ML18057A966

Review & Analysis of SRP Chapter 15 Events for Palisades W/ 15% Variable High Power Trip Reset.

ML18057A966
Person / Time
Site:	Palisades
Issue date:	11/09/1990
From:	Lindquist T, Welzbacker R SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML18057A964	List: ML18057A963 ML18057A965 ML18057A966
References
ANF-90-181, NUDOCS 9106180005
Download: ML18057A966 (72)

Text

I.

I ANF-90-181 I

I ADVANCED NUCLEAR FUELS CORPORATION I

I REVIEW AND ANALYSIS OF SRP CHAPTER 15 EVENTS I FOR PALISADES WITH A 15% VARIABLE HIGH POWER TRIP. RESET I

I I

I I

I I NOVEMBER 1990 I

I ---------;;:;- c* -i:::;;;;c>--,.-.\

- - - 9 t ()/-.1 :=:(1005 ';/.1 -'~*-.:* - -

A Siemens Company F*. 0*F*.- r-~r1n~K 050002~5 p ' -i - -

• F'DF<

!""----------------------------------~-

I I ADVANCED NUCLEAR FUELS CORPORATION

,. ANF-90-181 Issue Date:

11 /09/90 t

,,t

.,,/ REVIEW AND ANALYSIS OF SRP CHAPTER 15 EVENTS FOR PALISADES WITH A 15% VARIABLE HIGH POWER TRIP RESET I

Prepared by:

I I

I Prepared by:

T. R. Lindquist, earn t:.eader f, PWR Fuel Eng* eering Fuel Engineering and Licensing I

I November 1990

I CUSTOMER DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS ANO USE OF*THIS DOCUMENT I

PLEASE READ CAREFUU. Y Advanced Nuclear Fuels Corporation's warranties and representations con-

• ceming the subject matter of this document are those set forth in the Agreement between AdVanced Nuclear Fuels Corporation and the Customer pursuant to which this document is issued. Accordingly, except as otherwise expressly pro-vided in such Agreement, neither Advanced Nuclear Fuels Corporation nor any .

person acting on its behaJf mal<es any warranty or representation. expressed or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained in this document, or that the use of any information, apparatus, method or process disclosed in this document will not infringe privately owned rights: or assumes any liabilities with respect to the use of any information, ap-paratus, method or process disclosed in this document.

I The information contained herein is for the sole use of Customer.

In order to avoid impairment of rights of AdVanced Nuclear Fuels Corporation in patents or inventions which may be included in the information contained in this

• document, the recipient, by its acceptance of this document. agrees not to publish or mllke public use (in the patent use of the term) of such information until so authorized in writing by AdVanced Nuclear Fuels Corporation or until after six (6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless otherwise expressly provided in the l\greement .. No rights or licenses in or to any patents are implied by the furnishing of this docu-I ment.

I I

I I

ANF-3145.472A (12/87)

ANF-90-181 Page i TABLE OF CONTENTS

'I Section

,,' 1.0 2.0 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Summary and Conclusions ...... *. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1 2

Ii

,, 3.0 Disposition and Analysis of Plant Transients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

15.0 Accident Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

15.0.1 Classification of Plant Conditions . . . . . . . . . . . . . . . . . . . . . .

8 9

9 I 15.0.2 15.0.3.

. Plant Characteristics and Initial Conditions . . . . . . . . . . . . .

Power Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

11 11 15.0.4 Range of Plant Operating Parameters and States . . . . . . . . 13

~I 15.0.5 Reactivity Coefficients Used in the Safety Analysis . . . . . . . . 13

,, 15.0.6 Scram Insertion Characteristics ..................... 13

,, *15.0.7 15.0.8 Reactor Protection System Trip Setpoints and Time Delays.........................................

Component Capacities and Setpoints . . . . . . . . . . . *. . . . . .

13 14

,Ii 15.0.9 Plant Systems and Components Available for Mitigation of

,, 15.0.10 Accident Effects . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . .

Effects of Mixed Assembly Types and Fuel Rod Bowing .............................. : . . . . . . . . .

14 14 15.0.11 Plant Licensing Basis and Single Failure Criteria . . . . . . . . . 14 I 15.1 Increase in Heat Removal by the Secondary* System ........ ; . . . . . . . . 15

• 15.1.1 Decrease in Feedwater Temperature . . . . . . . . . . . . . . . . . . 15 15.1.2 Increase in Feedwater Flow .........................

• 16

.I*

ANF-90-181 Page ii I

TABLE OF CONTENTS Section

.\.'*

15.1.3 15.1.4 Increase in Steam Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Inadvertent Opening of a Steam Generator Relief or Safety Valve .................................... -......

16 18 15.1.5 Steam System Piping Failures Inside and Outside of Containment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ." . . . 18

\I 15.2 Decrease in Heat Removal by the Secondary System ..... * . . . . . . . . . . . 20 I\

15.2.1 15.2.2 Loss of External Load . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Turbine Trip . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

20 21 I/

15.2.3 Loss of Condenser Vacuum . . . . . . . . . . . . . . . . . . . . . . . . 22 15.2.4 Closure of the Main Steam Isolation Valves (MSIV) (BWR) . . 22 15.2.5 Steam Pressure Regulator Failure . . . . . . . . . . . . . . . . . . . . 23 I I

15.2.6 Loss of Nonemergency A.C. Power to the Station Auxiliaries ................. , . . . . . . . . . . . . . . . . . . . . 23 15.2.7 Loss of Normal Feedwater Flow . . . . . . . . . . . . . . . . . . . . . 24 15.3 15.2.8 Feedwater System Pipe Breaks Inside and Outside Containment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Decrease in Reactor Coolant System Flow . . . . . . . . . . . . . . . . . . . . . . . . .

26 28 15.3.1 Loss of Forced Reactor Coolant Flow . . . . . . . . . . . . . . . . . 28 1*

15.3.2 15.3.3 Flow Controller Malfunction . . . . . . . . . . . . . . . . . . . . . . . . .

Reactor Coolant Pump Rotor Seizure . . . . . . . . . . . . . . . . .

28 29 15.3.4 Reactor Coolant Pump Shaft Break . . . . . . . . . . . . . . . . . . . 29 15.4 Reactivity and Power Distribution Anomalies . . . . . . . . . . . . . . . . . . . . . . . 31

I I ANF-90-181 Page iii

,,, TABLE OF CONTENTS Section I{,,, 15.4.1 Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition . . . . . . . . . . . . . 31 15.4.2 Uncontrolled- Control Rod Bank Withdrawal at Power . . . . . . 31 I 15.4.3 15.4.4 Control Rod Misoperation . . . . . . . . . . . . . . . . . . . . . . . . . .

Startup of an Inactive Loop . . . . . . . . . . . . . . . . . . . . . . . . .

46 51 I 15,4.5 Flow Controller Malfunction . . . ; . . . . . . . . . . . . . . . . . . . . . 52 I 15.4.6 CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant . . . . . . : . . . . . . . . . . 52 15.4.7 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position .........* . . . . . . . . . . . . . . . . . . . . . . 53 15.4.8 Spectrum of Control Rod Ejection Accidents . . . . . . . . . . . . 53 15.4.9 Spectrum of Rod Drop Accidents (BWR) . . . . . . . . . . . . . . . 54 15.5 Increases in Reactor Coolant System Inventory , . . . . . . . . . . . . . . . . . . . . 55 15.5.1 Inadvertent Operation of the ECCS that Increases Reactor Coolant Inventory . . . : . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55 15.5.2 CVCS Malfunction that Increases Reactor Coolant Inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55 15.6 Decreases in Reactor Coolant Inventory . . . . . . . . . . . . . . . . . . . . . . . . . . 57 15.6.1 lnadvert~nt Opening of a PWR Pressurizer Pressure Relief Valve.......................................... 57 15.6.2 Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside of Containment . . . . . . . . 58 15.6.3 Radiological Consequences of Steam Generator Tube Failure .........................................

• 58

ANF-90-181 Page iv TABLE OF CONTENTS Section 15.6.4 Radiological Consequences of a Main Steam Line Failure Outside Containment (BWR) . . . . . . . . . . . . . . . . . . . . .. . . . 59 15.6.5 Loss of Coolant Accidents Resulting from a Spectrum of

• Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . *. 59 15.7 Radioactive Releases from a Subsystem or Component . . . . . . . . . . . . . . . 60 15.7.1 Waste Gas System Failure . . . . . . . . . . . . . . . . . . . . . . . . . . 60 15.7.2 Radioactive Liquid Waste System Leak or Failure (Release to Atmosphere) ......... ; . . . . . . . . . . . . . . . . . . . . . . . . 60 15.7.3 Postulated Radioactive Releases Due to Liquid-Containing Tank Failures ..............................* . . . . . 60 I

15.7.4 Radiological Consequences of Fuel Handling Accident . . . . 60 15.7.5 Spent Fuel Cask Drop Accidents . . . . . . . . . . . . . . . . . . . . . 60 4.0 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . 61 I

I t,,

1*

.1

ANF-90-181 Page v LIST OF TABLES I

'\

2-1 Disposition of Events Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2-2 Summary of Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 15.0.1-1 Accident Category Used for Each Analyzed Event . . . . . . . . . . . . . . . . . . . 1o I 15.0.3-1 Core Power Distribution . *. . . . . . . : . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

.( 15.4.2-1 Event Summary for .the Uncontrolled Rod Bank Withdrawal Event from 91.5% of Rated ...................................... ; . . . . . . .. . . 35 I

i.

l 1 I

• I)

I

~:*

/,!11.

ANF-90-181

-e Page vi LIST OF FIGURES Figure 15.4.2-1 MDNBR Values for Uncontrolled Bank Withdrawal from 91.5% of Rated I

15.4.2-2 Power ............... ~. ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Reactor Power Level for Uncontrolled Bank Withdrawal at Part Power . . . .

36 37

'I-: ;.:,

15.4.2-3 Core Average Heat Flux for Uncontrolled Bank Withdrawal at Part Power . . 38 1 <

.I 15.4.2-4 Pressurizer Pressure for Uncontrolled Bank Withdrawal at Part Power . . . . 39 15.4.2-5 P.ressurizer Liquid Level for Uncontrolled Bank Withdrawal at Part Power . . 40 15.4.2-6 Primary Coolant System Mass Flow Rate for Uncontrolled Bank Withdrawal at Part Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 I 15.4.2-7 Primary Coolant System Temperatures for Uncontrolled Bank Withdrawal at Part Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 15.4.2-8 15.4.2-9 Reactivities for Uncontrolled Bank Withdrawal at Part Power . . . . . . . . . . .

