ML20195H074

Nonproprietary Vol 1 of Palisades Modified Reactor Protection Sys Rept-Disposition of SRP Chapter 15 Events

ML20195H074
Person / Time
Site:	Palisades
Issue date:	06/13/1988
From:	Linquist T SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML18053A418	List: ML20195H030 ML20195H074 ML20195H327
References
ANF-87-150(NP), ANF-87-150(NP)-V01, ANF-87-150(NP)-V1, NUDOCS 8806280214
Download: ML20195H074 (167)

Text

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k A N F-8 7- 15 0(N P)

VOLUME 1

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l f ADVANCED NUCLEAR FUELS CORPORATION P ALIS ADES MODIFIED RE ACTOR PROTECTION SYSTEM REPORT -

DISPOSITION OF ST AND ARD REVIEW PLAN CH APTER 15 EVENTS JUNE 19 8 8

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ADVANCEDNUCLEAR FUELS CORPORATION ANF-87-150(NP)

Volume 1 Issue Date: 6/13/88

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PALISADES MODIFIED REACTOR PROTECTION SYSTEM REPORT -

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DISPOSITION OF STANDARD REVIEW PLAN CHAPTER 15 EVENTS i Prepared by f :cl T. R. Lindquist/Proje'ct Engineer

~

PWR Saffy Analysis a

June 1988 h 9f

i NUCLEAR REGULATORY COMMISSION REPORT DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This fechnical report was derived through research and development pro-grams sponsored by Advanced Nuclear Fuels Corporation. It is being submit-ted by Advanced Nuclear Fuels Corporation to the U.S. Nuclear Regulatory Commisafon as part of a technical contribution to facilitate safety analyses by licensees of the U.S. Nuclear Regulatory Commission which utilize Ad-vancoo Nuclear Fuels Corporation fabricated reload fuel of other technical services provided by Advanced Nuclear Fuels Corporation for light water power reactors and it is true and correct to the best of Advanced Nuclear Fuels Corporation's knowledge, information, and belief. The information con-tained herein may be used by the U.S. Nuclear Regulatory Commission in its review of this report, and under the terms of the respective agreements, by licensees or applicants before the U.S. Nuclear Regulatory Commission which are customers of Advanced Nuclear Fuels Corporation in their

• demonstration of compitance with the U.S. Nuclear Regulatory Commission's

• regulations.

Advanced Nuclear Fuels Corporation's warranties and representations con.

coming the subject matter of this document are those set forth in the agree-ment between Advanced Nuclear Fuels Corporation and the customer to which this document is issued. Accordingtv, except as otherwise expressly provided in such agreement, neither Advanced Nuclear Fuels Corporation nor any person acting on its behalf:

A. Makes any warranty, or representation, express or im-piled, with respect to the accuracy, completeness, or usefulness of the information contained in this docu-ment, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights, or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, ap-paratus, method. or process disclosed in this document.

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'l i ANF-87-150(NP)

Volume 1 1ABLE OF CONTENTS Section gitgg 1

1.0 INTRODUCTION

2.0

SUMMARY

OF DISPOSITION OF EVENTS . . . . . . . . . . . . . . . . 4 3.0 BASIS AND JUSTIFICATION FOR DISPOSITION OF EVENTS ....... 10 15.1- INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM . . . . . . . . 13 15.1.1 Decrease in Feedwater Temperature ............... 13 15.1.2 Increase in Feedwater Flow . . . . . . . . . . . . . . . . . . . 15 15.1.3 Increase in Steam Flow . . . . . . . . . . . . . . . . . . . . . 18 r

a 15.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve . 21 15.1.5 Steam System. Piping Failures Inside and Outside of Containment . 21

, 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM , . . . . . . . 37 l

' 15.2.1 Loss of External Load ..................... 37 L

15.2.2 Turbine Trip . . . . . . . . . . . . . . . . . . . . . . . . . . 38 f 15.2.3 Loss of Condenser Vacuum . . . . . . . . . . . . . . . . . . . . 40 Closure of the Main Steam Isolation Valves (MSIVs)(BWR) 41 15.2.4 ....

I 15.2.5 Steam Pressure Regulator Failure . . . . . . . . . . . . . . . . 42 15.2.6 Loss of Nonemergency A.C. Power to the Station Auxiliaries . . . 42 Loss of Normal Feedwater Flow 45

! 15.2.7 .................

Feedwater System Pipe Breaks Inside and Outside Containment 46 15.2.8 ..

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............ 65 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW 15.3.1 Loss of Forced Reactor Coolant Flow .............. 65

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L 11 ANF-87-150(NP)

Volume 1 TABLE OF CONTENTS (Con'.)

Section Eagg j 15.3.2 Flow Controller Malfunction .................. 67 15.3.3 Reactor Coolant Pump Rotor Seizure . . . . . . . . . . . . . . 67 15.3.4 Reactor Coolant Pump Shaft Break . . . . . . . . . . . . . . . . 68 15.4 REACTIVITY AND POWER DISTRIBUTICN ANOMALIES .......... 78 15.4.1 Uncontrolled Control Rod Bank Withdrawal From a Subcritical or low Power Startup Condition .................. 78 15.4.2 Uncontrolled Control Rod Bank Withdrawal at Power ....... 81 15.4.3 Control Rod Misoperation . . . . . . . . . . . . . . . . . . . . 82 15.4.4 Startup of an Inactive Loop .................. 88 15.4.5 Flow Controller Malfunction .................. 90 15.4.6 CVCS Malfunction That Results in a Decrease in the Boron Concentration in the Reactor Coolant . . . . . . . . . . . . . . 50 15.4.7 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position ....................... 32 15.4.8 Spectrum of Control Rod Ejection Accidents . . . . . . . . . . . 93 15.4.9 Spectrum of Rod Drop Accidents (BWR) . . . . . . . . . . . . . . 94 15.5 INCREASES IN REACTOR COOLANT SYSTEM INVENT 0RY . . . . . . . . 121 15.5.1 Inadvertent Operation of the ECCS That Increases Reactor Coolant Inventory ...................... 121 15.5.2 CVCS Malfunction That Increases Reactor Coolant Inventory . . 123 15.6 DECREASES IN REACTOR COOLANT INVENTORY . . . . . . . . . . . . 128 15.6.1 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve . 128

111 ANF-87-150(NP) f Volume 1 i

TABLE OF CONTENTS ;ont . )

Section ELqa

! 15.6.2 Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside of Containment . . . . . . . 129 15.6.3 Radiological Consequences of Steam Generator Tube Failure . . 129 l

15.6.4 Radioivgical Consequences of a Main Steamline Failure Outside Containment (BWR) . . . . . . . . . . . . . . . . . . . . . . 132 15.6.5 Loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure

' Boundary . . . . . . . . . . . . . . . . . . . . . . . . . . . 133 15.7 RADI0 ACTIVE RELEASES FROM A SUBSYSTEM OR COMP 0NENT , . . . . . 147

, 15.7.1 Waste Gas System Failure . . . . . . . . . . . . . . . . . . . 147 L 15.7.2 Radioactive Liquid Waste System Leak or Failure (Release to

- . Atmosphere) ............. . . . . . . . . . . . . 147 15.7.3 Postulated Radioactive Releases Dua to Liquid-Containing Tank Fa; lures . . . . . . . . . . . . . . . . . . . . . . . . . . . 147 l

15.7.4 Radiological Consequences of Fuel Handling Accident . . . . . 147

) 15.7.5 Spent Fuel Cask Drop Accidents . . . . . . . . . . . . . . . . 148

4.0 REFERENCES

. . . . . . . . . . . . . . . . . . . . . . . . . . 152 i

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f iv ANF-87-150(NP)

Volume 1 LIST OF TABLES Table P_itan 1.1 Reactor Operating Conditions for Palisades . . . . . . . . . . . 3 Disposition of Events Summary 5 f 2.1 .................

15.1.1-A Available Reactor Protection for Decrease in Feedwater Temperature E,ent .................. 23 15.1.1-B Disposition of Events for the Decrease in Feedwater Temperature Event .................. 24 15.1.2(1)-A Available Reactor Protection for the Increase 25 Feedwater Flow Event ....................

15.1.2(1)-B Disposition of Events for the Increase in Feedwater Flow Event .................... 26 f 15.1.2(2)-A Available Reactor Protection for the Increase in 27 Feedwater Flow Event ....................

15.1.2(2)-B Disposition of Events for the Increase in Feedwater Flow Event .................... 28 15.1.2(3)-A Available Reactor Protection for the Increase 29 in Feedwater Flow Event . . . . . . . . . . . . . . . .,. . .

15.1.2(3)-B Disposition of Events for the Increase in Feedwater Flow Event .................... 30 15.1.3-A Available Reactor Protection for the Increase in Steam Flow Event . . . . . . . . . . . . . . . . . . 31

/ 15.1.3-B Disposition of Events for the Increase in Steam Flow Event . . . . . . . . . . . . . . . . . . . . .... 32 h 15.1.4-A Available Reactor Protection for the Inadvertent Opening of a Steam Generator Relief or Safety Valve Event ............................. 33 15.1.4-B Disposition of Events for the Inadvertent Opening of a Steam Generator Relief or Safety Valve Event ....... 34 1 ,

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v ANF-87-150(NP)

Volume 1 4

i LIST OF TABLES (Cont.)

i Table Eagg 15.1.5-A Available Reactor Protection for Steam System Piping Failures Inside and Outside of Containment Event ............ 35 s 15.1.5-B Disposition of Events for Steam System Piping Failures Inside and Outside of Containment Event . . . . . . . . 36 15.2.1-A Available Reactor Protection for the loss of External Load Event .................. 49 15.2.1-B Disposition of Events for the Loss of External Load Event . . . . . . . . . . . . . . . . . . . . . 50 r

15.2.2-A Available Reactor Protection for the Turbine Trip Event . . . . . . . . . . . . . . . . . . . . . . . 51 15.2.2-8 Disposition of Events for the Turbine Trip Event . . . . . . . . . . . . . . . . . . . . . . . 52 a

15.2.3-A Available Reactor Protection for the {

Loss of Condenser Vacuum Event . . . . . . . . . . . . . . . . . 53

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15.2.3-8 Disposition of Events for the Loss of Condenser Vacuum Event . . . . . . . . . . . . . . . . . 54 15.2.4-A Available Reactor Protection for the Closure of the MSIVs Event . . . . . . . . . . . . . . . . . . . 55 15.2.4-B Disposition of Events for the 1 Closure of the MSIVs Event . . . . . . . . . . . . . . . . . . . 56  ;

15.2.5-A Available Reactor Protection for the Steam Pressure Regulator Failure Event . . . . . . . . . . . . . 57 15.2.5-B Disposition of Events for the ,

Steam Pressure Regulator Failure Event . . . . . . . . . . . . . 58 (

15.2.6-A Available Reactor Protection for the loss of Nonemergency A.C. Power to i the Station Auxiliaries Event ................. 59 1 I

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f vi ANF-87-150(NP)

Volume 1 i

LIST OF TABLES (Cont.)

P_a.ulg Table 15.2.6-8 Disposition of Events for the Loss of Nonemergency A.C. Power to the Station Auxiliaries Event . . . . . . . . . . . . . . . . 60 15.2.7-A Available Reactor Protection for the Loss of Normal Feedwater Flow Event .............. 61 15.2.7-B Disposition of Events for the Loss of Normal Feedwater Flow Event .................. 62 15.2.8-A Available Reactor Protection for the Feedwater System Pipe Breaks Inside and Outside Containment Event . . . . . . . . . . . . . . 63 15.2.8-B Disposition of Events for the ,

j Feedwater System Pipe Breaks Inside and Outside Containment Event ................. 64 l 15.3.1-A Available Reactor Protection for the Loss of Forced Reactor Coolant Flow Event ........... 70 15.3.1-B Disposition of Events for the loss of Forced Reactor Coolant Flow Event . . . . . . . . . . . . . . . . . . . . . . . . . . . 71 15.3.2-A Available Reactor Protection for the Flow Controller Malfunction Event ............... 72 15.3.2-B Disposition of Events for the Flow Controller Malfunction Event ............... 73 f 15.3.3-A Available Reactor Protection for the Reactor Coolant Pump Rotor Seizure Event . . . . . . . . . . . . 74 f 15.3.3-B Disposition of Events for the Reactor Coolant Pump Rotor Seizure Event . . . . . . . . . . . . . . . . 75 15.3.4-A Available Reactor Protection for the Reactor Coolant Pump Shaft Break Event . . . ......... 76 15.3.4-8 Disposition of Events for the Reactor Coolant Pump Shaft Break Event . . . . . . . . . . . . . . . . . 77 3

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vii ANF-87-150(NP)

Volume 1 LIST OF TABLES (Cont.)

Table P_a_qq 1

15.4.1-A Available Reactor Protection for the Uncontrolled Control Rod Bank Withdrawal from a Subcritical or Low Power Startup Condition Event . . . . . . . . . . . . . . 95 15.4.1-B Disposition of Events for the Uncontrolled Control Rod Bank Withdrawal from a Suberitical or low Power Startup Condition Event . . . . . . . . . . . . . . 96 ,

k 15.4.2-A Available Reactor Protection for the Uncontrolled Control Rod Bank Withdrawal at Power Event . . . . . . . . . . . 97 15.4.2-B Disposition of Events for the Uncontrolled Control '

Rod Bank Withdrawal at Power Event . . . . . . . . . . . . . . . 98 15.4.3(1)-A Available Reactor Protection for the Dropped Control Rod / Bank Event ............... 99 15.4.3(1)-B Disposition of Events for the Dropped Control Rod / Bank Event . . . . . . . . . . . . . . 100 .

15.4.3(2)-A Available Reactor Protection for the Dropped Part-Length Control Rod Event . . . . . . . . . . . 101  ;

15.4.3(2)-B Disposition of Events for the Dropped Part-Length Control Rod Event . . . . . . . . . . . 102  ;

15.4.3(3)-A Available Reactor Protection for the Malpositioning  ;

of the Part-Length Control Red Group Event . . . . . . . . 103 l, 15.4.3(3)-B Disposition of Events for the Ma1 positioning  !

of the Part-Length Control Rod Group Event . . . . . . . . 104 15.4.3(4)-A Available Reactor Protection for the Statically Misaligned Control Rod / Bank Event . . . . . . . . . . . . . 105 l 15.4.3(4)-B Disposition of Events for the Statically Misaligned Control Rod / Bank Event . . . . . . . . . . . . . 106 ,

i 15.4.3(5)-A Available Reactor Protection for the Single l Control Rod Withdrawal Event . . . . . . . . . . . . . . . 107 l

i viii ANF-87-150(NP)

Volume 1 LIST OF TABLES (Cont.)

l E!Lgg Table f 15.4.3(5)-B Disposition of Events for the Single Control Rod Withdrawal Event . . . . . . . . . . . . . . . . . . . 108 L

15.4.3(6)-A Available Reactor Protection for the

' Core Barrel Failure Event . . . . . . . . . . . . . . . . . 109 7 15.4.3(6)-B Disposition of Events for the Core Barrel Failure Event . . . . . . . . . . . . . . . . 110 j

g 15.4.4-A Available Reactor Protection for the Startup of an inactive Loop Event . . . . . . . . . . . . . . . . . . . . 111 15.4.4-B Disposition of Events for the Startup of an .

Inactive Loop Event ................. . . . , 112 l -

15.4.5-A Available Reactor Protection for the Flow I

Controller Malfunction Event . . . . . . . . . . . . . . . . . 113 15.4.5-B Disposition of Events for the Flow Controller Mal function Event . . . . . . . . . . . . . . . . . 113 9 15.4.6-A Available Reactor Protection for the CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant Event . . . . . . . . . . . . . . . . . 114 15.4.6-B Disposition of Events for the CVCS Malfunctior, that Results in a Decrease in the Boron Concentration in the Reactor Coolant Event . . . . . . . . . . . . . . . . . 115 15.4.7-A Available Reactor Protection for the Inadvertent Loading and Operation of a Fuel Assembly in an

.! Improper Position Event . . . . . . . . . . . . . . . . . . . 116 15.4.7-B Disposition of Events for the Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position Event . . . . . . . . . . . . . . . . . . . . . . . 117 15.4.8-A Available Reacto'r Protection for the Spectrum of Control Rod Ejection Accidents . . . . . . . . . . . . . . . . 118 15.4.8 B Disposition of Events for the Spectrum of Control Rod Ejection Accidents . . . . . . . . . . . . . . . . 119 i

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i ix ANF-87-150(NP) '

Volume 1 LIST OF TABLES (Cont.)

