation, Four Pumps Coasting Down t - 3.3-8 DNBR versus Time for Four Loops in Operation, Four Pumps 3-59 Coasting Down i 3.3-9 Flow Transients for Three Loops in Operation, One Pump 3-60 Coasting Down (N-1 Loop Operation) 06s7v;1D/081187 Vii

-LIST OF FIGURES (Continued)' I Figure-Title Page s 3.3-10 Nuclear Power and Pressurizer Pressure Transients for Three 3-61 Loops in Operation, One Pump Coasting Down { (N-1 Loop Operation) 3.3 Average and Hot Channel Heat Flux Transients for Three Loops . 3-62 in. Operation, One Pump Coasting.Down -(N-1 Loop Operation) - 3.3-12 DNBR.versus Time for Three Lecos in Operation, One' Pump 3-63' Coasting Down '(N-1 Loop Operation) -3.3-13 ' Core Flow Coastdown for'Three Loops in Operation, Three 3-64 Pumps Coasting Down (N-1 Loop Operation). 3.3-14 Nuclear Power and Pressurizer Pressure Transients for Three 3-65 Loops in Operation, Three Pumps Coasting Down (N-1 Loop Operation) d 3.3-15 Average and Hot Channel Heat Flux Transients for Three Loops 3-F4 in Operation, Three Pumps Coasting Down (N-1 Loop Operation) 3.3-16 DNBR versus Time for Three Loops in Operation, Three Pumps 3-67 Coasting Down (N-1 Loop Operation) 3.4-1 Flow Transients for Four Loops in Operation, One Locked Rotor 3-68 3.4-2 Peak Reactor Coolant Temperature for Four Loops in Operation, 3-69 l One Locked Rotor 06s7v:1D/o81187 Viii l'

1 ? LIST OF FIGURES (Continued) ' Figure Title Page i-t- L 3.4-3 Nuclear' Power, Avg Channel'and Hot Channel Heat Flux 3-70 Transients for Four Loop in Operation, One Locked Rotor i 3.4-4 Maximum Clad Temperature At Hot Spot for Four Loops 3-71 in Operation, One Locked Rotor 1 - 3.4-5 Flow Transients for Three Loops in Operation, One 3-72 Locked Rotor 1 3.4-6 Peak Reactor Coolant. Pressure for Three Loops in 3-73 Operation, One Locked Rotor 3.4-7 Nuclear Power, Avg Channel and Hot' Channel Heat Flux 3 ~ Transients for Three Loops in Operation, One Locked Rotor 3.4-8 Maximum. Clad Temperature At Hot Spot for Three Loops 3-75 in Operation, One Locked Rotor 3.5-1 Pressurizer Pressure, Water Volume and Nuclear Power 3-76 for Turbine Trip With Pressure Control and Minimum Reactivity feedback i 3.5-2 Core Inlet Temperature, Coolant Average Temperature 3-77 I and DNBR for Turbine Trip With Pressure Control and Minimum Reactivity feedback 3.5-3 Pressurizer Pressure, Water Volume and Nuclear Power 3-78 for Turbine Trip With Pressure Control and Maximum Reactivity Feedback 0657v;1o/081187 iX

LIST OF FIGURES (Continued) 1

Figure, Title Page 3.5-4 Core Inlet Temperature. Coolant Average Temperature 3-79 l

and DNBR for Turbine Trip With Pressure Control and Maximum. Reactivity Feedback.- 3.5-5 Pressurizer Pressure, Water Volume and Nuclear Power 3-80 'for Turbine Trip Without' Pressure Control and Minimum Reactivity Feedback 3.5-6 -Core' Inlet Temperature, Coolant Average Temperature 3-81 and DNBR for Turbine Trip Without Pressure Control and Minimum Reactivity Feedback 3.5-7 . Pressurizer Pressure, Water Volume and Nuclear Power 3 for' Turbine Trip Without Pressure Control and Maximum' Reactivity Feedback -3.5-8 ' Core Inlet Temperature, Coolant Average Temperature 3-83 and DNBR for Turbine Trip Without Pressure Control and t Maximum Reactivity Feedback 3.5-9 Pressurizer Pressure, Water Volume and Nuclear Power 3-84 for Turbine Trip With Pressure Control and Minimum Reactivity Feedback (N-1 Loop Operation) 3.5-10 Core Inlet Temperature, Coolant Average Temperature 3-85 7 and DNBR for Turbine Trip With Pressure Control and Minimum Reactivity Feedback (N-1 Loop Operation) I- }.. 06s7v.1D/CB1187 x I __________a

' LIST OF FIGURES (Continued) Figure. Title' Page 3.5-11' Pressurizer Pressure, Water Volume and Nuclear Power 3-86 .for Turbine Trip With Pressure Control and Maximum i Reactivity' Feedback L(N-1 Loop Operation) 3.5-12 Core Inlet Temperature, Coolant Average Temperature 3 and DNBR for Turbine Trip With Pressure Control and Maximum Reactivity Feedback. (N-1 Loop Operation) 3.5-13~ Pressurizer Pressure, Water Volume and Nuclear Power 3-88.- for Turbine Trip Without Pressure Control and Minimum Reactivity Feedback' (N-1-LoopOperation) 3.5-14 Core Inlet Temperature, Coolant Average Temperature- .3-89 and DNBR for. Turbine Trip Without Pressure Control and Minimum Reacthity Feedback (N-1 Loop Operation) l 3.5-15 Pressurizer Pressure, Water Volume and Nuclear Power 3-90 for Turbine. Trip Without Pressure Control and Maximum Reactivity Feedback' (N-1 Loop Operation) 3.5-16 Core Inlet Temperature, Coolant Average Temperature 3-91 and DNBR for Turbine Trip Without Pressure Control and Maximum Reactivity Feedback i r. [. (N-1 Loco Operation) \\ I s i 0657v.1D/081187 Xi i 1

I LIST OF FIGURES (Continued) c Figure-Title Page 3.6-1 . Nuclear Power' Transient for an RC0A;Eje: tion Accident 3-92 0 3.6-2 Hot _ Spot Fuel and Clad Temperature Transients for an 3-93 RCCA Ejection Accident 3.7-1 Pressurizer Pressure and Water Volume for a' Loss of 3-94 -Feedwater n i 3.7-2 Nuclear Power and Core Heat Flux for a Loss of Feedwater 3-95' '3.7-3 RCS Hot and Cold leg Temperatures for a Loss of Feedwater-3-96 3.7-4 Steam Generator Pressure and Mass for a Loss of Feedwater 3-97 l 3.7-5 RCS Flow and. Pressurizer Relief for a loss of Feedwater 3-98 .3.7-6 Pressurizer Pressure and Water Volume for a loss of 3-99 Feedwater (N-1 Loop Operation)- 3.7-7 Nuclear Power and Core Heat Flux for'a Loss of Feedwater 3-100 ,(N-1 Loop Operation) 3.7-8 RCS Hot and Cold Leg Temperatures for a. Loss of Feedwater 3-101 i. (N-1 Loop Operation) 3.7-9 ' Steam Generator Pressure and Mass for a Loss of Feedwater 3-102 (N-1 Loop Operation) g_ /. 3.7-10 RCS Flow and Pressurizer Relief for a Loss of Feedwater 3-103 (N-1 Loop Operation) xii 0657v.1o/081187 I e ._________9

=. _ _ _ 1 l LIST OF FIGURES (Continued) Figure Title Pace 3.7 Pressurizer Pressure and Water Volume for a Loss of 3-104 Feedwater.Without Offsite Power Available i 3.7-12 Nuclear Power and Core Heat Flux for a Loss of Feedwater 3-105 Without Offsite~PoweriAvailable 3.7-13 RCS' Hot and Cold Leg Temperature for a Loss of Feedwater 3-106 Without Offsite Power Available -3.7-14 Steam Generator Pressure and Mas's for a Loss of feedwater 3-107 . Without Offsite Power Available 3.7-15 RCS Flow and Pressurizer Relief for a Loss of feedwater 3-108-Without Offsite Power Available 3.7-16 '. Pressurizer Pressure and Water Volume for a Loss of Feedwater 3-109 Without Offsite Power Available (N-1 Loop Operation) 3.7-17. Nuclear Power and Core Heat Flux for a Loss of Feedwater 3-110 Without Offsite Power Available (N-1LoopOperation) 3.7-18 RCS Hot end Cold Leg Temperatures for a Loss of feedwater 3-111 Without Offsite Power Available I (N-1 Loop Operation) i 3.7-19 Steam Generator Pressure and Mass for a loss of Feedwater 3-112 Without Offsite Power Available (N-1 Loop Operation) 0657v:1o/081187 xiii

] w !. I j P LIST OF FIGURES'(Continued) .\\ g =) ~ Title Page

Figure f

3-113 3.7-20 RCS Flow and Pressurizer Relief'for a Loss of Feedwater Without Offsite Power Available 1 (N-1 Loop _0peration)- I .3-114 3.8-1 Nuclear Power and Coolant Average Temperature for

\\

Inadvertent Opening-of a Pressurizer Safety or Relief Valve e 3-115'- ) 3.8-2 Pressurizer Pressure and DNBR for Inadvertent Opening of a Pressurizer Safety or Relief Valve I (N-1 Loop Operation). .j 3-116 '3.8-3 ' Nuclear Fower and Coolant Average. Temperature for Inadvertent: Opening of a Pressurizer Safety or Relief Valve- ~I -(N-1 Loop Operation). 3-117' l 3.8-4 Pressurizer Pressure and DNBR for Inadvertent Opening j .of a Pressurizer Safety or Relief Valve (N-1 Loop Operation) 4-16 1 4.3 Post LOCA RCS/ Sump Boron Concentration Versus Pre-Trip RCS Boron Concentration l 4-17 4.4-1 -Corrosion Versus pH.at Several Temperature for Zinc 4-18 l 4.4-2' Corrosion Versus pH et Several Temperatures of Aluminum XiV 06s7v:1D/081187 J

l'. 0 INTRODUCTION' w. The present Millstone Unit 3 Technical Specifications require the moderator temperature coefficient (MTC) to be O pcm/*F* or less at all times while the reactor is critical. A positive coefficient at reducee , power levels results in a significant increase in fuel cycle flexibility, ) while having only a minor impact on the safety analyses for the accident. events presented in the FSAR. The proposed Technical Specification change would allow a +5 pcm/*F MTC below 70 percent of rated power, ramping down to O pcm/*F at 100 percent This MTC is shown in Figure 1.0-1 and is applicable to both N and power. N-1 loop operation. A power-level dependent MTC was chosen to minimize the effect of the specification on postulated accidents at high power levels. Moreover,-as the power level is raised, the average core water temperature becomes higher as allowed by the programmed average temperature.for the plant, tending to produce a more negative moderator coefficient. Also, the boron concentration can be reduced as xenon builds into the core. Thus, there is less need to allow a positive coefficient as full power is approached. As fuel burnup is achieved, boron is further reduced and the moderator coefficient will become negative over the entire operating power range. As a result of the cycle 2 fuel design (PMTC and longer cycle length), an increase in the minimum boron concentration requirements for the Refueling Water' Storage Tank (RWST) and Accumulators is proposed to meet l post-LOCA Shutdown requirements. An evaluation has been performed to show the impacts of increasing the boron concentration range to 2300-2600 ppm for the RWST and 2200-2600 ppm for the accumulators. The impacts on reactor vessel boron precipitation, dose analysis, sump and spray pH, 1 equipment qualification, and the FSAR Safety Analyses have been consider.ed. (See Section 4.0) 4

• 1 pcm = 10 Ak/k i

1~1 D6s7v:1D/081187 ._-_-__-___a

,3 The impact of a positive moderator temperature coefficient (PMTC) on the lE ' accident analyses presented in Chapter'15 of the Millstone Unit 3 Final' . Safety Analysis Report (FSAR) for both N and N-1-loop operation has been . assessed. Those incidents'which were found to be. sensitive to positive j or minimum moderator temperature coefficients were reanalyzed. In h general, these incidents are limited to transients which cause' reactor coolant. temperature to increase. The analyses presented in Section 3 that rely on.the LOFTRAN computer code as a primary analytical tool, with the exception of the locked rotor l accident, were based'on a +5 pcm/*F moderator temperature coefficient over the entire power range. The coefficient-was assumed to remain l const' ant for variations in-temperature. The only Section 3 events not using LOFTRAN'were rod ejection and rod withdrawal from suberitical, which are addressed in the next paragraph. The N loop control rod ejection and rod withdrawal from suberitical i analyses were based on a coefficient which was at least +5 pcm/*F at zero power nominal average temperature, and which became less positive for -higher temperatures. The N-1 loop analysis for rod ejection used a coefficient that was at least +5 pcm/*F at the assumed N-1 loop full power'(75% rated power) and nominal. average temperature, becoming less positive for higher temperatures. This was necessary since the TWINKLE computer code (Reference 2), on which the analyses are based, is a diffusion-theory code rather than a point-kinetics approximation and the moderator temperature feedback cannot be artificially held constant with temperature. For all accidents which were reanalyzed, the assumption of a por,itive moderator temperture coefficient existing at 100% power or a +5 pcm/*F at 75% power is conservative since as diagrammed in Figure 1.0-1, the proposed Technical Specification requires that the coefficient be linearly ramped to zero above 70 percent power. 1-2 0657v;1o/081187

p ~ ~ L In general, reanalysis was based on the identical analysis methods, ! computer: codes, and assumptions employed in the FSAR;'any exceptions are ~ noted in the discussion of'each incident. Accidents not reanalyzed tincluded those which.were not'significantly impacted by:a PMTC or those where the' assumption of a--large negative moderator coefficient is: conservative. Table 1.0-1 gives a list of accidents presented in the Millstone Unit 3 FSAR,'and denotes those events reanalyzed or evaluated: -for a positive moderator coefficient. The following sections provide discussions for each of the FSAR events. I f 06s7v;1D/081187 13

'N i SCDDIATO9 TEMPERAnBiE +G \\ COEFFICIENT (pca/*F) I l I I 70 $00 rowEn (K) l \\ i h f FIGURE 1.0-1 l MODERATOR TEMPERATURE COEFFICIENT VS. RATED THERMAL POWER 0657v:1D/070787 1-4 j _____________ _ __ _ j

{ 7 !~ TABLE 1.0-l' FSAR' ACCIDENTS EVALUATED FOR POSITIVE MODERATOR COEFFICIENT EFFECTS

FSAR,

' Acciifent Time in Life 15.1.1/1.2. Feedwater System Malfunction EOC 15.1.3 Excessive Increase in Secondary Steam Flow BOC/EOC 15.1.4/1.5 Steam Line Depressurization/ Break EOC

• 15.2.2/2.3 Loss of Load / Turbine Trip BOC/EOC
• 15.2.6/2.7 Loss of AC Power / Loss of Normal Feedwater BOC 15.2.8 Feedwater System Pipe Break EOC
• 15.3.1/3.2 Loss'of Flow BOC'
• 15.3.3 Locked Rotor BOC
• 15.4.1-RCCA Withdrawal from'Suberitical BOC
• 15.4.2 RCCA Withdrawal at Power

-BOC/EOC 15.4.3 RCCA Misalignment BOC 15.'4.4 Startup of an Inactive Loop' EOC

• 15.4.8 RCCA Ejection BOC/EOC 15.5.1 Inadvertent ECCS Operation at Power BOC
• 15.6.1 Inadvertent Opening of a Pressurizer Safety or BOC Relief Valve 15.6.3 SGTR BOC 15.6.5 LOCA BOC j
• Accidents reanalyzed BOC - Beginning of Cycle EOC - End of Cycle 0657v:10/081187 1-5

.) i,. ' / J2.0 Transients-Evaluated For a Positive Moderator Coefficient f The.following' transients were not reanalyzed since the impact of'a PMTC on the results is small, or the results~are sensitive to.a negative. ~ moderator temperature coefficient, or the results are not affected by a p W tive' moderator temperature coefficient..The eval.uations cover both N s f ~ and N-1-loop operation. 2.1 RCCA Misalignment Only the cases considering single or multiple dropped RCCAs that are presented in Section 15.4.3 of the FSAR are.potentially affected by a positive moderator temperature coefficient. Use of a positive coefficient;in'the analysis would result in a larger reduction in core power level following the RCCA drop, thereby increasing the potential of. a reactor trip. For the return to power automatic rod control cases with~ ' lower worth dropped RCCAs, a positive coefficient.would result-in a small increase'in the power overshoot; The limiting conditions for this transient occur.for the N loop analyses at or near 100% power..-These results bound those.for the N-1 loop cases,'which are initiated from a . reduced power level. Since the' moderator temperature coefficient must be j close to zero or negative at 100% power, the limiting case is unaffected I by the proposed Technical Specification and reenalysis'was not performed. The limiting analysis presented in the FSAR remains bounding' and the associated conclusions remain valid. j 1 'l 2.2 Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature 1 1: An inadvertent startup of an idle reactor coolant loop at an incorrect h. temperature results in a decrease in core average temperature. As the l' most negative values of moderator reactivity coefficient produce the greatest. reactivity addition, this accident is unaffected by the proposed Technical Specification and thus reanalysis was not required. Therefore, 1 l ~the analysis presented in FSAR Section 15.4.4 remains limiting and the i conclusions reached remain valid. L o6s7v:1o/081187 2-1 J

p..

B, x. p4 Additionally, at Millsto'ne Unit 3 the Technical Specifications for-N-1 loop operation preclude the actual occurrence of the startup.of an. inactive; reactor coolant pump event _in Operating Modes 1, 2, 3, and.4 by

restricting l opening.of the RCS loop stop:. valves.