Secondary Pressure for Uncontrolled Bank Withdrawal at Part Power . . . .

43 44

.I\

15.4.2-10 Steam Generator Liquid Level for Uncontrolled Bank Withdrawal at Part Power ........................................ *............ 45

\I I

I 4*'

I .

ANF-90-181 Page 1 i.,

~

1.0 Introduction This report documents the results of an FSAR Chapter 14 disposition of events review and analysis for Palisades operation with a 15% variable high power (VHP) trip reset margin. The 12 1* review and analysis were performed using NRG-approved ANF methodology( * ) and .fC?llows the 3

format of the Standard Review Plan (SRP) Chapter 15. events ( ). Each SAP Chapter 15 event was reviewed and dispositioned into one of the following categories:

~

,f 1. The event initiator or controlling parameters have been changed from the analysis of record so that the event needs to be reanalyzed for the current licensing action;

2. The event is bounded by another event which is to be reanalyzed;

1* 3. The event initiating mechanism(s) and principal variables which control the results of the event are unchanged from or bounded by the analysis of record; or

4. The event is not in the licensing basis for the plant.

• ~/

This report specifically addresses the change in VHP trip reset margin from 10% to 15%

relative to the licensing basis for Palisades Cycle 9. References 4 and 5 document the licensing l analyse.s supporting Palisades* Cycle 9 operation and lists the plant and fuel related changes for that cycle. The impact of steam generator replacement prior to Cycle 9 operation is addressed I in Reference 6.

• I Section 2.0 of this report presents a summary of the analyses performed to support the subject licensing action. Section 3.0 presents the conditions employed and the results of these analyses. Each event is numbered in accordance with Chapter 15 of the SRP to facilitate review.

• 11 I

ANF-90-181 Page 2 2.0 Summary and Conclusions

~\

A summary of the Disposition of Events for the subject licensing action is given in Table YI; 2-1 . This table lists each SAP Chapter 15 event, indicates whether or not the event requires reanalysis as a result of the subject licensing action and provides a reference to the bounding event or analysis of record. The only events that were reanalyzed as a result of increasing the I" VHP trip reset to 15% are the Uncontrolled Control Rod Withdrawal at Power (Event 15.4.2) and the Single Control Rod Withdrawal (Event 15.4.3). Table 2-2 summarizes the results of this I.

reanalysis. For both events, the acceptance criteria are met.

II I\

I

,f, I

I

\'t

ANF-90-181 Page 3

I Table 2-1 Disposition of Events Summary 1* Event Classifi-SRP Event Desig-Bounding Event or UFSAR cation nation Name Dis12osition Reference Desig.

I 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.1 f,

Decrease in Feedwater Temperature Bounded 15.1.3 14.9.4

,, 15.1.2 15.1.3 Increase in Feedwater Flow 1) 2)

Power Startup Increase In Steam Flow Bounded Bounded Bounded 15.1.3 15.1.3 Ref. 5 14.9.6 14.9.5 14.10

t 15.1.4 Inadvertent* Opening of a Steam Generator Relief or Safety Valve

1) Power Bounded 15.1.3

2) Scram Shutdown Margin Bounded 15.1.3 15.1.5 Steam System Piping Failures Inside and Outside of Containment Bounded<al Ref. 5,7 14.14

I*

\.,

. 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.1 Loss of External Load Bounded Ref. 5 14.12 l

15.2.2 Turbine Trip Bounded 15.2.1 15.2.3 Loss of Condenser Vacuum Bounded 15.2.1 15.2.4 *Closure of the Main Steam Isolation Valves (MSIV) Bounded 15.2.1 I

15.2.5 Steam Pressure Regulator Failure Not applicable; BWR*Event 15.2.6 Loss of Nonemergency A.C. Power Short term to the Station Auxiliaries bounded 15.3.1

'I*i Long term bounded 15.2.7 a

The SRP Chapter 6 steam line break containment pressure analysis was performed to confirm current limits with the replacement steam generators. CPCo furnished this analysis.

I

ANF-90-181 Page 4 Table 2-1 Disposition of Events Summary, Continued I\

SRP I

Event Classifi-cation Event Desig-nation Name Disposition Bounding Event or Reference UFSAR Desig.

15.2.7 Loss of Normal Feeclwater Flow 1) 2)

3)

Maximum PCS pressure Maximum Primary/Secondary pressure difference Minimum SG inventory Bounded Bounded Bounded Ref. 14 Ref. 14 Ref. 14 14.13 14.13 ,,

I*

15.2.8 Feeclwater System Pipe Breaks Cool down Inside and Outside Containment bounded 15.1.5 Heatup bounded 15.2.1 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW .1:

15.3.1 Loss of Forced Reactor Coolant Flow Bounded Ref. 5 14.7 15.3.2 Flow Controller Malfunction Not Applicable 14.7 15.3.3 Reactor Coolant Pump Rotor Seizure Bounded Ref. 5 14.7 I\

'\*

15.4 15.3.4 Reactor Coolant Pump Shaft Break REACTIVITY AND POWER DISTRIBUTION ANOMALIES Bounded 15.3.3 14.7

\I 15.4.1 Uncontrolled Control Rod Bank Withdrawal from a Subcritical or Low Power Condition Bounded Ref. 5 14.2.2.2

.I 15.4.2 Uncontrolled Control Rod Bank Withdrawal at Power Operation

.. j

""-.r Conditions Analyze 14.2.2.3 15.4.3 Control Rod Misoperation 1) 2)

Dropped Control Bank/Rod Dropped Part-Length_ Control Rod Bounded Bounded Ref. 5 15.4.3(1) 14.4 14.6 i**

3) Malpositioning of the Part-Length Control Group Not Applicable 14.6

4) Statically Misaligned Control Rod/Bank Bounded Ref. 5

5) Single Control Rod Withdrawal Analyze 14.2.2.4

6) Core Barrel Failure Bounded Ref. 5 14.5

,f\'

ANF-90-181 Page 5 I: Table 2-1 Disposition of Events Summary, Continued

.\II Event Classifi-SRP Event Desig-Bounding Event or UFSAR cation nation Name Disposition Reference Desig.

I'

'~ 15.4.4 Startup of an Inactive Loop Bounded Ref. 16 14.8 15.4.5 Flow Controller Malfunction Not applicable; I No Flow Con-troll er

,1, 15.4.6 CVCS Malfunction that Results in a Decrease in the Boron Con-centration in the Reactor Coolant

1) Rated and Power Bounded Ref. 5 14.3

• 1 2) 3)

Operation Conditions Reactor Critical, Hot Standby and Hot Shutdown Refueling Shutdown Con-Bounded Bounded Ref. 5 Ref. 5 14.3 14.3 dition, Cold Shutdown Condition and Refueling Operation

.r~

15.4.7 Inadvertent Loading and Operation Administrative of a Fuel Assembly in an Improper Procedures Position Preclude this I 15.4.8 Spectrum of Control Rod Ejection Accidents Event Bounded Ref. 5 14.16 I\* 15.4.9 Spectrum of Rod Drop Accidents (BWR)

Not applicable; BWR Event I

15.5 INCREASES IN REACTOR COOLANT INVENTORY 15.5.1 Inadvertent Operation of the Overpressure 1.

ECCS that Increases Reactor Bounded 15.2.1 Coolant Inventory Reactivity Bounded 15.4.6

,,' 15.5.2 CVCS Malfunction that In-creases Reactor Coolant Inventory Overpressure Bounded Reactivity Bounded 15.2.1 15.4.6

ANF-90-181 Page 6 Table 2-1 Disposition of Events Summary, Continued

• 1*-

Event Classifi-SRP Event Desig-Bounding Event or UFSAR l

1 1

cation nation Name Dispositfon Reference Desig.

15.6 DECREASES IN REACTOR COOLANT INVENTORY

~-

15.6.1 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve Bounded Ref. 5 I. /

\,,

15.6.2 Radiological Consequences of the Bounded 15.6.5 14.23 Failure of Small Lines Carrying Primary Coolant Outside of Containment 15.6.3 Radiological Consequences of Bounded (b) 14.15 15.6.4 Steam Generator Tube Failure Radiological Consequences of a Main Steam Line Failure Outside Not applicable; BWR Event 1.

Containment 15.6.5 Loss of Coolant Accidents Bounded Ref. 7, 15 14.17 Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary 14.18 14.22 I'.

15.7 RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.1 Waste Gas System Failure Bounded Ref. 7 14.21

\I 15.7.2 Radioactive Liquid Waste System Leak or Failure (Release to

'I.

Atmosphere) Bounded Ref. 7 14.20 15.7.3 Postulated Radioactive Releases Bounded Ref. 7 14.20 .I -....

due to Liquid-Containing Tank Failures ,,{

15.7.4 Radiological Consequences of Fuel Bounded Ref. 7 14.19 Handling Accidents 15.7.5 Spent Fuel Cask Drop Accidents Bounded Ref. 7 14.11

'I b

This event was reanalyzed for the replacement of the steam generators. CPCo furnished this analysis.

Table 2-2 Summary of Results Maximum Core Maximum Maximum Power Average Heat Flux Pressurizer Event Level (MWt) (Btu/hr-ft2) Pressure (psia) MDNBR(a) 15.4.2 Uncontrolled Control Bank Withdrawal at Power 2900.9 178,867 2267.14 1.420 15.4.3 Control Rod Misoperation

• Single Rod Withdrawal 2900.9 178,867 2267.14 1.197 a )>

A bounding MDNBR analysis was performed based on the radial peaking assumptions given in Table 15.0.3-1 and the XNB correlation (95/95 limit = 1.170).

MDNBRs for Reload M fuel with HTP spacers were calculated with the ANFP correlation (95/95 limit = 1.154). A 2% mixed core penalty was Included in "U I

,,z the MDNBR analyses such that the effective DNB correlation limits are 1.190 and 1.174 for the XNB and the ANFP correlations, respectively. The reported D.> CD values bound all fuel types in the Cycle 9 core. co

(!)

er:>

())

.I ANF-90-181 Page a 1

3.0 Disposition and Analysis of Plant Transients

'I The following section discusses the rationale used to disposition each of the SRP Chapter 15 events. The event initiator and sequence of events is compared to the reference analysis to I assess the impact of increasing the VHP trip reset margin. If the reference analysis remains bounding of the proposed licensing action, the event will not require further analysis. If, however, the licensing action results in a more severe transient, or the impact on the severity of the I,,,

transient is .uncertain, the event is categorized as requiring further analysis. Event numbering and '-

• 11 nomenclature are consistent with the SRP to facilitate review.