Table EiLqn 15.4.9-A Available Reactor Protection for the Spectrum of Rod Drop Accidents (BWR) . . . . . . . . . . . . . 120 15.4.9-8 Disposition of Events for the Spectrum of Rod Drop Accidents (BWR) . . . . . . . . . . . . . . . . . 120 15.5.1-A Available Reactor Protection for the Inadvertent Operation of the ECCS that Increases Reactor Coolant Inventory Event . . . . . . . . . . . . . . . . . . . . . . . 124 15.5.1-B Disposition of Events for the Inadvertent Operation of the ECCS that Increases Reactor Coolant Inventory Event . . . . . . . . . . . . . . . . . . . . . . . . . . . . 125 15.5.2-A Availabl'e Reactor Protection for the CVCS Malfunction that Increases Reactor Coolant Inventory Event . . . . . . . . . . 126 15.5.2-B Disposition of Events for the CVCS Malfunction that Increases Reactor Coolant Inventory Event . . . . . . . . 127 15.6.1-A Available Reactor Protection for the Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve Event . . . . . . . 138 ,

15.6.1-B Disposition of Events for the Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve Event . . . . . . . . . 139

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15.6.2-A Available Reactor Protection for the Radiological Consequences of the Failure of Small Lines Carrying -

Primary Coolant Outside of Containment Event . . . . . . . . . 140 15.6.2-B Disposition of Events for the Radiological Consequences ,

of the Failure of Small Lines Carrying '

Primary Coolant Outside of Containment Event . . . . . . . . . 141 15.6.3-A Available Reactor Protection for the Radiological Consequences of Steam Generator Tube Rupture Event . . . . . . 142 15.6.3-B Disposition of Events for the Radiological Consequences of Steam Generator Tube Rupture Event . . . . . . . . . . . . . . . . . . . . . . . . . . . . 143

x ANF-87-150(NP)  ;

Volume 1 l s

LIST OF TABLES (Cont.)

Table hag 15.6.4-A Available Reactor Protection for the Radiological Consequences of a Main Steamline Failure Outside Containment (BWR) Event . . . . . . . . . . . . . . . . . . . 144 15.6.4-8 Disposition of Events for the Radiological Consequences of a Main Steamline Failure Outside Containment (BWR) Event . . . . . . . . . . . . . . . . . . . . . . . . . 144 15.6 5-A Available Reactor Protection for the Loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary . . . . . . . . . . . . . . . . . . . . . . . . . . . 145 15.6.5-B Disposition of Events for the Loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks

' Within the Reactor Coolant Pressure Boundary . . . . . . . . . 146 15.7.3-A Available Reactor Protection for the Postulated Radioactive Releases Due to Liquid-Containing Tank Failures Event . . . . . . . . . . . . . . . . . . . . . 149

, 15.7.3-8 Disposition of Events for the Postulated Radioactive Releases Due to Liquid-Containing Tank Failures Event . . . . 149 15.7.4-A Available Reactor Frotection for the Radiological i Consequences of Fuel Handling Accidents . . . . . . . . . . . 150 15.7.4-B Disposition of Events for the Radiological Consequences of Fuel Handling Accidents . . . . . . . . . . . 150 15.7.5-A Available Reactor Protection for the Spent Fuel Cask Drop Accidents . . . . . . . . . . . . . . . . . . . . . 151 15.7.5-B Disposition of Events for the Spent Fuel Cask Drop Accidents . . . . . . . . . . . . . . . . . . . . . 151 i

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1 ANF-87-150(NP)

Volume 1 i

1.0 INTRODUCTION

This report provides a review of the Standard Review Plan (I) Chapter 15 Events for Palisades operation with the modified reactor protection system (RPS). In accordance with Advanced Nuclear Fuels Corporation (ANF) methodology (2) , 3) ;

events described in the Standard Review Plan (SRP) have been reviewed and placed (dispositioned) into one of the following four categories:

(1) The event initiator or controlling parameters have been changed from the i

analysis of record so that the event needs to be reanalyzed for the current licensing action; i

(2) The event is bounded by another event which is to be reanalyzed; I

f (3) The event causes 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.

1 The revied of the current plant safety analysis for the event disposition process incorporated the following additions and modifications to the plant:

(1) Addition of a variable overpower trip; (2) Revised thermal margin / low pressure trip function; and f (3) Increased level of steam generator tube plugging to a total of 29%.

Items (1) and (2) constitute the modified reactor protection system additions and revisions to the current RPS system.

Section 2.0 provides a summary of the event disposition. In order to facilitate review, the events are numbered in accordance with the SRP and cross-referenced to the pertinent updated FSAR sections. Section 3.0 presents

2 ANF-87-150(NP)

Volume 1 l

the results of the analysis and justifications. The results of simplified bounding calculations are also included in Section 3.0. Section 4.0 presents l the references used in this report.

l In the event disposition, all of the reactor operating conditions allowed by the plant Technical Specifications are examined to insure that the bounding  ;

subevents are identified for each SRP event category. This insures that the subsequent safety analysis will support the complete range of allowable operating conditions. The reactor operating conditions allowed for Palisades by the plant Technical Specifications are listed in Table 1.1.

l Cycle 7 has been chosen as a reference cycle for the purposes of this report.

l Core kinetic parameters, temperature coefficients, and control rod characteristics are taken to be representative or bounding of Cycle 7.(3)

=

Volume 2 of tnis report documents the analysis of those events dispositioned to be analyzed. '

Table 1.1 Reactor Operating Conditions for Palisades Reactor Operating Rated Average Coolant Conditions Reactivity Condition Therpli Pcwer TemDerature

1. Ra.ted Power Critical 2530 MWt 2 525*F

2. Power Operation Critical 2 2% 2 525*F

-4

3. Reactor Critical Critical 2 10 % 2 525*F

4. Hot Standby Any of the control rods are withdrawn s 2% > 525*F ,

~4 "'

5. Hot Shutdown Shutdown margin of < 10 % > 525*F l

2 2% Ap

6. Refueling Shutdown Shutdown margin of at 0 s 210*F least 5% Ap with all control rods withdrawn

7. Cold Shutdown Condition k 5 98 5 210*F eff

8. Refueling Operation Any operation involving movement of core components when the vessel head is unbolted or removed.

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4 ANF-87-150(NP) f Volume 1 l'

2.0

SUMMARY

OF DISPOSITION OF EVENTS I

ll Table 2.1 presents a sumary of results of the event disposition. In accordance with ANF methodology, the events are placed in one of the four categories identified in Section 1.0.

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Table 2.1 .Dispo:.ition of Events Summary SRP Event Event Bounding Updated Classifi- Desig- Event or FSAR cation nation Name DisDosition Reference Desianation 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.1 Decrease in Feedwater Temperature Bounded 15.1.3 14.9.4 f

1 15.1.2 Increase in Feedwater Flow

1) Power Bounded 15.1.3- 14.9.5 & 14.9.6 I 2) Startup Bounded 15.1.3 15.1.3 Increase in Steam Flow Analyze 14.10 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 Ref.5, 6 & 22 14.14

, 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 4 15.2.1 Loss of External Load Analyze 14.12 15.2.2 Turbine Trip Bounded 15.2.1 3 o

3 15.2.3 Loss of Condenser Vacuum Bounded 15.2.1 h

~Gc) 15.2.4 Closure of the Main Steam j

Isolation Valves (MSIVs) Bounded 15.2.1 3 15.2.5 Steam Pressure Regulator Failure Not applicable; BWR Event

. _ __.____# _ , _ _ _ _ _ _ . . _ _ _ _ _ _ _ ~ , _ _

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Table 2.1 Dispnsition.of Events Summary (Cont.)

SRP Event Bounding Updated Event Classifi- Desig- Event or .FSAR cation nation Name DisDosition Reference Desianation 15.2.6 Loss of Nonemergency A.C. Power Short term bounded 15.3.1 to the Station Auxiliaries Long term bounded 15.2.7 15.2.7 Loss of Normal Feedwater Flow Analyze 14.13 15.2.8 Feedwater System Pipe Breaks Cooldown Bounded 15.1.5 Inside and Outside Containment Heatup Bounded 15.2.7 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW 15.3.1 Loss of Forced Reactor Coolant Flow Analyze 14.7 ,,

15.3.2 Flow Controller Malfunction Not Applicable 14.7 15.3.3 Reactor Coolant Pump Rotor Seizure Analyze 14.7 15.3.4 Reactor Coolant Pump Shaft Break Bounded 15.3.3 14.7 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1 Uncontrolled Control Rod BanR cf EE Withdrawal from a Subcritical -

or Low Power Condition Analyze 14.2.2.2j[][

74 15.4.2 Uncontrolled Control Rod Bank Withdrawal at Power Analyze 14.2.2.3 S 15.4.3 Control Rod Misoperation l Dropped Control Rod Bank Analyze 14.4 1)

Dropped Part-Length Control Rod Bounded 15.4.3(1) 14.6 2)

Table 2.1 Disposition of Events Summary (Cont.)

SRP Event Event Bounding Updated Classifi- Desig- Event or FSAR cation nation Name Disposition Reference Desianation

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

4) Statically Misaligned Control Rod / Bank Analyze

5) Single Control Rod Withdrawal Analyze 14.2.2.4

6) Core Barrel Failure Bounded 14.5 14.5 15.4.4 Startup of an Inactive Loop Bounded 14.8 14.8 I

15.4.5 Flow Controller Malfunction Not applicable; No Flow Con-troller 15.4.6 CVCS Malfunction that Results in a Decrease in the Boron Con-centration in the Reactor Coolant

1) Rated and Power Operation Conditions Analyze 14.3 l 2) Reactor Critical, Hot

< Standby and Hot Shutdown Analyze 14.3

3) Refueling Shutdown Condition, J Cold Shr:tdown Condition and Ea Refueling Operation Analyze 14.3 ]

15.4.7 Inadvertent Loading and Operation Administrative of a Fuel Assembly in an Improper Procedures ~

G 1 Position Preclude this Event

_u _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ - _ _ _ _ _ _ _ _ . . ___ _._ -_ _-- = - 1 -N

-M. _ _ _ - _ _ - __ WD e _

. . ~ _ - _ __. _ _ .

Table 2.1 Disposition of Events Summary (Cont.)

SRP Event Event Bounding Updated Classifi- Desig- Event or FSAR cation nation Hamg Disposition Reference Desianation-15.4.8 Spectrum of Control Rod Ejection Analyze 14.16 Accidents 15.4.9 Spectrum of Rod Drop Not applicable; Accidents (BWR) BWR Event 15.5 INCREASES IN REACTOR COOLANT INVENTORY 15.5.1 Inadvertent Operation of the Overpressure ECCS that Increases Reactor Bounded 15.2.1 Coolant Inventory Reactivity Bounded 15.4.6 00 -

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

15.6 DECREASES IN REACTOR COOLANT INVENTORY 15.6.1 Inadvertent Opening of a PWR.

Pressurizer Pressure Relief Valve Bounded 15.6.5 15.6.2 Radiological Consequences of the Bounded 15.6.5 Failure of Small Lines Carrying Primary Coolant Outside of Con- g- 2; tainment gn B OD 15.6.3 Radiological Consequences of Bounded Ref. 4; 14.15 14.15 7 Steam Generator Tube Failure {j 15.6.4 Radiological Consequences of a Not applicable; 55 Main Steamline Failure Outside BWR Event Containment

Table 2.1 Disposition of Events Summary (Cont.)

. SRP Event Event Bounding Updated Classifi- Desig- Event or FSAR cation nation Name DisDosition Reference Desianation 15.6.5 Loss of Coolant Accidents Bounded Ref. 11-20; 14.17 Resulting from a Spectrum of Ref. 4; 14.17, 14.18 Postulated Piping Breaks within 14.18 & 14.22 the Reactor Coolant Pressure 14.22 Boundary 15.7 RADI0 ACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT i

15.7.1 Waste Gas System Failure Deleted

• 14.21 15.7.2 Radioactive Liquid Waste System

• Leak or Failure (Release to Atmosphere) Deleted

• 15.7.3 Postulated Radioactive Releases Bounded Ref. 4; 14.20 14.20 due to Liquid-Containing Tank i Failures 15.7.4 Radiological Consequences of Fuel Bounded Ref. 4; 14.19 14.19 Handling Accidents 15.7.5 Spent Fuel Cask Drop Accidents Bounded Ref. 4; 14.11 14.11

<= 2, J

!h=

O

• This section of the Standard Review Plan has been deleted. jg m

• 4 . s ,

I 10 ANF-87-150(NP)

Volume 1 l

3.0 BASIS AND JUSTIFICATION FOR DISPOSITION OF EVENTS This section presents the basis and justification for disposition of the events. The section numbers and event names are in accordance with those events described in the SRP.

) Each event described in the SRP (and events in the FSAR but not in the SRP) is considered in accordance with the plant licensing basis and dispositioned into one of the four categories described in Section 1.0. Events which are not bounded by other events or by existing accepted analysis, and are in the plant licensing _ basis, are dispositioned to be analyzed. In the event disposition f process, the event initiator is identified for each event. The magnitude of the initiator for each event is calculated and compared to the magnitude of the initiator for other events. The comparison basis includes all the plant operating conditions. This allows, in several cases, a ranking of the event initiators as to severity, allowing the lesser events to be dispositioned as bounded by the greater event. Similar logic is applied in determination of the applicability and bounding nature for existing accepted analysis.

The licensing basis for Palisades is as stated in the Final Safety Analysis Report.(4) The formulation of event scenarios to be considered in the safety analysis depends on single failure criteria established by the plant licensing basis. Examination of the Palisades licensing basis yields the following

( single failure criteria:

l l

(1) The Reactor Protection System (RPS) is designed with redundancy and independence to ::ture that no single failure or removal from service of any component or %u of a system will result in the loss of the protection function.

1 (2) Each Engineered Safety Feature (ESF) is designed to perform its intended safety function assuming a failure of a single active component.

11 ANF-87-150(NP)

Volume 1 (3) The onsite power system and the offsite power system are designed such that each shall independently be capable of providing power for the ESF assuming a failure of a single active component in either power system.

The safety analysis is structured to demonstrate that the plant systems design satisfies these single failure criteria. The following assumptions result: ,

(1) The ESF required to function in an event are assumed to suffer a worst single failure of an active component.

(2) Reactor trips occur at the specified setpoint within the specified delay time assuming a worst single active failure.

(3) The following postulated accidents are considered assuming a concurrent loss of offsite power: main steam line break, control rod ejection, steam generator tube rupture, and LOCA.

(4) The loss of normal feedwater, an anticipated operational occurrence, is analyzed assuming a concurrent loss of offsite power.

The requirements of 10 CFR 50, Appendix A, Criteria 10, 20, 25 and 29 require that the design and operation of the plant and the reactor protective system assure that the Specified Acceptable Fuel Design Limits (SAFDLs) not be exceeded during Anticipated Operational Occurrences (A00s). As per the definition of A00 in 10 CFR 50, Appendix A, "Anticipated Operational Occurrences mean those conditions of normal operation which are expected to occur one or more times during the life of the plant and include but are not limited to loss of power to all recirculation pumps, tripping of the turbine '

generator set, isolation of the main condenser, and loss of all offsite power." The Specified Acceptable Fuel Design Limits (SAFDLs) are that: 1) the fuel shall not experience centerline melt (-21 kW/ft); and 2) the departure

- - , ~ . , , - - . . . - - - - - ~ r ,. , ,,. -- ,- -- - . - - - .

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12 ANF-87,150(NP)

Volume )

'from nucleate boiling ratio (DNBR) shall have a minimum allowable limit such that there is a 95% probability with a 95% ccnfidence interval that departure from nucleate boiling (DNB) has not occurred (XNB DN8R of 1.17).

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13 ANF-87-150(NP)

Volume 1 f

, 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.1 Decrease in Feedwater Temoerature 15.1.1.1 Event Initiator - A decrease in feedwater temperature may be caused by loss of one of several feedwater heaters. The loss could be due to the

< interruption of steam extraction flow or to an accidental opening of a feedwater heater bypass line. Each high pressure heater increases the feedwater temperature by approximately 55'F at full load; in order to lose this heating, two valves (one per extraction line) would have to be operated.