In Modes-5 and 6 the Technical Specifications limit any possible imbalance in boron concentration'and: temperature between the isolated loop and the rest of' 'the Reictor Coolant system to such an extent, that the associated reactivity insertion would be negligible. These assessments are p . unaffected by the moderator temperature coefficient,-so that the _ conclusions in the.FSAR remain valid. 2.3 Excessive Heat Removal Due to Feedwater System Malfunctions The addition of excessive feedwater or the reduction of feedwater -temperature are excessive heat removal incidents, and are consequently most sensitive to a negative moderator temperature coefficient. Results presented in Section 15.1.1 and 15.1.2 of the FSAR, based on a negative coefficient.. represent the limiting case. Therefore, this incident was not reanalyzed and the conclusions of the FSAR for both N and N-1 loep operation remain. valid. l 2.4 Excessive Increase in Secondary System Flow An excessive increase in secondary system steam flow (excessive load increase event) is defined as a rapid increase in steam flow that causes a power mismatch between the reactor core power and the steam generator load demand. This results in decreased reactor coolant system l temperature. With the reactor in manual control, the analysis presented in Section 15.1.3 of the FSAR shows that the limiting case assumes a large negative moderator coefficient. If the reactor is in automatic l control, the control rods.are withdrawn to increase power and restore the average temperature to the programmed value. The analysis of this case [ in the FSAR shows that the minimum DNBR is not sensitive to moderator j h f 1 0657v:10/081187 2-2 3 ') -__---_-_-_--____-_--__D

m 1 l 1 u l temperature coefficient. Therefore,' the results presented in the FSAR for both N and N-1 loop operation continue to be limiting and the conclu.sions presented remain valid. 2.5 Main-Steam Line Depressurization/ Steam System Piping Failure The steam line depressurization and steam system piping failure events are transients that result in' cooldown of the reactor coolant sys am. ~ Therefore, the associated analyses initiated from hot zero power 1 conditions (FSAR Sections 15.1.4 and 15.1.5) are performed assuming a strongly negative moderator temperature coefficient, which represents a limiting assumption for consideration of a post trip core power excursion. As a result these analyses are unaffected by the proposed i Technical Specification change in the allowable positive moderator temperature coefficient._ Therefore, the steam line depressurization/ steam system piping failure analyses of the FSAR for both N and N-1 loop operation remain limiting and the conclusions reached remain valid. I 2.6 Feedwater System Pipe Break The main feedwater pipe rupture accident (Section 15.2.8 of the FSAR) is analyzed to demonstrate the ability of the auxiliary feedwater system to remove decay heat from the reactor coolant system. For conservatism, the current FSAR analysis is performed at 102% of Engineered Safeguards power (104.5% NSSS rated) in order to significantly increase the total energy which eventually must be removed from the core, due to stored energy and decay heat. The N-1 loop FSAR analysis was performed at 77% of rated i i power, which represents the assumed full N-1 loop power of 75% including the calorimetric error. Both these analyses were performed using a large negative moderator temperature coefficient, and sensitivities have confirmed that this assumption continues to be appropriate. Therefore, the results for this accident are unaffected by the I-incorporation of a positive moderator temperature coefficient Technical l I-l h l 0657v:10/081187 2-3 I b. .________________U

E 1 --Specification.. Based on this,.the cases presented in the FSAR for both'N and N-l', loop operation remain limiting and the associated conclusions remainivalid. 2.7. Inadvertent Operation of the'ECCS at Power l l Analysic'of a spurious actuation of safety injection at power is presentedinSectioI) 15.5.1 of the FSAR. This transient results in a decrease in average. reactor coolant temperature'.and core power. The results are not sensitive-to moderator temperature coefficient. Therefore, this incident was not reanalyzed with a positive moderator coefficient. The analysis presented in the FSAR for both N'and N-1 loop operation continues to be limiting and the' conclusions reached remain valid. 2.8 Loss of Coolant Accidents (LOCA) Small Break LOCA The influence:of a positive MTC on peak clad temperatures calculated for the small break LOCA (SBLOCA) FSAR analysis described in section 15.6.5 has been evaluated with respect to the effect of a +5 pcm/*F MTC on SBLOCA response and the margin available to meet 10CFR50.46 ECCS acceptance criteria. In the SBLOCA analysis methodology, core kinetics calculations are not explicitly performed. The core power is maintained at initial conditions (102% power) until the reactor trip setpoint is reached and the rods reach the bottom of the core. This delay results in the generation of an additional few full power-seconds of heat, by not accounting for partial rod worth while the rods are falling into the core and the shutdown effect of voiding'for cores which have a negative moderator temperature coefficient. After trip, the power is calculated from a decay heat curve. This generic power decay curve is composed of three parts: residual fission heat, fission product decay, and actinide gamma decay. 0657olo/081187 2-4

1 The residual fission term is based upon the exponential' decay of the l fission-power for a low shutdown margin, and with the core full of' hot f water. This will be conservative for essentially all SBLOCAs,.since some net-voiding occurs coincident with reactor trip,'due to the sudden 7 depressurization. An evalua' tion has determined that the excess co're power generation which may be expected from explicitly modeling the +5 pcm/*F MTC would be much less than 1 full power-second' By reference to applicable' calculations an'd sensitivity studies it has been concluded that the positive MTC has only a small, third order effect on peak clad temperature (PCT). The small core power excursion induced in the initial few seconds of the transient will slow the depressurization negligibly, delaying reactor trip and SI initiation only slightly. These delays, plus the small excess power, in turn will have a small influence on loop clearing and subsequently the core boiloff uncovery transient, hundreds of seconds into the accident, during which time the clad PCT occurs. However, there j will be virtually no direct influence on decay heat generation during the-clad temperature excursion. The FSAR analysis parformed with the WFLASH (Reference 8) code, shows that a'very substantial margin exists between the results calculated and the 10CFR50.46 limits. Any impact from operation with a +5 pcm/*F MTC I will be very small compared to this margin. While this calculation was performed for 4-loop operation, analysis (Reference 9) has demonstrated that plant operation in the N-1 loop configuration at corresponding reduced power will result in less limiting SBLOCA PCTs than for the full power, N-loop configuration. Large Break LOCA i The large break LOCA analysis for Millstone Unit 3 is described in FSAR l_ Section 15.6.5. The current FSAR large break LOCA analysis was performed with the 1981 Evaluation Model. Currently, the large break LOCA analysis j i i i 06s7v:1o/081187 2-5 j _A

..a I l presented in-the FSAR does~not.take credit for the negative reactivity 1 s introduced by:the' soluble boron in the ECCS water in determining the-reactor power during the early phases of'a postulated.LOCA. The large break LOCA also does not take credit for the neg'ative reactivity-introduced by the control rods. During a break LOCA','the reactor is brought-to.a suberitical condition by the presence of voids in the core. The implementation of a:PMTC as shown in Figure 1.0-1 does not have any-adverse effect on the 4-loop operation large break'LOCA analysis'since. the PMTC'.is zero at 100%' power. This evaluation will focus on the effect-of a PMTC'on the'3-loop (N-1) large break LOCA analysis as presented in the FSAR. For large break.LOCA analyses, PMTC can currently only be modelled during blowdown. Once voids form during blowdown the negative reactivity.added from the voids is substantial compared to'the positive reactivity added by PMTC. Since the'effect during refill and reflood.will be n'egligible, PMIC is not currently modelled in the refill and reflood ph.ases of.a LOCA. The effect on the peak clad temperature, modelling'PM.TC'during L,iowdown, has been pra iously evaluated. Using this existing evaluation 't is determined, that for the PMTC case the clad temperature can be estimated to increase as much as 56*F, if a PMTC of +5 pcm/'F is modelled' during blowdown in'the 3-loop (N-1) large break LOCA analysis. The PCT for 3-loop (N-1) opera' tion reported in the FSAR is 1878'F. The new estimated PCT for 3-loop (N-1) operation is 1934*F. This is still well below the 4-loop PCT of 2132*F which is also reported in the FSAR. -There is no adverse effect on acceptance criteria in 10CFR50.46 for the large break LOCA analysis presented in the Millstone 3 FSAR as a resuit-of the proposed PMTC implementation shown in Figure 1.0-1. LOCA Mass and Energy Releases i The containment analyses described in FSAR Section 6.2 for N,-loop I operation are not impacted by PMTC since the PMTC as proposed in Figure 1.0-1 is zero at'100% power. For N-1 loop operation the subcompartment I i 0657v:1o/081187 2-6 { q

0. ,V a. .l 1 is l analysis is not affected due to'the short duration of.the transient...In f ) y addition,Lthelongt[mmassandenergyreleasesforN-1loopoperation F areboundedby'thetyloopreleases. Therefore there is no adverse affect ~ 3 on the containment ' analyses a d the conclusion in/the FSAR remain v'alid. ] n LOCA: Forces n I Theblowdownhydraulic.loadsresultingffomaLOCAdescribedinFSAR sections. 3.6.2.2.2, 3.9B.1.42.4,'3.9N.1.,4.3, and 3.9N.2.5 for N-loop - ( operation are not impacted by PMTC since the PMTC is zero at4100% power. -l The effect of PMTC on_'the N-1 loop analysis is to increase the fluid-I t pressure in the reactor vessel by approximately 1.4 psi. This results in an estimated increas6 of 0.4% for the calculated peak forces. Sufficient 4 margin exists. in the dynamic analysis presented in the FSAR to offset-this small increase. 2.9lSteamGeneratorTubeRupture(SGTR) .\\ The Steam Generator Tube Rupture (SGTR) event, deseHbedln!FSAR section 'p 15.6.3, was performed using the LOFTRAN program. The primary to, f) t ' secondary break ficw was assumed terminated at 30 minutes after initiation of the'SGTR event. Themajor.fhetorsthataffectthe M., a a radiological doses of an SGTR event are the amount of fuel failure, the amount of primary coolant transferred to the secondary side of the ruptured steam generator through the ruptured tube after reactor trip, j and the steam released from the ruptured steam generator to the atmosphere. A positive moderator temperature coefficient results in an i earlier reactor trip and thus a net increase'in the primary coclant

transferred to-the ruptured SG secondary through the ruptured tube after the reactor. trip.

Also, there is an increase in steam released to the atmosphere via'the ruptured SG safety valve. The increase in mass I 8 released through the tube rupture to the affected SG after trip is less than 4% of the total mass release via the-tube rupture for the current d, 4 [ FSAR analysis. Additionally, the increase 'in st'eam released from the m; >l i 0657v.lo/081187 2-7

i; r ruptured SG after trip is less than 5.5% of the. steam release from the ruptur'ed SG for.the current FSAR analysis. The net total increase in-mass' release via the tube rupture flow and ruptured SG safety valve steam ' flow w'11 result <in less than a 10% increase in offsite. radiation doses. ThIIincreasein't'eoffsitedosesisstillmuchlessthanthe10CFR100 h limits,- The. impact on minimum DNBR'for the SGTR analysis with positive moderator tem;ihrature coefficient has also been evaluated. The results show that ' the"$$nimum DN5R for the SGTR event remains above the DNBR limits. 1 Therefore, fuel failure will not occur for the SGTR event due to a positive; moderator temperature coefficient. 'l f Band on the above evaluation, it is concluded that a positive moderator temperature coefficient will not change.the conclusions reported in the i< FSAR SGTR analysis. Since'the offsite doses with the positive moderator . temperature coefficient are still much less than a small fraction of the 10CFR100 limits-and the DNBR limits for the SGTR will not be challenged, no reanalysis or FSAR changes are required. 2 i i f* i f(L, des 7w1D/081187 2-8 3 3x hO' {

i 4 I 3.0 Transients Analyzed for a Positive Moderator Coefficient The principal computer codes used for the reanalyses documcnted in this section are LOFTRAN (Reference 1), TWINKLE (Reference 2), FACTRAN l (Reference 3), and THINC (References 4 and 5). These codes are the same as those used in the Millstone Unit 3 FSAR Analyses. ' Summaries of these computer codes are presented in Sections 15.0 and 4.4 of the FSAR. For each event reanalyzed the basic assumptions regarding initie.1 conditions, instrumentation errors, and setpoint errors remain largely the same as those found in Chapter 15 of the FSAR. However, the current analyses do incorporate certain additional changes that should be noted. FSAR Section 15.0.3.2 specifies a 1 30 psi allowance on pressurizer pressure for steady state fluctuations and measurement penalty. The reenalyzed events include a more conservative 1 45 psi allowance for pressurizer pressure. Additionally, increased uncertainties have been applied to the pressurizer and steam generator water levels. These uncertainties have been increased from 5% to 5.73% and 5.53%, respectively, for the pressurizer and the steam generator. These increased uncertainties have been incorporated to bound calculated increases in the associated transmitter uncertainties. For analyses which rely on the overtemperature delta-T and overpower delta-T reactor trip functions to provide protection, an increased Resistance Temperature Detector (RTD) response time of 7.0 seconds and an increase in the lead-lag compensation of delta-T were considered to bound i i the effects of RTD Bypass Loop Elimination. For more detail on the effects of RTD Bypass Elimination, see Reference 7. 3.1 Uncontrolled RCCA Bank Withdrawal From a Suberitical Condition Introduction A RCCA bank withdrawal incidea when the reactor is suberitical results in an uncontrolled addition of reactivity leading to a power excursion (see Section 15.4.1 of the FSAR). The nuclear power response h DE5k10/081187 3-1 l i

)- ~ l {[- +

- " l y

characterized by a' very' fast rise tersinated by the reactivity feedback ' .-Q of.thenegativefuel' temperature;coefficientand.areactortrip'on-4 1 j source, intermediate or power ra'nge flux, or'high positive nuclear flux-r ' rate." The power. excursion causes a heatup.of.the' moderator and fuel. A. 1 positiveimoderator coefficient causes an increase in.the rate' of reactisityaddition,_.resultinginan: increase'.inpeak.heatflux.andpeaki fuel and clad temperature. 1-y, v Method of: Analysis; p The' analysis was performed in the FSAR for a reactivity insertion rate.of '65.pem/sec. The accident was reanalyzed for Cycle 2 with an'inseftion rate"of'58 pcm/sec. This insertion' rate is greater than that for' the simultaneous' withdrawal of the combination of the two sequential control bsnks having the' greatest combined-worth.at maximum speed (45 inches / minute). -The analysis used'a' moderator temperature coefficient of +5 pcm/*F.at hot zero. power initial conditions.. The computer codes, . initial conditions,,and other assumption's remain as noted in the FSAR. l-Results Figures 3.1-1 through 3.1-3'show the transient behavior for the uncontrolled ~RCCA bank withdrawal incident, with the accident terminated by reactor trip at 35% of nominal power. Figures 3.1-1 and 3.1-2 show the neutron flux and thermal flux I transients. The neutr;.. flux slightly overshoots the nominal full power value; however, due to the beneficial effect of the inherent thermal lag in the fuel, the peak' heat flux is much less than the full power nominal

value, t.

[' Figure 3.1-3 shows the hot spot fuel average and clad temperature transients. The fuel average temperature increases to a value lower than at nominal full power. The minimum DNBR at all times remains above the L limiting value. ' 0657v:1o/0811s7 3-2

g yg Q 3 r t The'.calcula'tedsequenceofeventsforthistransientisshowninTable m s 3.1-1. ,1 Conclusio'n [ i In the event of a RCCA bank withdrawal accident from a suberitical 1 condition,:the core and the RCSlare not adversely affected, since.the = combination of thermal power and the coolant temperature result in a DNBR-greater than the: limit'value and thus, no fuel or' clad damage is predicted. Therefore, the conclusions presented in the'FSAR. remain. . valid. s '3.2 Uncontrolled RCCA Bank Withd'rawal'at Power Iritroduction. 'The Uncontrolled RCCA' bank withdrawal at power event is described in Section 15.4.2 of the FSAR.. An uncont?c11ed RCCA bank withdrawal at c.- power causes-a positive reactivity insertion which results in 'an increase 1 in the core heat flux. Since the heat extraction from the steam . generator lags.behind the core power generation, there is a net increase 'in the reactor coolant temperature. With a PMTC this temperature increase could add positive reacthity. Unless-terminated by manual or automatic action, the' increase in coolant temperature and power could' result.in DNB.. For this event, the Power Range.High Neutron Flux and 'Overtemperature Delta-T reactor trips'are assumed to provide protection. against DNB. Therefore,'this event was reanalyzed with a +5 pcm/*F moderator _ temperature coefficient and with increased RTD time constants to show that the DNBR limit is. met. Methods With the exception of the items noted here and in Section 3.0, the assumptions used.are consistent with the FSAR. The transient is analyzed at 10%,.60%, and 100% power for N-loop operation and at 10% and 75% power 06s7v;1o/081187 3-3 a .--[

,j for'N-1 loop operation. Both minimum'and maximum reactivity' feedback -cases were reanalyzed'with the increased time response value. A constant. i moderator coefficient of +5'pcm/*F was used in.the analysis for cases -based on. minimum feedback. The assumption:that a-positive moderator i coefficient exists'at full powerLis conservative.since at full power the moderator coefficient will actually be ze.ro or negative. The analysis ~ was performed using the LOFTRAN Computer. Code. Results ForLboth' minimum and maximum reactivity insertions, at the various power levels analyzed,'the DNBR limit is met.. A calculated sequence of events for a fast and slew insertion rate from full power (N and N-1 Loop Operation).is presented on Table 3.2-1. The transient response for a fast insertion case and a slow insertion case from full power is shown in ~ Figures 3.2-1 through 3.2-8 (N-Loop) and Figures 3.2-9 through 3.2 (N-1 loop operation). The plots of minimum DNBR versus reactivity insertion rate at the analyzed power levels are shown as Figures 3.2-17 through 3.2-19 (N-loop) and Figures 3.2-20 and 3.2-21 (N-1 loop operation). Conclusions The limit DNBR is met, and therefore, the conclusions presented in the FSAR remain valid for both N and N-1 loop operation. 3.3 Loss of Forced Reactor Coolant Pump Flow Introduction The loss of. flow events presented in FSAR Sections 15.3.1 and 15.3.2 were reanalyzed to determine the effect of a +5 pcm/'F moderator temperature coefficient on the nuclear power transient and the resultant minimum DNBR reached during the incident. The effect on the nuclear power transient t e657v:1o/o81187 3-4

l. i L would be. limited to the initial stages of the incident.during which ) reactor coolant temperature increases; this increase is terminated-shortly after reactor trip. Method of Analysis i With the exception of'the moderator temperature coefficient and the-items -noted in Section 3.0, the methods and assumptions'used in the reanalysis . ere consistent with the FSAR. All the cases presented in the FSAR, w partial and complete loss of flow for N and N-1 loop operation, were reanalyzed. The computer' codes used in the reanalysis remained the same as those described in the FSAR, while a constant moderator-temperature coefficient of-+5 pem/*F was used in the reanalysis to reflect the revised Technical Specification. Results Figures 3.3-1through3.3-4showthetransientresponsefortilelossof i one reactor coolant pump with four loops in operation. Figure 3.3-4 shows the DNBR to be always greater than 1.30. Figures 3.3-9 through 3.3-12 show the transient response for the loss of one reactor coolant pump with three loops in operation. The minimum DNBR is greater than 1.30, as shown in Figure 3.3-12. .For both the partial loss of flow cases analyzed, since DNB does not occur, the ability of the primary coolant to remove heat from the fuel rod is not greatly reduced. Thus, the average fuel and clad temperatures do not increase significantly above their respective initial values. The calculated sequence of events tables for.the two partial loss of flow cases are shown in Table 3.3-1. The affected reactor coolant pump will l continue to coastdown, and the core flow will reach a new equilibrium i value corresponding to the number of pumps still in operation. With the reactor tripped, a stable plant condition will eventually be obtained. Normal plant shutdown may then proceed. o657v:10/081187 3-5 ~

u Figures 3.3-5 through'3.3-8 show the transient response for the loss of-power; to' all reactor coclant pumps with four loops in operation. The reactor is assumed to be tripped on an RCP underspeed signal.. Figure-3;3-8 shows'the DNBR to be'always greater than'1.30. Figures:3.3-13 through 3.3-16'show the~ transient response for the loss of power'to all' reactor coolant pumps with three loops in operation. The' reactor'is again assumed to be tripped'on an underspeed. signal. The-minimum DNBR is greater than 1.30, as shown in Figure 3.3-16'. For'both complete loss of flow cases' analyzed, since DNB does not-occur, the. ability of.the primary coolant to remove' heat from the' fuel' rod is not greatly reduced. Thus, the average fuel and clad temperatures'do not increase significantly above their respective' initial values. The calculated sequence of events for the two complete loss of flow cases ~ is'shown in Table 3.3-1. The reactor coolant pumps will continue to coastdown, and natural' circulation flow will eventually be established. As demonstrated in'Section 15.2.6'of the FSAR, with the reactor. tripped, a stable plant condition would be attained. Normal plant' shutdown may then proceed. Conclusions The'DNBR-design basis is met for-the partial and complete loss of' flow cases, both for N and N-1 loop operation. Therefore, the conclusions of the FSAR remain valid. - 3.4 Reactor Coolant Pump-Shaft Seizure Introduction .{ The case presented in the FSAR (Section 15.3.3) for this transient was reanalyzed. Upon initiation of a locked rotor incident, reactor coolant system temperature rises'until shortly after reactor trip. A positive moderator coefficient will not affect the time to DNB since DNB is 0657v:1o/081187 3-6

conservatively assumed to occur at the beginning of the incident. The transient was reanalyzed, however, due to the potential impact on the nuclear power' transient which would effect the peak reactor coolant system pressure end fuel and clad temperatures. Method of Analysis With the exception of the items noted in Section 3.0, the methods and assumptions used in the reanalysis were consistent with the FSAR. All cases in the FSAR, one locked rotor for N and N-1 loop operation both with and without offsite power available, Were reanalyzed using the same computer codes described in the FSAR. The reanalysis, which initiated the transient from full power conditions, employed a constant moderator temperature coefficient of +5.0 pcm/*F for peak pressure and clad temperature analyses. For the N loop rods in DNB analysis, a constant i moderator temperture coefficient of 0.0 pcm/*F was used. This case bounds results for cases initiated from a lower power level, but with a positive moderator temperature coefficient, as indicated by Figure 1.0-1. The locked rotor rods in DNB analysis for three loop operation employed a moderator temperature coefficient of +5 pem/*F, which is slightly conservative relative to the value indicated by Figure 1.0-1. Results i The transient results for the locked rotor event with four loops operating are shown on Figures 3.4-1 through 3.4-4. The results of these calculations are also summarized in Tables 3.3-1 and 3.4-1. The peak reactor coolant system pressure reached during the transient is less than that which would cause stresses to exceed the faulted condition stress limits. Also, the peak clad surface temperature is considerably less than 2,700*F and the amount of Zirconium-water reaction is small. It should be noted that the clad temperature was conservatively calculated j assuming that DNB occurs at the initiation of the transient. The percentage of fuel rods predicted to be in DNB for the locked rotor event initiated with four loops in operation, was 3.0%. i o65?v 1D/081187 3-7 i l