The licensing analyses supporting Palisades Cycle 9 constitute the basis for this evaluatiQn and are given in the References 4 and 5. This section .also provides .information or references for the following:

• Classification of plant conditions

• Event* acceptance criteria .:,\

• Single failure criteria

• j\

Plant operating modes

• Plant initial conditions Neutronics data

• Core and fuel design parameters

• Listings of systems and components available for accident mitigation, trip setpoints, time delays and component capacities. ..._,

• i*

rf

.1' I ANF-90-181 Page 9 ii' 15.0 Accident Analyses

,I/

-/

15.0.1 Classification of Plant Conditions Plant operations are placed in one of four categories. These categories are those i adopted by the American Nuclear Society (ANS). The categories are given in Section 15.0.1 in Reference 5.

II 15.0.1.2 Classification Of Accident Events By Category I

t Table 15.0.1-1 lists the accident category used for each event analyzed in this report.

This classification is used in evaluating the acceptability of the results obtained from the analysis .

• II

',,(

I .---

,f

ANF-90-181 Page 10 Table 15.0.1-1 Accident Category Used for Each Analyzed Event I, Event 15.4.2 Uncontrolled Bank Withdrawal at Power 15.4.3 Control Rod Misoperation Accident Category Moderate I/

• Single Control Rod Withdrawal Infrequent I

I,

• I"

'11.

I\

I\

• i t

j

ANF-90-181 Page 11

,, 15.0.2 Plant Characteristics and Initial Conditions Eight operational modes are considered in the analyses and are given in Section 15.0.2 in Reference 5. The nominal plant rated operating conditions and principal fuel design characteristics are presented in Tables 15.0.2-1 and 15.0.2-2 of Reference 5, respectively.

I 15.0.3 Power Distribution

\l The radial and axial power peaking factors used in the analysis are presented in Table I 15.0.3-1 . Figure 15.0.3-1 in Reference 5 shows the limiting axial shape used for the analysis of the uncontrolled rod withdrawal event in Section 15.4.2. This axial shape has an axial shape

1

index (ASI) of -0.135.

As in the transient analysis for Cycle 9, the limiting DNBR occurs on an interior pin of a 216 fuel rod assembly. The Technical Specification(S) Limiting Conditions of Operation (LCO)

_,,. ensure that the power distribution is maint~ined within these limits during normal operation.

However,* som*e events analyzed result in transient redistribution of the radial power peaking

'f. factors. Transient radial power redistribution is treated as described in Section 15.4.3.2 of Reference 1.

I\

,, The an~lyses for the inlet temperature LCO and for the thermal margin/low pressure (TM/LP) trip utilize axial PC?Wer distributions and associated ASls. These analyses are presented in Reference 5 for Cycle 9.

1,

ANF-90-181 Page 12 Table 15.0.3-1 Core Power Distribution II

-~./

Radial Peaking Factor1a):

208 rods/assembly Assembly(F rA) 1.48 Peak Rod (F rT) 1.92 I 216 rods/assembly(b>

• Reload L (and earlier) 1.66 2.03 II/

~,

• Reload M 1.57 1.92 Axial Peaking Factor See Figure 15.0.3-1 (Reference 5)

I\

Engineering Tolerance Uncertainty 1.03 -

Fraction of Power Deposited in the Fuel .974

/__ '

J' I

/,,

t, a

• For operation at less than rated power, the rated power radial peaking factors are multiplied by [1 + 0.3 * (1 - f)] for 0.5 ..:s. f .:s. 1.0 and 1.15 for f < 0.5, where f is the fraction of rated power (i.e., 2530 MWt). j,

• \.,./

b The Cycle 9 non-LOCA transient analysis was performed using a F A of 1.66 and a F T of 2.03 for Reload L fuel. The

• r r MDNBRs calculated under these conditions conservatively bound those for Reload M fuel with a F A of 1.57 and a F T r

of 1.92. The Cycle 9 Technical Specificati6n radial peaking limits for Reload L fuel are the same as for Reload M f i.e., FA= 1.57 and FT= 1.92.

r r

ANF-90-181 Page 13 15.0.4 Range of Plant Operating Parameters and States Table 15.0.4-1 of Reference 5 presents the range of key plant operating parameters considered in the analysis.

i 15.0.5

• Reactivity Coefficients Used in the Safety Analysis I. Table 15.0.5-1 of Reference 5 presents the reactivity coefficients used in the analysis. The ii set of parameters which most challenges the event acceptance criteria is used in each analysis.

As in Reference 5, the analysis conservatively supports the Technical Specification moderator 4

~--

temperature coefficient of < +0.5 x 1o- .I'.\ p/°F. The nominal Cycle 9 burnup is 9,400 MWd/MTU, 1" however, the safety analysis is applicable to an end-of-Cycle 9 exposure of 10,150 MWd/MTU.

\. ~*

15.0.6 Scram Insertion Characteristics

• I Figure 15.0.6-1 of Reference 5 presents the negative insertion used in the analysis for reactor trip. The insertion worth includes the most reactive rod stuck out.

• 15.0.7 Reactor Protection System Trip Setpoints and Time Delays

'.I Table 15.0.7-1 of Reference 5 presents the trip setpoints, uncertainties, and time delays used in the Cycle 9 transient analysis. The VHP trip setpoint with a reset margin of 15% is given*

below. The basis for the VHP trip uncertainty is given in Reference 5. All other trip setpoints remain unchanged from the values given in Table 15.0.7-1 of Reference 5.

I t

Trip Setpoint Uncertainty Delay Time Variable High 106.5% maximum + 8.5% of rated 0.4 seconds Power 30.0% minimum 15.0% above thermal power The effects of a 15% VHP trip reset were included in the generation of the TM/LP trip in I

J I

ANF-90-181 Page 14 Reference 5. The reset margin factors into the TM/LP analysis via the part-power radial peaking factors.

-'I, 15.0.8 Component Capacities and Setpoints Table 15.0.8-1 of Reference 5 presents the component setpoints and capacities used in I the analysis.

I*~,

15.0.9 Plant Systems and Components Available for Mitigation of Accident Effects

i,

\.

Table 15.0.9-1 in Reference 5 is a summary of trip functions, engineered safety features, and other equipment available for mitigation of accident effects. These are listed for all Chapter 1: \

15 SRP events.

15.0.10 Effects of Mixed Assembly Types and Fuel Rod Bowing The effects of a _mixed core and fuel rod bowing are given in Section 15.0.1 b of Reference

-I* i 5.

15.0.11 Plant Licensing Basis and Single Failure Criteria The licensing basis and single failure criteria for Palisades are as stated in Section 15.0.11 of Reference 5.

I I

j ii

I.

• 1 ANF-90-181 Page 15

,, 15.1 Increase in Heat Removal by the Secondarv System I The magnitude of the decrease in feedwater temperature (15.1.1), increase in feedwater flow rate (15.1.2), increase in steam flow (15.1.3) and inadvertent opening of a steam generator j* relief/safety valve (15.1.4) is not affected by the subject licensing action. Thus, the relative PCS cooldown rate and severity of each of the above events remains unchanged from previous event dispositions. For this category of events, the Increase in Steam Flow event (15.1.3) remains I bounding of events 15.1.1 , 15.1.2 and 15.1.4.

II

- )' 15.1.1 Decrease in Feedwater Temperature 15.1.1.1 Event Description A decrease in feedwater temperature event may result from the loss of one.*of several of the feedwater heaters. This loss may be due to the loss of extraction steam flow from the turbine

• I generator or due to an accidental opening of a feedwater heater bypass line.

I The event results in a decrease of the secondary side enthalpy leading to an increase in the primary-to-secondary side heat transfer rate. The steam generator outlet temperature on the .

primary side decreases causing the core inlet tempe~ature to also decrease. In the presence of a negative moderator coefficient, reduced core inlet temperature results in an increase in the core I

• power and a decrease in thermal margin.

II 15.1 .1 .2-- - Event Disposition and Justification Reference 4 disposed this event as being bounded by the Increase in Steam Flow event

' (Event 15.1.3). As noted in Section 15.1, the increase in VHP trip rest margin will not change this disposition .

• ANF-90-181 Page 16 15.1.2 . Increase in Feedwater Flow 15.1.2.1 Event Description The Increase in Feedwater Flow event is initiated by a failure in the feedwater system.

The failure may be a result of: 1) a complete opening of a feedwater regulating valve; 2) over-speed of the feedwater pumps with the feedwater valve in the manual position; 3) inadvertent startup of the second feedwater pump at low power; 4) startup of the auxiliary feedwater system; or, 5) inadvertent opening of the feedwater control valve bypass line.

The event results in an increase in the primary-to-secondary side heat transfer rate due to increased feedwater flow. The steam generator outlet temperature on the primary side decreases causing the core inlet temperature to also decrease. In the presence of a negative moderator coefficient, a reduced core inlet temperature results in an increase in the core power and a decrease in thermal* margin. ---

1 "15.1.2.2 Event Disposition and Justification Reference 4 disposed this event as being bounded by the Increase iri Steam Flow event '\*

{Event 1S.1.3). As noted in Section 15.1, the increase in VHP trip rest margin will not change this disposition.

15.1.3 Increase in Steam Flow

• I 15.1.3.1 Event Description This event is initiated by a failure or misoperation of the main steam system that results in an increase in steam flow from the steam generators.. The increased steam flow creates a mismatch between the heat being generated in the core and that being extracted by the steam generators. As a result of this power mismatch, the primary-to-secondary heat transfer rate

ANF-90-181 Page 17

• 1 increases and the primary system cools down. If the moderator temperature coefficient is negative, cooldown of the primary system coolant will cause an increase in reactivity and a I -~-:-- ....

subsequent decrease in thermal margin.

I 15.1.3.2 Event Disposition and Justification

,f The parameters controlling the severity of the transient include: 1) secondary steam flow (load), 2) primary/secondary heat transfer, 3) moderator reactivity coefficient, 4) Doppler reactivity coefficient, 5) reactor safety system setpoints and, 6) scram reactivity.

_-l This event was analyzed for rated power conditions in Reference 5. Increasing the VHP trip reset will not affect the outcome of the rated power event since the maximum high power trip setpoint of 106.5% remains unchanged. Thus, the Excess Load event from rated power is

'\~ounded by the analysis documented in Reference 5.