The loss of any of the low pressure heaters will produce a lesser effect due to the compensating effect of the high pressure heaters. The effects of any decrease in the feedwater temperature due to flow increases (Main or Auxiliary Feedwater) are discussed in Section 15.1.2.

15.1.1.2 Event Descriotion - Due to a malfunction in the feedwater heater system, the enthalpy of the feedwater being injected into the steam generators is reduced. The increased subcooling of the feedwater reduces the secondary system average fluid enthalpy and increases the energy removal rate from the i primary system. The increase in primary to secondary heat transfer causes the i

reactor coolant temperature at the outlet of the steam generator to decrease.

This causes a corresponding decrease in the core inlet coolant temperature, i With a negative moderator reactivity temperature coefficient, the reactor core power will begin to increase as the cooler moderator fluid reaches the core.

l 15.1.1.3 Reactor Protection - Reactor protection is provided by the variable overpower, thermal margin / low pressure, low pressurizer pressure, and low l

l steam generator pressure trips. Reactor protection for the decrease in feedwater temperature event is summarized in Table 15.1.1-A.

15.1.1.4 Q11p_gsition and Justification -

There is no extraction to the feedwater heaters for the following reactor operating conditions: hot standby,

14 ANF-87-150(NP)

Volume 1 hot shutdown, refueling shutdown, cold shutdown, and refueling operation. As such, there is not a credible event for these reactor operating conditions.

For the fol!owing reactor open. ting conditions: reactor critical, power operation, and rated power, the response of the nuclear steam supply system is governed by the magnitude of the overcooling introduced by the initiating event. As the initial reactor power (turbine load) ir, decreased, the amount of heat supplied by the turbine extraction steam is reduced. As a result, the magnitude of the overcooling introduced by the decrease in feedwater heating will also decrease. In addition to this, the initial steam generator inventory increases as the initial reactor power is decreased, which also decreases the effect of the overcooling introduced by the decrease in feedwater temperature. Therefore, the largest effect of the decrease in feedwater temperature will occur at the full load rated power conditions and will bound all other conditions.

Also, at rated power conditions, the initial thermal margin (DNBR) is minimized. Maximizing the reduction in feedwater temperature maximizes the load demand. This results in the maximum rate of moderator cooldown which, in the presence of a negative moderator temperature coefficient, results in the maximum challenge to the thermal margin (DNB) specified acceptable fuel design limits (SAFDLs).

Quantifying the magnitude of the reduction in feedwater heating due to ,

bypassing the feedwater heaters indicates that a conservative estimate of the 5 This cooldown rate is bounded by the cooldown rate is 2.0 10 Btu /sec.

cooldown introduced by the limiting event postulated ir. Section 15.1.3, which 6

produces a load increase of approximately 1.098 10 Btu /sec. As such, the consequences of the Increase in Steam Flow event (15.1.3) bound the consequences for the Decrease in Feedwater Temperature event discussed in this section (15.1.1) . The disposition of events for the Decrease in Feedwater Temperature event is summarind in Table 15.1.1-B.

15 ANF-87-150(NP)

Volume 1 15.1.2 Increase in Feedwater Flow 15.1.2.1 Event Initiator - This event is initiated by a failure in the feedwater system which causes an increase in the feedwater flow to the steam generators. The magnitude of the increase in feedwater flow was determined by considering possible failures in the feedwater system. An increase in feedr 'er flow may be caused by:

(1) Complete opening of a feedwater regulating valve: Complete opening of both feedwater regulating valves can increase feedwater flow by about 20%

above nominal. Each feedwater regulating valve is controlled by an error signal which is the sum of 1) flow mismatch (steam-feed), and 2) time integrated level error. The speed of both feed pumps (in automatic) is controlled by whichever valve error signal is greater. A single failure which would cause a large surge in feedwater flow is a sudden failure of one level control channel so as to give a maximum demand signal (i.e.,

instrument line rupture). This would cause one feedwater regulating valve to open wide and both main feed pumps to increase their speed.

(2) Overspeed of the feedwater pumps with feedwater valve control in manual:

The overspeed trip of the feedwater pump turbines is set 10% above 5,000 r/ min. Nominal operating speed is 4,000 r/ min. As such, an ove'rspeed of l 37.5% can be postulated. The corresponding flow increase is, however, limited by the capacity of the main feedwater regulating valves.

(3) Inadvertent start of the second feedwater pump at low power: The plant may be operated at a maximum of approximately 60% power with only one feedwater pump operating. Start of the second main feedwater pump, which I requires several operator actions to bring the pump up to speed and properly align the isolation and control valves or the failure of the feedwater controller could more than double the feedwater flow.

1 16 ANF-87-150(NP) l Volume 1 )

l (4) Two other possibilities that could result in an increase in feedwater flow are the startt;p of the auxiliary feedwater system and the l inadvertent opening of the feedwater control valve bypass line. However, each of these would provide less flow than failures (1) through (3) above, and hence these failures, (1) through (3), will bound the consequences of an increase in feedwater flow.

1 15.1.2.2 Event Descriotion - The increased flow to the steam generators causes an increase in the energy removal capability of the steam generators by reducing the average fluid enthalpy in the steam generators. The increased energy removal from the primary system causes the reactor coolant temperature at th0 outlet of the steam generator to decrease. The core inlet temperature will correspondingly be reduced, which will cause the core power to increase if the moderator temperature coefficient is negative.

Because this event is characterized as a primary system overcooling event, the primary system pressure initially decreases along with the core inlet temperature. There is also a possibility for a core power increase in the presence of a negative moderator reactivity feedback coefficient. Increased reactor power reduces the core DNB margin. A potential exists that the net effect of these three factors will represent a challenge to the core DNB margin.

1 15.1.2.3 Reactor protection - Reactor protection for the rated power and power operation conditions is provided by the variable overpcwer trip, low l pressurizer pressure, thermal margin / low pressure, and low steam generator pressure trips. l I

For the reactor critical and hot standby operating conditions, early protection is provided by the low pressurizer pressure trip, safety injection actuation signal, and a nonsafety grade high rate-of-change of power trip (10'4% to 16% power). Reactor protection for the Increase in Feedwater Flow )

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17 ANF-87-150(NP)

Volume 1 event is summarized in the following tables: Table 15.1.2(1)-A, Complete Opening of a Feedwater Control Valve; (2) Table 15.1.2(2)-A, Overspeed of the Feedwater Pumps; and (3) Table 15.1.2(3)-A, Inadvertent Start of the Second Feedwater Pump.

15.1.2.4 Disoosition and Justification - In the case of failure 1 (event initiator 1) (Complete opening of a feedwater regulating valve) and failure 2 (0verspeed of the feedwater pumps with the feedwater valve control in manual),

the event consequences at rated power operating conditions will bound the consequences from all other conditions.

At rated power operating conditions, the initial thermal margin (DNBR) is minimized. Maximizing the increase in feedwater flow (assuming that ksih main feedwater regulating valves fail to the full open position) maximizes the load demand. This results in the maximum rate of moderator cooldown which, in the presence of a negative moderator temperature coefficient, results in the maximum challenge to the specified acceptable fuel design limits (SAFDLs).

. Therefore, the limiting consequences of the increase in feedwater ficw due to failures 1 and 2, above, will occur at the full load rated power conditions and will bound all other conditions due to the initial steam generator inventory and initial margin to DNB. This event would be limited by the high l,

steam generator level override circuit that would close the regulating valve l to the affected steam generator, which should produce a thrust bearing high load trip on the driver of the steam turbine feedwater pump. If the feedwater transient were to continue, the high water level trip would trip the plant.

I For the following reactor operating conditions: power operation and hot standby, it is possible that only one feedwater pump could be in service. In i these cases for the failure modes (1) and (2), only a 10% increase in flow

( would be available, which is less than the 20% increase in flow which was conservatively assumed for failure mode 1.

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18 ANF-87-150(NP)

Volume 1 In the case of failure 3 (Inadvertent start of the second feedwater pump at low power), only those initial conditions between 5% and 60% of rated power need to be considered. This is due to the fact that 60% power is the highest power level with one feedwater train in operation, and 5% power is the lowest power level for the main feedvater pumps to be on line. Also for these l operating conditions, the variable overpower trip setpoint provides suffi.:ient margin to DNB that the consequences of these events are bounded by the increase in feedwater flow from rated power condition. Finally, it should be noted that a 50% increase in feedwater flow (symmetric case - steaming rate constant) would result in a primary cooldown of approximately 0.25'F/sec, as compared to 1.78'F/sec postulated in Event 15.1.3.

The disposition of events for the Increase in Feedwater Flow event is summarized in the following tables: (1) Table 15.1.2(1)-8, Complete Opening of a Feedwater Control Valve; (2) Table 15.1.2(2)-8, Overspeed of the Feedwater Pumps; and (3) Table 15.1.2(3)-8, Inadvertent Start of the Second Feedwater Pump.

15.1.3 Increase in Steam Flow 15.1.3.1 Event Initiator -

This event is initiated by a failure or misoperation of the main steam system which results in an increase in steam flow from the steam generators. This event could be caused by the rapid opening of the turbine control valves, the atmospheric steam dump valves, or the steam bypass to condenser valves.

15.1.3.2 Event Descriotion - This increase in steam flow creates a mismatch between the energy being generated in the reactor core and the energy being removed through the secondary system. This mismatch results in a cooldown of the primary system. A power increase will occur if the moderator temperature reactivity feedback coefficient is negative. This power increase will cause a decrease in the DNBR margin. .

/

L 19 ANF-87-150(NP)

Volume 1 15.1.3.3 Reactor Protection - The main steam system is designed to accommodate a 10% increase in load (step increase) or a 5% per minute load ramp for power levels between 15% and 100% of full power. Reactor protection against a main steam flow increase (greater than a 10% step or 5% per minute ramp) is provided by the following trip signals: variable overpower trip, thermal margin / low pressure trip, low steam generator water level trip, and low secondary pressure trip. Reactor protection for the Increase in Steam Flow event is summarized in Table 15.1.3-A.

15.1.3.4 Disoosition and Justification - The turbine control valves are sized to accommodate steam flow for a power 5% in excess of 2450 MWt. This area is equivalent to a 1.7% increase in power above the nominal 2530 MWt. The atmospheric steam dump valves are sized to accommodate 30% of the steam flow at 2450 MWt. Ihis represents a 26% increase in power above the nominal 2530 MWt. The stea'm bypass to condenser valve is sized to accommodate 5% of the steam flow at 2450 MWt. This area is equivalent to an increase in power, above the nominal 2530 MWt, of approximately 1.7%. As such, the simultaneous opening of the turbine control valves, atmospheric dump valves, and the steam bypass valves would result in an increased load of approximately 30% of the steam flow above the rated power operating condition of 2530 MWt. At r'ated power operating conditions, this will result in an increased energy removal rate from the secondary system of approximately (0.30)(10.97 10 6 lb/hr, FSAR Chapter 15, Table 14.1-3)(1201.14 Btu /lb)/(3600 sec/hr) - 1.098 10 6 Btu /sec.

This energy removal rate bounds the rated power operating conditions for Events 15.1.1, 15.1.2, and 15.1.4. Therefore, this event will be analyzed as part of the plant transient analysis for Palisades operation with the modified RPS. The consequences of this event for all other power operating conditions are bounded by the rated power operating condition due to the increased margin to DNB at the other power operating conditions.

In the case of the following operating conditions: reactor critical, hot

20 ANF-87-150(NP)

Volume 1 I

standby, hot shutdown, refueling shutdown condition, cold shutdown condition, and refueling operation, the turbine control valves are closeo. Since the hot shutdown operating condition has a higher average coolant temperature and a larger potential for cooldown than the refueling shutdown, cold shutdown, and refueling operation reactor operating conditions, in combination with a Technical Specification minimum shutdown margin of -2% 40, which is less than the hot standby and reactor ;ritical conditions due to the additional margin available when the withdrawn control rods are tripped, this condition represents the bounding event for the "zero power" initial conditions.

l For the above case, the capacity of the valves that are assumed to open is (1.098 106 Btu /sec)(0.28/0.30) - 1.0248 106 Btu /sec. This would result in a primary cooldown of approximately 1.78'F/sec. In combination with the most negative moderator temperature coefficient (3.5 . 10'4 Ap/'F), this would '

result in a reactivity insertion rate of 6.22 10~4 Ap/sec. This' reactivity i addition rate will be bounded by the spectrum of reactivity addition rates and 1 initial conditions considered in Event 15.4.1 and the steamline break events considered in Reference 5. However, as compared to Event 15.4.1, this event I

has the potential for emptying the pressurizer with a return to pcwer which j could result in a more severe thermal margin transient than the reactivity addition transients due to control rod withdrawals. Also, as compared to Event 15.1.5, which is a much lower probability event, this event should result in consequences which do not violate SAFDLs and thermal margin (DNBR) limits. As such, this event will be analyzed with the assumption of the simultaneous opening of the atmospheric dump valves and the steam bypass valves with the transient initiated from the Technical Specification shutdown margin limit of -2% Ap. The disposition of events for the Increase in Steam Flow event is summarized in Table 15.1.3-8.

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L f 21 ANF-87-150(NP) i Volume 1

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15.1.4 Inadvertent Ooenina of a Steam Generator Relief or Safety Valve 15.1.4.1 Event Initiator - This event is initiated by an increase in steam flow caused by the inadvertent opening of a secondary side safety or relief valve.

{

15.1.4.2 Event Descriotion - The resulting mismatch in energy generation and a removal rates results in an overcooling of the primary system. If the 1 moderator temperature coefficient is negative, the reactor power will increase.

15.1.4.3 Reactor Protection - Reactor protection is provided by the variable overpowar trip, thermal margin / low pressure trip, low secondary pressure trip, and low steam generator water level trip. Reactor protection for the Inadvertent Opening of a Stea'm Generator Relief or Safety Valve event is summarized in Table 15.1.4-A.

15.1.4.4 Disoosition and Justification - The inadvertent opening of a steam 4 generator safety valve would result in an increased steam flow of approximately 4.32% of full rated steam flow. Each dump (relief) valve is sized for approximately 8% steam flow with the reactor at full rated power.

As such, the consequences of any of these occurrences will be bounded by the events in Section 15.1.3. The disposition of events for the Inadvertent Opening of a Steam Generator Relief or Safety Valve event is summarized in Table 15.1.4-B.

15.1.5 Steam System Pioina Failures Inside and Outside of Containment 15.1.5.1 Event Initiator - This event is initiated by a rupture in the main steam piping which results in an uncontrolled steam release from the secondary system.

22 ANF-87-150(NP)

Volume 1 15.1.5.2 Event Descriotion - The increase in energy removal through the secondary system results in a severe overcooling of the primary sysem. In the presence of a negative moderator temperature coefficient, this cooldown -

causes a decrease in the shutdown margin (following reactor scram) such that a j return to power might be possible following a steamline rupture. This is a potential problem because of the high power peaking factors which exist, l assuming the most reactive control rod to be stuck in its fully withdrawn

( position.

1

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15.1.5.3 Reacte* Protection - Reactor protection is provided by the low steam generator pressure trip, variable overpower trip, thermal margin / low pressure

, trip, low pressurizer pressure trip, and safety injection actuation signal.

Reactor protection for the Steam System Piping Failures inside and Outside of Containment event is summarized in Table 15.1.5-A.

15.1.5.4 Disoosition and Justification - This event and its consequences were considered for the Palisades e.ycle 6 safety analysis report (Reference 7) and found to be bounded by the analyses presented in Reference 22. Additionally, steam line break analyses were also performed as documented in Reference 6. f Due to the increased steam generator tube plugging the effective primary to secondary heat transfer area has been reduced from that previously considered.