1 The transient results for the locked rotor event with three loops operating are shown on Figures 3.4-5 through 3.4-8. The results of these calculations are also summarized in Tables 3.3-1 and 3.4-1. The peak RCS pressure reached during the transient is less than that which would cause stresses to exceed the. faulted condition stress limits. The cladding temperature transient is still well below the 2,700*F limit and the amount of zirconium-water reaction is small. The percentage of fuel rods predicted to be in DNB for the locked rotor event initiated with three loops in operation, was 8.0%. Conclusions Since the peak reactor coolant system pressure reached during any of the transients is less than that which would cause stresses to exceed the faulted condition stress limits, the integrity of the primary coolant system is not endangered. Similarly, since the peak clad surface temperature calculated for the hot spot during the worst transient remains considerably less than 2,700*F, the core will remain in place and intact with no loss of core cooling capability. Therefore, the conclusions presented in the FSAR for both N and N-1 loop operation with respect to peak pressure and clad temperature remain valid. 3.5 Loss of Lead / Turbine Trip Introduction The FSAR explicitly analyzes the turbine trip event (FSAR Section 15.2.3) which bounds the loss of load (FSAR Section 15.2.2). All eight cases presented in the FSAR were reanalyzed. These consist of minimum reactivity feedback (beginning of life) and maximum reactivity feedback (end of life), each run for N and N-1 loop conditions, using both of tne following sets of assumptions regarding plant control systems: l I DE57v:10/081187 3-8

1. Full credit is taken for operation of the pressurizer spray and power operated relief valves. The pressurizer safety valves are also assumed to be available. 2. No credit is taken for the operation of the pressurizer spray and power operated relief valves. The pressurizer safety valves are assumed to be available. Method of Analysis The minimum reactivity feedback cases were run with a constant moderator temperature coefficient of +5 pcm/*F, which is conservative for full power. Although the maximum feedback cases i ve a negative moderator i temperature coefficient and are therefore unaffected by the proposed Technical Specification change, they were reanalyzed because of the items noted in Section 3.0. The other assumptions and the methodology employed were consistent with the FSAR. Results For all combinations of reactivity feedback and pressure control, the applicable safety limits are met. The results of these cases are presented as Figures 3.5-1 through 3.5-8 (N-loop) and Figures 3.5-9 through 3.5-16 (N-1 loop operation). A calculated sequence of events is shown in Table 3.5-1. Figures 3.5-1 and 3.5-2 (N-1 loop) and Figures 3.5-9 and 3.5-10 (N-1 loop operation) show the responses for a turbine trip event with minimum reactivity feedback assuming operability of pressurizer sprays and PORV's. The reactor is tripped by the High Pressurizer Pressure trip function. The DNBR increases throughout most of the transient and never drops below the design limit. The primary system pressure remains below the 110% design value. ? 1 1 1 i o6s7v 1o/081187 3-9 ) l

Figures 3.5-3'and 3.5-4.(N-loop) and Figures 3.5-11 and 3.5-12 (N-1 loop operation) show the respontes for a turbine trip with maximuu reactivity feedback and pressure control. The reactor is, tripped by the low-low-Steam: Generator Level trip function, and the DNBR never drops below the initial value. The primary. system pressure remains below the 110% design value. Figures 3.5-5'and 3.5-6 (N-loop) and Figures 3.5-13 and 3.5-14 (N-1 loop operation) show the responses for a turbine trip with minimum reactivity-feedback and without pressure control. The reactor is tripped by the High Pressurizer _ Pressure trip function,'and the DNBR never drops below the in'itial value. The primary system pressure remains below the 110% design value. Figures 3.5-7 and.3.5-8_(N-loop) and Figures 3.5-15 and 3.5-16 (N-1 loop l operation) show the responses for a turbine trip with maximum reactivity feedback'and without pressure control. The reactor is tripped by the High Pressurizer Pressure trip function, and the DNBR never drops below the initial value. The primary system pressure remains below the 110% design value. Conclusions The DNBR design basis is met and the system pressure remains below 110% of the design.value in all four cases for both N and N-1 loop operation, and therefore, the conclusions presented in the FSAR remain valid. The Overpressure Protection Report is also not impacted by these results, and thus, the conclusions presented in that document remain unchanged. 0657v:1D/081187 3-10 1

x 1. 3.6' RCCA Ejectior Introduct'ien The RCCA ejection transient is analyzed.at full power and hot. standby for bbth beginning and end.of life conditions in the FSAR. Since the moderator.. temperature coefficient is negative:at end'of life, only the beginning-of-life cases are affected by a positive moderator temperature coefficient. The high nuclear power. levels'and hotispot fuel temperatures resulting from a rod ejection are increased by-a positive moderat'or. coefficient.-'A discussion of this transient is presented.in Section 15.4.8 of.the'FSAR. Method of Analysis 1, The' digital computer codes.for analysis of the nuc' lear power transient and hot' spot 1 heat transfer are the same as those used in the FSAR. The ejected rod worths and transient peaking factors assumed are conservative with respect to the actual calculated values.for current fuel cycle. The . analysis.used a moderator temperature coeffi:ient which was consistent-with the +5 pcm/'F allowed at hot zero power conditions. Results i L 'The cases analyzed were beginning of life at hot full power and hot zero I power for N loop operation, and beginning of. life at hot full power for -1 N-1 loop operation. The maximum clad average temperature was reached in the hot zero power ] case. However, the peak hot spot value of 2419'F was below the limit specified.in the FSAR. r; l 1 I i o657v;1D/081187 3-11 i a:

The maximum fuel' temperature and fuel' enthalpy were associated with the : ~ hot' full power cases. Although the peak fuel centerline temperature at the hot spot ~ exceeded melting for the N loop operation cast, melting was restricted to less than the innermost 10% of the' pellet. No nelting. occurred in the N-1 loop operation. case. The peak fuel enthalpy in both . cases was-well below the limit specified in the FSAR. A' summary'of,the parameters and results for the N'and N-1 loop cases is presented in Table 3.6-2. The calculated sequence of events for each case is.shown in Table 3.6-1. The nuclear power.and hot spot fuel and clad temperature transients'for the worst case, hot full power for N' loop'. operation, are shown in Figures 3.6-1 and 3.6-2. Conclusions 'As fuel and clad' temperatures do not exceed the fuel and clad limits specified in the FSAR, there is no danger of sudden fuel dispersal into. the coolant,.or consequential damage to the primary coolant loop. Therefore, the conclusions ~ presented in'the FSAR remain valid. ' 3. 7 Loss of Normal Feedwater v -Introduction This accident is: described'in Sections 15.2.6 and 15.2.7 of the FSAR and is presented both with and without offsite power available. Since this i transient is' analyzed consistent with beginning of life conditions, it ] was reanalyzed with a positive moderator temperature coefficient. Methods A constant moderator temperature coefficient of +5 pcm/*F was assumed. A . conservative core decay heat model based on the 1979 version of ANS 5.1 l (Reference 6) was used. The pressurizer pressure control system (sprays l 1 0657v;1o/081187 3-12 ) ) a .-__-____--__-_-__---_____O

and power operated relief valves) was assumed to be available since a lower pressure results in greater system expansion. With the exception of the items noted in Section 3.0, all remaining assumptions are consistent with the analysis presented in the FSAR. For the case without offsite power available, the uncertainties and errors on the initial average temperature were subtracted from the nominal value, and power is assumed to be. lost to the reactor coolant pumps following rod motion. The reanalysis used LOFTRAN to obtain the plant transient following a loss of normal feedwater. Results The transient response of the RCS following a loss of normal feedwater is shown in Figures 3.7-1 through 3.7-5 (N loop) and Figures 3.7-6 through Figure 3.7-10 (N-1 loop) with offsite power available and Figures 3.7-11 through 3.7-15 (N loop) and Figures 3.7-16 through Figure 3.7-20 (N-1 loop) for the case without offsite power available. The calculated sequences of events are listed in Tables 3.7-1 and 3.7-2. The plots of pressurizer water volume clearly show that for all cases the pressurizer does not fill. Conclusions The reanalysis shows that a loss of normal feedwater does not adversely affect the core, the RCS, or the steam system, and the auxiliary feedwater system is sufficient to preclude water relief through the pressurizer relief or safety valves. For the case without offsite power available, the natural circulation capability of the RCS is sufficient to remove decay heat following a RCP coastdown to prevent fuel or clad damage. Therefore, since the pressurizer does not fill, the conclusions presented in the FSAR remair valid. oss7v;1otos11s7 3-13

p (. < : i[ c l w i3.8 ' Inadvertent Opening of a Pressurizer Safety or Relief Valve r Introduction This event is analyzed in Section.15.6.1 of the FSAR using a zero moderator coefficient in; order to minimize negative reactivity feedback. A positive moderat'or. temperature coefficient'can-also be considered as a y

negative density coefficient.and therefore~, the density reduction due'to the RCS depressurization causes a positive reactivity insertion and'ani

-increase in nuclear power.- Reactor trip.is generated by the ~

overtemperature delta-T function. The RCS depressurization. incident is

. reanalyzed.to determine:the impact on the nuclear po'wer transient and the ~ minimum DNBR.

Methods Assumptions made in the RCS Depressurization analysis include,a constant-moderator temperature coefficient (+5 pcm/*f) and a small (absolute!

'value) Doppler coefficient of reactivity such that the resultant amount - of positive feedback is conservatively high. The. rod' control-system is assumed to be in the manual mode in order-to prevent rod insertion due to 4 < an RCS' temperature and power mismatch prior to reactor trip. With'the exception' of the items noted here and in 3.0,: the method of ' analysis and - assumptions used were otherwise in accordance with those presented in the FSAR. -As in the FSAR, the reanalysis considers both N and N-1 loop l operation. Results A calculated sequence of events is presented in Table 3.8-1. Figures 3.8 and 3.8-2. (N-loop) and Figures 3.8-3 and 3.8-4 (N-1 loop opei ation) .show the nuclear power, average temperature, pressurizer pressure, and DNBR vs. time for the accidental depressurization of the RCS. The positive moderator coefficient causes nuclear power to increase as oss7w1o/os11s7 3-14

pressure decreases,until' reactor trip occurs on Overtemperature. Delta-T. The DNBR decreases initially, but increases rapidly following the trip.. 1 4 The DNBR remains aoove 1.30 throughout the transient. l' - I Conclusions The' analysis demonstrates that the DNBR remains ~ above 1.30 and therefore, the conclusions presented in the FSAR remain valid for both N and N-1 loop operation. 4 Y a 4 0657v;1o/081187 3-15

[. f' -:f - q. LJ . TABLE 3.1-l' p .. TIME-SEQUENCE 0F EVENTS FOR AN. UNCONTROLLED BANK WITHDRAWAL FROM SUBCRITICAL CONDITION p,.; , i:.. EVENT-TIME (sec) Initiation.of uncontrolled rod withdrawal

0. b from 10 N of nominal power i'

Power range high neutron; flux. low 12.3 ' setpoint reached , Peak nuclear power occurs 12.5 ( ' Rods begin to fall into core 12.8

Minimum DNBR occurs 14.0-Peak heat flux occurs 14.1 Peak clad average: temperature occurs 14.5 4

14.6 Peak fuel average temperature occurs 4 t 0657v:10/081187 3-16

4 TABLE 3.2-1 (page1of2) TIME SEQUENCE OF EVENTS FOR A J RCCA BANK WITliDRAWAL AT POWER j 4 N-LOOP N-1 LOOP ACCIDENT EVENT TIME (SECS) TIME (SEC) Case A Initiation of uncontrolled RCCA 0.0 0.0 withdrawal at a fast reactivity insertion rate (70 pem/sec) with minimum reactivity feeciback at full power Power range high neutron flux 1.5 1.5 reactor trip signal initiated Rods begin to drop 2.0 2.0 Minimum DNBR occurs 3.0 2.9 Peak wator level in the 4.3 4.3 pressurizer occurs Case B Initiation of uncontrolled RCCA 0.0 0.0 withdrawal at a low reactivity insertion rate (3 pcm/sec) with minimum reactivity feedback at full power Overtemperature AT reactor 21.9 trip signal initiated 29.8 Power range high neutron flux reactor trip signal initiated Rods begin to drop 23.4 30.3 Ninimum DNBR occurs 23.9 30.7 Peak water level in the 25.5 32.3 I l pressurizer occurs i i l l I f 3-17

TABLE 3.2-1 l (page2of2) 1 TIME SEQUENCE OF EVENTS FOR A RCCA BANK WITHDRAWAL AT POWER I N-LOOP N-1 LOOP ACCIDENT EVENT TIME (SECS) TIME (SEC) Case C Initiation of uncontrolled RCCA 0.0 0.0 withdrawal at a fast reactivity insertion rate (70 pcm/sec) with maximum reactivity feedback at full power j Pawer range high neutron flux 4.8 4.5-reactor trip. signal initiated 4 Rods begin to drop 5.3 5.0 Ninimum DNBR occurs 5.7 5.2 Peak water level in the 7.5 7.1 pressurizer occurs Case D Initiation of uncontrolled RCCA 0.0 0.0 withdrawal at a low reactivity insertionrate(3pcm/sec)with maximum reactivity. feedback at full power Overtemperature AT reactor 212.7 427.6 trip signal initiated Rods begin to drop 214.2 429.1 Ninimum DNBR' occurs 213.1 429.1 Peak water level in the 216.1 431.1 pressurizer occurs l L 3-18 S'

s ,J . TABLE 3.3-1 TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH RESULT IN A DECREASE IN REACTOR COOLANT SYSTEM' FLOW N-Loop N-1 Loop . Time Time-Accident Event- [sec) (sec) Partial loss of forced reactor coolant flow Four loops operating, one pump coasting down Coastdown'begins' 0.0 0,0 Low flow reactor trip 1.4 2.05 Rods begin'to drop 2.4 3.05 Minimum DNBR occurs 3.3 4,0 ) i Complete loss of forced reactor coolant flow All operating pumps 0.0 0.0 lose power and begin coasting down Reactor coolant pump 0. *e 0.7 '~ underspeed trip point reached Rods begin to drop 1.4 1.3 Minimum DNBR occurs 3.0 2.9 L Reactor. coolant pump shaft seizure ~ (locked rotor) Rotor on one pump. 0.0 0.0 locks l. Low flow trip O.05 0.07 point reached Rods begin to drop. 1.05 1.07 Maximum clad temperature 3.1 3.3 ) Maximum RCS pressure 2.9 3.0 occurs 3-19

P ' TABLE 3.4-1

SUMMARY

OF RESULTS FOR LOCKED ROTOR TRANSIDITS ~ Four Loops Three Loope ' Initially ' Initially operating' operating . Maximum' Reactor Coolant System 2583 2613 Pressure (psia) Maximum Clad Temperature (*F) 1778 '1828 Core Hot Spot Zr-H O Reaction at Core Hot Spot 2 0.3 - 0.4 (percent by weight) l i J 3-20

.i-TABl.E 3.5-1 (page1of2) i TIME SE0VENCE OF EVENTS FOR A TURBINE TRIP N-LOOP N-1 LOOP ACCIDENT. . EVENT TIME (SECS)~ TIME (SEC)- Case A Initiation of turbine trip, 0.0 0.0 loss'of main feedwater.. flow, minimum reactivity feedback with-pressure control Initiation of steam release 5.5 8.5 from S/G safety valves' High pressurizer pressure 9.2 12.5 reactor trip signal generated Rods begin to drop 11.2 14.5 Peak pressurizer pressure occurs 12.5 15.5 Minimum DNBR occurs 12.5 16.0 Case B. Initiation of turbine trip, 0.0 0.0 loss of main feedwater flow, maximum reactivity feedback with pressure control Initiation of steam release 5.5 8.5 from S/G safety valves Peak pressurizer pressure occurs 6.5 6.0 Lowelow steam generator water 50.8 77.6 j level reactor trip signal generated Rods begin to drop 52.8 79.6 Minimum DNBR occurs (1) (1) (1) DNBR does not decrease below its initial value. 3-21