• I Operating Modes 3 to 8 are not affected by the change in the VHP trip reset and, 4

consequently, are bounded by the full power case (Mode 1)( ). Mode 2, however, is affected l by the change in VHP reset margin because of the increase in radial peaking that occurs at part-power. The allowable peaking factor, as a function of power, is specified by:

I I ..

,, where:

Fr=

Fr,ated=

Allowed radial peaking factor Fr allowed at rated power ri*

P= Fraction of rated power I. . The most limiting initial condition during Mode 2 operation is that which maximizes the allowable radial peaking combined with the maximum trip setpoint of 106.5%. With the VHP trip reset of 15%, the limiting initial power level is 91.5% of rated. Because of the combined effects

.I ANF-90-181 I.

Page 18 of a core power increase that matches the degree of excess load and a depressurization of the ,~

primary system, the event initiated from 91 .5% power will tend to trip on a TM/LP signal. An I analysis of the full power case in Reference 5 demonstrated that the TM/LP trip adequately protects the plant from penetrating DNB limits. Thus, an Excess Load event from 91.5% power I will terminate on a TM/LP trip and results equivalent to those of Reference 5 will occur.

Consequently, the Excess Load event from 91.5% power does not require reanalysis for the I

.change in VHP reset margin.

f, 15.1.4 15.1.4.1 Inadvertent Opening of a Steam Generator Relief or Safety Valve Event Description I*

This event is initiated by an increase in steam flow caused by the inadvertent opening of a steam generator relief or safety valve. The increase in steam flow rate causes a mismatch between the heat generation rate on the primary side and the heat removal rate on the secondary side.

15;1 ..4.2

'I Event Disposition and Justification The increase in steam flow due to opening a steam generator valve is less than that considered in the Increase in Steam Flow event (Event 15.1.3), and therefore is bounded by Event 15.1.3. The increase in VHP trip margin does not change this disposition.

I 15.1.5 Steam System Piping Failures Inside and Outside of Containment I

15.1.5.1 Event Description ,,\

A steam line piping failure event, or steam line break (SLB), is initiated by a rupture of a main steam line pipe causing an uncontrolled steam release from the secondary system. As a result of the uncontrolled release of steam, the heat extraction rate from the primary side is no I

I.

I ANF-90-181 Page 19

'I longer equal to the core heat generation rate. This power mismatch increases the primary-to-

. secondary side heat transfer rate and, consequently, reduces the primary side temperatures.

I) When this overcooling on the primary side is coupled with a negative moderator temperature coefficient, the shutdown margin after scram can potentially. be eroded. Such an erosion of shutdown margin may result in a return-to-power which, in turn, challenges thermal margin.

I 15.1.5.2 Event Disposition and Justification I

,, The consequences of a main steam line break are not dependent on the activation of a specific reactor trip, but rather on the competing effects of shutdown reactivity as a result of a reactor trip and positive reactivity insertion due to moderator overcooling. Thus, changing the

'I VHP trip reset margin has no impact on the consequences of a main steam line break event and the reference analysis remains bounding.

I

'I 1"1 I

ii

'\._ .. '

.. .I ANF-90-181 I

Page 20 15.2 Decrease in Heat Removal by the Secondary System l'

The initiating mechanisms for a Loss of External Load event (15.2.1 ), Turbine Trip event (15.2.2), Loss of Condenser Vacuum event (15.2.3), Closure of the Main Steam Isolation Valves

\I (15.2.4) and the heatup phase of a Feedwater System Pipe Break event (15.2.8) are not affected by the change in VHP trip reset margin. Thus, the relative severity of these events as established

• 1 in prevfous dispositions remains valid. The Loss of External Load event, 15.2.1, as biased in previous analyses is bounding of events 15.2.2, 15.2.3, 15.2.4 and the heatup period of event I

15.2.8.

I.

15.2.1 Loss of External Load I'

15.2.1.1 Event Description

• A Loss of External Load event is initiated by either a loss of external electrical load or a turbine trip. Upon either of these two conditions, the turbine stop valve is assumed to rapidly

• close (0.1 second). Normally, a reactor trip would occur on a turbine trip. However, to calculate I

a conservative system response, the reactor trip on turbine trip is disabled. The steam dump system (atmospheric dump valves- ADVs) is assumed to be unavailable. These assumptions I) allow the Loss of External Load event to bound the consequences of the following events:

Turbine Trip (15.2.2), Loss of Condenser Vacuum (15.2.3) and, Closure of the MSIVs (15.2.4).

I The Loss of External Load event challenges the acceptance criteria for both primary I

system overpressurization and DNBR. The event results in an increase in the primary system temperatures due to an increase in the secondary side temperature. As the primary system

-I temperatures increase, the' coolant expands into the pressurizer causing an increase in the 1~

pressurizer pressure. The primary system is protected against overpressurization by the pressurizer safety and relief valves. Pressure relief on the secondary side is afforded by the steam line safety/relief valves. Actuation of the primary and secondary system safety valves limits the magnitude of the primary system temperature and pressure increase.

I

ANF-90-181 Page 21 With a positive moderator temperature coefficient, increasing primary system temperature results in an increase in core power. The increasing primary side temperatures and power challenge the DNBR and system pressurization acceptance criteria.

15.2.1.2 Event Disposition and Justification The Loss of External Load is credible only for rated power and power operation modes because there is no load on the turbine at other reactor conditions. The rated power conditions bound the consequences for other reactor power operating conditions because of the increased stored energy.

For the pressurization case, the reactor trip system acts to terminate the event by activating a high pressurizer pressure trip signal. Thus, increasing the VHP trip reset margin will not change the primary system pressure response for the pressurization case.

I For the minimum DNBR case, Reference 4 concludes that the consequences of a Loss of Load event are bounded by other Anticipated Operational Occurrences. Modifying the VHP

\I trip reset will not change this disposition since the rated power case is limiting and maximum VHP trip setpoint of 106.5% of rated power is unmodified for this initial operating state. Thus, I the Loss of External Load Event is bounded by previous analyses.

1. 15.2.2* Turbine Trip 15.2.2.1 Event Description I

This event is initiated by a turbine trip which results in the rapid closure of the turbine stop I valves. A reactor trip would occur on a turbine trip and the steam dump system would operate to mitigate the consequences of this event. The primary system is protected against overpressurization by the pressurizer safety and relief valves. Pressure relief on the secondary side is afforded by the steam line safety/relief valves.

ANF-90-181 Page 22 15.2.2.2 Event Disposition and Justification

• The assumptions made in the Loss of External Load event bound the consequences of a turbine trip. Specifically, the Loss of External Load event considers: a conservatively fast turbine stop valve closure time, no reactor trip on a turbine trip and, the unavailability of the atmospheric dump valves. Thus, the Turbine Trip event is disposed as being bounded by the Loss of .External Load event (Event 15.2.1 ).

15.2.3 Loss of Condenser Vacuum 15.2.3.1 Event Description I

This event is initiated by a reduction in the circulating water flow or an increase in the circulating water temperature which can impact the condenser back pressure. This condition can result in a turbine trip without the availability of steam bypass to the condenser. The primary system is protected against overpressurization by the pressurizer safety and relief valves. 1.\

Pressure relief on the secondary side is afforded by the steam line safety/relief valves.

-15.2.3.2 Event Disposition and Justification I)

I The assumptions made in the Loss of External Load event bound the consequences of a Loss of Condenser Vacuum. Therefore, this event is disposed of as being bounded by the Loss of External Load event (Event 15.2.1 ).

15.2.4 Closure of the Main Steam Isolation Valves (MSIV) (BWR) I

. 15.2.4.1 Event Description I Closure of the Main Steam Isolation Valves event is initiated by the loss of control air to II'_

the MSIV operator. The valves are swinging check valves designed to fail in the closed position.

I

'~

I ANF-90-181 Page 23 The inadvertent closure of the MSIVs is primarily a BWR event, however, the closure of these

'I valves in a PWR can drastically reduce the steam load.

I 15.2.4.2 Event Disposition and Justification I' The closure time of the MSIVs is less than 5 seconds, but greater than the value used in Event 15.2.1 (0.1 seconds). A MSIV closure event will progress in a similar fashion as a Loss of II External Load (Event 15.2.1 ), but at a slower rate. The consequences Of Event 15.2.1 will bound those for Event 15.2.4 because of the more rapid valve closure time. The increase in VHP trip I reset margin does not affect this dispositton.

'I 15.2.5 Steam Pressure Regulator Failure

. Palisades does not have steam pressure regulators. Therefore, the Steam Pressure Regulator Failure event is not considered in this analysis.

15.2.6 Loss of Nonemergency A.C. Power to the Station Auxiliaries 15.2.6.1 Event Description A Loss of Nonemergency AC. Power to the Station Auxiliaries event may be caused by a complete loss of the offsite grid together with a turbine generator trip or by a failure in the onsite AC. power distribution system.

The loss of AC. power may result in the loss of power to the primary coolant pumps and condensate pumps which, in turn, results in the loss of the main feedwater pumps. . The combination of the decrease in primary coolant flow rate, the cessation of main feedwater flow and trip of the turbine generator compounds the event consequences. The decrease of both primary. coolant flow and main feedwater decreases the primary-to-secondary. system heat transfer. rate resulting in the heatup of the primary system coolant. The increase in primary I

ANF-90-181 Page 24 system coolant temperature increases the overpressurization potential and increases the threat of penetrating DNB.

The event is most limiting when initiated from full power conditions. During this mode of operation the stored heat in the fuel rods is maximized and the margin to DNB is minimized.

15.2.6.2 Event *Disposition and Justification

• This event can be separated into two distinct phases: the near-term and the long-term.

• The near-term phase is characterized by the loss of power resulting in the coastdown of the primary coolant pumps, the coastdown of the main feedwater pumps and the trip of the turbine generator. The coastdown of the primary coolant pumps causes an immediate reduction in thermal margin. The trip of the reactor and the subsequent insertion of control rods terminates I

the challenge to DNB limits.

The near-term phase of the event is similar to that of a Loss of Forced Reactor Coolant Flow transient (Event 15.3.1). The near-term *consequences of this event are bounded by Event 15.3.1.

The long-term consequences of a Loss of A.C. Power event are determined by the heat removal capacity of the auxiliary feedwater system .. The long-term portion is similar to' the Loss of Normal Feedwater Flow transient (Event 15.2.7). The long-term effects are, therefore, addressed by the Loss of Normal Feedwater Flow event. Chan.ging the VHP trip reset does not alter this disposition.