In addition to this, the decreased steam generator operating pressure will slightly reduce the initial blowdown energy removal. As such, the results will be less severe than those presented in Reference 22. Therefore, the

~

analyses presented in Refere,1ce 22 will bound the consequences of this event for Palisades operation with the modified RPS and 29% steam generator tube plugging. The disposition of events for the Steam System Piping Failures Inside and Outside of Containment event is summarized in Table 15.1.5-B.

f 23 ANF-87-150(NP)

Volume 1 Table 15.1.1-A Available Reactor Protection for Decrease in Feedwater Temperature Event Reactor Operating Conditions Reactor Protection 1 Variable Overpower Trip Thermal Margin / Low Pressure Trip Low Pressurizer Pressure Trip Low Steam Generato'r Pressure Trip 2, 3 Same as above.

h Rate-of-Change NonsafetyGradeHig%to15%

of Power Trip, 10' Power, 1

No credit taken 4-8 Not a credible event for these reactor operating conditions since there is no extraction steam to the feedwater heaters

}

s 24 ANF-87-150(NP)

Volume 1 Table 15.1.1-B Disposition of Events for the Decrease in Feedwater Temperature Event Reactor Operating Conditions Disoosition l

1 Bounded by Event 15.1.3, Increase in Steam Flow Event 2, 3 Bounded by the above 4-8 No analysis required; not a credible event

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i 25 ANF-87-150(NP)

Volume 1 Table 15.1.2(1)- A Available Reactor Protection for the Increase Feedwater Flow E';ent i

Comolete Ooenina of a Feedwater Reaulatina Valve Reactor Operating Conditions Reactor Protection 1 Variable 0:lesporer Trip Low Pressurizer Pressure Trip t

Therma) s Pressure Trip i Low

• 4 ..cor Pressure Trip Safety .s ection Actuation Signal 2, 3 Same as above i

h Rate-of-Change Nonsafety Grade of Power Trip, 10'Hig% to 15% Power, No credit taken  ;

4-8 Not a credible event; Main Feedwater Pumps not on line 1

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- - - . - - - , -m -__-----._---.,-~,,c. c.,3,m._ _,,~7- - , - , . _ _ - - - ,, , , _ ,

l 26 ANF-87-150(NP)

Volume 1 Table 15.1.2(1)-B Disposition of Events for the Increase in Feedwater Flow Event Comolete Ooenina of a Feedwater Reaulatina Valve 1

l Reactor Operating l Conditions Disoosition l

1 Bounded by Event 15.1.3 (Increase in Steam Flow) 2, 3 Bounded by the above 4-8 No analysis required; not a credible event i

27 ANF-87-150(NP)

Volume 1 Table 15.1.2(2)-A Available Reactor Protection for the Increase in Feedwater Flow Event l Oversneed of the Feedwater Pumos Reactor Operating Conditions Reactor Protection 1 Variable Overpower Trip Low Pressurizer Pressure Trip Thermal Margin,' Low Pressure Trip Low Steam Generator Pressure Trip Safety injection Actuation Signal 2, 3 Same as above.

Nonsafety Grade High Rate-of-Change of Power Trip, 10-4% to 15% Power, No credit taken 4-8 Not a credible event; Main Feedwater Pumps not on line l

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28 ANF-87-150(NP)

Volume 1 Table 15.1.2(2) B Disposition of Events for the Increase in Feedwater Flow Event Oversoeed of the Feedwater Pumos Reactor Operating Conditions Disoosition i 1 Bounded by 15.1.2(1), Complete Opening of a Feedwater Control Valve 2, 3 Bounded by the above.

4-8 No analysis required; not a credible event l

l 1

29 ANF-87-150(NP) f Volume 1 Table 15.1.2(3)-A Available Reactor Protection for the Increase in Feedwater Flow Event Inadvertent Start of an Additional Main Feedwater Pumo .

Reactor Operating Conditions Reactor Protection 1 Not a credible event.

2, 3 Variable Overpower Trip Low Pressurizer Pressure Trip Jhermal Margin / Low Pressure Trip Low Steam Generator Pressure Trip Safety Injection Actuation Signal h Rate-of-Change NonsafetyGradeHig%to15%

of Power Trip, 10' Power,No credit taken t

4-8 No analysis required; not a credible event s

30 ANF-87-150(NP)

Volume 1 Table 15.1.2(3)-B Disposition of Events for the Increase in Feedwater Flow Event 3

Inadvertent Start of an Additional Main Feedwater Pumo Reactor Operating Conditions Disoosition 1 No analysis required; not a credible event.

2, 3 Bounded by Event 15.1.2(1) 4-8 No analysis required; not a credible event.

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h 31 ANF-87-150(NP) l Volume 1 f

Table 15.1.3-A Available Reactor Protection for the Increase in Steam Flow Event Reactor Operating Conditions Reactor Protection 1 Low Steam Generater Pressure Trip Thermal Margin / Low Pressure Trip Variable Overpower Trip Low Pressurizer Pressure Trio

. Safety Injection Actuation Signal 2,3,4 Saree as above.

' h Rate-of-Change NonsafetyGraceHig%to15%

of Power Trio, 10 Power,

No credit taken i

j 5 Low Pressurizer Pressure Signal Safety Injection Actuation Signal l 68 No analysis required; not a l

credible event.

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Volume 1 l

Table 15.1.3-B Disposition of Events for the Increase in Steam Flow Event I

t Reactor Operating l Conditions Disoosition i

j 1 Analyze 2,3,4 Bounded by the above 1

5 Analyze 4

6-8 ,

No analysis required; not a credible event -

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33 ANF-87-150(NP) f Volume 1 Table 15.1.4-A Available Reactor Protection for the Inadvertent Opening of a Steam Generator Relief or Safety Valve Event Reactor Operating Conditions Reactor Protection 1 Low Steam Generator Pressure Trip-1 Variable Overpower Trip Thermal Margin / Low Pressure Trip Low Pressurizer Pressure Trip Safety Injection Actuation Signal, 2,3,4 Same as above.

NonsafetyGrag%to15%

of Power, 10- Power,e High Rate-of-Ch l

No credit taken 5 Low Pressurizer Pressure Signal Safety Injection Actuation Signal r

! 6-8 No analysis required; not a l credible event 1

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34 ANF-87-150(NP) ,

Volume 1 l

Table !!<.1.4-B Disposition of Events for the Inadvertent Opening of a Steam Generator Relief or Safety Valve Event i

l Reactor Operating Conditions __ Disoosition l

1 Bounded by analyses presented for Event 15.1.3 2,3,4 Bounded by the above ,

5 Bounded by the analyses presented  ;

for Event 15.1.3 i 6-8 Not a credible event; no analysis required ,

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Volume 1 I

Table 15.1.5 A Available Reactor Protection for Steam System Piping Failures

Inside and Outside of Containment Event i

Reactor Operating Conditions Reactor Protection 1 Low Steam Generator Pressure Trip Variable Overpower Trip Thermal Margin / Low Pressure Trip Low Pressurizer Pressure Trip Safety Injection Actuation Signal 2,3,4 Same as above.

r e High Rate of-Change NonsafetyG'ag%to15%

of Power, 10' Power,No '

credit taken 5 Low Pressurizer Pressure Signal Safety Injection Actuation Signal 6-8 No analysis required; not a credible event i

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36 ANF87-150(NP)

Volume 1.

Table 15.1.5-8 Disposition of Events for Steam System Piping Failures Inside and Outside of Containment Event Reactor Operating Conditions Discosition 1 Bounded by analyses presented in ,

References 5, 6 and 22 L 2,3,4 Bounded by the above r 5 Bounded by analyses presented in References 5, 6 and 22 6-8 Not a credible event; no analysis required ,

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37 ANF-87-150(NP)

Volume 1 i 15.2 DECREASE IN HEAT REJ40 VAL BY THE SECONDARY SYSTEM 15.2.1 Loss of External load 15.2.1.1 Event Initiator - A major loss of load can be initiated as the result of a loss of external electrical load or a turbine trip. Turbine stop valve closure is assumed as the initiator of this event because this is the fastest load rejection which can be postulated which will challenge the plant overpressure and SAFDL protection features. The assumed fast valve closure time (0.1 sec.) and the assumed unavailability of the steam dump system allow this event to bound the effects of Events 15.2.2 (Turbine Trip - Steam bypass system available), 15.2.3 (Loss of Condenser Vacuum - Steam bypass system unavailable), and 15.2.4 (Closure of the MSIV - Valve closure time >0.1 sec.).

15.2.1.2 Event Descriotion - For a full load reduction at power, the primary to secondary heat transfer would be severely diminished because of the increase in secondary side temperature. Initially, in response to the load reduction and diminished energy removal through the secondary system, the I primary system temperatures begin to increase. The increasing primary system ,

average temperature causes an insurge into the pressurizer due to the expanding primary fluid. The primary system pressure increases as the pressurizer steam space is compressed by the ins rging liquid. Primary system overpressure protection is afforded by the pressurizer power operated relief i valves (the block valves are closed during power operations) and the primary ,

safety valves. Eventually, the secondary system pressure reaches the opening setpoint of the secondary side safety valves and steam discharge occurs to I

limit the secondary side pressure rise. Energy removal through the steam l generator and pressurizer safety valves mitigates the consequences of the load reduction. However, in analyzing the overpressurization aspects of this event, no credit is taken for the power operated relief valves on the primary  :,

or secondary systems.

6 38' ANF-87-150(NP)

Volume 1 15.2.1.3 Reactor Protection - Reactor protection is provided by the high pressurizer pressure, variable overpower, thermal margin / low pressure, and low steam generator water level trips. If the turbine is tripped at the initiation of this event, a direct reactor trip signal would be generated and the effects of this event would be mitigated. However, no credit is taken for a direct reactor trip on turbine trip. Additionally, reactor protection is provided by the primary and secondary safety valves. Because of the potential for increasing the primary system temperatures, with small increases in pressure, this event can challenge the SAFDLs as well as the overpressure criteria mentioned above. Reactor protection for the loss of External Load event is summarized in Table 15.2.1-A.

15.2.1.4 Disposition and Justification - This event is only credible for rated power and power operating conditions because there is no load on the turbine at other reactor conditions. The consequences of this event for rated power operation bound the consequences for other reactor conditions because of the increased stored energy. The higher the stored energy in the primary system, the more severe the consequences of this event.

1 This event will be analyzed from rated power conditions for Palisades operation with the modified RPS. l l

The disposition of events for the Loss of External Load event is summarized in Table 15.2.1-B.

l 15.2.2 Turbine Trio ,

15.2.2.1 Event Initiator - This event is initiated by a turbine trip which results in closure of the main steam stop valves and a rapid reduction in energy removal through the steam generators.  ;

15.2.2.2 Event Descriotion - The reactor protection system is designed to l

i

I e

39 ANF 87-150(NP)

Volume 1 generate a reactor trip signal automatically when the turbine is tripped.

Following reactor trip, there would be a rapid decrease in the energy being generated in the primt 7 stem. This would mitigate the consequences of the turbine trip event. Primary and secondary system overpressurization protection is provided by the code safety valves on both the primary and secondary systems and the secondary atmospheric dump valves. Also, if the condenser was available, the steam bypass system would be activated to reduce the secondary system pressure.

15.2.2.3 Reactor Protection - Geactor protection is provided by the high pressurizer pressure trip, variable overpower trip, thermal margin / low pressure trip, low steam generator water level trip, and a nonsafety grade ,

reactor trip on turbine trip. Additional protection is also provided by the primary and secondary side safety valves. Reactor protection for the Turbine Trip event is summarized in Table 15.2.2-A.

15.2.2.4 Disoosition and Justification This event is only credible for rated power and power operating conditions since the turbine will either be in a tripped condition or there will be no load on the steam generators for other reactor operation conditions. The consequences of this event for rated power operation bound the event consequences for other operating conditions because of the higher initial stored energy in the p'rimary system and the reduced SAFDL margin for rated power operation. Because of the limiting assumptions used in the analys4s of the consequences of the Loss of External Load event (15.2.1), the consequences of the Turbine Trip event are bounded by the consequences of Event 15.2.1. The major assumptions used in Event 15.2.1 are the conservatively rapid turbine stop valve closure time, the failure to trip the reactor on turbine trip, and the assumed unavailability of the atmospheric steam dump system. The disposition of events for the Turbine Trip event is sunnarized in Table 15.2.2 B.

(b 40 ANF-87-150(NP)

Volume 1

'Jgf Condenser Vacuum IS 13.2. ..en t Initiator - A reduction of circulating water flow or an increi e .o the circulating water temperature can impact the condenser back pressure (condenser vacuum conditions), and result in a turbine trip without the availability of steam bypass to condenser.

15.2.3.2 Event Descriotion - Loss of coqdenter vacuum and subsequent turbine trip results in a rapid increase in secondary side temperatures with severe reduction of primary to secondary heat transfer for the rated power and power operating conditions. The primary and secondary side temperature and pressure increase will be limited by the primary and secondary side safety valves.

Trip of the reactor on turbine trip due to the loss of condenser vacuum will terminate the challenge to vessel overpreusurization limits and the DNB SAFDL.

In the case of the following reactor operating conditions: reactor critical, hot standby, and hot shutdown (where steam is being bypassed directly to the condenser), the loss of condenser vacuum would result in a loss of primary and secondary temperature control . This would be a long term effect due to the low power levels and the operator would have sufficient time to initiate a controlled cooldown using the power operated relief valves.

15.2.3.3 Reactor Protection -

Reactor protection is provided by the high pressurizer pressure trip, variable overpower trip, thermal margin / low pressure trip, low steam generator water level trip, and a nonsafety grade reactor trip on turbine trip (on loss of condenser vacuum). Additional protection is provided by the primary and secondary side safety valves. In the case of the following reactor operating conditions: reactor critical, hot standby, and hot shutdown, reactor protection is provided by the primary and secondary safety valves and the operator controlled primary power operated relief valves and secondary power operated atmospheric relief valves, and the

I. l I

l l

l 41 ANF-87-150(NP) I Volume 1 i l

l pressurizer sprays. Reactor protection for the Loss of Condenser Vacuum event is summarized in Table 15.2.3-A.

15.2.3.4 Disoosition and Justification - The consequences of this event are exacerbated by the unavailability of the steam bypass system to release the excess steam following turbine trip because of the loss of condenser vacuum.

As such, the consequences of thic event are similar to those of the 15.2.1 event; however, a reactor trip on turbine trip is assumed. Therefore, the consequences of this event are bounded by Event 15.2.1. No credit for the operation of the steam dump and bypass system is assumed in 15.2.1.

For reactor operation conditions other than rated power a*1d power operation, the loss of condenser vacuum will have little impact since the operator has sufficient time to initiate a controlled cooldown. This disposition of events for the Loss of Condenser V'acuum event is summarized in Table 15.2.3-B.

15.2.4 Closure of the Main Steam Isolation Valves (MSIVs)(BWR) 15.2.4.1 Event Initiator - The event postulated is the loss of control air to the MSIV valve operator. The valves are swinging disc type check valves, l_

installed in a reversed position and held open against steam flow by a I

pneumatically operating cylinder assembly. The valves are spring-loaded to the c!osed position.

15.2.4.2 Event Descriotion - The inadvertent MSIV closure is primarily of concern in boiling water reactors, but closure of the MSIVs in a PWR would cause a drastic reduction in tile load on the reactor. As such, the consequences of this event are similar to the consequences of Event 15.2.1.

Although the valve closure time for the MSIVs is less than 5 seconds, this is much longer than the turbine stop valve closure time assumed in Event 15.2.1 (0.1 seconds); as such, the transient events will proceed somewhat slower and be less severe than in the case of Event 15.2.1.

l l

42 ANF-87-150(NP)

Volume 1 15.2.4.3 Reactor Protection - Reactor protection is provided by the high pressurizer pressure trip, variable overpower trip, thermal margin / low pressure trip, low steam generator water level trip, and a nonsafety grade reactor trip on turbine trip. Additional protection is provided by the primary and secondary side safety valves. Reactor protection for the Closura of the Main Steam isolation Valves event is summarized in Table 15.2.4-A.

15.2.4.4 Disoosition and Justification - This event is not a concern for the following reactor operating conditions: reactor critical, hot standby, hot shutdown, refueling shutdown, cold shutdown, and refueling since the turbine will either be in a tripped condition or there will be no load on the steam generators and the MSIVs will be closed. As_in Event 15.2.1, the consequences of this event from rated power conditions will bound the consequences for all other power operating conditions, and these consequences will be bounded by the consequences of Event 15.2.1. The disposition of events for the Clcsure of the Main Steam Isolation Valves event is summarized in Table 15.2.4-B.

15.2.5 Steam Pressure Reculator Failure Palisades does not have any steam line pressure regulators, so this event is not credible for this plant. No analysis needs to be considered for this event.