TABLE 3.5-1 (page 2 of 2) 1 TIME SEQUENCE OF EVENTS FOR A TURBINE TRIP I-N-LOOP N-1 LOOP l ACCIDENT EVENT TIME (SECS) TIME (SEC) = l l Case C Initiation of turbine trip, 0.0 0.0 loss of main feedwater flow, minimum reactivity feedback i' j without pressure control High pressurizer pressure 5.1 6.7 reactor trip signal generated Initiation of steam release 5.5 8.5 from S/G safety valves Rods begin to drop 7.1 8.7 Feak pressurizer pressure occurs 8.5 10.0 Ninimum DNBR occurs (1) (1) Case D Initiation of turbine trip, 0.0 0.0 loss of main feedwater flow, maximum reactivity feedback without pressure control High pressurizer pressure 5.0 6.7 reactor trip signal generated Initiation of steam release 5.5 8.5 from S/G safety valves Rods begin to drop 7.0 8.7 Peak pressurizer pressure occurs 7.5 9.5 Ninimum DNBR occurs (1) (1) (1) DNBR does not decrease below its initial value. 3-22

TABLE 3.6-1 TIME SEQUENCE'0F EVENTS FOR AN RCCA EJECTION ACCIDENT .. l. ACCIDENT EVENT N-Loop, N-1 Loop Time (sec) Time-(sec) 1. Beginning of life Initiation'of rod ejection 0.0 0.0 full power-Power range high neutron 0.05 0.05 flux setpoint reached Peak nuclear power occurs' O.14 0.70 Rods begin to fall'into core 0.55-0.55 l Peak fuel average temperature 1.84 1.85 occurs Peak clad temperature occurs 1.90 1.95 Peak heat flux occurs 1.94 1.98 2. Beginning of life Initiation of rod ejection 0.0 'zero power-Power range high neutron flux 0.24 low setpoint reached Peak nuclear power occurs 0.29 Rods begin to fall into core 0.74 i Peak clad temperature occurs 1.90 Peak heat flux occurs 1.93 h Peak average fuel temperature 2.09 occurs u f-3-23 0657v:1o/081187

l. TABLE 3.6-2 l ' PARAMETERS USED IN THE ANALYSIS OF THE ROD CLUSTER CONTROL ASSEMBLY EJECTION ACCIDENT _ N-1 LOOP N-LOOP Time in Life Beginning. Beginning Beginning 77 t 102 0 Power Level, % 0.23 0.78 0.23 LEjected rod worth, % delta-k Delayed neutron fraction, % 0.55 0.55 0.55 Feedback reactivity 1.3 2.07 1.3 weighting-Trip reactivity, % delta-k 4.0 2.0 4.0 2.75 LF before rod ejection 2.50 q 'F after rod ejection 5.90 11.5 5.90 q Number of operational pumps 4 2 3 3778 3305 3271 Max.. fuel pellet average temperature ('F) 4335 Max. fuel center temperature 4939 3796 ('F) Max. clad average temperature 2057 2419 1911 (*F) Max.' fuel stored energy 163 139 137 (cal /gm) -r = 1 3-24 0657v:1D/081187 1 l

i; 1 1 1. , TABLE 3.,7 <a .I ..a .d TIME' SEQUENCE 0F EVENTS'FOR A LOSS OF NORMAL'FEEDWATER j i (with offsite power ~available) f N-LOOP-N-1. LOOP: - EVENT TIME-TIME'

Main feedwateriflow stops 10-10 Low-low steam generator water'. level-

- trip,setpoint reached-61 59 L - Rods;begin to drop. 63 61- . Peak water-level _in the pressurizer: occurs: 65 64; Two motor driven _ auxiliary feedwater< 121 _ pumps _.. start and_ supply 520 gpm to four steam generators 119-Two. motor driven ~ auxiliary feedwater . pumps start'and supply-490 gpm to three. steam generators ' Cold auxiliary feedwater.is delivered 467 384 to the. steam. generators Core decay heat plus pump heat decreases Approx. 2700 Approx. 1800 to the auxiliary feedwater heat removal t capacity p. l i. o w L-k e 3-25 - 0657v:1D/081187 ___w____.m__ ____.m_

L. ' ' y TABLE 3.7-2 . TIME SEQUENCE OF EVENTS FOR A LOSS OF NORMAL'FEEDWATER l L .(without offsite power available) I t N-LOOP N-1 LOOP EVENT TIME TIME I. Main'feedwater flow. stops 10 10 .l Low-low steam generator water level 62 60 trip setpoint reached Rods begin to drop' 64 62 Reactor coolant pumps begin to'coastdown 66 64 Peak water level in the pressurizer occurs 71 66 Two' motor driven auxiliary feedwater 122 pumps start and supply 520 gpm to four steam generators 120 Two motor driven auxiliary feedwater pumps start and supply 490 gpm to three steam l generators ~ Core decay heat decreases to the Approx. 400 Approx. 350 4 auxiliary feedwater heat removal capacity Cold auxiliary feedwater is delivered 468 384 to the steam generators y t' i i '-26 i' oss w o/osit u

---

____---____j

. if y. m v, a f y,,,s. . \\1 U TABLE 3.8-1 7 c. 0 1 .oj

TIME SEQUENCE OF EVENTS FOR AN

./., y INADVERTENT OPENING OF A p t PRESSURIZER SAFETY OR RELIEF VALY2 0; N-LOOP N-l' LOOP. l ? EVENT-liff(SECS)_ TIME (SEC) Safety valve-0.0 0.0 .) j fully opens t ~, Overtemperature AT reactor trip 15.5 31.1 i signal initiated Rods begin to-drop' 17.0 32.6 Ninimum DNBR. occurs 17.4 33.2 i 4 a l 4 I 1 i ) 4 4 l k 1 p 3-27 i

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c, ,e . W - gge, O-E T... .g: M f le 1, 1 18 29. 5. 19. 15, 20, 25. 50. t N.s TIME (SECONDS) ) l f FIGURE 3.1-1 I NEUTRON FLUX TRANSIENT UNCONTROLLED ROD WITHDRAWAL FROM A SUBCRITICAL CONDITION-MILLSTONE NUCLEAR POWER STATION UNIT 3 3-28 i =__ - - - - - - a

mm '.i.! I" 'I ' ..,'j. t .,, I2 f Y.- ,a +

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d'.6 -

1 a -e

g. s.

4 .5 ,p. .1 J ' e, e. 5. te. 15. 2e, 25. se.

TIME (SECONDS) 4' FIGURE'3.1-2 THERMAL FLUX TRANSIENT.

UNCONTROLLED ROD WITHDRAWAL-FROM A SUBCRITICAL CONDITION lf MILLSTONE NUCLEAR POWER STATION UNIT 3 1 I h-6'- l 3-29 ..,,j, _.__.____..m._________

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• 1,
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4. l r S.5 e

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• S.

S. 80. It. 30. 36.' W. flE eKts FIGURE 3.2-4 RCCA BANK WITHDRAWAL AT FULL POWER WITH MINIMUM REACTIVITY FEEDBACK (3 PCM/SEC RATE) 3-34

= 846 ' g.. g,, l I,.. .4 1.. .a S.S. 3. 4 5. S. IB. St. l..

38. St. 30.

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36. St. 30.

11 5 sECCl 'T ga:8. IESS.< I 54. l 5e8.' ] sis.' i l Y l D sss. FIGURE 3.2-5 j L RCCA BANK WITHDRAWAL AT FULL POWER WITH MAXIMUM REACTIVITY ste.. - 8. 3. e. s. s. 1s. 32. 14 is. se. as. FEEDBACK (70 PCM/SEC RATE) tag sEC s 3-35 j

E 85=- E 1 edH. 7688.

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36. 38. 30.

flME ISCC: . ~ IdN. b u 1500. 1888. 1862. j = iste. i 1 40. 5 g Sat.' { ?N. SN. S. 2. d. S. S.

18. 32. 14

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18. 32.

Id. 35. Is. M. TifE eECCI FIGURE 3.2-6 RCCA BM K WITHDRAWAL AT FULL POWER WITH PAXIMUM REACTIVITY FEEDBACK (70 PCM/SEC RATE) 3-36

= 8.6 ' 8.4 b.a E,, t 1 4 8 f S. 68. 898. .858. 388. M8. $88. l flE INC8 8.8 [ 8.a 8.t' b g e. - .9 .6 1' s. -58. 380. 358. 388. 368. See. 114 . ISCC l. 639.' ~

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4. -

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55. "48.

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---

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sis., E==- E 34W. o

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1. EC 8 l

1 5.31 8. 3.0 3.8 38 l>.> 3. 8.8 i t.8 l 84 8.3

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t, g, g, g, y, g, g, 33. nc inc. FIGURE 3.2-14 RCCA BANK WITHDRAWAL AT FULL l POWER WITH MAXIMUM REACTIVITY FEEDBACK (70 PCM/SEC RATE) i (N-1LOOPOPERATION) i l \\ 3-44 l A

-I. 1.4 b 1.3 .K. 3. i i I.. a B i.4 A i s.

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38 4

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68. 180. l W. 388. 353. Sec. 554. 488. 8 W. W e fl4

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1.. -

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88. IN. IM. age. 758. Sas. 854, ees. 450. Sas.

18 4 DECs

1. asas E.

,i 3.. I353. IlE. 1888. e. 0 011 r m. .a.8. SS. ISS.158. 700. 758. Bas. 354. das. #58. Eas. TIE IECs 5.2 5. _l e.. } 2.8 3.# 8.0 g,, t.4 8.3 w.in.iv.m.m. .iv. II* IECs l FIGUR". 3.2-16 RCCA BANK WITHDRAWAL AT Full F0WER WITH MAXIMUM REACTIVITY FEEDBACK (3 PCM/SEC RATE) (N-1 LOOP OPERATION) 1 3-46 - - - _ -. = _ _ _.. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _,, _

o

7.

• I q.

I g i j ,i E \\ k E 4

• [

\\ B i g e \\ w \\ m R / N / p / < ~ l f 1 I I g M I .R i I ) t. Y ^* g t 5 3 ~ l w I E ( FIGURE 3.2-17 RCCA BANK WITHDRAWAL AT FULL POWER: MINIMUM DNBR VS. REACTIVITY INSERTION RATE 3-47 pg Gr4 P- '"g" l

1 as M g g N ] \\ E-s E I N yt k s E L {W b m 4 2 g M 3 / ~ h / d / / i d. E O E e w "I l l f 3 [ l b w I a a i i i i g g ^$ k k k I k _$5..% a i FIGURE 3.2-18 RCCA BANK WITHDRAWAL AT 60% 3-48 POWER: MINIMUM DNBR VS REACTIVITY INSERTION RATE

1 O- .l. l x + ~, [ ' s. s. s ~- y g e E 3 n s r 1 e W Sa ) g t SC . O, a .x ~ 1 ~ i ~ ~, s s [. f g \\ u \\ y y 5 a 1 io r O w i t 1 i lI I L T-( a 3 ) I j i i i a R 8 4 9 M i 3 ) a a a q i FIGURE 3.2-19 1 "N RCCA BANK WITHDRAWAL AT 10% ) POWER: MINIMUM DNBR VS l REACTIVITY INSERTION RATE

I I ~ I I i .\\ l_ \\ I I "o $yE / h I

~

~ E I g g \\ k w g 1 5 S ~ I)r g \\ \\ N \\ N \\ o y i t E g \\ E \\< 3 5 n 1 ~ l ~ l l I O g g i 3 I s a a R. 8! 3 a EEE FIGURE 3.2-20 RCCA BANK WITHDRAWAL AT 75% POWER: MINIMUM DNBR VS REACTIVITY INSERTION RATE ("~1 3-50

l C2 f; 1 \\ 7 y l s e, .N S Nys k i E ) E su j "E a w g 5 e ( w s g 5 \\ l l E \\ Y\\ 5 o w h \\ ~ \\ y \\ a m T a I I I s a a R. E 4 4 1 a a a. FIGURE 3.2-21 RCCA BANK WITHDRAWAL AT 10% IN POWER: MINIMUM DNBR VS. ( REACTIVITY INSERTION RATE. l (N-1 LOOP OPERATION) 3-51 l

l' f l 1.4 " 1 l 1 1.2 o l .J BE I* ' I O *"* J$ J .s " Wo E. I; 8$ .6 " g .2 ' m g,. 3 1.4 " O J k l.2 ' (3 ooo .J E !.' N zL to ga .e-C<E E6 .6 " t *"' kg .2 " 8. O. 1. 2. 5. 4. 5. 6. 7. 9. 9. le. TIME ISEC) FIGURE 3.3-1 FLOW TRANSIENTS FOR 4 LOOPS IN OPERATION, ONE PUMP C0ASTING l DOWN 3-52

t.a ' c 1.2< - .2 4' E E-8.< - WX Eo E= q k. go. WE

• G- -

Jo UE 'Do E4 4 E ^ g,- 2400. = ~ W E' 2588. " W .{ 2200. ' i g,< 1 E =M Wgb. 2180.- l lg 2002.- E i l I908.a-1882. g. 1. 2. 5. 4. 5. 6. 7. S. 9. 18. TIMC (SECl l l i FIGURE 3.3-2 NUCLEAR POWER AND PRESSURIZER PRESSURE TRANSIENTS FOR 4 LOOPS l IN OPERATION, ONE PUMP i l C0ASTING DOWN 3-53 i j

. 4 \\l: W ,N 8.4 < - l y. D 8.2< a. .J L' W.. I.. 1 q j m3 EO: i E .4< - .J 1 Wk 3D i i' 3 .8< 40 34-1 UE .s. g. W., t .a. x g g.- 8.t < " x: D' .J I*< L -. ] F3 40

• g.

Wa 3 6 1 =JO .g. hJEd Eg. (g 4< Sk u >= .2< k S.- 8. l. 2. 5. 4 5. s. 7 3. 9. 13. Titt ISCCI i ) i j i 1 I~ FIGURE 3.3-3 AVERAGE AND HOT CHANNEL HEAT FLUX TRANSIENTS FOR 4 LOOPS IN OPERATION, ONE PUMP COASTING DOWN 3-54

5.. .. ?";t ; 'r v ,f ;.b [ :, [: o (-

9.

i 0 ~ 2.s 1 2.4 I - 2.2- - 2.< 3.0-9 t 1.s-l.4 < - i< 1.2^ t k, g g, 2-5. 4 E'. s'. e, p. i' I TIME (SEC) f' b b. FIGURE 3.3-4 DNBR VS.11ME FOR F09R LOOPS IN OPERATION, ONE PUMP C0ASTING DOWN r-I: 3-55 g. i

g> a !.4 < .J 1.2< .E ~ 3 1.< Bo I i oa .J6 k

• g, O

WE E O O g p .6 < q U4 { .4< .2< ) 1 1 i 8.- s. 1. 2. 5. 4 5. E. 7. 8. it. 18. l TIME (SEcl l l l j I FIGURE 3.3-5 CORE FLOW COASTDOWN FOR FOUR I l LOOPS IN OPERATION, FOUR PUMPS C0ASTING DOWN l 3-56 (. I

3.2 < 'O 4 EI e k. +8' gO WE .J O Uy 4< D U E<g .2< 8. 2489. " W E 2500." R= k q'2200. g $ 2198. i .= S. 2000. " W k 1900. gggg,_ 8. 3. 2. 5. 4 5. S. 7. 8. 9. 18. TIME ISECl FIGURE 3.3-6 NUCLEAR POWER AND PRESSURIZER PRESSURE TRANSIENTS FOR FOUR LOOPS IN OPERATION, FOUR PUMPS 3-57 C0ASTING DOWN

M t.2 " DJ- ' d, r i.. <= wa zoz .e a j wA zo E .6 o g 5i u wU 4' e< $E \\ .t v l I ^ S. ^ t.2 " x _J D4 .J gE

1. "

3 g 4o wz z .t " 1o wzE s o zo 4 EE oy ,4 Fz EU I 8. ^ ^ ^ ^ ^ e. 1. 2. 5. 4 s. 6. 7. e. 9. 19. j TIME ISEC) l FIGURE 3.3-7 l. AVERAGE AND HOT CHANNEL HEAT FLUX TRANSIENTS FOR FOUR (. LOOPS IN OPERATION, FOUR PUMPS [ C0ASTING DOWN l 3-58

I i ) i i 2.s < - i t.e- - d 3.3< 3.< ~ i i l.8 i t 1.6-I.e < - let^ ^ ^ 8. I. 2. 5. 4 5. s. 7. 8. 9. 18. T!WE (SEC) FIGURE 3.3-8 DNBR VS, TIME FOR FOUR LOOPS IN OPERATION, FOUR PUMPS COASTING DOWN 3-59

,' J' 4

1. d "

1 I 1.2 l 1 f 1.--- a* Ls E .g. ,,J M to g . g0 .6 " 8.U

• g 4

4-I w E .2- >0., 1,4 1.2*' 9 gi 1.* oO J Z 64 .6-O d 4 ,s o E .J u. .t.. .2= D. 8. 1. 2. 5. 4 5. 6. 7. S. 9 Ie. i TIME (SEC) i i FIGURE 3.3-9 FLOW TRANSIENTS FOR THREE LOOPS IN OPERATION. ONE PUMP C0ASTING DOWN 3-60 L

1. A < -

c

1. 2 "

1.- Em M .9" g k. $O, .6 < WU .J 4g .2 " A g.- 2400. a-Wg' 2523. " m me... 4 E5 WN CL 2123.- 5 m( 2000 a-E A. 1922.-- 18C3. l 2. 1. 2. 5. 4 5. 6. 7. 8. 9 10. TIME (SEC) I. \\: ( FIGURE 3.3-10 NUCLEAR POWER AND PRESSURIZER PRESSURE TRANSIENTS FOR THREE LOOPS IN OPERATION. ONE PUMP OASTING DOWN. 3-61 l. C__.---_._-._

1 I .1 0

1. d "

I 1.2< q m a

1. "

b d .8 v AVERAGE CHANNEL k. ~ .6 " M3 .4 - 4 W .2 '

1. e '

? 1.2 ' E B

i..

d 4 HO' CHANNEL oc 'g.. k. w x36 k. >q .a Wz .2 " 1. 2. 5. 4 5. 6. 7. 8.

• l.

10. 0's. TIME (SEC) L E i FIGURE 3.3-11 [- AVERAGE AND HOT CHANNEL HEAT FLUX! f TRANSIENTS FOR THREE LOOPS IN 'i ? OPERATION. ONE PUMP C0ASTING DOWN. I-3-62 i L i

_,e'l I i .d I J f O >. i._ 2.6 o ? - 2. 4 < 2.2 p.. E. m 1.8 < . g. o J 1.6 " 1.4 <. g,p - 2. 1. 2. 5. 4 5. 6. 7. B. 9 10. TIME (SEC) 4 i FIGURE 3.3'12 DNBR VS. TIME FOR THREE LOOPS IN OPERATION. ONE PUMP F. C0ASTING DOWN l 3-63

I e l' s

1. 4 <

1.2-- BC D4 1 <~ E 'W"8O .8 g mM L go .6-- N N E 6 >U U44 WE .2-E.E. B.- .9. 1. 2. 5. 4 5. 6. 7. 6. 9 10. TIM E (SEC) i FIGURE 3.3-13 CORE FLOW C0ASTDOWN FOR THREE LOOPS IN OPERATION, THREE PUMPS C0ASTING DOWN. 3-54

I i 4 'O I. 4.. ( W i.2 1.- g E- .8 " a. 8 .6 " 3: 4WD d *E s. .2 ' g, 2422. 25C3.-- m m W 2223 l 2 .: a I WE 2100.- g m reco.. W tt'E i 18t23. l 1833.- 2. 1. 2. 5. 4 5. s. 7. 8. 9. 10. l 4 1 TIME (SEC) L- ) L k FIGURE 3.3-14 i L NUCLEAR POWER AND PRESSURIZER I I PRESSURE TRANSIENTS FOR THREE I LOOPS IN OPERATION. THREE PUMPS C0ASTING DOWN. 3-65 1 1 ] E

{ I.4 - f 1.2 - - 1.< - m . x h

k. k.