15.2.7 Loss of Normal Feedwater Flow

. 15.2.7.1 Event Description A Loss of Normal Feedwater Flow transient is initiated by the trip of the main feedwater I

ANF-90-181 Page 25 pumps or a malfunction in the feedwater control valves. The loss of main* feedwater flow decreases the amount of subcooling in the secondary-side downcomer which diminishes the primary-to-secondary system heat transfer and leads to an increase in the primary system coolant temperature. As the primary system temperatures increase, the coolant expands into the pressurizer which increases the pressure by compressing the steam volume.

The opening of the secondary-side safety valves controls the heatup of the primary-side.

The long-term cooling of the primary system is governed by the heat removal capacity of the auxiliary feedwater flow. The auxiliary feedwater pumps are automatically started upon a steam generator low liquid level signal.

15.2.7.2 Event Disposition and Justification A Loss of Normal Feedwater

• Flow event is only credible for rated power and power operating conditions. The worst consequences occur when the feedwater is lost during rated I power operation since more stored heat is contained in the fuel than in other modes of operation.

I *For the initial PCS heatup phase of the transient,. both the DNB and the primary system overpressurization acceptance criteria are challenged. The DNB challenge is maximized when I it is assumed that pumps.

offsit~ *power is lost which results in a coastdown of the primary coolant After the reactor trip system is activated, the core power *is drastically reduced I alleviating the challenge to DNB. The loss of forced reactor coolant flow event (Event .15.3.1) bounds the short term DNB consequences of a loss of normal feedwater transient.

I For the longer term phase of the transient, the slow PCS heatup threatens both the overpressurization of the primary system by filling the pressurizer with liquid, and the dryout of I the steam generators. The parameters influencing the severity of the long term phase of the I .I transient include: 1) decay heat generation, 2) secondary safety/relief valve settings, 3) primary coolant pump operation, 4) auxiliary feedwater flow rate, an~ 5) steam generator secondary *side mass at the time of reactortrip.

I

ANF-90-181 Page 26 Consequently, the revision of the VHP trip reset margin has no effect on the consequences of the long term phase of the Loss of Normal Feedwater transient. Thus, the long I ,

term transient is bounded by previous analyses ..

.I 15.2.8 Feedwater System Pipe Breaks Inside and Outside Containment

• 1.

.15.2.8.1 Event Description I

A Feedwater System Pipe Break event occurs when a main feedwater system pipe is

• ruptured. The ruptured pipe will cause a blowdown of the affected steam generator if the break I occurs upstream of the *feedline check valve. If the rupture occurs downstream of the check valve, the event would behave much like the Loss of Normal Feedwater Flow transient. Since the auxiliary feedwater flow is injected into the steam generators via a separate piping network I

than the main feedwater, the delivery of auxiliary feedwater will not be interrupted by the pipe rupture.

The event results in both a primary system cooldown and a heatup. Initially, the event I

results in a cooldown of the primary-side coolant due to the energy removal during the-blowdown stage of the event. The eventual depletion of secondary-side inventory and lack of main I

. feedwater will cause the primary system to heatup much like a Loss of Normal Feedwater Flow event.

I 15.2.8.2 Event Disposition and Justification I Reference 4 disposed this event as being bounded during rated power operation for the I following two reasons: 1) The cooldown aspect of the event is bounded by the steam line break event (Event 15.1.5) and, 2) The heatup effects are bounded by the Loss of External Load event I (Event 15.2.1) for the primary system overpressurization and the Loss of Normal Feedwater Flow event (Event 15.2.7) for the long-term cooling requirements. Feedwater pipe breaks from modes I other than rated power result in a primary system*cooldown and are bounded by the steam line I

ll I

,. ANF-90-181 Page 27 I break accident. Revising the VHP trip reset margin has no effect on this disposition. Thus, this event is bounded by Events 15.1.5, 15.2.1and15.2.7.

II I

I I

I I

II I

"I,,

ANF-90-181 I

-I Page 28 15.3 Decrease in Reactor Coolant System Flow I

15.3.1 Loss of Forced Reactor Coolant Flow

,I 15.3.1 .1 Event Description I

The Loss of Forced Reactor Coolant .Flow transient is initiated by a disruption of the electrical power supplied to or a mechanical failure in a primary coolant system (PCS) pump.

These failures may result in a complete or partial loss of forced coolant flow.

I The impact of losing a PCS pump or pumps _is a decrease in the active flow rate in the reactor core and, consequently, an increase in core temperatures. Prior to reactor trip on low PCS flow, the combination of decreased flow and increased temperature poses a challenge to I*

DNB limits.

15.3.1.2 Event Disposition The most limiting loss of flow transient is initiated from rated power(S). The. transient is initiated by tripping all four primary coolant pumps. As the pumps coast down, the core flow is reduced, causing a reactor scram on low primary flow. Since the event is terminated by the.low flow trip, revising the VHP trip reset margin has no effect on the consequences of this event.

. Thus, this event is bounded by the Reference 5 analysis .

15.3.2 Flow Controller Malfunction There are no flow controllers on the PCS at Palisades. Therefore, this event is not credible.

I

I **

I I ,- ANF-90-181

• ~ Page 29

,- 15.3.3 Reactor Coolant Pump Rotor Seizure

(**- ....

15.3.3.1 Event Description I

This event is initiated by a seizure of a PCS pump rotor. The seizure causes an I immediate reduction in PCS flow rate. As in the Loss of Forced Coolant Flow event {Event

'II 15.3.1 ), the impact of losing a PCS pump is a decrease in the active flow rate in the reactor core and, consequently, an increase in core temperatures. As in Event 15.3.1, the PCS flow is reduced rapidly, and the combination of decreased flow and increased temperature poses a challenge to DNB limits. The event is terminated by the PCS low flow trip.

I 15.3.3.2 Event Disposition The controlling parameters for the pump seizure event are identical to those for the loss of flow event. Therefore, revising the VHP trip reset has r:io impact on this event and the

• 1 Reference 5 analysis remains bounding.

I 15.3.4 Reactor Coolant Pump Shaft Break I 15.3.4.1 Event Description This event is initiated by a failure of a PCS pump shaft which results in a free-wheeling I ..

impeller. The impact of a coolant pump shaft break is a loss of pumping power from the affected

• 1 pump and a reduction in the PCS flow rate. The flow reduction due to the seizure of a pump .

rotor is more severe than that for a shaft break; however, the potential for flow reversal is greater for the shaft break event. The event is terminated by the low reactor ~oolant flow trip.

I 15.3.4.2 Event Disposition The event is most limiting at rated power conditions because of a minimum margin to I

.I I

ANF-90-181 Page 30 DNBR limits. The initial flow reduction for this event is bounded by that for the Reactor Coolant Pump Rotor Seizure event (Event 15.3.3). The potential for greater reverse flow due to a shaft I.

break is accounted for in the seized rotor analysis by decreasing the rotor inertia to zero at the time of predicted reversed flow. The severity of the pump shaft break event is bounded by Event 4

t 15.3.3( ). Thus, this event is bounded by Event 15.3.3 as in previous dispositions.

I I

I I

I I

I I

I

.1

I.

I

,. ANF-90-181 Page 31 I 15.4 Reactivity and Power Distribution Anomalies I 15.4.1 Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition I 15.4.1.1 Event Description I This event is commenced by an uncontrolled withdrawal of a control .rod bank. This withdrawal adds positive reactivity to the core which leads to a power excursion. Event 15.4.1 I considers the consequences of the control bank withdrawal at subcritical or low power startup initial power levels.

I As the control bank is withdrawn, the positive reactivity insertion causes a significant core power increase as the reactor approaches prompt criticality. As the core power increases, the core average and hot leg temperatures also increas~. Due to the increasing power and I temperatures, the DNB limits are challenged.

I, 15.4.1.2 Event Disposition and Justification Increasing the margin oh the VHP trip reset margin will have no effect on the core power I response of this event. The event is terminated by the minimum VHP trip setpoint (i.e., 30% of rated power) which is unaffected by the change in reset margin. Thus, this event is bounded by I the reference analysis.

I

,, 15.4.2 15.4.2.1 Uncontrolled Control Rod Bank Withdrawal at Power Event Description As with Event 15.4.1 , this event is initiated by an uncontrolled withdrawal of a control rod bank. This withdrawal adds positive reactivity to the core which leads to potential power and I

.I ANF-90-181 I'

Page 32

• temperature excursions. Event 15.4.2 considers the consequences of control bank withdrawals at rated and power operating initial power levels.

I As the control bank is withdrawn, the positive reactivity insertion causes an increase in

.I core power and in primary coolant system temperatures. Due to the increasing power and temperatures, the DNB limits are challenged. In most cases, the transient will terminate on a I

VHP, a TM/LP or a high pressurizer pressure trip; however, some cases do not activate a reactor protection system trip. I 15.4.2.2 Event Disposition and Justification I The uncontrolled rod withdrawal from rated power (Mode 1) is not affected by the I increase in the VHP trip reset margin because the maximum trip setpoint remains at 106.5% of rated power. Mode 2 operation, however, is affected by the increase in VHP trip reset since increased part-power radial peaking must be considered. The allowable peaking factor as a function of power is specified by Reference 8: I Fr= Frrat8d*(1.0+0.3*(1-PJ) .I where:

Fr = Radial peaking limit I

Frrated p

=

=

Rated power Fr Fraction of rated power I

The most limiting part-power initial condition is that which maximizes the allowable Fr.

I while still allowing the maximum trip setpoint of 106.5%. With the VHP trip reset of 15%, the initial power level of 91.5% of rated satisfies these c~nditions. Thus, this event initiated from 91.5%

I power will be analyzed for the increase in VHP trip reset margin.

The objective of the analysis is to demonstrate the adequacy of the trip setpoints to I

I, I ANF-90-181 Page 33 I assure meeting the acceptance criteria. To assure this objective, the analysis considers a spectrum of reactivity insertion rates. Since neutronic feedback as a function of cycle exposure I and design also influences the results, these. effects are also included in the analysis.

• 1 This event is classified as a moderate frequency event (Table 15.0.1-1). The acceptance criteria are as described in Section 15.0.1.1 of Reference 5. The single failure criteria are given in Section 15.0.11 of Reference 5. The safety systems challenged in this event are redundant I and no single active failure will adversely affect the consequences of the event.

I 15.4.2.3 Analysis I The analysis is performed using the PTSPWR2 code(9) and XCOBRA-lllC( 1O).

PTSPWR2 code models the salient system components and calculates neutron power, fuel The thermal. response, and fluid conditions. The fluid conditions and rod surface heat transport at the time of MDNBR are input to the XCOBRA-lllC code for calculation of the MDNBR. Systems which minimize DNBR are enabled in the* analysis.