15.2.6 Loss of Nonemeraency A.C. Power to the Station Auxiliaries 15.2.6.1 Event Initiator - The loss of nonemergency A.C. power to the station auxiliaries may be cLused by a complete loss of the offsite grid in conjunction with a turbine generator trip or by a failure in the onsite A.C.

power distribution system.

15.2.6.2 Event Descriotion - A complete loss of nonemergency A.C. power may

i t

43 ANF-87-150(NP)

Volume 1 result in the loss of all power to the station auxiliaries, including the main reactor coolant pumps. This event is more severe than the turbine trip event becaues the decrease in heat removal by the secondary system is accompanied by a larger increase in the core coolant temperatures due to the reduction in primary system flowrate.

Also, in this case the main feedwater pumps will trip on low suction pressure due to the loss of the condensate pumps due to lack of A.C. power. Therefore, a total loss of main feedwater flow will be superimposed at this point 'n time. However, the auxiliary feedwater system will be available to provide feedwater for the long term removal of decay power from the core.

This event is comprised of two distinct phases. A near-term phase in which -

the system response is dominated by the flow coastdown due to the loss of power to the main reactor coolant pumps. The second phase concerns the long term event consequences which are dominated by the large reduction in main feedwater flow in conjunction with the loss of flow.

The flow coastdown will result in a rapid reduction of the DNB . margin.

However, the consequences are mitigated by the core power reduction following reactor trip and control rod insertion.

The loss of flow in conjunction with the turbine trip and the reduction of main feedwater flow will cause a reduction in the primary to secondary heat transfer through the steam generators. The reduction in energy removal capability will result in an increase in the primary side steam generator outlet temperature during the initial phase of the event before the reactor power has decreased appreciably. Core coolant temperatures will al so initially increase due to the flow coastdown. In this initial phase, there will be an insurge into the pressurizer due to the expansion of the primary system fluid. Over the course of the transient, the primary system temperature response will be dependent upon the capability of the auxiliary i

l

. __ - ___ _ _ ___-_ ____ _ _ ._ __ _ _ _ . _ _________ __ ____ A

44 ANF-87-150(NP)  !

Volume 1 i feedwater system to insure that sufficient water is available for the retroval of decay energy through the steam generators.

15.2.6.3 Reactor Protection - Reactor protection is afforded by a reactor trip on one of the following: low reactor coolant flow, high pressurizer pressure, thermal margin / low pressure, and low steam generator water level. i Overpressurization protection for the primary and secondary system is provided by the primary and secondary safety valves. Reactor protection for the loss of Nonemergency A.C. Power to the Station Auxiliaries event is summarized in Table 15.2.6-A.

15.2.6.4 Disoosition and Justification - The consequences of this event at rated power operation will bound the consequences of all other reactor conditions. At rated power operation, the challenge to both the SAFDLs and overpressure limits is maximized because the' initial stored energy in the system is maximized and the initial margin to the SAFDLs is minimized.

The near-term thermal margin (DNB) response of the primary system is similar to that resulting from a four pump loss of flow such as is postulated in Event i 15.3.1. Because of the early reactor trip in response to the loss of nonemergency A.C. power, the consequences in the near term will be bounded by l the consequences of the Loss of Forced Reactor Coolant Flow event postulated in 15.3.1.

1 The long-term consequences of this event are similar to those of the Loss of Normal Feedwater Flow event with a concurrent loss of offsite power. Because of the earlier reactor trip in the Loss of Nonemergency A.C. Power event, the I long-term consequences of this event will be bounded by the consequences of the Loss of Normal Feedwater Flow event with a concurrent loss of onsite and offsite power, which is discussed in Event 15.2.7. ,

i The disposition of events for the loss of Nonemergency A.C. Power to the 1

l 45 ANF-87-150(NP)

Volume 1 Station Auxiliaries event is summarized in Table 15.2.6-8.

15.2.7 Loss of Normal Feedwater Flow 15.2.7.1 Event Initiator - The Loss of Normal Feedwater Flow transient is initiated by a trip of the main feedwater pumps or a malfunction in the feedwater control valves.

15.2.7.2 Event Descriotion - The Loss of Normal Feedwater Flow event results in a total loss of all main feedwater flow to the steam generators. Because the main feedwater system is supplying subcooled water to the steam generators, the loss of main feedwater flow will result in a reduction of the secondary system heat removal capability. The decrease in energy removal rate from the primary system causes the primary system fluid temperature to increase. The resulting primary system fluid expansion results in an insurge into the pressurizer, compressing the steam space and causing the primary system pressure to increase.

1 l Long-term cooling capability through the steam generators is assurcd by the feedwater supplied to the steam generators through the Auxiliary Feedwater Sys teia. The motor driven auxiliary feedwater pumps (two) and the turbine l

l driven auxiliary feedwater pump are automatically started on the following signal: low level in any one steam generator. The motor driven auxiliary feedwater pumps ara powered by the diesels if a loss of offsite power occurs.

The steam for the turbine driven pump is provided through the plant secondary system.

15.2.7.3 Reactor Protection - System overpressure protection is provided by the primary and secondary system safety valves. A reactor trip occurs on low steam generator level with additional reactor protection provided by the high l pressurizer pressure trip and the thermal margin / low pressure trip. Reactor protection for the Loss of Normal Feedwater Flow event is summarized in Table l .

t

46 ANF-87-150(NP)

Volume 1 15.2.7-A.

15.2.7.4 Discosition and Justification - This event is only credible for rated power and power operating conditions because the main feedwater system is not required to provide feedwater to the steam generators for other reactor operating conditions, The consequences of this event for rated power operation bound the consequences for other conditions because of the higher initial stored energy in the primary system and the greater impact of the loss of feedwater flow on the secondary system. The consequences of this event will be analyzed for rated power operation to assess the long term consequences of increased steam generator tube plugging. The disposition of events for the loss of Normal Feedwater Flow event is summarized in Table 15.2.7-B.

15.2.8 Feedwater System Pioe Breaks Inside and Outeide Containment 15.2.8.1 .rgent Initiator - This event is initiated by a rupture in the main feedwater piping network which is sufficient to prevent the addition of main feedwater to the steam generators to maintain shell-side fluid inventory in the steam generators. If the break is postulated in the feedline between the steam generator and the feedline check valve, then fluid may be discharged from the steam generators out through the postulated break. "On each steam generator the auxiliary feedwater is injected directly into the stea.m generators. As such, any break in the main feedline will not affect the subsequent addition of auxiliary feedwater into the affected steam generator.

A break upstream of the check valve would result only in a loss of feedwater  !

flow to the steam generator and not in any fluid discharge from the steam generator, and the transient would proceed similar to a loss of normal {

feedwater.

15.2.8.2 Event Descriotion - The main feedwater for the Palisades plant is injected into the upper regions of the steam generator downcomer. Depending on

/

47 ANF-87-150(NP)

Volume 1 the size of the break and the plant operating conditions, the primary system may respond to the postulated break as a cooldown event or as a heatup event.

The cooldown event results due to excessive energy removal capability via flow out the break. The heatup of the reactor coolant system results from a i decrease in the energy removal capability of the secondary system due to a 1

combination of the following:

~

l (1) Reduction in the addition of subcooled feedwater to the steam generators.

l (2) Loss of shell-side inventory in the steam generators by discharge through the break.

15.2.8.3 Reactor Protection - The primary concern with this event is the i overpressurization potential. Reactor overpressure protection is provided by the power operated pressurizer relief valves (if available) and the pres-surizer safety valves. The reduction in core power on reactor trip (low steam generator water level, thermal margin / low pressure, high pressurizer pres-sure, and low steam generator pressure) alleviates the primary system overpressurization.

Long-term cooling protection is provided by the auxiliary feedwater system which provides a source of feedwater to the steam generators for decay energy removal. Upon detection of low level in either steam generator, Auxiliary Feedwater Actuation Signal (AFAS) signals are generated to automatically initiate the Auxiliary Feedwater System (AFWS). Reactor protection for the Feedwater System Pipe Breaks Inside and Outside Containment event is I

summarized in Table 15.2.8-A.

15.2.8.4 Discosition and Justification - This event is of primary concern for the rated power operation. The consequences of this event for other reactor operating conditions are bounded by the rated power operating condition due to the higher initial stored energy in the primary system which maximizes the

1 48 ANF-87-150(NP)

Volume 1 overpressurization potential. Also, the rated power operating condition maximizes the decay heat which must be removed to insure the long-term coolability of the primary system. The cooldown effects of this event at rated power are bounded by the cooldown resulting from a rupture in the main steam piping. The heatup effects following the initial cooldown of this transient (overpressurization) are bounded by the Loss of External Load Event 15.2.1. The long-term cooling aspects are essentially equivalent to those due to the Loss of Normal Feedwater Event 15.2.7. For those reactor operating conditions with little or no load on the reactor and the system is at temperature 2525*F (reactor critical, hot standby, and hot shutdown), the event is primarily a cooldown event and is bounded by the cooldown associated with the main steamline break (15.1.5). The disposition of events for the Feedwater System Pipe Breaks Inside and Outside Containment event is summarized in Table 15.2.8-B.

l

k 49 ANF-87-150(NP)

Volume 1 Table 15.2.1-A Available Reactor Protection for the Loss of External Load Event Reactor Operating Conditions Reactor Protection 1 High Pressurizer Pressure Trip Variable Overpower Trip

' Thermal Margin / Low Pressure Trip Low Steam Generator Water Level Trip 2 Same as above.

3-5 No significant consequences for these reactor operating conditions.

{

) 6-8 No analysis required; not a credible event.

I, i

i

~

!=

1 1

50. ANF-87-150(NP)

Volume 1 i

l' i Table 15.2.1-B Disposition of Events for the Loss i of External Load Event l

l^

Reactor Operating Conditions Disoosition

(

l Analyze I I

2 Bounded by the above, no analysis I required.

3-5 No analysis required; operator has sufficient time to initiate controlled cooldown.

f 6-8 No analysis required; not a j credible event, i I

i i

51 ANF-87-150(NP)

Volume 1 1

Table 15.2.2-A Available Reactor Protection for the Turbine Trip Event-Reactor Operating Conditions Reactor Protection 1 High Pressurizer Pressure Trip Variable Overpower Trip Thermal Margin / Low Pressure Trip low Steam Generator Water Level Trip 2 Same as above.

). 3-5 No significant consequences for' these reactor operating conditions.

6-8 No analysis required; not a credible event.

/

I I

i 52 ANF-87-150(NP) ,

Volume 1 Table 15.2.2-8 Disposition of Events for the Turbine Trip Event Reactor Operating Conditions Disoosition 1 Bounded by Event 15.2.1.for the rated power operating condition (#1).

2 Same as above.

i 3-5 No analysis required; operator has sufficient time to initiate l

controlled cooldown.

6-8 No analysis required; not a g credible event.

k l

1 i

8 l

l i

. . , - . , . - , - - . . , - - . . - - . . - - , - . . . - . . . - - - . , . - , - - , . . , . - - - , - . , , . , - . , . , , . . - , . , . . - , , , . , . - . . .....c.- ,

[

53 ANF-87-150(NP)

Volume 1 B

Table 15.2.3-A Available Reactor Protection for the loss of Condenser Vacuum Event Reactor Operating Conditions Reactor Protection i

1 High Pressurizer Pressure Trip Variable Overpower Trip Thermal Margin / Low Pressure Trip Low Steam Generator Water Level Trip 2 .

Same as above.

3-5 No significant consequences for

' these reactor operating conditions.

6-8 No analysis required; not a credible event.

r l

i i

i I

I

(

54 ANF-87-150(NP).

Volume 1 c

Table 15.2.3-B Disposition of Events for the Loss of Condenser Vacuum Event l

Reactor Operating Conditions .DiscositiqD 1 Bounded by Event 15.2.1 for the f rated power operating condition (#1).

2 Bounded by above.

3-5 No analysis required, operator has sufficient time to initiate controlled cooldown.

6-8 No analysis required; not a .

credible event. -

h

. J l

i

)

i

, _ _ . - . . ym_, .._. .-.._. . , - _ _ _ , , . _ . _ . _ . _ . - _ , , _ _ - . . _ _ , , - , , _ . , . . _ ~ , _ , . , , . . . . , , ,

r 55 ANF-87-150(NP)

Volume 1 1,

Table 15.2.4-A Available Reactor Protection for the Closure of the MSIVs Event Reactor Operating Conditions Reactor Protection 7

1 High Pressurizer Pressure Trip Variable Overpower Trip Thermal Margin / Low Pressure Trip Low Steam Generator Water Level Trip ,

2 Same as above.

J 3-5 No significant consequences for these reactor operating conditions.

0 6-8 No analysis required; not a credible event.

1 l

(

I 1

. .,, , , - . - - , . . - - - - - - - , . , , ,,- ,. , .,,...v. ..,,n.,, ,, . . . , - - - - - - - - - ,-,

V l-'

l 56 ANF-87-150(NP)

i. Volume 1

(

Table 15.2.4-8 Disposition of Events for the-Closure of the MSIVs Event '

L.  !

l .

f I

l l Reactor Operating ~

l Conditions Disoosition .I

\

1 Bounded by Event 15.2.1 for the rated power operating condition (#1).

2 Same as above, i l 3-5 .No analysis required; operator has sufficient time to initiate j ,

controlled cooldown.

\ \

6-8 No analysis required; not a f I

credible event. I L

i

'l i

l i

c

(

57 ANF-87-150(NP)

Volume 1

(

l Table 15.2.5-A Available Reactor Protection for the f Steam Pressure Regulator Failure Event Reactor Operating Conditions Reactor Protection 1-8 None required, not a credible event for this plant.

L i

J l

l 58 ANF87-150(NP) I Volume 1 J l

Table 15.2.5-B ' Disposition of Events for the .

Steam Pressure' Regulator Failure Event- l i

l Reactor Operating Conditions Discosition l t

1-8 No analysis required.

9 l

j i

l 1

9 a

l

, m._ -. - , - __c-,.-,, , , , , _ . . - - , - , - , - . ~ , . , - - ,,-.__,.._,,- . . ... . .. __... ..--_... ,,_ , -_, . _ ,, ,,,-..., ,

__m., , - - - .

f 59 ANF-87-150(NP)

L Volume 1 I-Table 15.2.6-A Available Reactor Protection for the loss of Nonemergency A.C. Power to the Station Auxiliaries Event 1

Reactor Operating Conditions Reactor _ Protection 1 Low Reactor Coolant Flow Trip High Pressurizer Pressure Trip Thermal Margin / Low Pressure Trip Low Steam Generator Water Level Trip 2 Same as the above.

3-8 No significant cons'equences for these reactor operating conditions.

I 1

60 ANF-87-150(NP)

Volume 1 Table 15.2.6-B Disposition of Events for the -

Loss of Nonemergency A.C. Power to the Station Auxili. ries Event i Reactor Operating Conditions Disoosition ,

1 Bounded, short term by Event 15.3.1, ,

Four Pump Loss of Flow for reactor operating condition (#1).

• Bounded, long term by Event 15.2.7, Loss of Normal Feedwater Flow for reactor operating condition (#1).

2 Bounded by the above, no analysis required. .

3-8 No analysis required.

I l

l a

v - - . . . . , -- , er-n-- ,,-,,--m,o ,e, - , , - e, -

v ---- - ---~,,,-wc- --

I' if I 61 ANF-87-150(NP)

Volume 1 Table 15.2.7-A Available Reactor Protection for the

(- Loss of Normal Feedwater Flow Event t

Reactor Operating Conditions Reactor Protection 1 Low Steam Generator Water Level Trip

' High Pressurizer Pressure Trip Thermal Margin / Low Pressure Trip 2 Same as the above.

3-8 No significant_ consequences for these reactor operating conditions.

e 4

i 1

1

1 1

62 ANF-87-150(NP)' .1 Volume 1 )

i i

Table 15.2.7-8 Disposition of Events for the-Loss of 1 Normal Feedwater Flow Event J i

Reactor Operating <

l Dispositian j Conditions 1 enalyze r

2 Bounded by the above, no analysis ,

required.

3-8 No analysis required.

q h

l

)

I 1

l l

1 l

i 63 ANF-87-150(NP)

Volume 1

c.

Table 15.2.8-A Available Reactor Protection for the .