- k. O AVERAGE CHANNIL N <g .6< 4mh4

k

.2-2.-

1. 4.-

g.2, a e I 2 Xo j 2= ,J h, 40 8" HOT CHANNEL g q u-1 4 i E .6" b .a. .2-B * ', ^ ^ ^ ^ ^ e t. 2. 5. 4 5. 6. 7. B. 't. 30-TIME (SEC) \\ l \\' FIGURE 3.3-15 AVERAGE AND HOT CHANNEL HEAT FLUX TRANSIENTS FOR THREE 1 LOOPS IN OPERATION. THREE PUMPS COASTING DOWN. 3-66 ~

p .it-e' ( .4' L t 2.6

2. 4 <>

'2.2 l 2.< l R l. 8 1.'s.. E' O 1.6 - t.A< -

1. 2 8.

1. 2. 5.' 4 5. 6. 7. 8. 9. te. TIME (SEC) l FIGURE 3.3-16 j DNBR VS. TIME FOR THREE LOOPS j IN OPERATION. THREE-PUMPS C0ASTING DOWN. 3-67 ___-_-_-________a

I.2 " WITH l DFFSITE ~ \\ POWER g \\

• 3

.8' O Oz Nk O .6<- ggE ~ OS WI1HOUT "g d" 0FFSrTE POWER E .u. 3.4 3.2 " .J 1.< R 4 .e.\\ O x aA 2 O E g .6 OO k .J O O E W O .2' WITHOUT b 5 A FSITE g U C.' POWER y 4 4 .2< k 4<- m WITH OFFSZTE POWERA .6 ~ ~~~. 4 5. 6. 7. 9. 9 12 8. l. 2. 5 TIME (SEcl FIGURE 3.4-1 FLOW TRANSIENTS FOR IN OPERATION, ONE LO 3-68

l 2800. " 2700,a 2600.- - W wrTNovT OFFSITE POWER-2500.- - A w E-s $ g< 24 e 0. - wrm g a. OFFSrTE \\ P0htiR \\ ^u~ 2500.- - z i Q 2200.- = l 2!00. - i I 2030.- B. 1. 2. 5. 4. 5. 6. 7. 8. 9. 10. TIME ISECl l 1 FIGURE 3,4-2 PEAK REACTOR COOLANT PRESSURE FOR FOUR LOOPS IN OPERATION, ONE LOCKED ROTOR 3-69

] 'e j { i. ,j E s, O l E E .S< 4 N i .4 < - 5

i d'

q i E l E 3 g( l.2 ' WE N I gj s., O. l d2 .e < ES 'j EO .s. 4 4' =5 (g E h s. l l \\ 8.>- g4 8.< N gg ea EEE (- Eg U .s. - E s. E" 8. 8. 8. S. 4. S. 6. 7. 8. 88. 3 I TIME (SEC) l I FIGURE 3.4-3 NUCLEAR POWER AVERAGE CHANNEL AND HOT CHANNEL HEAT FLUX TRANSIENTS FOR FOUR LOOPS IN l l ' NOTE: OPERATION, ONE LOCKED ROTOR l $ WITH AND WITHOUT OFFSITE POWER AVAIL ABLE 3-70 \\

); t i l l l I i

j i

1 l 5828. 2758.' 2582. L' WITHOUT h 2250. < DFFS%TE n eowa gy 2800. < We EW 1750. E==.c

• % 'e %

1588. yyy g arrszre J ~~~'~ n U 1258* pagm 1800. 750. o 500. 8.- 1. 2. 5. 4 5. 4. 7. s. 9. Ig. TIME (SEC) I i FIGURE 3.4-4 1 MAXIMUM CLAD TEMPERATURE AT 1 HDT SPOT FOR FOUR LOOPS IN { CPERATION, ONE LOCKED ROTOR 3-71 m

l 1.- l< l-1.d

1. 2 < -

D 1.<- I od9 WZTH POWER y f _ ,g,, WE mm4 WO .6 > u. w E4 -==., OE 4-pg u ~' WITHOUT POWER qy .2 < 1.4 1.2-- l E 9 I.- 2E O .8-u, E O 4 WITH G WZTHOUT POWER N .6, uE D k. 2" 4 .2-1 2.- 8. 1. 2. 5. 4 5. 6. 7 e, 9 10. TIME (SEC) i FIGURE 3.4-6 FLOW TRANSIENTS FOR THREE LOOPS IN OPERATION. ONE LOCKED ROTOR. i l 3-72

l l o 1 ^ b d i 2003.<> 2723.. 2600.-- WITH3UT POWER 2500..-> '* s g2 N-N g E 2dC3. - s N .- 8 s 2500. - NZTH POWER 22C2.- f 2 sea.- acc:. 2. 1. 2. 5. 4 5. 6. 7. 6. 9 10. TIME (SEC) FIGURE 3.4-6 PEAK REACTOR COOLANT PRESSURE FOR THREE LOOPS IN OPERA 1 ION. ONE LOCKED ROTOR-l l 3-73

l.d " e 1.2-1.< WZTH C WZTHOUT POWER x$ .. ~ =g !E m yo .s. WN AVERAGE CHANNEL $[ 4< - U~ .2 ' 0.- i. e.. 1.2 ' 5, 9a s.. ko WITH C WITHOUT POWER i 8 p .B o Wk HOT CHANNEL U~ .A .2-g, l E. 1. 2. 5. 4 5. 6. 7 6. 9. 10. TIME (SEC) FIGURE 3.4-7' NUCLEAR POWER AVERAGE LHANNEL AND HOT CHANNEL HEAT FLUX l TRANSIENTS FOR THREE LOOPS IN OPERATION. ONE LOCKED ROTOR. 3-74 I L_______ i

= us 5022. " 2750.- 2522. " h. 225E. ' E NZTHOUT P0nER { 1752. 1520.- wZm ponen = i,5... E-O 1220.- 4 J 752.- o ^ ^ 500,- 2. 1. 2. 5. 4 5. 6. 7. 8. 9. 12. TIME (SEC) 4 4 I i ) FIGURE 3.4-8 li MAXIMUM CLAD TEMPERA 1uRE AT HOT 2 f SPOT FOR THREE LOOPS IN OPERATION. ONE LOCKED ROTOR. l i f 1 3-75 l b-

') i { amm. ' d s.a < - I's.< A l .g 15 4 " k .2 g s.- +. I = 2423. P6DB. ' ' i& ^ ^ E 2222.' \\ 2822. 1882. ~ ^ 1 j 1488. " 'b"l 1288. " kl t583. " h ly ^ ^ ^ ^ ~ ^ see. " 682.

58. 88. 78. 82.

'W.*ISB. 8. 18. =2.

58. 43.

TIME ESCCI FIGURE 3.5-1 TURBINE TRIP EVENT WITH PRESSURE CONTROL MIMINUM REACTIVITY FEEDSACK 3-76 i I

l 1 i iC 600. ' $4. I E. 580.< 578. " Est.< Og lit.< Eng. 425. o i j W 828.< l k. j 0e ses. < W D E* 582.< f N f Est.< I Eng. 1

4. v 5.5 <.

5.' g e 2.5 a D t. 1.5 <. j T.- , S.- 38, 22. 58. 48, le. St. 73. 33. g,

ggg, TIME e5ECs 1

i. l FIGURE 3.5-2 TURBINE TRIP EVENT WITH PRES 5URE CONTROL MINIMUM REACTIVITY FEEDBACK 3-77

i ) 3 ll..if Il d< 2' ~ 1 3 ~ 8.~ l 2622.<- 2422.< - 2222.* { 2222.' ^ 1822.^ 1882.=- ~ 1422.' 1282.' ISta.a 828. " $22.= ^ ^ ^ 8.

18. 22. 82. 45. St. 42. 78. St.

4. 128.

?!ME REECs FIGURE 3.5-3 TURBINE TRIP EVENT WITH PRES 5URE CONTROL 3-78 MAXIMUM REACTIVITY FEEDBACK

i 688. e

54. <

[ g L g 5 y ses,.- x gs7s.- ~ 548.- I ] 458.< gag,- 34 3... stb.- E g g 822.< f gW D h,EE2. ~ Est.< Edg.N ^ ^ q l

4. o 5.5 '

5.- tE 3.5 E-O 1.s ~ g,- S. ts. as. se. 4s. se. se. Ts. es. w. ise. l TIME I5CC FIGURE 3.5-4 ( TURBINE TRIP EVENT WITH PRESSURE CONTROL 3-79

I "1.t" 3.4 ) 3.' 9 Igl.6< 1 ,.7 ' - k l i 7620. " 2482. " { 2228.' E j , 2888.' t 3322., ^ ^ ^ 1488. o l288.< 1888. 0 l 388. t s sea. 'S.

18. 29. 58. 48. 58. Se, 78, 33. ens, leg.

TIMC 45tc) FIGURE 3.5-5 i TURBINE TRIP EVENT WITHOUT PRESSURE CONTROL MINIMUM REACTIVITY FEEDBACK 3-80

s,.... i sup g sat.a gtiitsys.. I 568. " s- !I f s58.< E! sas.- 4 842. " tg 6tt.< 682.' :., [g.g .D, 588.' D 558.< - ses. ^ ^ e.<- 5.5 < 5.< t.5 < 3.< ~ f.s c e ~ g.. 5. IS. 28. 58. 49. ss. St. 78. es. 4. Iss. l TIMC tEEC FIGURE 3.5-6 TURBINE TRIP EVENT I WITHOUT PRESSURE CONTROL MINIMUM REACTIVITY FEEDBACK i 3-S1

l.4 < > fle 'I" I... .8 ' 4< .2 ' ~~ S. l M88. o r i 3488. " lE ^ ^ ^ 2288.= g 8 w H88.< 1838.- 1488. " 1 1398,a g - 1988. < i l,;g33,:,' 808. o -m S. St. 30. 58. 48. 58. SS. 78. St.

4. 198.

Tipt ISCC FIGURE 3.5-7 l TURBINE TRIP EVENT WITHOUT PRESSURE CONTROL l MAXIMUM REACTIVITY FEEDBACK l 3-82

l i i l 592. ' - y l sg." W f 588s ' W S 578.' - u8. < 558.' ~ 548.J. 1 549. o 1 $g $25.< C, 1 E ses.' EO WW &O 588.' E-588.' 1

4. e 1

I 5.5 ' ' 5.' ' g s E l 3.< 1.5 ' ' ~ ) h, 43, 59. 30. 78. St. ag, 333. la,* gg, gg, TIrg 85CCI FIGURE 3.5-8 TURBINE TRIP EVENT WITHOUT PRESSURE CONTROL MAXIMUM REACTIVITY FEEDBACK 3-83

3.4 < l E.t v

s. J l

..S< # gg i .s. d 4 E, t 1 w C. 1 2488. 2688. < a I_2288.< l 2888.< 8888. I480. < 1 8. IB .88.. 488. ^ ^ ^ ~ ^ ^ 8.

18. 28. 58. 48. 58. SS. 78. 88. age, 3g3, 11tC ISECl FIGURE 3.5-9 N-1 LOOP I

TURBINE TRIP EVENT WITH PRESSURE CONTROL MINIMUM REACTIVITY FEEDBACK 3-84

648. < W E29.< ; i E t e ass. " l W 8 9 E,588.< 549. SS8. < SM. < lE Ste. ' e0 578. " w E$* 568.< 558. e 548. 4,<- 5.5 - g

5. "

e E 2.5 " D 4 .5 8. 8.

10. 29, 58. 48. 55. SS. 78. 88.

4. 188. I1MC ISCCI FIGURE 3.5-10 N-1 LOOP TURBINE TRIP EVENT WITH PRESSURE CONTROL MINIMUM REACTIVITY FEEDBACK l l 3-85

4 e 4 648.<' 588.< ~ g 1 "' '/ i 588.' 588.< gag,- $88. EM. < lg 588.' g ,,8.. E N 568.< - ss8.a ^ ^ ^ E48.' l

4. <

5.5a 5.< - t* 3.E ' - g.-

a. n s.s' 8.-

8.

38. 28. 58. 48. 58. 88. 78. 53. etB. 388.

TIPE 8SCCI FIGURE 3.5-11 N-1 LOOP TURBINE TRIP EVENT WITH PRESSURE CONTROL MAXIMUM REACTIVITY FEEDBACK 3-86

i E E E.t < 3*' .8< 4< .2' - g, 1.8488.<- 2688. " N E Iw&, E 2288.< 2988.< ^ r l 1498. ~ b 1298.</ 1988.< g ,U,,, 888.a n = ^ as. . u. 48. u. u. n. a. =. sa. 11tE ISECB FIGURE 3.5-12 N-1 LOOP TURBINE TRIP EVENT WITH PRESSURE CONTROL MAXIMUM REACTIVITY FEEDBACK 3-87

ij t-

• ti. <

u I.s- .~ .9< L I*2$38.<> { ^ 8.- 1 ~ 2488.< l 5 4 2288.< g

• 2888.<

,4 1988.- ^ ^ ^ ^ ^ ~ t488.<- 3298.< 1988. " l,333.., ^ ^ ~* 0 ^ 588.- 9.

38. 28. 58. 48. 58. 88. 78. es. eig 333.

TIPE (SCC 8 FIGURE 3.5-13 N-1 LOOP TURBINE TRIP EVENT ) WITHOUT PRESSURE CONTROL 1 l MINIMUM REACTIVITY FEEDBACK 3-88

r-i H 548. o td t- $28.< a E h 588.' a. N s98. " l l

1. .(

ss8. " 1 ~ ^ ^ ^ ^ ^ 548. 888. o gg 5%. < ~ bg 588.<

g C. s t J

~E .t 578.* t 3* 5s8.< ' ggs,.. s48.- ^ ^ ^

4. "

) 5.5 " r

5. "

2.5 " [

a. "

l.E ' ,"g.-

18. 29. 58. 48. 58. 88. 78. 88. ate. 188.

TIME ISCC) FIGURE 3.5-14 N-1 LOOP j TURBINE TRIP EVENT WITHOUT PRESSURE CONTROL MINIMUM REACTIVITY FEE 0BACK 3-89

l l \\ h I.4 < - I< I 4* .t* g,. = 88.. 3488.< I_g ar88.- \\ f 3888.< $888. t488.<- 1288.< O 3888.< I.,

• 888...

i ~ 388.- SW. 78. 88.

• 18.

188. S.

18. 38. Es. 48. 58.

TIPC ISCCI I FIGURE 3.5-15 N-1 LOOP TURBINE TRIP EVENT WITHOUT PRESSURE CONTROL MAXIMUM REACTIVITY FEEDBACK 3-90 1 1

o 548.<- g

a...

0 48 E 758.a k " h. 5.... I r E* N 568.. 'l 'gsg.- Sea. 4 5%. = !";y It: 3g< $38.' E t me sys.<. ] ( 3-589.- N ss... 54 8. --- { t } 4.<- s.s ' s.- 1 E i E 3.< 1.5' L g. IS. 29. 58. 45. 58. Si. 75. Si. eig. gg's. Tilt IStc) i. l I 1 ( FIGURE 3.5-16 N-1 LOOP TURBINE TRIP EVENT WITHOUT PRESSURE CONTROL l MAXIMUM REACTIVITY FEEDBACK l 3-91 lL_

ja i; }' ,1: ? 5. 2.5 ac W 2. on. ' 5 5 1.5 i. .5 8. B. - 1. 2. 5.. 4. 5. 6. 7. 8.. 9.

18..

11MC 1SCCONDSI l FIGURE 3.6-1 1 NUCLEAR POWER TRANSIENT BOL ETP ROD EJECTION ACCIDENT i MILLSTONE NUCLEAR POWER STATION UNIT 3 1

q i 'Il S

E J

C 6000. < MELTING 4900*F 5000.- C FUEL CENTER-TEMPERATURE g 4eBe. g Er ' 5t 50ee. w -FUEL AVERAGE ~ TEMPERATURE '2000.- 1980.- CLAD OUTER TEMPERATURE 8. 2. 5. 4. 5. 6. 7. 8. 9. 19. 8. 1. . TIME ISECONDSI 'I )'i. I L FIGURE 3.6-1 I HOT SPOT FUEL AND CLAD l< TEMPERATURE TRANSIENTS BOL HFP ROD EJECTION ACCIDENT MILLSTONE NUCLEAR POWER STATION UNIT 3 .3-93 CEE._ ___ _.

1 2800. *l d l c G 2600. i a l l ~ w w,2800. ' / G Y, n.' m 2200. H._ L $ 2000. i N j a-1800. 1 j 1600. l l 2000 1 i i G. 1803. b I 1600. i (1400. ao 1 a 1200. w y,1000. a i M 800. ~w5 600. m j wt 400. 200. O.- 100 101 1:2-105 3c4 TIME ISEC l'IGURE 3.7-1 PRESSURIZER PRESSURE AND WATER VOIIFE FCR A IDSS OF NORMAL FEEDWATER 3 4 3-94 1

1. 4 ' -

e }1.2 g 1. m E a-- l .6 e ~ sw S' oE. .6 E~ 0 4 .2 0. 1.4 6 1.2 z 'o M

1. '

a 6 '~ .8 w p .J w .6 'Y W 4 u .2 4 3ge gg1 tc2 Ig5 10 TIMC ISCCI l FIGURE 3.7-2 mza nERne = HEAT FLUX FGt A IDSS OF 1Ge%L FE! DETER l 3-95 1.

4 11 703. <

• 600.

'660. 640. IE7f I 620. l 600. 580. ) H560'

• t" 1

5 0. 520.. d gc0 101 102 gg5

c itME.