I The analysis evaluates the consequences of an uncontrolled control rod bank withdrawal from 91.5% of rated power. A spectrum of reactivity insertion rates was evaluated in order to I bound events ranging from a slow dilution of the primary system boron concentration to the fastest allowed control bank withdrawals. Specifically, the analysis encompasses reactivity 6 5 insertion rates from 1. x 1o- to 5. x 1o- ll p/sec.

I Figure 15.4.2-1 shows MDNBR versus Reactivity Insertion Rate for this event. MDNBR I

,, versus insertion rates are shown for both positive (BOC) and negative (EOC) feedback. MOC kinetics are bounded in the analysis by considering conservatively bounding BOC and EOC kinetics, along with a comprehensive range of reactivity insertion rates. The range of insertion rates was conserVatively c*alculated based on control rod worth and withdrawal speed.

The limiting uncontrolled control rod bank withdrawal at 91.5% power and BOC kinetics I

I ANF-90-181 I

• Page 34 5

occurred at an insertion rate of 2.25 x 1o- /!t. p/sec. The reactor tripped on a TM/LP signal. The 11 bounding MDNBR for this event is 1.420 using the XNB( ) correlation. For comparative purposes, the MDNBR calcu.lated using the ANFP correlation(12*13) is 1.586.

The maximum peak pellet LHR occurs in the case which uses BOC kinetics. The I:

maximum peak pellet LHR is calculated to be 18.1 kW/ft. The sequence of events for the Uncontrolled Bank Withdrawal transient is given in Table 15.4.2-1. The transient responses of I.

• key system variables are given in Figures 15.4.2-2 to 15.4.2-10.

I 15.4.2.4 Conclusion I

Reactivity insertion transient calculations demonstrate that the DNB correlation limit will not be penetrated during any credible reactivity insertion transient. The maximum pea_k pellet linear heat generation" rate for this event is less than the fuel centerline melt criterion of 21 kW/ft.

Applicable acceptance criteria are therefore met and the adequate functioning of the reactor protection system is demonstrated.

.1 I

I I

I 1*

I

.I, I

,. ANF-90-181 Page 35 I Table 15.4.2-1 Event Summary for the Uncontrolled Rod Bank Withdrawal Event from 91 .5% of Rated Power I

Event Time (sec)

I Start Rod Withdrawal 0.00 Letdown flow valve open 0.00 I Minimum DNBR Reactor Scram (TM/LP Trip) 1.420 22.95 24.82 I Turbine Stop Valve closed Peak Power Level 2900.9 MWt 25.00 25.38 I Peak Core Average Heat Flux Peak Core Average Temperature 178,867 Btu/hr-ft 576.53 °F 2

25.45 25.77 Peak Pressurizer Pressure 2267.14 psia 27.00 Steam Line Safety Valves open 29.50 I Peak Steam Dome Pressure 1039.58 psia 31.69 I

I

'I I

'I, I

---

BOC 1.7 ... -~-me

- - - 1.19 MONBR LIMIT 1.6 1.5 fU EOC CASES TRIPPED ON TM/LP TRIP Cl:::

m z

Cl I: 1.4 HIBH PRESSLfiE TRIP HIGH TM/LP TRIP J.3 - POWER TRIP 1.1 -

I ,___ _ _......__ _..__...._......__.__.__.__._..__*_ _ __.__ _...___..___._---L---L--'--L......1 10-(1 101 10-i REACTIVITY INSERTION RATE, DEL.TA RHO/SEC flGLRE 15.1.2-1 MONBR Vfl.UES fOR UNCONTROLLED BAl'I< MITtDRAMAL INITIATED fROM 91.SZ Of RATED POWER

)>

z

-u Ill 71 co

<O 0 (D

I W CD m ......

3500 3000

..., 2500 3:

l::

Q:'.

w 2000 3:

0 0....

1500 1000 500 0 .__....._._,__.__.__.___._......._......._...._....__...........___.__.__.__.___._......._......_..._....._...___.__.__.__.__.___._....1-......._..._....._.i.....J 0 5 10 15 20 25 30 35 TIME:, SE:C FIGURE 15.i.2-2 REACTOR POWER LEVEL FOR )>

z UNCONTROLLED BANK WITHDRAWAL AT PART POWER "U "Tl O> cb co CD 0

I w Q)

..0

.: 20.0 18.0 N

(-4 L...

16.0 I

~

I:

) 14.0 m

x

) 12.0

...J L...

cc w 10.0

I:

w

(.!)

cc

~ 8.0 w

cc 6.0 1.0 2.0 0 5 10 15 20 25 30 35 TIME, SEC flGURE 15.4.2-3 CORE AVERAGE HEAT fLUX fOR z>

"'U -;n UNCONTROLLED BANK WITHDRAWAL AT PART POWER Ill co

<O 0 (1)

I

(,.) CX>

Cl> _..

2400 a:

(/)

a... 2200 w

0::

(/)

(/)

w 0::

a... 2000 1800.

1600 L.....,j.....__.__,__.__,__._......___,_...._"--...__...........__,_-J..._,__.__,__,_..L-..__.__..._.___.__.___.__._.......__,_...._..__.__.___J 0 5 . 10 15 . 20 25 30 35 TIME:, SEC

)>

flGURE 15.4.2-4 PRESSURIZER PRESSURE fOR z UNCONTROLLED BANK WITHDRAWAL AT PART POWER -a -;n D> <O (Q 0 CT> .....

I w

<O OJ

13

_J w

> 12 w

_J II 10 9 ...__.__.__.____._....__._....._....................._.__.____.______.._....__.__.__.____.__,_....._....._....................._.__.___.__.__.

0 5 10 15 _20 25 30 35 TIME, SEC

)>

z flGURE 15.4.2-5 PRESSURIZER LIOUIO LEVEL "'U 7" II> <O fOR UNCONTROLLED BANK WITHDRAWAL AT PART POWER <O 0 (l) I

~ O>

0 .......

...a

-:.: so.a I I I I I -, .

15.0 ... -

(.)

w

(/)

c iO.O ... -

co

_J w

f-<

cc a:: 35.0 ... -

J::

a_J L...

(/)

(/)

cc

c 30.0 ... -

25.0 ...

20.0 ..._.__....___.___._,_*_.___.___.___._~*___._,__,__.__.__.,...___.___._.__...__..__....___.___.__._._,__.__,____._-L-_._.....___,__.L--1 0 5 10 15 20 25 30 35 TIME, SEC

)>

f IGURE 15 .* 1. 2-6 PR INARY COOLANT SYSTEM MASS fLOW z

-u -;n

<O RATE fOR UNCONTROLLED BANK WITHDRAWAL AT PART POWER Q) co 0

(!) _..

I

~ OJ.

• • • • • • • • • • • • ** PCS AVERAGE

- - - COLO LE:G

-~-HOT LE:G 625

---

CORE INLET w ******************

a::::

l f-t 575 ******************************************************************************************************* *****

cc a::::

w a...

I:

w f-t 5~ L------~~~---------- . . . .,....,.,=-~-=-=--=-=-~-=--~-~-~--:::-:::-.

525 500 .___.__.,___.___.___.__.__.__.....__._..._.__..___.__.,___,__.___.__.__.___.__._..._..........._.__.,___.__.___.__.__.__.....__._..___.

0 5 10 15 20 25 35 TIME:, SEC

)>

flGURE 15.4.2-7 PRIMARY COOLANT SYSTEM z

-a "Tl TEMPERATURES fOR UNCaNTROLLED BANK WITHDRAWAL AT PART POWER Q) I

<O ~,

CD

~CX>

I\) ..... -

- -~ ---- - --- .. .. -

2.0 a::::

a: .......... ..................

_J . 0.0 ......................................................... ::~::=-- ... ~::*..: __ -- - - - -- -

_J 0

Cl

> -2.0 f-<

(.)

a:

w a::::

-i.O

---TOTAL

---

MODERATOR

• • • • • • • • • • • • • • DOPPLER

-6.0

-8.0 .__..._.__.._.___.__.__,__,_...._..._...._._.....___.__.___.____.__...__,_...L-_._...._..._.--1._.__.__,__._...J..__..__,_...._i__J 0 5 10 15 20 25 30 35 TIME, SEC

")>

fIGURE 15.4.2-8 REACTIVITIES fOR z UNCONTROLLED BANK WITHDRAWAL AT PART POWER "U -;n I>> <O

<O Cl) 0 I

.:;... OJ w .....

STEAM GENERATOR wl STEAM GENERATOR *2 1100 a: 1000 (J")

Q_

w a:::

> 900 (J")

(J")

w a:::

a....

BOO 700 600 0 5 10 15 20 25 30 35 TIME SEC 1

)>

flGURE 15.i.2-9 SECONDARY PRESSURE fOR z

-u .,,

I)) I UNCONTROLLED BANK HITHORAWAL AT PART POWER co ~

(1)

I

~ OJ.....

- - .. - - - - - - .. - - - -a- -

-2.0 '*"---------------------------1(

~

. -1.0 L...

_J w

w

_J

-6.0

- - - STEAM GENERATOR

• l

- - * - " - STEAM GENERATOR *2

-8.0

-10. 0 ...._...___._....__.__.__.__._........__.__..__..........__..__.__.__.__,__........__.__..__..........__..__.___.__._......_........__.__..__..........__.___.___.

0 5 10 15' 20 25 30 35 TIME, SEC

)>

fIGURE 15.4.2-10 STEAM GENERATOR LIQUID LEVEL z fOR UNCONTROLLED BANK WITHDRAWAL AT PART POWER ""U Ill 7'co1 co 0 (I) __.I

~ CX>

01 __.

I I

ANF-90-181 Page 46 15.4.3 Control Rod Misoperation I

The control rod misoperation event considers a number of different event initiators. These include: 1) Dropped control rod or bank; 2) Dropped part-length control rod; 3) Malpositioning I

of a part-length control rod group; 4) Statically misaligned control rod or bank; 5) Single control rod withdrawal; and 6) Core barrel failure.

I Each of the above events includes a redistribution of power which leads to a local I augmentation of the peaking factor in the affected region ofthe core.

I 15.4.3.1 Event Description I

(1) . Dropped Control Rod/Bank A control rod drop event is initiated by a de-energized control rod drive mechanism .

(CROM) or another failure in the control rod system. With the insertion of negative reactivity due to the dropped rod, the core power decreases. Moderator and Doppler temperature feedback.