Feedwater System Pipe Breaks Inside and Outside Containment Event Reactor Operating Conditions Reactor Protection 1 High Pressurizer Pressure Trip Thermal Margin / Low Pressure Trip Low Steam Generator Water Level Trip Low Steam Generator Pressure Trip I

2 Same as above.

3-8 No significant consequences for these reactor operating conditions.

t

s 64 ANF-87-150(NP)

Volumo 1 e Table 15.2.8-B Disposition of Events for the I Feedwater System Pipe Breaks Inside and Outside Containment Event i

i Reactor Operating Conditions Disoosition 4 l

1 The initial cooldown is bounded by' .

Event 15.1.5 from rated power -

operating conditions (#1). The subsequent heatup is bounded by  ;

Events 15.2.1 and 15.2.7 from rated power operating conditions (#1).

2 Bounded by the above, no analysis required.

3-8 No analysis required.

l 4

2

(

i i

f 4

r -

9r -

m--ewe- , ww.s*,we, -n. -*.,---e ---

,r-- -- + r.---e-- --- --- - -= < - - - - - - - - - ----e.--., ,.-wy- ,e -.+-r # .r--- 5 -

k 65 ANF-87-150(NP)

Volume 1

{

15.3 DECREASE IN REACTOR C0OLANT SYSTEM FLOW 15.3.1 Loss of Forced Reactor Coolant Flow 15.3.1.1 Event Initiatar - The loss of forced reactor coolant flow in the primary system may result from a mechanical or electrical failure in a main reactor coolant pump or in the power supply to these pumps. Forced coolant flow may be completely or partially lost.

15.3.1.2 Event Descriotion - The immediate result of the loss of forced coolant flow is an increase in the coolant temperature as it flows through the reactor core. If the reactor is at power, this temperature increase could challenge the specified acceptable fuel design limits. DNB could result with subsequent fuel damage if the reactor is not tripped.

15.3.1.3 Reactor Protection - Reactor protection is provided by the following reactor trips:

(1) Low reactor coolant flow; (2) Thermal margin / low pressure; and (3) High pressurizer pressure trip.

Reactor protection for the Loss of Forced Reactor Coolant Flow event is summarized in Table 15.3.1-A.

I 15.3.1.4 Discosition and Justification - The power sources for the main reactor coolant pumps are the most likely initiator for a loss of flow event l involving more than one pump. A mechanical or electrical fault in one of the pumps will only result in a single pump loss of forced coolant flow transient.

The normal power supplies for the pumps are from two buses which receive power from the main generator. Two pumps, in opposite loops, are powered from each bus. If there is a generator trip, the pumps are automatically transferred to

_--__--___._.-------____-__j

66 ANF-87-150(NP)

Volume 1 a bus supplied from the external power lines. A generator trip with the failure of this transfer could result in a loss of power to all four pumps.

In the case of four pump operation, two situations must be considered: two pump coastdown and a total loss of forced coolant flow. Considering first the total loss of flow cases, the consequences of this postulated event are bounded by rated power operation. The rated power case is bounding because of the reduced DNB margin for this initial state compared to other reactor power operating conditions. For the two pump loss of flow cases, the magnitude of the coastdown is less severe than the four pump coastdown, and the consequences of this event are bounded by the four pump loss of flow event.

For the two pump flow coastdown cases, there is always some degree of forced reactor coolant flow. These events are, therefore, not as challenging as the four pump coastdown events. A comparison of the governing parameters indicates that these events are bounded by the four pump loss of flow event from full rated power conditions.

In the case of less than four pump operation, infrequent operation with three pumps is permitted by Technical Specifications to provide a limited time for repair / pump restart, to provide for an orderly shutdown, or to provide for the conduct of reactor internals noise monitoring test measurements. The maximum power level allowed for three pump operation'is 39% of rated power. As a result of the increased flow to power ratio relative to full power four pump operation, loss of flow from the three pump configuration is less severe than the loss of flow from a four pump configuration.

In summary, the four pump loss of flow event is the bounding event for the 15.3.1 events. In the 15.3.3 event category, the SAFDL challenge posed by the reactor coolant pump rotor seizure event is greater than the four pump loss of ,

flow event and will bound the consequences of the four pump loss of flow event. However, the reactor coolant pump rotor seizure event (15.3.3) is a Condition IV event and can only be used to bound the 15.3.1 events if no fuel

+ -

--- . ,-. -,,---n, - - - - - - - - - - - - - - - - - , . - . . . , - . - - - g

I 67 ANF-87-150(NP) l Volume 1 damsge is predicted for the reactor coolant pump rotor seizure event. Limited fuel damage is anticipated for the reactor coolant pump rotor seizure event

and as such the four pump loss of flow will be analyzed.

The disposition of events for the Loss of Forced Reactor Coolant Flow event is summarized in Table 15.3.1-B.

15.3.2 Flow Controller Malfunction There are no flow control devices on the primary reactor coolant system of the l

Palisades plant. This event is therefore not credible and need not be analyzed.

15.3.3 Reactor Coolant Pumo Rotor Seizure 15.3.3.1 Event Initiator - This event is initiated by an instantaneous seizure of a reactor coolant pump rotor.

15.3.3.2 Event Descriotion - Flow in the affected loop will be rapidly reduced causing the core flow to also decrease rapidly. As in the 15.3.1 events, the reduction in primary reactor coolant flow will result in the increase in primary coola'nt temperatures and a challenge to the DNB margin. A pressurization of the primary system will also occur due to the heatup of the primary coolant which causes a rapid insurge into the pressurizer. A low reactor coolant flow trip will be generated. The reactor coolant pump rotor

seizure event represents the most rapid reduction in total core flow, bounding the flow coastdown response of all of the postulated 15.3 events during the very early coastdown period when there is the greatest challenge to the DNB margin. The mir,imum DNB typically occurs during the first 1 to 3 seconds following the initiation of the event.

15.3.3.3 Reactor Protection - Reactor protection for the reactor coolant pump J

i 68 ANF-87-150(NP) l Volume 1 rotor seizure event is provided by the low reactor coolant flow trip and the high pressurizer pressure trip. Additional reactor protection is provided for three-Primary Coolant Pump (PCP) operation by the increased margin to DNB reouired by the Technical Specification for three-PCP operation. Reactor protection for the Reactor Coolant Pump Rotor Seizure event is summarized in Table 15.3.3-A. 1 15.3.3.4 Disoosition and Justification - This event is a concern for i- .

rated power and power operating conditions because for other reactor opera..ng conditions there is sufficient thermal margin so there will not ha a challenge to the fuel design limits. The core heat flux to flow ratio is an excellent indicator of the potential DNB challenge for a loss of flow event. The highest ratio ~s for this event are predicted to occur during the fu:t few seconds of the transient from full power rated operating conditions. The consequences of this event'will therefore be bounded by a pump rotor seizure event initiated from full power rated conaitions. The consequences of a reactor coolant pump rotor seizure were analyzed from full power rated conditions for Palisades, Reference 5. Although this event is insensitive to transient secondary side effects, the event will be analyzed due to the increased coolant inlet temperature and increased tube plugging (decreased flow) .

The disposition of events for the Reactor Coolant Pump Rotor Seizure event is summarized in Table 15.3.3-B. 1 I

15.3.4 Reactor Coolant Pumo Shaft Break 15.3.4.1 Event Initiator - This event is initiated by a failure in the pump shaft of a primary reactor coolant pump causing the pump impeller to spin freely.

15 3 .2 Event Descriotion - This will cause the reactor coolant flow to l 1

l

69 ANF 87-150(NP)

Volume 1 decrease rapidly. The rate of flow reduction will be less than that predicted for the pump shaft seizure event (15.3.3). The characteristics of this event will be very similar to those of the Reactor Coolant Pump Shaft Seizure Event (15.3.3), except for the fact that the initial flow reduction will be slower in this event (15.3.4); however the steady-state core flow will be lower because of the potential for higher reverse flow in the affected cold leg.

15.3.4.3 Reactor Protection - Reactor protection for the Reactor Coolant Pump Shaft Break event is provided by the low reactor coolant flow trip and the high pressurizer pressure trip. Additional reactor protection is provided for three-PCP operation by the increased margin to DNB required by the Technical Specification for three-PCP operation.

15.3.4.4 Disoosition and Justification - The consequences of this event are only a concern at rated power and power operating conditions because of the reduced reactor core power and increased thermal margin at other reactor operating conditions. As in Event 15.3.3, the consequences of this event are bounded by the rated power operating condition. The flow reduction is more ,

rapid for the pump rotor seizure event (15.3.3), although the shaft break event will ultimately result in a lower steady-state core flow because of the potential for reverse flow in the affected cold leg. Therefore, it is difficult a priori to bound the .:onsequences of the shaft break event by those of the pump rotor seizure event. However, in the analytical methodology for the pump rotor seizure event, the characteristics of the affected pump are adjusted at the time reverse flow is predicted such that the endpoint core

) flow reached in both the pump rotor seizure event and the shaft break event is analytically identical. Therefore, the consequences of the Reactor Coolant

)

Pump Shaft Break are bounded by the Reactor Coolant Pump Rotor seizure event.

The disposition of events for the Reactor Coolant Pump Shaft Break event is summarized in Table 15.3.4-B.

70 ANF-87-150(NP)

Volume 1 Table 15.3.1-A Available Reattor Protection for the loss of Forced Reactor Coolant Flow Event i

Reactor Operating Conditions Reactor Protection 1 (4 pump operation) Low Reactor Coolant Flow Trip.

Thermal Margin / Low Pressure Trip High Pressurizer Pressure Trip 2.(4,3 pump operation) Same as above.

3-6 (4,3 pump operation Low Reactor Coolant Flow Trip and 3-pump operation for reactor operating {

condition 86) 1 7-8 (RHR operation) No Forced flow by reactor coolant pumps. In the case of an inter-ruption in the residual heat removal operation, the operator has sufficient time to take the appropriate action.

(

I 71 ANF-87-150(NP)

[ Volume 1 Tatle 15.3.1-8 Disposition of Events for the Loss of Forced Reactor Coolant Flow Event Reactor Operating Conditions Discosition 1

Analyze 2 (All Technical Bounded by the above, no analysis Specifications allowed required.

pumpconfigurations) 3-6 (All Technical No analysis required.

Specifications allowed pump configurations) 7-8 No analysis required.

4

I 72 ANF-87-150(NP) l Volume 1 t

l l

Table 15.3.2-A Available Reactor Protection for the l Flow Controller Malfunction Event Reactor Operating Conditions Reactor Protection 1-8 No analysis required, event not credible.

9 t

l 1

J l

i I

l

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[_'

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Volume 1 A

f Table 15.3.2-B. Disposition of Events for the

Flow Controller Malfunction Event i

Reactor Operating Conditions Disoosition 1-8 Not a credible event, no analysis required.

?

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i t

l l

I 1

i

+ - -, , . - - , - - , - - - - -

l J

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i 74 ANF-87-150(NP)

Volume 1 l l

1 Table 15.3.3-A Available Reactor Protection for the i Reactor Coolant Pump Rotor Seizure Event l l

1 Reactor Operating Conditions Reactor Protection 1 Low Reactor Coolant Flow Trip High Pressurizer Pressure Trip 2 Same as above.

3-6 Low Reactor Coolant Flow Trip i

7-8 Not a credible event; no forced flow by reactor coolant pumps. ,

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Table 15.3.3-B Disposition of Events for the Reactor Coolant Pump Rotor Seizure Event Reactor Operating.

Conditions Disoosition 1 Analyze 2 Bounded by the above.

3-6 No analysis required.

7-8 No analysis required.

( .

t i

76 ANF-87-150(NP) l l

Volume 1 l

Table 15.3.4-A Available Reactor Protection for the l'

, Reactor Coolant Pump Shaft Break Event l

l Reactor Operating Conditions Reactor Protection 1 Low Reactor Coolant Flow Trip High Pressurizer Pressure Trip 2 Same as above.

3-6 Low Reactor Coolant Flow Trip 7-8 Not a credible event; no forced flow by reactor coolant pumps.

l i

l I i

l

\

r

r. 77 ANF-87-150(NP) l- Volume 1 f

Table 15.3.4-B Disposition of Events for the Reactor Coolant Pump Shaft -Break Event Reactor Operating Cor.ditions Disoosition 1 Bounded by Event 15.3.3, Reactor Coolant Pump Rotor Seizure Event due to analytical methodology.

2 Bounded by the above.

3-6 No analysis required.

7 No analysis required.

i i

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I Volume'l f

15.4 REAC11VITY AND POWER DISTRIBUTION ANOMALIES 15.4.1 Uncontrolled Control Rod Bank Withdrawal From a Suberitical or

' Low Power Startuo Condition 15.4.1.1 Event Initiator - Initiated by the uncontrolled withdrawal of a control rod bank, this event results in the insertion of positive reactivity and consequently a power excursion. This event could be caused by a malfunction in the reactor control or rod control systems. The consequences of a single bank withdrawal from reactor critical, hot standby, and hot shutdown (subcritical) operating conditions are considered in this event category; the consequences at rated power and power operating initial conditions are considered in Event 15.4.2.

The control rods are wired together into preselected bank configurations.

These circuits prevent the control rods from being withdrawn in other than their respective banks. Power is supplied to the banks in such a way that no more than two banks can be withdrawn at the same time and in their proper withdrawal sequence.

15.4.1.2 Event Descriotion - The neutron flux rises very rapidly in response I to the continuous positive reactivity insertion. The initial rapid rise is terminated by the reactivity feedback effect of the negative Doppler coefficient. The number of reactor coolant pumps in operation can significantly affect the peak transient heat flux.

15.4.1.3 Reactor Protection - The power transient is eventually terminated (as well as the control rod withdrawal) by the reactor protection system on one of the following signals:

(1) Nonsafety grade high rate-of-change of power trip, 10'4%

to 15% power, no credit taken; (2) Variable overpower trip; (3) Thermal margin / low pressure trip;

79 ANF-87-150(NP)

Volume 1 (4) High pressurizer pressure trip; or (5) High rate-of-change of power alarms, which initiate Rod .

Withdrawal Prohibit Action.

Reactor protection for the Uncontreiied Control Rod Bank Withdrawal from a Suberitical or Low Power Startup Condition event is summarized in Table 15.4.1-A.

l l 15.4.1.4 Disoosition and Justification - Combustion Engineering analyzed the uncontrolled rod withdrawal from source range to 10% of full power (2220 MWt) j for the original FSAR, Reference 10, and the results are considered to apply to the present 2530 MWt core for four pump operation, j However, for the reactor critical and particularly the hot shutdown operating condition, these analyses did not consider operation with less than four reactor coolant pumps in operation. Technical Specification 3.1.lb requires that four primary coolant pumps shall be in operation whenever the reactor is operated continually above 5% power. Therefore, the analysis for Palisades will be concerned with power levels less than 57 and less than four reactor coolant pumps in operation. Technical Specification Table 2.3.1 allows three pump operation to provide a limited time for repair / pump restart, to provide for an orderly shutdown, or to provide for the conduct of reactor internals noise monitoring test measurements. As such, an Uncontrolled Control Rod Bank Withdrawal from Subcritical or less than 5% Power with less than four reactor coolant pumps in operation should be classified as a Condition III or Condition IV event with the corresponding criteria applicable.

Additionally, Technical Specification 3.1.lc states in part that startup (above hot standby) with less than four pumps is not permitted. Also, Technical Specification 3.1 ld requires that both steam generators shall be capable of performing their heat transfer function whenever the average temperature of the primary coolant is above 325'F. This eliminates one steam

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Volume 1 l

generator operation from the set of rod withdrawal transients under consideration.

Reference 11, lesson 4, page 14 states that: No more than three primary coolant pumps should be operated et the same time when the primary coolant temperature is less than 400'F. Therefore, uncontrolled control rod bank withdrawal events will be analyzed for the following type initial conditions:

(1) Three reactor coolant pumps operational (2) Subcritical to 10~4% of rated power (3) Average coolant temperature 2 525'F.

The disposition of events for the Uncontrolled Control Rod Bank Withdrawal from a Subcritical or low Power Startup Condition event is summarized in Table 15.4.1-B.

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s 81 ANF-87-150(NP)

Volume 1 15.4.2 Uncontrolled Control Rod Bank Withdrawal at PoWLt-i 15.4.2.1 Event Initiator - This event is initiated by an uncontrolled control rod bank withdrawal'from power operating conditions.