I SEC I. m FIGURE 3.7-3 RCS }DI AND CDID IB3 TDGUATURES FOR A I.OSS OF NORMAL FEED &TE 3-96 . 1

J ( 1 i 1600. _a l G 1400. ,1 4 1 -{1200. m .m 1 W 1000. 1 c. g 800. 4i C i j aw$ 600. J, e 1 r. 400.< -m 200. D.- .14C 6 < 2 12E 6 co - 2 a I f m.10E 6 e l E l ) E.80E 5 y ) mw 5 60E 5 e rcw.40E 5 .m / LOOP WITH l .20E 5 PRESSURIZER i D.- 4 100 101 102 gg5 gg TIME ISEcl I .) I 1' l l l FIGURE 3.7-4 STEAM GENERATOR PRESSURE l AND MASS KR A IDSS OF NORMAL FEEDaTER l l 3-97 Q

s. 1.a 'd. I 1.2 z g 1. ~ ve. g. .O 6 >o 8 .6 u mu .,d. ac O '~ .2 3. 1

25. <

8 20. m i s sn

• =

b I5. uw ~ l0. { w i u 1 &w 5* L 5 i m O. i e uaA .-5. ~ gge g31 102 105 10' TIME ISEC1 j o. i i ) J 'l FIGURE 3.7-5 RG FIDW AND PRESSURIZER l Frm PTR A i IDSS OF NCRRL FEEDETER 3-98

2600. E. Di2600. w 2400. m$' / I4 .E a 2200. O_ %g 2000. E' N o. 1800. 1600. \\ 2003. ~ 1800. 6 ~ 1600. u i 5 sdee. a S '4 a 1200. \\ U .J $ 1000. j u J H.900. J ~ j 8 600. l] O '.. g 400. i 200. D. 100 103 102 105 10 TIME ISEC1 1 FIGURE 3.7-6 PRESSURIZER MtESSURE AND WATER VOILNE FIR A IDSS OF NDRMAL FEEDWATER J (N-1 IDOP OPERATICE) l 3-99 1 __.______________________J

1 I ? t.d i .p h 1.2 '2 6o I,

1..

W= .M g .8 u y!o-S .6 acw s 4 .2 1 0. 1.4 j { 5 1.2 2: L 4 o i 1* u ] e" 6 .8 J ~ MD s 6 .6 ,_c Y w 4 &Ou .2 \\g 0 100 101 102 105 10 j d TIME ISEC) J J . s J l J FIGURE 3.7-7 NOCEAR POWER AND CDRE HEAP FUJX FIR A ILSS OF NORMAL FEEDUCER (N-1 ICOP OPERAT15 ) 3-100 l

9 1 700. < \\, 680. 660. F

• 640.

ICI I620. 600.

500, CDID 560.

SAB. l 520. 4 100 101 102 1g5 10 ~ TIME (SEC1 FIGURE 3.7-8 BCS IUr AND CDID IJE TDPERA2URES FOR A i-IDSS OF NOR91 FEEDWATER (N-1 ILOP OPERATICN) 3-101

1600. -e-m 1400. I E b g1200. i m m W tee 0. a. ! 900. c a $ 600. o r I a 400. .y w 200. 1 0. l .14E 6 < E.12E 6 5 i m.10E 6 m 3 a i r \\ i E.BOE 5 f E La $.60E 5 o x 6 40E 5 i x m 3 w .20E 5 ) l gg0 gg: ig2 gg5 ted j TIME ISEC1 FIGURE 3.7-9 srEm coexxa msSuaE 1 AND MASS FtR A IDSS OF NORMAL FEEDOCER { (N-1 IDOP OPERATION) l 3-102 l l t

l 1.4 _y \\ ~ 1.2 ~ g 1. ~ E .e ( ~ 2 C) d .6 d

• A 5

.2 O. l l

25. <

d 20. m E 15, ww ~3 10. U w M 5. )g e. u Of CL -5. 100 301 102 gg5 10' TIME ISECl FIGURE 3.7-10 RCS FICM AND M 9URIZER RELTEF FOR A LOSS OF }ORMAL FEEDETER (N-1 LOOP OPERATICN) 3-103

1 i 1 20:30. 'l - .j I G 2E00. n u $ '2400. ' a s ) c. a 2200. ' 4 { w$ $ 2000. 8l a.- 1800. i 1600. 2000. ~ 1900 b ~ 1600. u 5'i400. S a 1200. w @ 1000. a M 000. h 600. m ug-400. 200. g'- 4 tg0 tel 102 105 tg TIME ISECl FIGURE 3.7-11 PRESSURIZER PRESSURE AND WATER VOI1EE FtR A IIES OF NORf%L FEEDWATER W1'DEUr OFFSITE PCHER AVAIIAEE 3-104

1. 4 '

) E i.2 z N E'I. = 6 I ~~ .6 =- Mo S- .6 'g. 1*g. 4 x i .2 3. 1.4 E I.2 z o.y u 6 ~ .8 u 3 w-6 c Y g .4 O U .2 l 4 t' gg0 . Ig! 102 gg5 30 TIME ISECl l-I I FIGURE 3.7-12 NUCIEAR PNER AND CDRE (. MFAT FIUX FIR A (. I.OGS OF NORMAL FEEDWATER i MWI' OFfSITE PCMER AVAIIABLE 3-105 1

l 4 i s. a 700 <

602.

I 662. F l ~ 640. wr l I 622..+ -i 600. 500. 560. 540. COID 520. d 500. ]c0 101 102 gc5 .!C i TIME ISECl l J l FIGURE 3.7-13 RCS wr AND CDID IE TMPERNIURES PIR A IDSS OF NORfRL N WrncJr OFESTrE 3%'ER AVAIIABE 3-106

1 1600. ~c l G 1400. c y1200.' wm W 1000. a. I 5_ 800. El e EE $'600. e i r 6 400. G 200. 1 g,,__ 1 i .idC 6 2 12C 6 3_ m.10E 6 E 1 l E.BOE 5 5 cm k.I 5 60E 5 e r 8.40E 5 LOOP WITH [ PRESSURIZER ) .20E*5 1 1g0 1g1

02 tg5 1e

{ d g*-- TIME ISCEI l i k t FIGURE 3.7-14 STEAM GDERATOR PRESSURE AND MASS FCR A IDSS OF NORMAL FEEDETER WIDUJT OFTSITE POWER AVAIIABIE i 3-107

1 ] 3 1.4 E o ~ 1. 2 -z g 1. y .t. .8 g u >c 3 .6-w U

)

4 4 ) .s l c -o i ~ .2 D. l j -n J 1 . 25. < i S 20. w %m 6-15. 1 u 1 w 3

10. <

wa a-u 5*< ~ E w 0. u r v u D. S. i 4 l 100 101 3g2 gg5 10 TIME ISECl 1 l FIGURE 3.7-15

f. '

RCS FIDW AND PRESSURIZER Imf M FtR A '] LOSS OF HORMAL PEEDWATER i WrDOUT OFTSITE ICWER AVAIIABLE I 3-108 I J

r_ I' 2800. V 9' .7600. l o. i f w l 5 2d00. E - O l u 2200. k N_ 4 .i o 2000.< O E n. I800. J 1600. 2000. ~ l800. Le ~ 1600. 1400. g w 1200. s u.- % 1000, u M 800. ~ m 5 600. m-ug 400. 200. B. 100 le! 102 gg5 3c4 TIME ISEC3 I; FIGURE 3.7-16 I PRESSURIZER PRIESURE AND WATER VOI1NE FtR A { IDSS OF loi@RL TEEDWATER 1-WrDCUT OFFSTIE POWER AVAIIAEE (N-1 IDOP OPERATICN) 3-109

l t I ' g l.4 n l.2 b - I - g 1. i a 6 ~ .G J l a i W 1, 3 . o -t 3 3 n. .6 , \\ :t i-1 4 g .,1,. -( ) 1 s w g 4 't 4 \\.' s .2 \\ 4

\\

-1 ..._..w 'W h 4 8. $..s .. \\ s 3 ^ g h ; . \\ \\

• \\'

.) \\ e y\\ i.4 - p,e 1.2 . 4I 3 ,3 l i 1 4 g 4 x' i-s g a g x ? = 6 g s g '\\[ o [8 - [' x. f t.t. - h r ~v s s s J s - .d 6 .6 s, g1 rh 1 g. a 1 '4' - \\. R.. g \\ i ) .4 4 w x g s u t i t y4 8 ( ,r . 2 "L 3 a_. m.=seman m, q i g, '( ggd .t 100 103 ' s. [ '!B2 gg5 .? h. .t 3 s

ISECl, a,

p' tilME 1, ~ s a = 5 (+ s m. .c ,t h-j PYME 3,7-17 = F RAMMI 7%??; AND CCF7 s I' e, ICAT P107.PCP A ) 'j IMS OF NXT%L FEEDhWA "' M"DIXE OFMITE IOiEk A\\%IIANJ. 4 i (N-1 ISOP "nERATILW). 4 i 3 z 4 3-110 s t I r g' i 7 s 4 .3 [ ] t s

t i T e 700. t 683. k-660. [6,o. 623. }gyr t 600. 1 500. l l s- ' i' ,1,_ 56 h,0. s s s.. CDID S20-503. 8 g i ig2 135 10 ? TIME ISCCI a .-)[ 's,. 's l/. q4b FIGURE 3.7-18 ? ' l' b s 343 }ct AND CDID IEG TEMPERNIURES PCR A IDSS OF NORMAL FEED ETER l 6 WITHCETT OFFSITE POWER A\\M N (N-1 IDOP OPERATICH) 4 I i 1 \\ l 1 1 1 3-111 g e <o l YL, f t Q;_,

L l 1600.< i! e i G 1400. b y1200. M W 1000. a. B. 800. .a CE E' 600. bi b 400, a 200. i 0. I i .14C 6 i l 2 12C 6 i } e ~ g.10E 6 E E.00E 5 E cxw $.GOE 5 t w ] l r 5.4et.5 E .20C 5 t-l 0. 100 101 102 gg5 ggd j TIME ISEC1 r FIGURE 3.7-19 STEAM GDERAICR PRESSURE l AND MNSS FOR A ] IDSS OF lOFe%L FEEDGGER WIT 10l1T OFFSITE POWER AVAllABE (N-1100P OPERATIm) 6 3-112

e 1.4 d

s.2

'z~ $ g

1. -

W. 5..e >o d .6 wu" Y .4 5 .2 0. 25. g 2a. v> m E 15. L t.J

• 3 10.

Y a 5. M.- f~ = U. u n. 5. 1 l 100 101 102 105 gg4 {

10..

TIME ISEC) l FIGJRE 3.7-20 RCS FLOW AND IEESSUPl2D, REU!IF PGt A LOSS OF IGMAL ITEDW7CER WID00T OFTSITE PCHER AVAIIABLE (N-1 IDOP OPERATIQi) i 3-113 -l j

,1.4 8.9 b I. j s .9 I.. .3 ( '8.

5. Is. 35. M. 35. M. 95. 40. 45. Es. 85. 88.

1IPC GSECl ese.' SN. (S00. 5 I. Sat.

s. 5. SS.15. N. 35. as. 55. W. 85. 88. 55. M.

. gift S$[C FIGURE 3.8-1 INADVERTENT OPENING OF A PRESSURIZER SAFETY VALVE 3-114

i SM. l SNC. E $388. I1838.' !l 1 gless.- i a i 1302. 9988.< 988.8.

5. 88. 15. 30. M. 58. M. 48. 45. 58. 55. 88.

TIE IKts 1 1 I l / 4

4. <

5.5 5. 1.5 2 3. I.5 - S. 5. Is.15. n.35. M. M.as. 85. 58. 85. es. TIE ' SEC8 FIGURE 3.8-2 INADVERTENT OPENING 0F A PRESSURIZER SAFETY VALVE 3-115

p i 4.a < B 3.3-I a. < .g 4< g .t< k S. S.

5. le. 15. M. 35. as, gl. 40. 45. 60. 55. 80.

11 4 iMCs Sat. 580. E560.< saa.< 548. 529. See.8.

5. 18. 15. 30. 35. 54. 55. as 45. 58. 85 es.

fl* IECCs FIGURE 3.8-3 INADVERTENT OPENING OF A l PRESSURIZER SAFETY VALVE (N-1 LOOP OPERATION) j 3-116

l 1 P-l I i f E ** ' ases. < l _. u 1 sees. i ises.' l 48. j "'s.

5. IB.15. 38. 36. 98. 55, es. al. 54. 55 11C IECl I

I 5.2 ' 5.< 4 i R.t ' 3.6 ' l s.a 2.I' -t I.t ' 8.8 J 8.a ) s.t

5. 38.85.38.35.as.35.as.45.Es.85.88.

I* S. Tit (KC 8 i 1 i FIGURE 3.8-4 INADVERTENT OPENING 0F A PRESSURIZER SAFETY VALVE (N-1 LOOP OPERATION) 3-117 i )

RWST/ Accumulator Boron Concentration Increase Evaluation 4.0 4.1 Introduction The large break Loss-of-Coolant Accident (LOCA) Analysis for Millstone Consequently, to Unit 3 takes no credit for control rod insertion. w maintain the validity of the analysis, it must be demonstrated for each cycle that the core can be maintained suberitical via boron addition fr 2 3.0 ft. This post-LOCA the ECCS in the unlikely event of a LOCA 3 However, as a result of shutdown requirement has been met for cycle 1. the Cycle 2 core design (PMTC and longer cycle length) this requirement In order to achieve is not assured with the present plant design. l adequate design flexibility for future cycles, an increase in the boron concentration range to 2300 - 2600 ppm for the RWST and 2200 - 2600 ppm The following evaluation is provided for the accumulators is proposed. 1 to address tha impacts of this proposed change. 4.2 Non-LOCA Safety Analysis The RWST, accumulators, and the Safety Injection System (SIS) are Upon actuation of the subsystems of the Emergency Core Cooling System. SIS, borated water from the RWST is delivered to the reactor coolant system in order to provide adequate core cooling as well as provide sufficient negative reactivity following steamline break transients to The accumulators are a passive system prevent excessive fuel failures. and provide borated water to the RCS when the system pressure drop I approximately 600 psig. The only non-LOCA safety analyses in which boron from the RWST or accumulators is taken credit for, or assumed to be present, are those inj which the SIS is actuated. These analyses are: l [ Inadvertent Operation of the Emergency Core Cooling System During Power Operation (FSAR Section 15.5.1) l 4-1 0657v;1o/081187 _______-__a

REACTIVITY CONTROL SYSTEMS ( $ASES 80 RAT!0N SYSTEMS (Continued) MARGIN'from expected operating conditions of 1.6% Ak/k after xenon decay and cooldown~to 200*F. The maximum expected boration capability requirement occurs at E0L from full power equilibrium menon conditions and requires 21,020 gallons of 6300 ppe borated water from the boric acid storage. tanks or 1,166,000 galions C. Minimum RWST volume ofof 9999 ppa borated water free the refueling water storage tank (RW A 1,166,000 gallons is specified to be consistent with y o ECCS requirement. .With the RCS temperature below 200*F, one Boron Infection System is accept-able without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTER-ATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable. The limitation for a maximum of one centrifugal charging pump to be OPER-ABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 350*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. The boron capability required below 200*F is sufficient to provide a O SHUTDOWN MARGIN of 1.6% Ak/k after xenon decay and cocidown from 200'F to 140*F. This condition requires either 4100 gallons of 6300 ppe borated water from the beric acid storage tanks or 250,000 gallons of 4000 ppe borated water from the RWST. Asoo The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics. The limits on contained water volume and beron concentration of the RWST also ensure a pH value of between 7.0 and 7.5 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of icdine and einfaizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The minimum RWST solution temperature for M DES 5 and 6 is based on analysis assumptions in addition to freeze protection considerations. The tainimum/maxiom RWST solution temperatures for MODES 1, 2, 3 and 4 are based on analysis assugtions. The OPERABILITY of one Boron Injection Systes during REFUELING ensures that this system is available for reactivity control while in M00E 6. 3/4.1.3' MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distri-( bution limits are maintained, (2) the minimum SHUTDOWN MARGIN is saintained, and (3) the potential-effects of rod misalignment on associated accident analyses are limited. OPERABILITY'of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control MILLSTONE - UNIT 3 8 3/4 1-3 A-9 1

9 Inadvertent Opening of'a Steam Generator Relief or Safety Valve Causing a Depressurization of the Main Steam System (FSAR Section 15.1.4) Steem System Piping failure (FSAR Section 15.1.5) Feedwater System Pipe Break (FSAR Section 15.2.8) Rupture of a Ccntrol Rod Drive Mechanism Housing /RCCA Ejection (FSAR Section 15.4.8) Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures (FSAR Section 6.2) The effect of an increase in the minimum RWST boron concentration from the current level of 2000 ppm to 2300 ppm on each of these transients for both N and N-1 loop operation is discussed below. The effect of an increase in the minimum accumulator boron concentration from the current level of 1900 ppm to 2200 ppm was evaluated for transients where the accumulators were actuated (see Sectins 4.2.3 and 4.2.6). 4.2.1 Inadvertent Operation of the Emergency Core Cooling System During Power Operation i Spurious actuation of the emergency core cooling distem while at f J power would result in a negative reactivity excursion due to the injected boron from the RWST. The decreasing reactor power causes a drop in the core average temperature and coolant shrinkage. If reactor trip on SIS actuation is assumed not to occur, the reactor will ultimately trip on low pressurizer pressure. DNBR never drops below the initial value. If the RWST boron concentration were increased from the current minimum value of 2000 ppm to 2300 ppm the negative reactivity excursion would occur at a faster rate causing a more rapid drop in the core average temperature and coolant shrinkage. The reactor will trip 4-2 .oss7v:1o/os1187 i

k p Lg h L on low pressurizer pressure as before, though at an earlier time

in the transient. As before the DNBR will never decrease below c

the initial value. Thus, the conclusions in the FSAR remain valid for both N and N-1 loop operation. .4.2.2 Inadvertent Opening of a SG Safety Valve r 0 An accidental depressurization'of the main steam system due to the inadvertent opening of a steam generator safety or relief valve results in-a cooldown of the RCS which, in the presence of a negative moderator temperature coefficient, causes a positive 0 reactivity excursion. Borated water from the RWST enters the core following actuation of the SIS on low pressurizer pressure or low steamline pressure. The negative reactivity provided by the'2000 ppm water from the RWST limits the return to power to an acceptable l'evel so that the minimum DNBR remains above the limiting value. As the transient proceeds and more water from the RWST reaches the RCS, the boron concentration in the RCS gradually increases, ultimately causing the sore to become suberitical. If the RWST boron concentration were increased to 2 2300 ppm more negative ' reactivity would be available to terminate the return to power sooner and at a reduced peak power level. l Thus, the maximum core heat flux reached will be reduced. Additionally, the core would become suberitical earlier in the transient. Thus, the minimum DNBR would be highor than for the case currently analyzed with 2000 ppm in the RWST and the conclusions in the FSAR remain valid for both N and N-1 loop operation. l 4'.2.3 Steam System Piping Failure c A major rupture of a main steam line results in a rapid cooldown f of the RCS which, in the presence of a negative moderator temperature coefficient, causes a positive re6ctivity excursion. h Borated water from the RWST enters the core following actuation 4-3 0657v;1o/081157

l l of the SIS on' low steam line pressure or' low pressurizer pressure. Berated water from the accumulators.will enter the RCS i ] whenever system pressure drops below approximately 600 psia. The negative reactivity provided by the 2000 ppm water from the RWST i and accumulators limits the return to power to an acceptable level so that the minimum DNBR remains above the limiting value, y As.the transient proceeds and more water from the RWST and accumulators reaches.the RCS, the boron concentration in the RCS gradually increases, ultimately causing the core to become ) suberitical. Increasing the boron concentrations' would add more j negative reactivity to terminate the return to power. sooner and reduce the peak power level. Thus, the maximum core heat flux reached will be reduced. Additionally, the core would become' suberitical earlier in the transient. Thus, the minimum DNBR would be higher than for the cases currently analyzed and the-conclusions in the FSAR for both N and N-1 loop operation remain valid. l 4.2.4 Feedwater System Pipe Break Following the rupture of a main feedwater line actuation of the SIS may. occur. Although boron from the RWST is not required to q maintain the reactor in a suberitical condition following a feedwater line break, the cold SIS water serves to reduce the RCS I temperatures and pressures. An increase in the minimum RWST I boron concentration from 2000 ppm to 2300 ppm will increase the negative reactivity insertion rate without affecting the reduction of the RCS temperatures and pressures. Thus, an increase in the RWST boron concentration to 2300 ppm will have no l adverse impact on the feedwater line break analysis and the L conclusions in the FSAR for both N and N-1 loop opera 6 ion will remain valid. 1 1 0657v.1D/081187 - 4-4 -_-__________-__a