I driven by a constant turbine generator load, cause the power to increase to its initial state.* A localized increase in the radial peaking results from power redistribution due to the dropped rod.

I This event is a challenge to DNB limits because of radial peaking augmentation together with near full power*operating conditions.

I (2) Dropped Part-Length Control Rod Part-length control rods are not used during power operation and are maintained in a I

withdrawn state. A failure of the rod brake mechanism could result in a part-length control rod drop. I I

I.

I ANF-90-181

~

Page 47 I (3) Malpositioning of a Part-Length Control Rod Group Use of part-length control rods is not allowed during power operation. The part-length I control rods are maintained in a fully withdrawn. state; therefore, this event is not credible.

I (4) Statically Misaligned Control Rod/Bank I A static misalignment occurs when a malfunction in the CROM causes a control rod to be out of alignment with its bank or a control. group to be in violation. of its. Power Dependent I Insertion Limits.

I In the case of a static misalignment of a control rod, one control rod is positioned out of the core while the balance of the control bank is inserted. This situation causes a localized increase in radial peaking in the affected region of the core. The increased radial peaking, together with the initial core power level, can significantly reduce the margin to DNB. The reverse I condition, i.e. one control rod fully inserted with its bank fully withdrawn, is .essentially the same as a dropped control rod event..

I . (5) Single qontrol Rod Withdrawal I The withdrawal of a single control rod results in a reactivity insertion and a localized increase in radial peaking. The degradation of core conditions characteristic of a reactivity I insertion transient, combined with an increase in local radial peaking, poses a challenge to DNBR limits.

I (6)

• Core Barrel Failure I

This event is initiated by the circumferential rupture of the core support barrel. The core stop supports serve to support the barrel and the reactor core by transmitting all loads directly to the vessel. The clearance between the core barrel and the supports is approximately one-half I

I I

ANF-90-181 Page 48 inch at operating temperatures. The worst possible axial location of the barrel rupture is at the midplane of the vessel nozzle penetrations so that a direct flow path is formed between the inlet I

and exit nozzles in parallel with the path that goes through the core. The core sustains a smal_I reactivity transient induced by the motion of the core relative to the inserted rod bank(s).

I Reactor protection for the Core Barrel Failure event during hot shutdown, refueling 1*

shutdown, cold shutdown, and refueling operating conditions is provided by Technical Specification Shutdown Margin requirements. For the reactor critical and hot standby operating I

conditions, reactor protection *is provided by the VHP trip and a norisafety grade high rate-of-change power trip. For the rated power and power *operating conditions, reactor I protection is afforded by the VHP and TM/LP trips.

I 15.4.3.2 Event Disposition and Justification (1) Dropped Control Rod/Bank I

The licensing analysis(S) performed at rated power initial conditions conse~atively disregarded the VHP trip. Consequently, adjusting the VHP trip reset margin has no affect on the results of this event. The consequences of this event at rated power initial conditions bounds I

those for other modes of operation. Thus, this event is bounded ~y the Reference 5 analysis. I (2) Dropped Part-Length Control Rod I

A dropped part-length control rod will not be as severe as a dropped full-length control rod and is, therefore, bounded by Event 15.4.3(1 ).

I (3) Malpositioning of the Part-Length Control Rod Group I

Use of part-length control rods is not allowed during power operation. The part-length control rods are maintained in a fully withdrawn state; therefore, this event is not credible.

I

ANF-90-181 Page 49 I (4) . Statically Misaligned Control Rod/Bank I The consequences of a statically misaligned control rod or bank are evaluated at rated power conditions with an augmented radial peaking factor to account for possible power redistribution. This event does not rely on any specific reactor trip; thus, increasing the VHP trip I* reset margin has no effect on the results of this event. Thus, the statically misaligned control rod is bounded by the *Reference 5 analysis.

I (5) Single Control Rod Withdrawal I

The withdrawal of a single control rod results in a reactivity insertion and a localized I increase in radial peaking. The degradation of core conditions characteristic of a reactivity insertion transient, combined with an increase in local radial peaking, poses a challer:ige to DNBR

. limits.

I The disposition of this event is controlled by the same parameters as Event 15.4.2, and thus, will require reanalysis for the increase in VHP trip reset margin. The *single control rod.

I withdrawal is actually a continuation of the respective reactivity insertion rate curves generated for Event 15.4.2.

• Consequently, as was the disposition of Event 15.4.2, the part power rod I withdrawal initiated from 91.5% power is dispositioned to be analyzed. The rated power single rod withdrawal event is bounded by the Reference 5 analysis.

I The conseq1.Jences of a single rod withdrawal from Modes 3, 4, and 5 are either bounded or the event does not challenge the acceptance criteria. Mode 3 operation* (Reactor Critical) is I 4 defined as having a power greater than 1o- % and Tave greater tha,n 525°F. Since the peak power obtained. during a low power reactivity insertion increases with increasing insertion rate, I the results for a single rod withdrawal are bounded by the results for a bank withdrawal (Event 15.4.1) where the insertion rate is much larger(S). Mode 4 operation (Hot Standby) applies when 4

the power is between 1o- % and 2% and any of the control rods are withdrawn. The peak heat flux following a rod withdrawal decreases with increasing initial power level. Since Mode 3 I

I I

ANF-90-181 Page 50 4

includes 1o- % power, Mode 4 is bounded by the results of Mode 3. Finally, Mode 5 operation I

4 (Hot Shutdown) applies when the power is less than 1o- % and Tave is greater than 525°F. The most reactive rod worth is less than the required shutdown margin; therefore, the reactor could I

not become critical by the withc;jrawal of any single rod.

• I (6) Core Barrel Failure I

A core barrel failure is initiated by a circumferential rupture of the core barrel support.

During this event, the core experiences a small reactivity insertion due to motion of the core I relative to the control rods. The event is established to be incredible during hot shutdown, refueling shutdown, cold shutdown and refueling operation due to the Technical Specification I shutdown margin requirements. The event initiated from rated power bounds the power operating, reactor critical and hot standby operating modes. At rated power, this event is bounded by the consequences of the Control Rod Ejection event (15.4.8) which has a more limiting reactivity insertion rate and radial power redistribution.' Adjusting the VHP trip reset I

margin will not affect this disposition.

15.4.3.3 Analysis I

In the analysis of the single rod withdrawal event, the core boundary conditions of I

average heat flux, temperature, pressure and flow are selected to conservatively bound the consequences of this event at 91.5% of rated power. *The bank withdrawal analysis {15.4.2)

.. I 6

considers reactivity insertion rates down to 1.00 x 1o- il p/sec which is bounding of a single rod.

The boundary conditions used in the calculation of MDNBR were obtained from the limiting I

transient response from Event 15.4.2. Those conservatively biased core boundary conditions are combined in an XCOBRA-lllC calculation with a radial augmentation peaking factor calculated I

to bound the possible single rod withdrawal radial power redistribution. A radial peaking augmentation factor of 1.08 was used. Based on the peaking factors given in Table 15.0.3-1 for Reload L fuel, the bounding MDNBR for this event is 1.197 using the XNB correlation. For I

I I ANF-90-181

~

Page 51 I comparative purposes, the MDNBR calculated for Reload M fuel using the ANFP correlation is 1.325. The peak LHR for this event is 19.6 kW/ft.

I 15.4.3.4 Conclusion I For the single control rod withdrawal, the MDNBR for this event is greater than the 95/95 DNBR limit for the XNB correlation. The peak LHR is less than the 21 kW/ft limit for centerline I melt. Thus, all applicable acceptance cri~eria are met for this infrequent event.

I 15.4.4 Startup of an Inactive Loop I 15.4.4.1 Event Description This ev~nt is initiated by the startup of an inactive primary coolant pump. The startup of ari inactive pump can lead to an introduction of .colder primary coolant into the reactor core. The I lower coolant temperature, together with a negative moderator temperature coefficient, can cause an increase in core power and a degradation of DNB margin. Sufficient protection is available I to reduce the consequences of this event.

I 15.4.4.2 Event Disposition and Justification Continuous power operation with less than four primary coolant pumps is not allowed by I the Technical Specifications. Additionally, startup with less than four primary coolant pumps

• above hot shutdown is not allowed. Thus, this event is most limiting for an initial condition of I three operating primary coolant pumps with the corresponding reduced power level and VHP trip setpoint.

I For operation with one inoperative pump, the low flow trip setpoint and the VHP trip setpoint are changed to the allowable values for the selected pump condition. The maximum VHP trip setpoint under these conditions is 49% of rated power and is unchanged by the I

I I

ANF-90-181 Page 52 adjustment to the trip reset margin. Thus, this event is not reanalyzed for the increase in VHP trip reset margin.

I 15.4.5 Flow Controller Malfunction I

There are no flow controllers on the PCS at Palisades. Therefore, this event is not I credible.

I 15.4.6 CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant I

15.4.6.1 Event Description I

A borori dilution event can occur when primary grade water is added to the primary coolant system via the Chemical Volume and Control System (CVCS) or the accidental transfer of the contents of the iodine removal system during cold shutdown or refueling shutdown conditions.

I The dilution of primary system boron adds positive reactivity to the core. This event can I lead to an erosion of shutdown margin for subcritical initial conditions, or a slow power excursion for at-power conditions. A boron dilution at rated or power operating conditions behaves in a I manner similar to a slow uncontrolled rod withdrawal transient (Event 15.4.2).

I For Operating Modes 2-7, the boron dilution event must not cause the core to become critical in less than 15 minutes from the time of initiation. For the refueling mode, the core must I not become critical prior to 30 minutes from the time of initiation.

15.4.6.2 Event Disposition and Justification I

Changing the VHP trip reset margin will not affect the events initiating from subcritical I

I I ANF-90-181 Page 53 I configurations. *The consequences of the boron dilution event for power operation are addressed by Event 15.4.2.

I 15.4.7 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position I 15.4.7.1 *. Event Description I An inadvertent loading of a fuel assembly in an improper position can result in. an I alteration* of the power distribution in the core which can adversely affect thermal margin.

15.4.7:2 Event Disposition *and Justification I

The event is precluded due. to the administrative c~mtrols and procedures,* including startup testing, that ensure aproperly loaded core. The change in the VHP trip reset margin will not invalidate this disposition; consequently, this event does not require reanalysis.

I 15.4.8 Spectrum of Control Rod Ejection Accidents I '15.4.8.1 Event Description I This event i~ initiated by a failure in the CROM pressure housing causing a rapid ejection I of th.e affected control rod. The ejection of the control rod inserts positive reactivity causing an increase in core power. Because of the increase in core power and the radial power distribution I in the vicinity .of the ejected control rod, this* event threatens both DNBR . and PCS overpressurization acceptance criteria. In addition, deposited fuel enthalpy is evaluated.