15.4.2.2 Event Descriotion - Positive reactivity is added to the reactor core due to the uncontrolled bank withdrawal resulting in a power transient. The-increase in core power results in an increase in the core heat flux creating a challenge to the DNB margin. The challenge to the DNB margin is further ,

accentuated by the mismatch between the energy removal from the steam .

generators and the power produced in the core. This mismatch in power causes the. primary system temperatures to rise, reducing the DNB margin. ,

15.4.2.3 Reactor Protection - The challenge to the fuel design limits is j terminated by the automatic action of the reactor prc,tection system which terminates the bank withdrawal and inserts negative reactivity to terminate ,

the power transient. The automatic action of the reactor protection sisteni is initiated as the result of one of the following signals: ,

(1) Variable overpower trip; f (2) Thermal margin / low pressure trip; or (3) High pressurizer pressure trip.

Reactor protection for the Uncontrolled Control Rod Bank Withdrawal at Power event is summarized in Table 15.4.2-A. j

\

15.4.2.4 Disoosition and Justification - This event is designed to address the safety challenge posed by an uncontrolled control rod bank withdrawal transient from power conditions. This event addresses all the power operating I conditions and the rated power operating conditions. This event will be j analyzed at several operating power levels, for conditions ranging from BOC to j E0C, for several reactivity insertion rates for Palisades, j j

i l

_ . , . , _ . . . _ _ _ - _ _ _ _ - - - . _ _ - - . . _ _ - . . - - - - _ - - _ . ~ _ _ _ . _

82 ANF-87-150(NP)

Volume 1 The disposition of events for the Uncontrolled Control Rod Bank Withdrawal at Power event is summarized in Table 15.4.2-8.

15.4.3 Control Rod Miscoeration 15.4.3.1 Event Initiator - The control rod misoperation event encompasses a number of transients resulting from different event initiators. The specific l events addressed under this event category include the following:

(1) Dropped control rod or control rod / bank; (2) Dropped part-length control rod;

(3) Malpositioning of the part-length control rod group; (4) Statically misaligned control rod / control rod bank; and

(5) Single control rod withdrawal.

l

. (1) Droceed Control Rod / Bank l

15.4.3.l(1) Event Initiator - The dropped control rod and dropped control bank events are initiated by a de-energized control rod drive mechanism or by a malfunction associated with a control rod bank.

l l 15.4.3.2(1) Event Descriotion - In the dropped control rod event, the reactor power initially drops in response to the insertion of negative reactivity.

However, the local peaking increases due to the local effect on the power distribution. The reactor core will attempt to return tc a new equilibrium at i the original power level as a result of moderator and Doppler reactivity feedback. Because of the increased peaking and the potential return to the initial power level, the dropped control rod event poses a severe challenge to the DNB margin, j 15.4.3.3(1) Reactor Protection If the amount of reactivity is large enough l

l 83 ANF-87-150(NP)

Volume 1 to cause a significant reduction in core power, a reactor trip would be generated by the low pressurizer pressure trip, low steam generator water level trip, or thermal margin / low pressure trip. Reactor protection for the Control Rod Misoperation (Dropped Control Rod / Bank) event is summarized in i Table 15.4.3(1)-A.

15.4.3.4(1) Disoosition and Justification - In general, a bank drop will cause a reactor trip and, as such, poses no challenge to the DNB margin.

Because of the concern over the challenge to the DNB margin, the dropped control rod is analyzed at rated power operating conditions for which the initial DNB margin is at a minimum. The consequences of the dropped control rod from rated power conditions bound the other power operating conditions.

An analysis of the consequences of,this event from rated power conditions will be performed. The disposition of events for the Control Rod Misoperation (Dropped Control Rod / Bank) event is summarized in Table 15.4.3(1)-B.

(2) Drocoer part-lenoth Control Rod 15.4.3.1(2) Event Initiator - The part-length control rods are not connected to any reactor trip circuit and will not drop into the core on a reactor trip or loss of power, but a mechanical failure in a rod brake mechanism could cause an individual rod to drop into the lower region of the core. If the drop is caused by a mechanical failure in the brake mechanism, the rod position lower limit switch may be actuated.

15.4.3.2(2) Event Descriotion - Due to the reduced worth of the part-length control rod, relative to a full length rod, the transient due to the dropped part-length control rod from a fully withdrawn position will not be as severe  !

as in the case of a dropped full length control rod. The local peaking factor j increases and the return to power will be less and subsequently pose less l challenge to the DNB SAFDLs. i

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15.4.3.3(2) Reactor Protection - If the amount of reactivity is large enough to cause a significant reduction in core power, a reactor' trip would be generated by the low pressurizer pressure trip, low steam generator water level trip, or thermal margin / low pressure trip. Reactor protection for the Dropped Part-Length Co:strol Rod event is summarized in Table 15.4.3(2)-A.

15.4.3.4(2) Disoosition and Justification - Due to the reduced worth of the part-length control rod, relative to a full length rod, in conjunction with the fact that during power operation the part-length control rods are maintained in the fully withdrawn position, the consequences of a part-length rod drop will be bounded by the dropped control rod event (15.4.3(1)) which will be analyzed for Palisades. The disposition of events for the Dropped Part-Length Control Rod event is summarized in Table 15.4.3.(2) B.

(3) Naloositionino of the Part-Lenath Control Rod Group Current Technical Specifications do not allow the use of the part-length control rods during power operation. The part-length control rod group is maintained in the fully withdrawn position and is not used; therefore, mispositioning of these rods is not a credible event.

(4) Statically Misalianed Control Rod / Bank 15.4.3.l(4) Event initiator - The statically misaligned control rod / bank i event is defined as a malfunction in the Control Rod Drive mechanism which causes a control rod to be out of alignment with its bank or ene/two (rated power / power operating condition) control group (s) to be in violation of the I

Power Dependant Insertion Limit (PDIL). The reactor is basically at steady state for thi: rated or power operating (77% to rated) condition. For the power operation conu: tion, rod banks 2, 3 and 4 will be of concern since rod bank I will be completely withdrawn. For the rated power condition, only rod bank 4 can be inserted. The shutdown groups are the first to be withdrawn on f

85 ANF-87-150(NP).

Volume 1 startup and they remain withdrawn throughout power operation. Part-length rods are fully withdrawn during normal operation and are only used during Physics testing.

15.4.3.2(4) Event Descriotion - In the case of a statically misaligned control rod (one control rod in a bank) event, a control bank (all other banks fully withdrawn) is inserted but one of the control rods remains in a fully withdrawn condition. This creates a local power distribution disturbance resulting in an increased radial peaking factor. The increased power peaking results in a reduction in the DNB margin. The most severe misalignment occurs at the rated power operating condition, with one bank inserted beyond its PDIi., and one of the control rods in the bank in the fully withdrawn cor.dition. The radial power distribution consequences of the reverse of this situation, one control rod inserted while the bank remains withdrawn, are essentially the same as the dropped control rod event which is to be analyzed.

In the case of a statically misaligned control rod bank (control rod bank 4) at the rated power condition, one control rod bank is inserted beyond the power dependent insertion limit and all other rod banks are withdrawn in addition to one control rod in the inserted bank withdrawn. This is identical to the case for the statically misaligned control rod at the rated power condition. At the power operating condition, the situation of concern will be the power range (35%-65%) in which control rod banks 3 and 4 are both inserted and then control rod bank 4 is inserted beyond the PDil or control rod group 3 is withdrawn in violation of the required control rod bank overlap criteria (40%), or the case where control rod bank 4 is inserted beyond the PDIL and control rod bank 3 does not insert.

15.4.3.3(4) Reactor Protection - Static misalignment of a control rod / bank would be detected by rod / bank position monitors and by rod / bank deviation monitors. For example, there is an upper and lower rod stop for the regulating rods. These rod stops are not necessarily at the same position for

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all groups but are the same for all rods within a regulating group. The Position Indication Primary (PIP) supplies rod position limit signals for the upper and lower rod stops. The Secondary Position Indication (SPI) will supply rod position limit signals independently that will initiate alarms at .

preset rod positions. These two position indication schemes operate simultaneously.

Upper and lower electrical limits of rod travel are detected by limit switches. These limit switches prevent rods from being driven beyond these limits by interrupting power to each rod drive motor as its respective limit switch is actuated. Both primary (PIP) and secondary (SPI) control rod position indicating schemes provide the following regulating control rod information. The positions of all rods are logged hourly on a typewriter. In addition to this, the following information is available at any time: a) the positions of all rods in each regulating group; and b) the position of any particular rod. Reactor protection for the Statically Misaligned Control Rod / Bank is summarized in Table 15.4.3(4)-A.

15.4.3.4(4) Disoosition and Justification - The statically misaligned single control rod event and the statically misaligned control rod bank evont will be analyzed. The disposition of events for the Statically Misaligned Control Rod / Bank event is summarized in Table 15.4.3(4)-B.

(5) Sinole Control Rod Withdrawal 15.4.3.l(5) Event Initiator - In the single control rod withdrawal event a single control rod is withdrawn from the reactor core causing an insertion of positive reactivity which results in a power transient.

15.4.3.2(5) Event Descriotion - The movement of a single control rod out of sequence from the rest of the bank results in a local power distribution disturbance which causes localized increased power peaking. The combination

87 ANF-87-150(NP)

Volume 1 of these two factors results in a challenge to the DNB margin.

15.4.3.3(5) Reactor Protection - Reactor protection for t.1e Single Control Rod Withdrawal event is summarized in Table 15.4.3(5)-A.

15.4.3.4(5) Disoosition and Justification - The concern of this event is the potential DNB challenge. At rated power operating conditions the initial DNB margin is at a minimum. At other power operating conditions the change in peaking factors is a concern. As such, the consequences of this event will be analyzed for both the rated power and the power operating condition. The disposition of events for the Single Control Rod Withdrawal event is summarized in Table 15.4.3(5)-B. }

(6) Core Barrel Failure 15.4.3.l(6) Event Initiator - This event is initiated by the circumferential rupture of the core support barrel.

15.4.3.2(6) Event Description - 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 approxirtatoly one-half 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 and exit nozzles in parallel with the path that goes through tne core. The core sustains a srall reactivity transient induced by the motion of the core relative to the inserted rod bank (s).

15.4.3.3(6) Reactor Protection -

Reactor ':rotecti9n for the Core Barrel Failure event during hot shutdown, ref 64tdor:0 cold thutdown, and refueling operating conditions is provit '-l Sp:.ificttion Shutcown Margin requirements. Fnr the reacto s kot star,ab; operat!ng

~

H D i dR% 4

88 ANF-87-150(NP)

/ Volume 1 conditions, reactor protection is provided by the variable overpower trip and a nonsafety grade high rate-of-change of power trip. For the rated power and power operating conditions, reactor protection is affordea for the variable overpower and thermal margin / low pressure trip. Reactor protection for the Core Barrel Failure event is summarized in Table 15.4.3(6)-A.

15.4.3.4(6) Disoosition and Justification - For the hot shutdown, refueling shutdown, cold shutdown, and refueling operating conditions, the reactivity transient induced by a failure of the core barrel will be less than the shutdown margin required by Technical Specifi .ations. As such, the transient will have no significant consequences for these reactor operating conditions.

The rated power operating condition bounds the power operating condition, reactor critical, and hot standby operating conditions due to the initial thermal margin which is minimized at rated power conditions and the Technical Specification Power Dependent Insertion limits (Figure 3-6, p 3-62, Palisades Technical Specifications). The analysis presented in Reference 4 assumed a high estimate of the total rod worth in the case of rated power conditions (2%), whereas the table of reactivity assignments showed the maximum possible rod worth in the core at full power to be 0.7%. For the current reference cycle, the maximum rod worth that would be inserted at rated power is 0.19%.(3) Thus, even though steam generator tube plugging has resulted in slightly less core flow and higher coolant inlet temperatures, these effects are easily compensated for by use of the expected rod worth inserted in the core. As such, the event is bounded by the analysis presented in Reference 4.

The disposition of events for the Core Barrel Failure event is summarized in Tabl a 15.4.3(6) B.

15.4.4 Startuo of an Inactive looo 15.4.4.1 Event Initiator - This event is initiated by the startup of in inactive reactor coolant pump.

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Volume 1 1 15.4.4.2 Event Descriotion - Each primary coolant loop is equipped with two single suction centrifugal pumps one per cold leg which are located between the steam generator outlet and the reactor vessel inlet nozzles. A f

nonreversing mechanism is provided to prevent reverse rotation of the pump rotor. This feature also limits backflow through the pump under non-operating conditions. If a reactor coolant pump is inactive, then the cold leg temperature will be slightly cooler than that in the operating cold legs.

Note: there is no backflow in the hot leg associated with the side of the plant that has the inactive reactor coolant pump. The inadvertent actuation of an inactive pump would therefore lead to a decrease in moderator temperature and, with a negative moderator coefficient, an increase in core reactivity with a resultant increase in core power level. }

l 15.4.4.3 Reactor Protection -

The start of an idle pump in the Primary Coolant System without first reducing the reactor power to a low value is in violation of administrative procedures and, in addition, plant operation with less than all four primary coolant pumps is not permitted by plant Technical Specifications except for very short periods of time and at reduced power levels (Tech. Spec. Table 2.3.1). The permissible part loop operating l conditions are given in the Technical Specifications (Table 2.3.1). Reactor protection for the Startup of an Inactive Loop event is summarized in Table ,

15.4.4-A.

15.4.4.4 Discosition and Justification -

For operation with one pump inoperative, the low flow trip setpoint and the variable overpower trip setpoint are simultaneously changed to the allowable values for the selected pump condition. The reduced variable overpower and low flow trip setpoints are effected by a manual trip switching arrangement which provides positive means of assuring that the more restrictive settings are used. Under this arrangement, the variable overpower trip will terminate any transient  ;

resulting from the inadvertent activation of an idle pump before any significant decrease in thermal margin. The disposition of events for the

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Volume 1 Startup of an Inactive Loop event is summarized in Table 15.4.4-8.

15.4.5 Flow Controller Malfuncticn Palisades does not have any flow control devices on the primary reactor coolant loops so this event is not credible and does not need to be analyzed.

15.4.6 SVCS Malfunction That Results in a Decrease in the Boron Concentration in the Reactor Coolant 15.4.6.1 Event Initiator - A dilution of the primary system boron concentration can occur as a result of adding primary grade water into the reactor coolant system via the Chemical Volume and Control System (CVCS) or by the accidental transfer of the contents of the iodine removal systems to the primary coolant system (cold shutdown and refueling operation conditions).

The iodine removal system contains two unborated water storage tanks, the Hydrazine Storage Tank (270 17 gal.) and the Sodium Hydroxide Storage Tank (maximum 6000 gals, Technical Specification minimum 3900 gals).

15.4.6.2 Event Descriotion - Dilution of the primary coolant Boron concentration results in the insertion of positive reactivity. For the following reactor operation conditions, reactor critical, hot standby, and hot shutdown the event results in a slowly evolving power excursion. In the case of a boron dilution at rated power and power operation reactor operating conditions, the consequences are very similar to the consequences of a slow control rod withdrawal.

15.4.6.3 Reactor Protection - Reactor protection for the boron dilution event during refueling shutdown, cold shutdown, and refueling operating conditions is provided by Technical Specification Shuto.own Margin requirements, Administrative procedures, and sufficient time for the operator to take the appropriate action in the unlikely event that a boron dilution should occur.

For the hot standby and hot shutdown operati 1 conditions, reactor protection

- ___ _ U

91 ANF-87-150(NP)

Volume 1 is basically provided by the Technical Specifications and the operator response time. Reactor protection for the reactor critical, power operation, and rated power operating conditions is provided by various trips and operator f response time. Reactor protection for the CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant event is summarized in Table 15.4.6-A.

15.4.6.4 Discosition and Justification - For the refueling shutdown, cold shutdown, and refueling operating conditions, the event will be analyzed. For the following reactor operation conditions: reactor critical, hot standby, and f hot shutdown, the event is bounded by the consequences of the reactor critical operating condition because the reactor is critical at this point and does not have subcritical margin to provide the operator additional response time. A spectrum of control rod withdrawal reactivity addition rates is considered for l Event 15.4.1 which considers these initial conditions. A calculation of the l dilution reactivity rate has been made (Ref. 4) and the pre trip sequence of

! events can be shown to be bounded by the spectrum of cases considered for Event 15.4.1. Following reactor trip, the operator would have to terminate ,

l the dilution prior to loss of Tech. Spec, shutdown margin requirements. This  ;

l response time will be calculated for the reference cycle.