4.2.5 Rupture of'a Control Rod Drive Mechanism Housing /RCCA Ejection c Following the ejection of a contro'1 rod the rapid nuclear power excursion causes the RCS to experience a large pressure rise due to the energy released into the coolant. The RCS pressure then drops as fluid inventory is lost through the break (2.75 inch diameter) in'the control rod housing.. As the RCS pressure continues to drop actuation of the SIS.on low pressurizer pressure will inject borated water from the RWST into the RCS. An increase in the RWST boron concentration from the current minimum of 2000 ppm to 2300 ppm will result in more rapid negative reactivity insertion to the core and no interference with the core cooling capability. Thus, the conclusions in the FSAR for both N and N-1 loop operation remain valid. 4.2.6 Mass and Energy Release Analysis for Postulated Secondary Steam System Pipe Ruptures A major rupture of a main steam line results in a rapid cooldown of the RCS which, in the presence of a negative moderator j temperature coefficient, causes a positive reactivity excursion. ] Borated water from the RWST enters the core following actuation of the SIS on low steam line pressure, low pressurizer pressure, or Hi-1 containment pressure. Borated water from the accumulators will enter the RCS whenever system pressure drops i below approximately 600 psia. The negative reactivity provided by the borated water from the RWST and accumulators limits the return to power to an acceptable level so that the minimum DNBR remains above the limiting value. Additionally, by limiting the return to power, the berated water reduces the total energy that is dissipated via steam release through the ruptured steam line. As the transient proceeds and more water from the RWST and accumulators reaches the RCS, the boron concentration in the RCS I l gradually increases, ultimately causing the core to become subcritical. Increasing the QWST anc ocev=u1= tor boren l o6s7v:1o/081187 4-5

n l.f ' ? o 'I concentrations would add more negative reactivity.to terminate the return to power' sooner and reduce the peak power level. Thus, the maximum. core heat flux reached will be reduced, and the core would become suberitical earlier in the transient. Over the course of the' transient, the reduced peak power and earlier return to suberiticality would reduce the integral mass and energy re16dses, as a function of time,. relative to those for cases analysed assuming 2000 ppm boron in the RWST. Therefore, the conclusions in the FSAR for both N and N-1 loop operation remain valid. 4.2.7 Conclusion An' increase in the minimum RWST boron concentration from 2000 ppm to 2300 ppm along with an increase in accumulator boron 1 concentration 1 from 1900 ppm to 2200 ppm, will have no adverse impact upon the non-LOCA accident analyses. The conclusions as stated'in the FSAR will remain valid. 4.3 FSAR LOCA Analysis (10CFR 50.46) For the full spectrum of LOCA postulated breaks, the ECCS is designed.to limit the consequences of an accident to within the acceptance criteria of 10CFR50.46. The analysis takes credit for pumped safety injection from the RWST and passive injection of accumulator water to prevent or mitigate the resulting pee.k clad temperature excursion. Also post-LOCA long term core cooling takes credit for the available water in the RWST and accumulators in determining the post-LOCA RCS/ sump boron concentration and the hot leg switchover time to prevent boron I precipitation. The effect of an increase in the boron concentrations to 2300-2600 ppm for the RWST and 2200 to 2600 ppm for the accumulators on these aspects of the LOCA analysis is discussed below. I f 06s75:1DID81187' 4-6 l j.

y:, 4.3.1 Small Break LOCA .The small break LOCA analyses described in FSAR Section 15.6.5 were performed with the'WFLASH Evaluation Model. The analyses assume that the reactor core is. bro'ught to a.suberitical-condition by.the trip reactivity of the control rods. There is no assumption requiring the presence of. boron in the ECCS water or the need for negative reactivity provided by the soluble boron. Thus the changes in the RWST and Accumulator boron concentrations. do.not alter the conclusions of the FSAR small break LOCA 1 analysis.- l i 4.3.2 Large Break LOCA The large break LOCA analysis described'in FSAR Section 15.6.5' was performed with the 1981 Evaluation Model. The analysis does-not take credit for the negative reactivity introduced by the soluble boron-in the ECCS water in determining reactor power; level during the early phases of a postulated large break LOCA. In addition, no credit is taken.for.the negative reactivity introduced by.the control rods. During this time period the reactor is kept suberitical by the voids present in the core. Thus the changes ir. the RWST and Accumulator boron concentrations do not alter the conclusions of the FSAR large break LOCA analy.,e s. 4.3.3 LOCA Mass and Energy Releases The containment analyses for Millstone Unit 3 are described in l the FSAR Section 6.2. This section considers the containment subcompartments, Mass & Energy releases for postulated LOCAs, and containment heat removal systems. For containment subcompartment analyses, an increase in the RWST and accumulator boron f [ concentrations, would have no effect on the calculated results, since the short duration of the transient (< 3 seconds) does ) l 1 4~7 o657v:1o/081187 L___-_-_-_-____

not consider any safety injection flow taken from the RWST. The long term mass and energy release calculations do not take credit c for the soluble boron present in the safety injection from the ] RWST supplied to the RCS. This is similar to the LOCA analyses assumptions, and therefore an increase in ECCS water boron concentration, would have no effect on the long term mass and energy releases calculated for Millstone Unit 3. 4.3.4 SGTR The Steam Generator Tube Rupture (SGTR) accident for Millstone Unit 3 is presented in FSAR Section 15.6.3. For the SGTR accident, the low pressurizer pressure SI signal is actuated due to the decrease in the reactor coolant inventory shortly e.fter { reactor trip. The borated water from the RWST is delivered to l the reactor coolant system to provide adequate core cooling as l well as sufficient negative reactivity to prevent return to criticality. If the RWST and accumulai;er boron concentrations q are increased, this results in more negative reactivity available ] to provide additional shutdown margin after trip in the SGTR accident. Therefore, the higher RWST and accumulator boron concentrations will have no adverse effect on the FSAR SGTR analysis. 4 4.3.5 Long-Term Cooling - Post LOCA Shutdown Long-term cooling is discussed in FSAR Section 15.S.5. The Westinghouse licensing position for satisfying the requirements of 10CFR50.46 Paragraph (b) Item (5) "Long-term cooling" is defined in WCAP-8339 (page 4-22). The Westinghouse evaluation model committment is that the reactor remain shutdown by the borated ECCS water. Since credit for the control rods is not taken for large break LOCA, the borated ECCS water provided ay 4 the RWST and Accumulators must have a concentration that, w',en mixed with other sources of water, will result in the reac'.or r 0657v;1o/081187 4-6

v m 1 4 -l l ' core remaining s'uberitical assuming ~.al.1 control rods.out.(AR0). .The'effect on the post-LOCA RCS/SumpLboron concentration as a-result'of changing the minimum Tech-Spec boron concentration from-2000;to:-2300 for-the RWST and from 1900 to'2200 for the i 1 ' Accumulators.is an increase of about 197 ppm in the RCS/ Sump 1 boron concentration. Figure 4.3-1 presents the'.new calculated RCS/ Sump-Boron Concentration, using the proposed minimum baron. concentrations.for the RWST and accumulators. Confirmation that i this proposed increase provides enoughl margin to keep the core.

suberitical for long term cooling requirements has been verified

-) through the normal reload evaluation process for Cycle 2.- I J 4.3.6 Long-Term Cooling - Boron Precipitation .A discussion-'on the hot. leg switchover time-is presented'in FSAR Section 15.6.5. An analysis has been performed to determine the .ma'ximum boron concentration in the reactor vessel following a. -hypothetical LOCA. This analysis considered a proposed maximum' boric acid concentration of 2600 ppm.in the RWST and accumulators i ] and 2200 ppm in the RCS, The analysis considers the increase in boric acid concentration in the reacter vessel during the long term cooling phase of a LOCA, assuming a conservatively small effective vessel volume. This volume includes only the free' volumes of the reactor core and upper plenum below the bottom of the hot leg nozzles. This assumption conservatively neglects the mixing of boric acid ) solution with directly connec'ed volumes, such as the reactor

l 1

vessel lower plenum. The calc dation of boric acid concentration i in the reactor vessel considers a cold. leg break of the reactor ] coolant system in which steam is generated in the core from decay l J heat while the boron associated with the boric acid solution is l completely separated from the' steam and remains in the effective vessel volume. Procedures philosophy assumes that it would be )j l' i 1. o657v:1D/081187 4-9 L ____n_

e very'difficul' for the operator to' differentiate between break t sizes and locations. Therefore one hot leg switchover time i.s used to cover the complete break spectrum. The results 'of the analysis show that the maximum allowable boric I acid concentration'of'23.53 weight percent which is the boric-acid solubility limit less 4 weight percent, will not be exceeded in the vessel if hot leg injection is initiated 11 hours after -i 'the inception of a LOCA. The operator should reference this switchover. time against the reactor trip /SI actuation signal. The typical time interval between the accident inception and the reactor trip /SI actuation signal is negligible when compared to l the switchover time.. i 4.3.7. . Rod Ejection Mass and Eneray Releases for Dose Calculations The~ Rod Ejection mass and energy releases for Millstone Unit 3 ) are presented in FSAR Section 15.4.8.4. The increase in the RWST l and' accumulator boron concentrations will be-negligible on the-I 1 Rod Ejection Accident analysis. Since the SI flow taken from the l RWST is modeled under similar assumptions as in the large break and small break LOCA analyses, there will be no adverse effect on the FSAR Rod Ejection accident. l 4.3.8 LOCA Hydraulic Vessel and Loop Forces The blowdown hydraulic loads resulting from a loss of coolant i accident are considered in the FSAR, Section 3.6.2.2.2, Section j 3.9B.1.4.2.4, Section 3.9N.1.4.3 and Section 3.9N.2.5 for Millstone Unit.3. The increase in the RWST and accumulator boron concentrations will have no effect on the LOCA blowdown hydraulic l loads since the maximum loads are' generated within the first few l seconds after break initiation. For this reason the ECCS, t [' including the RWST, is not considered in the LOCA hydraulic l 1 06s7v:1D/081187 4-10

T f '1 ' forces modeling and thus.the; increase in RWST.and accumulator -boron concentrations will have no effect on the results of the a LOCA hydraulic forces calculations. 3, 4.3.9 Conclusions .o The increase in the RWST boron concentration from a range of 2000 - 2200 ppm to a range of 2300 - 2600 ppm and Accumulator boron concentration from a' range of 1900 to 2200.to a range of 2200 to 2500 ppm does not;have a negative effect on the FSAR LOCA related analyses. Current margin.to the post-LOCA shutdown requirement.is increased with continued conformance verified through the normal reload evaluation process. The higher concentrations decrease-n the allowable time for operator -action to initiate' hot leg recirculation to 11 hours. The.resulting time requirement of 11. hours is still more than adequate to assure those operator actions can be accomplished. Therefore, there is no adverse T .effect on FSAR LOCA related accidents for.the proposed boron concentration-increases. These calculations are applicable.to both.N and N-1 loop operations.

j 4.4~ LOCA Related Design Consideration increasing the' maximum boron concentration in the Refueling Water Storage Tank (RWST) and accumulators, from 2200 to 2600 ppm, decreases the pH of the containment spray and recirculating core cooling solutions. A decrease in pH can decrease the elemental iodine spray removal coefficient and decontamination factor (DF), increase the rate of

] hydrogen production due to corrosion of zine (galvanize and zine based j paint)'and can increase the potential for chloride induced stress j corrosion cracking of stainless steel.- l -l e L i j 0657v:10/cB1187 4-11 1 1 a _----____=___-_.-.---_____.___Q

l 1 1 I i 4.4.1 Post LOCA Sump and Spray pH .) ~ The RWST/ Accumulator boron increase decreases the minimum surap 'j

solution pH from.7.0 to 6.65. This new minimum pH falls outside l

-the range of 7 to 9.5'specified in BTP-MTEB-6-1l(Reference 10) for minimizing chloride stress corrosion cracking of' stainless si steel Hence, the. CAT volume and hydroxide requirements.were reevaluated to raise the minimum pH into the required range. The l following are the revised CAT volumes, hydroxide concentrations. and resulting spray and sump pH values: j Minimum Contained Volume (Gal.) 18,000 Maximum Contained Volume (Gal.) 19,000 Minimum Concentration (% by weight) 2.41 Maximum Concen.tration (% by weight) 3.'10 With these limitations the following are values calculated for sump and spray pH: Minimum sump pH (long term) 7.00 Maximum sump pH'(long term) 7.35 Minimum spray pH 7.20 Maximum spray pH 8.70 1 4.4.2 Radiological Consequences The radiological consequences of the large-break LOCA are uneffected by the boron increase. This is based primarily on the fact that the radiological analysis presented in Section 15.6 of f the FSAR does not utlize containment sprays for fission product f removal. t [ 4-12 0657v:1D/081187

4.4.3 Hydrocen Production Hydrogen produced by the corrosion of aluminum and zine is a function of solution pH. The corrosion rates incorporated in the FSAR' Chapter 6 combustible gas analysis appear to be based on the corrosion rate information presented in Reference 11, for a pH of approximately 8. Corrosion rates for this evaluation are found in Figures 4.4-1 and 4.4-2 which correspond to figures 5 and 6 from the reference. Zinc corrosion, Figure 4.4-1, is approximately the same at pH 7.5 and 10.5 and a minimum at pH 9. Aluminum corrosion, Figure 4.4-2, monotonically decreases with decreasing pH. Hence, the revised long term sump pH's, presented in section 4.4.1, will result in hydrogen production rates, due to aluminum and zine corrosion, that are equal to or less than the rates assumed in the FSAR analysis. Boron concentration effects on zine corrosion were also investigated using the method of Reference 12. A corrosion rate constant comparison was made for the FSAR condition (pH approximately 8, 2000 to 2200 ppm B) versus the new pH/ increased boron condition (pH 7.0 to 7.35, 2200 to 2600 ppm B). The comparison showed a rate constant change, for the new condition, of approximately +1 to - 0.5 percent, depending on temperature. This variation is concluded to have a negligible impact on the aggregate hydrogen generation rate. To summarize, the rates of hydrogen generation due to corrosion of aluminum and zinc, for the revised boron /pH condition, will be comparable to the rates specified in the FSAR and, the FSAR analysis remains valid. j l. 0657v:1D.381187 4-13

q 4 if '4.4.4; ; Equipment Qualification .*\\ The ' primary concerns of equipment qualification are protection of t> the~' stainless ~~ steel components of the emergency core cooling system from chloride, induced stress corrosion cracking,. failures of electrical components required to operate post-accident, and-7' W failures ~of containment coatings which could jeopardize the ECCS by flaking or pealing off, clogging the emergency sump and other flow paths, and thus restrict the flow of emergency core cooling . water. 3 t Protection of Stainless Steel ] To minimize the occurrence of chloride stress corrosion cracking of stainless steel, the NRC recommends a solution pH in the. range. q of.7 to 9.5 (Reference 11). The minimum-revised sump solution.pH. of 7.0.is consistent with this recommendation. Electrical Components Electrical equipment is tested to determine the ability of component seals to exclude the containment environment from the ' interior of the component. To maximize the. challenge to the seal materials, high pH sprays, in the range of 8 to 11, have-traditionally been used. The quench spray pH(short term) and the recire pH (long term) with the revised CAT solution are comparable to those currently specified in the FSAR. Specifically, FSAR Section.6.1.1.2 q specifies short term pH in the range of 5 to 8.3 and long term pH j y in the range of 7 to 7.5. The revised short and long term pH j ranges are 5 to 8.7 and 7 to 7.5, respectively. Hence, there j J will be no adverse effect on electrical components. 3 4-14 0657v:1D/081187. I 1

1 i 1 Containment Coatings l Coatings are used in the containment to provide corrosion protection for metals and to aid in decontamination of surfaces I during normal operation. Like electrical equipment, coatings are tested with a high pH solution to maximize the potential deterioration of the coating. The solution pH's associated with the CAT revision are comparable j to those of the FSAR and, hence, there will be no adverse effect I on containment coatings. ] 4.5 Summary and Conclusions The proposed increase in the RWST and accumulator allowable boron concentration limits to 2300 to 2600 ppm for the RWST and 2200 to 2600 ppm for the accumulators has been assessed from a safety standpoint. Based on these results, it is concluded that the proposed boron concentration increases will have no adverse impact on the non-LOCA Accident Analysis, the LOCA Analysis or LOCA Related Design Consideration and is thus acceptable for implementation at Millstone Unit 3 beginning with Cycle 2. Confirmation that the boron concentration increases f provide enough margin to meet post-LOCA shutdown requirements has been confirmed through the normal reload evaluation process. I l l 0657v:1D/081187 4-li) i

l' } f. 2200' I I I I l' I i I i l' I I I I I I I i F l l 1 I I I I I 1.! E 2100 I I I I E I l' I I 4.- 'l i I I I l 4 l. i I I 2000 d M l l l l 1 1 I I l i I I g I I I i 1 1-c l l l l if l l l I .I I I I I I I I I I I 8 _1800 q3 l I I I o d, 1 I I I I I I I I I I I i 1 I i I I I I 1700 500 700 900-1100 1300 1500 Pre-Trip RCS Baron corw/w&tian, HFP, Peak XE, ARD (PEM) FIGURE 4.3-1 POSI-IDCA RCS/ SUMP BORON CWCENIFATION VERSUS PFI-@IP RCS ICRON OECDTM. TION i 4-16

a r l l ') WCAP-6774 i 12 t l Ii l ( .i 9-l a E 7 soo* g W s E 5 g 5-a B 1 1 1 3 l ise' 2 l20* j 7 ( i i l [ 1 i i o 6 7 8 9 10 il 12 l u i pH 1 i FIGURE 4.4-1 CORROSION VERSUS ph AT 'l SEVERAL TEMPERATURES ( FOR ZINC 4-17 i J

l' j u 'h' .~ ; ' s ,'f y / p, . wchP-877G, a' i nio0 & go .e .3 a ,h]C(M; -{ t g <e ,z-( , ('. X j m== /. p' goh 8 ;

,1,.