I 15.4.8.2 Event Disposition and Justification T.he parameters affecting the severity of the transient are: 1) the reactivity of the ejected control rod, 2) the reactivity insertion rate, 3) the radial peaking augmen.tation factor, 4) the I

I ANF-90-181 I

Page 54 Doppler reactivity coefficient and, 5) the VHP trip setpoint. Reference 5 states that the hot full power control rod ejection event is more limiting than the event initiated from hot zero power.

I

• • The Reference 5 analysis was performed in a manner that bounds all modes of operation. The maximum pressurizer pressure response for the Control Rod Ejection transient is bounded by I

the Loss of External Load event. Deposited fuel enthalpy at hot full power bounds that at part-power conditions. Thus, changing the VHP trip reset margin does not affect the consequences I

of this event.

I 15.4.9 Spectrum of Rod Drop Accidents (BWR)

I This event is not applicable to Palisades since it is not a BWR. .

I I

I I

I' I

I I

I I ANF-90-181

~

Page 55 I 15.5 Increases in Reactor Coolant System Inventory 15.5.1 Inadvertent Operation of the ECCS that Increases Reactor Coolant Inventory I

15.5.1 .1 Event Description I

This event is caused by an inadvertent actuation of the ECCS that results in an increase I in the primary system inventory. The primary challenge is to the primary system overpressurization criteria.

I 15.5.1.2 Event Disposition and Justification I The previous disposition concluded that the PCS pressurization for this event is bounded by the Loss of External Load event (15.2.1 ). Since modification of the VHP trip reset margin does not affect the pressurization rate from the charging pumps or the relief capacity of the PCS safety I valves, Event 15.2.1 remains bounding for this event. The potential boron dilution consequence of this event is bounded by the boron dilution event (15.4.6).

I 15.5.2 CVCS Malfunction that Increases Reactor Coolant Inventory I 15.5.2.1 Event Description I A malfunction in the CVCS could result in the inadvertent operation of the charging system pumps. If the letdown system is not operating, the result leads to an increase in the I primary system coolant inventory .and, potentially, an overpressurization of the primary system and/or a dilution of the primary system boron concentration.

I 15.5.2.2 Event Disposition and Justification The event initiators and significant parameters remain unchanged for operation with the I

I I

ANF-90-181 Page 56 VHP trip reset margin increased.

increase in VHP trip reset margin.

Therefore, the event does not require. reanalysis for the I

I I

I I

I I

I I

I I

I I

'~

I ANF-90-181 Page 57 I 15.6 Decreases in Reactor Coolant Inventory 15:6.1 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve I

15.6.1.1 Event Description I

An inadvertent opening of a pressurizer pressure relief valve or safety valve causes a I decrease in the primary system pressure resulting in a loss of both thermal margin and primary coolant inventory.

I For previous cycles, the pressurizer relief valves at Palisades have been blocked closed I during power operation by upstream isolation valves. Therefore, an inadvertent opening of a relief valve could not result in a loss of primary coolant inventory. For Cycle 9 operation, the impact of operable PORVs is considered. For a stuck open relief valve, the loss of coolant accident (LOCA) mitigating procedures will begin.

I 15.6.1.2 Event Disposition and Justification I The event is principally of concern in the short term because of the DNBR challenge due I to depressurization before scram. The depressurization has little effect on core power or primary temperatures. Consequently, increasing the VHP trip reset margin does not affect the event severity since the power excursion, resulting from a positive moderator coefficient, is of a limited I

• degree. Thus, the Inadvertent Opening of a Pressurizer Pressure Relief Valve event is bounded by the analysis in Reference 5.

I I

I

,, >I I

ANF-90-181 Page 58 15.6.2 Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside of Containment I 15.6.2.1 Event Description I This event occurs when a small line carrying primary coolant outside of containment ruptures leading to a depletion of primary system coolant and a release of contaminated liquid.

I The charging and HPSI systems provide sufficient coolant to replenish that which is lost.

Consequently, no fuel failures would be predicted assuming a reactor trip on low pressurizer I

pressure, TM/LP or Safety Injection Signal (SIS). The radiological consequences are limited by the maximum primary coolant activity level allowed by the Technical Specifications.

I 15.6.2.2 Event Disposition and Justification I

Reference 5 disposed this event as being bounded by the small break LOCA (Event 15.6.5). Changing the VHP trip reset margin will not change this disposition.

I 15.6.3 Radiological Consequences of Steam Generator Tube Failure I

15.6.3.1 Event Description I

This incident occurs when a steam generator tube f~ils causing a leakage of coolant from the primary system to the secondary system. The leakage may deplete the primary coolant I inventory thus reducing the PCS pressure. The tube failure will result in release of fission products from the PCS coolant to the main steam system. The controlling features of the analysis are the tube break size and the magnitude of the radiological source term.

I 15.6.3.2 Event Disposition and Justification I

This event was analyzed for the replacement of the steam generators. This analysis was I

f 1.

I

,. ANF-90-181 Page 59 furnished by Consumers Power. Since this is primarily a depressurization event, modification to I . .

the VHP trip reset margin will not impact the consequences of this event. Therefore, this event is bounded by the analysis performed for the replacement of the steam generators.

I 15.6.4 Radiological Consequences of a Main Steam Line Failure Outside Containment I (BWR)

I This event pertains to BWRs and is, therefore, not applicable to Palisades.

'I '15.6.5 .Loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary I 15.6.5.1 Event Description A loss of coolant accident is initiated by a breach in the primary system pressure boundary. The event initiators vary from relatively small break LOCAs to complete ruptures of .

the PCS piping for large break LOCAs. The limiting features of LOCA analyses are the peak clad temperature (PCT) and the time at elevated temperatures which influences the extent of localized and core-wide zircaloy oxidation reaction.

15.6.5.2 Event Disposition and Justification The coi;itrolling parameters for the transient are: 1) initial fuel stored energy, 2) decay heat,

3) radial and axial power profiles, 4) fuel rod/PCS coolant heat transfer with time and, 5)

I operating conditions for the ECCS systems. Consequently, adjustment of the VHP trip reset margin will not affect this disposition. Thus, the current LOCA analysis of record remains applicable for this licensing action.

I I

11 I

I

15. 7 Radioactive Releases from a Subsystem or Component ANF-90-181 Page 60

15. 7.1 Waste Gas System Failure I

15.7.2 Radioactive Liquid Waste System Leak or Failure (Release to Atmosphere)

I 15.7.3 Postulated Radioactive Releases Due to Liquid-Containing Tank Failures I

15.7.4 Radiological Consequences of Fuel Handling Accident I

15. i.5 Spent Fuel Cask Drop Accidents I

The results of these events are not dependent on the modification of the VHP trip reset margin. Therefore, the reference analyses(?) remain bounding for the conditions of this disposition.

I I

I I

I I

I

I. .

I ANF-90-181 Page 61 4.0 References I

1. Advanced Nuclear Fuels Methodology For Pressurized Water Reactors: Analysis of I Chapter 15 Events, ANF-84-73(P), Appendix B, Revision 4, Advanced Nuclear Fuels Corporation, Augu~t 1989.

• I 2. Letter, Mr. Ashok C. Thadani (USNRC) to Mr. A. A. Copeland (ANF), dated July 13, 1990,

• "Acceptance for Referencing of Topical Report ANF-84-73(P), Revision 4, Appendix B.

'Advanced Nuclear Fuels Corporation Methodology for Pressurized Water Reactors:

I Analysis of Chapter 15 Events"'.

3. Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants.

I NUREG-0800, LWR Edition, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, July 1981.

I 4. Disposition of Standard Review Plan Chapter *15 Events for Palisades Cycle 9, ANF 041, Revision 2, Advanced Nuclear Fuels Corporation, August 1990.

5. Palisades Cycle 9: Analysis of Standard Review Plan Chapter 15 Events, ANF-90-078, Advanced Nuclear Fuels Corporation, September 1990.

6; Palisades Steam Generator Replacement Disposition of Standard Review Plan Chapter I 15 Events, ANF-90-040, Revision 1, Advanced Nuclear Fuels Corporation, June 1990.

.7. Palisades Final Safety Analysis Report, Revision 9, Consumers Power Company, I September 1989.

8. Palisades Plant Technical Specifications; Consumers Power Company, Appendix A to I License No. DPR-20.

9. Description of the Exxon Nuclear Plant Transient Simulation Model for Pressurized Water I Reactors (PTS-PWR), XN-NF-74-5(A), Revision 2, Exxon Nuclear Company, October 1986, and Supplements 3-6.

I 10. XCOBRA-lllC: A Computer Code to Determine the Distribution of Coolant During Steady-State and Transient Core Operation, XN-NF-75-21 (A), Revision 2, Exxon Nuclear Company, January 1986.

I 11. Exxon Nuclear DNB Correlation for PWR Fuel Design, XN-NF-621 (A), Revision 1, Exxon Nuclear Company, April 1982.

12. Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel, ANF-1224(A) and Supplement 1, Advanced Nuclear Fuels Corporation, April 1990.

I

ANF-90-181 Page 62

13. Justification of the ANFP DNB Correlation for High Thermal Performance Fuel in the Palisades Reactor, ANF-89-192{P), Advanced Nuclear Fuels Corporation, January 1990.

14. Palisades Modified Reactor Protection System Report: Analysis of Chapter 15 Events,

• ANF-87-150{NP), Volume 2, Advanced Nuclear Fuels Corporation, June 1988.

15. Palisades Large Break LOCNECCS Analysis with Increased Radial Peaking, ANF-88-107, Revision 1, Advanced Nuclear Fuels Corporation, February 1990. '

16. Palisades Cycle 8: Disposition and Analysis of Standard Review Plan Chapter 15 Events, ANF-88-108, Revision 1, Advanced Nuclear Fuels Corporation, September 1988. I I

I I

I I

.. I I

I I

'~

I ANF-90-181 Issue Date: 11 /09/90 I

I REVIEW AND ANALYSIS OF SRP CHAPTER 15 EVENTS I FOR PALISADES WITH A 15% VARIABLE HIGH POWER TRIP RESET I

I I Distribution RA Copeland JD Gale RC Gottula I JW Hulsman TR Lindquist I JN Morgan KC Segard I RT Welzbacker CPCo/HG Shaw (20)

Document Control (3)

I I

I I