-r j i !V .{ '/ 7,. q g von E l Y 600 a, = '( i a j 500 u 5. un \\- .a 300 ,e E / 200 iso * '20' 107 T' 70 'il k i 1 3 -= o = ^ ^ 6 7 8 9 10 Ii 12 pH '\\ FIGURE 4.4-2 CORH3510N VERSUS!ph AT - dEVERAL TEMPERATURE.S FOR ALUMitPJM s jy/, 4-18

y v

t_

i h 2 5.0.. Control. System Evaluation T A One.of the more important parameters in defining the NSSS response to'a-j ' temperature and/or' power transient is the moderatoritemperature-h u Lreactivity. feedback coefficient to nuclear power. With the new core .{ designLthere is'a potential for this parameter to be positive for the j first time. Theeffectofapositive(oratleastlessnegative) 'j 2 moderator: temperature coefficient (MTC) is to potentially change the response of the.NSSS and in turn the response of the steam dump', feedwater, and rod. control systems towards a'less stable configuration. However, the extent and time during which the new core design is expected ~ to:have.a positive MTC is limited and should not significantly effect the q 4

response ofLthe control systems. Therefore, there is no apparent need to

,] revise.'any of the control' system'setpoints. L ~

The'need to modify. control system setpoints will be determined during the m

7 plant startup following the installation of the new' core by observing the ifhb. fresponse of:the control systems. If necessary, sign'al compensators and Ig . function generators in the. control systems could be adjusted to obtain a i{f;' ' more optimum system response. Also, control' system responses are not assumed'in the PMTC transient reanalyses, hence changing control system- ] g, I' 'setpoints will not impact the results reported in Section 3.0 of this ] . report. llf,w l?, L wd~, 49 ~ 3 W s ?f j;.~ .D p i y; U

3..,..

r (l pr .) j. X 24/ ' D 1

j-fyQ

MO). a . 4 _ i i .-u\\ .1 = 'MEsN.1o/ostic7 5-1 l

';,y%$ U f ww. y l p . - _,. @ tg i = s.f Q yS %.4.. 4,yv p. ' (sy }8 $L q 'g ,7 v .y. a o z .,.x n. . j g 7g s t ,4 5 L 1 a i. ,4s 4' [ '/ Q-Y 0, \\ '\\l l', N h 'r _, . \\( d 4,,,' 6.0 Technighgeifidation Changes c e +. t. 1!,j s' y TechnicalSpdeificatienjhangesacireq1tofaIoditivemodhrator, ' [ "\\ s I il w s + ~ e y, u,< s C '.temperatureconflitientandthe.increaseinthen.ir%umbjten J g s concentraticdrsquirementsfortheRWSTiendaicemuiakerssaIecontaiacedn ^ 1g o 3 4 Appendif Ap The::gtior.s below summarite thesel changes. > {'. ' 'V Oi ,1 /> t. s. w n. '9 6.I' Specification,3.f.~2'.3]clines the NTC.as +5 p$te/T helow 70% power j ['* .3. r linearly rancing 'doww to.0 pcm/*F at 100% poWeh '" ..ys : )'h k i - 6.2.Specifie nion 3.1.'2.5 increases the minimum boren'concrntrstion from 2000 t,' i-2 ppm to 2300 ppe,for(the R.PSfin Koce 5 and 6. Sp;cificalonk.1.226and, h T,3. n 3.5.4 change theLberon concentration ran9F for thr: RdSI.(inModes1 through4from,2000-2200ppmto2300-J600opm. Specificatidi 3.9 a increases the miniasm required 6croreconcehtratihn from 2000 ppm tcL2300-ppmforallf'lledportiohsof.,th'aRCSarfiefuelingcanalinMode'6. i m Sases '3/4.1,2 char.ges the borcn concen'tration ior the WST to ?F00 ppm. i h s

)

L 6.3 - Specification 3.51. changes the boron rodentration ?rarse b thNRCS m ~ 4 s 6 ~ 0-i ppm tb PE0 -2E00 ppm. accumulators frcm 1900-2200 + 6.4 Specification 3,0.2d changes the allowable CAT contained volume and concentration range in orcier to naintaia. the minimum cump pH at its ' existing 7.0 limit.. 4 i '~ s. ,3 ) + ~ '~ y s a >s, 3 't m., s l t l p o657v.1o/081187 6-1

\\,,' 4 A + / .g dN r.. h f ?.li '?, I " 3 .:( A[ 7,.0 FSAR Chance,s_ p-FSAR cNanges have been made to reflect use of a positive moderator temperature coefficient in the core design (Chapter 4) and safety snelyses((Chapter 15); the results of the transients analyzed in Section 3 (Chapte 15); and the increase in the boron concentration range for the WST and Accunahters (Chapters 6 and 15). These changes to the o Hillstene Uait 3 FSAR are contained in a separate document. 't \\,', [i, t. t 1. . i, k 3 s. \\ -. l' s } r i 0657v.1D/051187 7-1 i l

8.0. References. 'l. .Burnett, T.W.T et al.,'"LOFTRAN Code Description",.WCAP-7907-P-A (Proprietary), WCAP-7907-A (Non-Proprietary), April 1984. 2. Risher, D. H., Jr. and Barry, R. F., " TWINKLE.,A Multidimensional Neutron Kinetics' Computer Code," WCAP-7979-P-A (Proprietary) and WCAP-8028-A (Non-Proprietary), 1975. 3. Hargrove, H. G., "FACTRAN - A Fortran-IV Code for Thermal Transients in a 00 Fuel Rod," 1972. 2 4.. Hochreiter, L. E.; Chelemer, H.; and Chu, P. T., "Subchannel Thermal Analysis of Rod' Bundle Cores,"'WCAP-7015, Revision 1, 1969. 5. Chelemer, H.; Weisman J.;.and Tong L. S., "Subchannel Thermal Analysis of Rod Bundle Cores," WCAP-7015, Revision 1, 1969. 6. "American National Standard for Decay Heat Power in Light Water l Reactors," ANSI /ANS-5.1-1979, August, 1979. j 7. W. R. Pice and R. J..Sterdis, "RTD Bypass Elimination Licensing Report for Millstone Unit 3", WCAP-11496 (Proprietary), WCAP-11497 (Non-Proprietary), June.1987. 8. WCAP-8261 Revision 1, "WFLASH, a FORTRAN-IV Computer Program for Simulation of Transients in a Multi-Loop PWR", V. J. Esposito, K. Kesavan, B. A. Maul, 1974. 1 '9. WCAP-8904-A, "W Emergency Core Cooling System Evaluation Model for [ Analyzing (N-1) Loop Operation of' Plants with Loop' Isolation Valves", R. l M. Kemper, 1979. ] j p r i-i 0657v;1o/C81187 6-1 i

10., Branch Technical Position MTEB 6-1, "pH for Emergency Coolant Water for g PWR's" .11. '" Corrosion Study for Determining Hydrogen Generation from Aluminum, and l Zine During Post Accident Conditions", WCAP-8776, April, 1976 1.

12. "The relative Importance of Temperature, pH and Boric Acid Concentration on Rates of Hz Production From Galvanized Steel Corrosion",

NUREG/CR-2812, November, 1983. 9 i 1 8-2 0657v:1o/0811s7

u t i 9.0 Conclusion To assess the effect on accident analysis of operation of Millstone Unit L 3 with a positive moderator temperature coefficient of +5 pcm/'F evaluation and transient-reanalyses were performed. Discussions of the 7 transient evaluations are presented in Section 2. The results of the transients reanalyzed with a PMTC are presented in Section 3. These evaluations / reanalyses indicate that'the proposed technical specification of Figure 1.0-1 does not result in the violation of safety limits for any of the transients analyzed. Except as noted,'the analyses employed a constant moderator coefficient of +5 pcm/*F, independent of power level. The results of this study are conservative for the accidents investigated at full power, since the-proposed Technical Specification diagrammed in Figure 1.0-1 requires that the~' coefficient linearly decrease from +5 pcm/*F to O pcm/*F from 70 percent to 100 percent of rated power.. In addition, evaluations were performed to assess the impact of an increase in the boron concentration range for the RWST and accumulators as a. result of the cycle 2 fuel design. The results of these evaluations show that the increases in allowable boron concentration have no adverse impact on safety. Proposed technical specifications and FSAR changes as a result of PMTC and the RWST and accumulator boron concentration have been provided. 1 l 0 i L j i I i oss7v:10/os11s7 9-1 l

e a APPENDIX A TECHNICAL SPECIT! CATION MODIFICATIONS l A i I CS23v.1D/062287 A-1

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT C i.IMITING CONDITION FOR OPERATION 3.1.1.3 The moderator' temperature coefficient (MTC) shall be: Q +O SX/OM Less positive than g ok/k/*F for the all rods' withdrawn, beginning -CnMvh a. M n. c T _ A -0=I;. w = :-.., x 7 of cycle life (BOL)/sy pousr /cvel$ y 7'o 70'/o KMED 74EM LPn-eg wia W b. Less negative than -4.0 x 10 4 Ak/k/*Ffortheallrodswithdrawn,dd"/@ end of cycle life (EOL), RATED THERMAL POWER condition. nt /007p Specification 3.1.1.3a. - MODES 1 and 2* only**. 7ec 7W ' APPLICABILITY: Specification 3.1.1.3b. - MODES 1, 2, and 3 only**. gygg ACTION: With the MTC more positive than the limit of Specification 3.1.1.3a. a. above, operation in MODES 1 and 2 may proceed provided: l l Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than C k/k/"f 8'k. 1. l"" within 24 hours or be in HOT STANDBY within the next 6 hours. These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6; The control rods are maintained within the withdrawal limits 2. established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods l ) withdrawn condition; and 'I A Special Report is prepared and submitted to the Commission, 3. pursuant to Specification 6.9.2, within 10 days, describing the j value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition. With the MTC more negative than the limit of Specification 3.1.1.3b. b. above, be in HOT SHUTDOWN within 12 hours. l l i

• With K,fy greater than or equal to 1.

l

• • See Special Test Exceptions Specification 3.10.3.

1 l l r MILLSTONE - UNIT 3 3/4 1-4 I A-2 1

' REACTIVITY CONTROL-SYSTEMS .504ATED WATER SOURCE - $Ht1TDOWN LIMITING CONDITION FOR OPERATION i 1 l h ll be

3.1.2.5 As a minimum, one of the following borated water sources s a L

.0PERABLE: I A Boric Acid Storage System with: 1 a. A minimum contained borated water volume of 6700 gallons, f 1) 2)

A boron concentration between 6300 and 7175 ppa, and i

3) A minimum solution temperature of 67'F. The refueling water. storage tank-(RWST) with: b. ,1) A minimum contained borated water volume of 250,000 gallons, tsoo 2) A minimum boron concentration of-9000: ppe, and 3) A ' minimum solution temperature of 40*F. r l APPLICABILITY: MODES 5 and 6. ACTION: I With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. .. SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE: At least once per 7 days by: a. 1) Verifying the boron concentration of the water,

2)

. Verifying the contained borated water volume, and Verifying the Boric Acid Transfer Pump Room temperature and the beric acid storage tank solution temperature when it is 3) the source of borated water. At least once per 24 hours by verifying the RWST temperature when it is the source of borated water and the outside air temperature is b. less than 35'F. r A-3 3/4 1-11 ' MILLSTONE - UNIT 3 L

t-REACTIVITY CONTROL SYSTEMS { BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a sinimum the feifowing borated water source (s) shall be OPERABLE 'l as. required by Specification 3.1.2.2: A Boric Acid Storage System with- . a.

1) A minimum contained borated water volume of 23,620 gallons,

2) A boren concentration between 6300 and 7175 ppe, and

3) A minimum solution temperature of 67'F.

b. The refueling water storage tank (R'WST) with:-

1) A minimum contained borated water volume of 1,166,000 gailons, 2300 2600

2) A boron concentration between tect and 28tNlL ppe,

-i

3) A minimum solution temperature of 40'F, and f

4) A maximum solution temperature of 50*F.

SDPLICABILITY: MODES 1, 2, 3, and 4. .sN: With the Boric Acid Storage System inoperable, restore the system to a. OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN equivalent to at least 1.6% Ak/k at 200'F; restore the Boric Acid Storage System to OPER-ABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours, b. With the RWST inoperable, restore the tank to OPERABLE status within I hour or be in at least HOT STANDBY within the next l 6 hours and in COLD SHUTDOWN within the following 30 hours. l j 1 A-4 MILLSTONE - UNIT 3 3/4 1-12

f 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS ( ] I,IMITING CONDITION FOR OPERATION i 3.5.1 Each Reactor C*oof ant fystem (RCS) accumulator shall tie OPERABLE with: s. The isolation valve open and pawer removed, b. A contained borated water voluge of between 6618 and 6847 gallons, 22o0 2 600 c. A boron concentration of between 1990 and N ppe, and d. A nitrogen cover-pressure of between 636 and 694 psia. EPLICABILITY:. MODES 1, 2, and 3*. ACTION: a. With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulated to OPERABLE status within 1 heur or be in at least HOT STANDBY within the next i 6 hours and reduce pressurizer pressure to less than 1000 psig ) within the following 6 hours. b. With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours and reduce pressurizer pressure to less than 1000 psig within the following 6 hours. I wRVEILLANCE REQUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERACLE: a. At least once per 12 hours by:

1. ~

1) Verifyfog the contained borated water volume and nitrogen cover pressure in the tanks to be within the above limits, and l 1 2) Verifying that each accumulator isolation valve is open. j b. At least once per 31 days and within 6 hours after each solution volume increase of greater than or equal to 1% of tank volume by verifying the boron concentration of the accumulator solution; and 4 ) (

• Pressurizer pressure above 1000 psig.

A-5 MILLSTONE - UNIT 3 3/4 5-1 l

JMERGENCY CORE g LfMG SYSTEMS _ 1 1'I 3./.4 5.4 _#EFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with: i I a. A contained borated water volume between 1,166,000 and 1,207.000 l

gallons, 2300*

2 hoo b. A boron concentration between 3000 and 3304 ppa of boren, c. A minimum solution temperature of 40'F, and 1 d. A maximum solution temperature of 50*F. ] ) APPLICABILITY: PCDES 1, 2. 3, and 4. ACTION With the NWST inoperable, restore the tank to CPERABLE status within I hour or be in at least HOT STAND 8Y within 6 hours and in COLD SHUTDOWN within the following 30 hours. c SURVEILLANCE REQUIREMENTS ) 1 4.5.4 The RWST shall be demonstrated OPERABLE: a. At least once per 7 days by: 1 1) Verifying the contained borated water volume in the tank, and 2) Verifying thi boren concentration of the water. b. At least once per 24 hours by verifying the RWST temperature. I ( A-6 ~ MILLSTONE - UNIT 3 3/4 5-9

q g l. t 1

l s

l - -CONTAINMENT' SYSTEMS-l ~ t 1 SPRAY' ADDITIVE SYSTEM LIMITING CONDITION FOR OPERATION l 1.6.2.3. The Spray' Additive System shall be OPERABLE with: 18oco' a. A chemical addition tank containing a volume of between 2*iWMF and _ ,20,1:: gallons of between 338 and 47:004 by weight NaOH solution, /9000 and R.s t 3.to b. Two gravity feed paths each capable of adding NaOH solution from I the chemical addition tank to each Containment Quench Spray subsystem l pump suction, j ' APPLICABILITY: MODES 1, 2, 3, and 4. l ACTION:. J l With the Spray Additiv's System inoperable, restore the system to OPERABLE status within 72 hours or be in at least NOT STANDBY within the next 6 hours; restore the Spray Additive System to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours. ~ L SURVEILLANCE REQUIREMENTS 4.6.2.3 The Spray Additive System shall be demonstrated OPERABLE: At least once per 31 days by)verifyin0 that each valve (manual,in the flow p a. power-operated, or automatic t sealed, or otherwise secured in position, is in its correct position; b. At least once per 6 months by: 3 1 1) Verifying the contained solution v61ume in the tank, and ^ 2) Yerifying the concentration of the Na0H solution by chemical analysis is within the above limits. l c. At least once per 18 months, during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a CDA test signal. { A-7 MILLSTONE - UNIT 3 3/4 6-14

r7 3/4. fr L REFUELING OPERATIONS = 3/4.9.1 80RON CONCENTRATION LIMITING CONDITION FOR OPERATION m 3.9.1 The boron concentration of all filled portions of the Reactor Coolant

Systas and the refueling canal shall be maintained uniform and sufficient to cnsure that the more restrictive of the following. reactivity conditions is met;

) cither: A K,ff of 0.95 or less, or a. 2Joo b. A boron concentration of greater than or equal to :2000:pps. APPLICABILITY: MODE 6.* ACjToN: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity { changes and initiate and continus boration at greater than or equal to 33 gpa of a solution containing greater than or equal to 6300 ppe boron or its i l is reduced to less than or equal to 0.95 or the boroni equivalent until K concentrationisrNoredtogranterthanorequalto2000ppe,whicheveris ) ~ the more restrictive. .2300 1 $ SURVEILLANCE REQUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to: Renoving or unbolting the reactor vessel head, and a. b. Withdrawal of any full-length control rod in excess of 3 feet from its fully inserted position within the reactor vessel.

4. 9.1.1 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours.

4. 9.1. 3 Valve 3CHS-Y30$ shall be verified closed and secured in position by mechanical stops or by removal of air or niectrical power at least once per

37. days.

'The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed. 4 A-8 PILLSTONE - LHIT 3 3/4 9-1

REACTIVITY CONTROL SYSTEMS {. ' BASES 80 RATION SYSTEMS (Continued) MRGIN free expected operating conditions of 1.6% Ak/k after menon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs at E0L free full power equilibri m xenon conditions and requires 21,020 gallons of 6300 ppe borated water from the borie acid storage tanks or 1,166.000 galions MnfausRWSTvolumeofof 9000 ppe borated water from the. refueling water storage tank (RWST). A E 1,166,000 gallons is specified to be consistent with ,po ECCS requirement. . With the RCS temperature below 200*F, one Boron Injection Systes.is accept-able without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTER-ATIONS and positive reactivity changes in the event the single Boron Injection Systes becomes inoperable. 1 The limitation for a maximum of one centrifugal charging Icump to be OPER-ABLE and the Surveillance Requirement to verify all charging pia;ps except the required OPERABLE pump to be inoperable below 350*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single p0RV. The boron capability required below 200'F is sufficient to provide a h SHUTDOWN MARGIN of 1.6% Ak/k after xenon decay and cooldown from 200'F to 140*F. This condition requires either 4100 gallons of 6300 ppe borated water from the beric acid storage tanks or 250,000 gallons of 4004 ppe borated water from the RWST. Asco The contained water volume liafts include allowance for water not available because of discharge line location and other physical characteristics. The limits on contained water volume and baron concentration of the RWST also ensure a pH value of between 7.0 and 7.5 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on nechanical 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 oinimum/maxion RWST solution temperatures for MODES 1, 2, 3 and 4 are based on analysis assumptions. l The OPERABILITY of one Boron Injection Systes during REFUELING ensures that this system is available for reactivity control while in M00E 6. 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distri-1 ( bution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of rod afsalignment on associated accident analyses are i limited. OPERABILITY of the control rod position indicators is required to i determine control rod positions and thereby ensure compliance with the control MILLSTONE - UNIT 3 5 3/4 1-3 i A-9 L -}}