1 i

For a boron dilution at rated power and power operating reactor conditions, the pre-trip sequence of events is very similar to the conse- quences of 6 slow control rod withdrawal and can be bounded by the consequences of a control rod withdrawal event as analyzed for event 15.4.2. A spectrum of control rod withdrawal reactivity addition rates is considered for Event l 15.4.2, so the range of reactivity addition rates will be established to encompass the predicted reactivity addition rate for a boron dilution event at rated power and power operating conditions. Following reactor trip, the l operator would have to terminate the dilution prior to reaching Tech. Spec.

shutdown margin requirements and/or losing minimum shut.down margin. These response times will be calculated. The disposition of events for the CVCS 1

92 ANF-87-150(NP)

Volume 1 l

Malfunction that Results in a Decrease in the Boron Concentration in the Reector Coolant event is summarized in Table 15.4.6-B.

15.4.7 Inadvertent Loadina and Ooeration of a Fuel Assemb1v in an Imoroner Position l

15.4.7.1 Event Initiator - The inadvertent loading of a fuel assembly in an improper core location could result in a perturbation of the local power distribution which could adversely affect the DNB margin, 15.4.7.2 Event Descriotion - To reduce the probability of core loading errors, each fuel assembly is marked with an identification number and loaded in accordance with a core loading diagram. In addition, the power distribution distortion due to the misloading of a pair of fuel assemblies would result in increased peaking and would be detected by the in-core flux monitors. Finally, determining the measured peaking factors after each fuel loading prior to exceeding 50% of rated power provides additional assurance -

that the core is properly loaded (Technical Specification 3.23.2 Radial

Peaking Factors, pp. 3-111).

15.4.7.3 Reactor Protection - Reactor protectio.i for the misloading of a fuel assembly event depends upon administrative procedures or measurements required by Technical Specifications which preclude occurrence or provide early l

detection of this event prior to reaching rated power operating conditions.

l Reactor protection for the Inadvertent Loading and Operation of a Fuel l Assembly in an Improper Position event is summarized in Table 15.4.7-A.

l 15.4.7.4 Disoosition and Justification The disposition of events for the Inadvertent loading and Operation of a Fuel Assembly in an Improper Position event is summarized in Table 15.4.7-B. The event will not be analyzed for the rated power operating condition and power operating condition due to administrative procedures or measurements required by Technical Specifications l

t (

93 ANF-87-150(NP)

Volume 1 j i

1 which preclude occurrence or provide early detection prior to reaching rated I power operating conditions.

15.4.8 Soectrum of Control Rod E.iection Accidents -

I 15.4.8.1 Event Initiator - This accident is initiated by a failure in the control rod drive pressure housing which could result in the rapid ejection of a control rod. I 15.4.8.2 Event Descriotion - Ejection of the control rod from the reactor core results in a rapid loss of negative reactivity causing a nuclear power  ;

transient. In addition to the power transient, the ejected rod results in a j highly perturbed power distribution which, coupled with the power transient, l could possibly lead to localized fuel damage. Also, the rapid nuclear power l

excursion can result in a significant short term heatup of the coolant with a resultant reactor coolant system pressure increase, although on the long term the reactor coolant system will depressurize due to the break in the reactor coolant pressure boundary, j I

15.4.8.3 Reactor Protection - Reactor protection for the Spectrum of Control l

Rod Ejection Accidents is summarized in Table 15.4.8-A. Doppler feedback j inherent in the fuel also limits the nuclear power excursion. l

(

15.4.8.4 Disoosition and Justification The governing factors of this event  !

are the ejected rod worth, post-ejection power peaking factor, and Doppler l feedback. The fuel energy content is maximized by starting from rated power initial conditions, so the consequences of this event are bounding for power operating initial conditions. However, because of the complex interaction of the ejected rod worth, and ejected peaking factor (which are maximized at hot critical operating conditions), and Doppler feedback effects, it is difficult to a priori bound the consequences of the event for either rated power or hot critical operating conditions. Therefore, the consequences of this event are

94 ANF-87-150(NP)

Volume 1 r.

l analyzed for both rated power and hot critical operating conditions.

The disposition of events for the Spectrum of Control Rod Ejection Accidents is summarized in Table 15.4.8-B.

15.4.9 Soectrum of Rod Droo Accidents (BWR)

The Palisades plant is not a Boiling Water Reactor (BWR) and as such this event is not applicable.

l 95 ANF 87-150(NP)

Volume 1 t

Table 15.4.1-A Available Reactor Protection for the Uncontrolled Control Rod Bank Withdrawal from a Subcritical l or low Power Startup Condition Event i

l Reactor Operating Conditions E.us. tor Protect 193 e

1, 2 Not considered in this section 3 Thermal Margin / Low Pressure Trip i

Variable Overpower Trip h Rate-of Change NonsafetyGradeHig%to15%

of Power Trip, 10' Power, No credit taken High Pressurizer Pressure Trip 4 and 5 High Rate-of Change of Power <

Alarms, which initiate Rod Withdrawal Prohibit Action High Pressurizer Pressure Trip 6,7,8 No significant con' sequences for these reactor operating conditions I

i i

I i

t Id

- 96 ANF-87-150(NP)

Volume 1

(

Table 15.4.1-B Disposition of Ev9nts for the Uncontrolled Control Rod Bank Withdrawal -from a Suberitical or low Power Startup Condition Event Reactor Operating Conditions Disposition 1

Not considered in the section; no analysis required 2 Bounded; FSAR; Reference 10 3,4,5 Analyze, with 3-pump operation 6,7,8 No analysis . required 4

O 97 ANF-87-150(NP)

Volume 1 Table ib.4.2-A Available Reactor Protection for the Uncontrolled Control Rod Bank Withdrawal at Power Event Reactor Operating Conditions Reactor Protection 1, 2 Variable Overpower Trip Thermal Margin / Low Pressure Trip High Pressurizer Pressure Trip 3-8 Not considered in this section l

l l

l l

l

,ev- y m- _ , . - - - - _ , - , , . . . - . - . - - ~ , . - -- -, ,,.-%,y,...m-,,.-, .-y.,.-~,. , ,-.,.c.,, - . . . - - - , , , . + _ - . . . .,,m . ,, , -

- 98 ANF-87-150(NP)

Volume 1

(

Table 15.4.2-B Disposition of Events for the Uncontrolled Contro!

Rod Bank Withdrawal at Power Event Reactor Operating Conditions Disoosition i

1 Analyze 2 Analyze 3-8 No analysis required; not con-sid9 red in this section

99 ANF-87-150(NP)

Volume 1 Table 15.4.3(1)-A Available Reactor Protection for the Dropped Control Rod / Bank Event Reactor Operating Conditions Reactor Protection 1 Low Pressurizer Pressure Trip ,

Thermal Margin / Low Pressure Trip Low Steam Generator Water level Trip Safety Injection Actuation Signal 2 Same as above 3-8 No significant consequences for these reactor operating -

conditions

100 .ANF-87-150(NP)

Volume 1 Table 15.4.3(1)-B Disposition of Events for the Dropped Control _ Rod / Bank Event Reactor Operating Conditions Disoosition l

1 Analyze 2 Bounded by the abovc; no analysis required 3-8 No analysis required l

i l

l

101 ANF-87-150(NP)

Volume 1 Table 15.4.3(2)-A Available Reactor Protection for the Dropped Part-Length Control Rod Event Reactor Operating Conditions Reactor Protection 1 Thermal Margin / Low Pressure Trip Low Pressurizer Pressure Trip Low Steam Generator Water Level Safety Injection Actuation Signal 2 Same as above 3-8 No significant consequences for these reactor operating conditions s-, ...,___-.,...,,_-.y. .....,p.,,- _ __ ,, , ,,_ . - , _,.-,,,,,,_,_,y , , . , , , , . _ , _ , . _ , , . , _ ,-,_y_,.. ,___,,r,m

102 ANF-87-150(NP)

Volume 1 Table 15.4.3(2)-B Disposition of Events for the Dropped Part-Length Control Rod Event Reactor Operating Conditions Disoosition 1 Bounded by the Dror. ped Control Rod Assembly Event, 15.4.3(1) 2 Same as above; no analysis required 3-8 No analysis required -

i 1

l l

i i

1

103 ANF-87-150(NP)

Volume 1 Table 15.4.3(3)-A Available Reactor Protection for the Ma1 positioning of the Part-Length Control Rod Group Event Reactor Operating . .

Conditions Reactor Protection 1-8 Not a credible event O

t

. . , , , .-- - - - - - - . . . , . , - - -- - - - - - . . , - - ,.,e . - - - . , - - --.--aww, --,,

104 ANF-87-150(NP)

Volume 1 Table 15.4.3(3)-B Disposition of Events for the Ma1 positioning of the Part-Length Control Rod Group Event h' Reactor Operating Conditions Disuosition 1-8 Not a credible event; no analysis required l

1 l

i l

l

105 ANF-87-150(NP)

Volume 1 Table 15.4.3(4)-A Available Reactor-Protection for the Statically Misaligned Control Rod / Bank Event Peactor Operating i Conditions Reactor Protection 1 Control Rod and Bank Deviation Alarms 2 Same as Above 3-8 No significant consequences for

. these reactor operating conditions l

l l

1 l

l I

i 1

l

. . _ . . . _ _ _ . . _ _ . - , - - . _ , _ _ _ _ , _ - . - . - , , - . . _ _ . _ , . , _ , - - . - _ ~ . . . .

106 ANF-87-150(NP)

Volume 1 Table 15.4.3(4)-B Disposition of Events for the Statically Misaligned Control Rod / Bank Event h

Reactor Operating Conditions Disposition 1 Analyze 2 Analyze 3-8 No analysis required f

L h

l I

• --w ,. .- ,,,,__,y , ,

i 107 ANF-87-150(NP)

Volume 1 l

Table 15.4.3(5)-A Available Reactor Protection for the Single Control Rod Withdrawal Event l I

J Reactor Operating ,

Conditions Reactor Protection '

1 Variable Overpower Trip Thermal Margin / Low Pressure Trip High Pressurizer Pressure Trip 2, 3 Variable Overpower Trip Nonsafety Grade of Power Trip, 10' Hig% to 15% Power,h Rate No credit taken High Pressurizer Pressure Trip 4, 5 High Rate-of-Change of Power j Alarms which initiate Rod Withdrawal Prohibit Action High Pressurizer Pressure Trip l

6,7,8 No significant consequences for .

these reactor operating conditions l l

1 I

I j

l

l 108 .ANF-87-150(NP)

. Volume l' f Table 15.4.3(5)-B - Disposition of Events for the Single Control Rod Withdrawal Event Reactor Operating Conditions Disoosition 1 Analyze 2 Analyze 3,4,5 Analyze for 3-pump operation 6,7,8 No analysis required

\

9

)

t

)

109 ANF-87-150(NP)'

Volume 1 Table 15.4.3(6)-A Available Reactor Protection for the Core Barrel Failure Event Reactor Operating f Conditions Reactor Protection 1 Variable Overpower Trip Thermal Margin / Low Pressure Trip 2 Same as above

{

3-4 Variable Overpower Trip NonsafetyGradeHig%to15%

of Power Trip, 10' Power,h Rate-of-t No credit taken High Rate-of-Change of Power Alarms 5-8 No significant consequences for these i reactor operating conditions l

1

1 I

{. 110 ANF-87-150(NP)

Volume 1 Table 15.4.3(6)-B Disposition of Events for the Core Barrel Failure' Event

\

Reactor Operating Conditions Disoosition i

1 Bounded by analysis presented in Reference 4 f-2 Bounded by the above k Bounded by the above 3-4 5-8 No analysis required b

\

i

}

}

1

111 ANF-87-150(NP)

Volume 1 ,

Table 15.4. , A Available Reactor Protection for the Startup of an Inactive Loop Event J

Reactor Operating Conditions Reactor Protection 1-4 Low Reactor Coolant Flow Trip Variable Overpower Trip Thermal Margin / Low Pressure Trip 5-8 No significant consequences for j these reactor operating conditions 1 f

. 4 1

1 i

--,-.e-- -

--.,w.., ,i--m-r-,---- . - .v,-r - - -y- r g w -

112 ANF-87-150(NP)

Volume 1 p

' Table 15.4.4-B Disposition of Events for the Startup of an Inactive loop Event i

Reactor Operating Conditions Disoosition b l-4 Bounded by' Reference 4, Section 14.S; no analysis required 5-8 No analysis required k'

f v

h f-p

)

i 113 ANF-87-150(NP)

Volume 1 Table 15.4.5-A Available Reactor Protection for the Flow Controller Malfunction Event Reactor Operating Conditions Reactor Protection 1-8 Event is not credible

(

b Table 15.4.5-8 Disposition of Events for the Flow j Controller Malfunction Event 1 Reactor Operating Conditions Discosition 1-8 Event is not credible; no analysis /

required j 1

(

)

114 ANF-87-150(NP)

L Volume 1 g..

l Table 15.4.6-A Available Reactor Protection for the CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant Event 1

Reactor Operating Conditions Reactor Protection

}

1, 2, 3 High Rate-of-Change of Power Trip f Variable Overpower Trip Thermal Margin / Low Pressure Trip High Pressurizer Pressure Trip 4, 5 Nonsafety Grade High Rate-of-Change of Power Trip, 10" Y,to 157. Power, No credit taken Variable Overpower Trip Thermal Margin / Low Pressure Trip j

High Pressurizer Pressure Trip h

6,7,8 Technical Specification Shutdown Margin Requirements Administrative Procedures Operator Response Time 1

I 115 ANF-87-150(NP)

Volume 1 ,

t Table 15.4.6-B Disposition of Events for the CVCS Malfunction that 'W

-, Results in a Decrease in the Boron Concentration in the Reactor Coolant Event ,

Reactor Operating Conditions Disoosition 1, 2 Analyze v

I 3,4,5 Analyze 6,7,8 Analyze .

I I

y I

i 0

h

k L 116 ANF-87-150(NP)

Volume 1 l'

i- Table 15.4.7-A Available Reactor Protection for the Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position Event R_ actor Operating Conditions Reactor Protection 1, 2 Thermal Margin / Low Pressure Trip i Administrative Procedures Measurements required by Technical Specifications 3-8 No significant consequences for these reactor operating conditions

[ a l

4 I

I i

117 ANF-87-150(NP)

Volume 1

'i Table 15.4.7-B Disposition of Events for the Inadvertent Loading (

and Operation of a Fuel Assembly in an Improper Position Event 1 i

i Reactor Operating Conditions p,Lsoositigf1 1, 2 Bounded, Technical Specification Measurement Requirement and (

Administrative Procedures 3-8 No analysis required f j

(

1

(

l

{

1 l

l 4

118 ANF-87-150(NP)

[' Volume 1 Table 15.4.8-A Available Reactor Protection for the Spectrum of

}

Control Rod Ejection Accidents Reactor Operating Conditions Reactor Protection 1, 2 Variable Overpower Trip

{

Thermal Margin / Low Pressure Trip h Rate-of-Change NonsafetyGradeHig%to15%

of Power Trip, 10' Power, No credit taken 3, 4 h Rate-of-Change NonsafetyGradeHig%to15%

of Power Trip, 10- Power, No credit taken Long Term, Low Pressurizer Pressure Trip Long Term, Safety Injection Actuation Signal

{

5 No reactor protection required; i

ejected rod worth less than the Technical Specification minimum shutdown margin. No significant consequence for this operating condition.

0 No significant consequences for 6-8 these operating conditions. Not a credible event since the reactor coolant system is not pressurized.

.=m i.i i .o . . . _ _ , _ -

i I

l 119 ANF-87-150(NP)- i 1

Volume 1

)

Table 15.4.8-8 Disposition of Events for the Spectrum of -

Control Rod Ejection Accidents l l

1 Reactor Operating i Conditions Disoosition 1 Analyze I Bounded by reactor operating 2

condition #1, rated power f 3 Analyze 4 Bounded by reactor operating condition #3, Reactor Critical 5-8 No analysis required i

W l

Note: Analytical methodology used in energy deposition calculation is not affected by tne number of reactor coolant pumps assumed to be operational.

l I

9

i 120 ANF-87-150(NP)

{'- Volume 1 Table 15.4.9-A Available Reactor Protection for the Spectrum of Rod Drop Accidents (BWR)

Reactor Operating 4 Conditions Reactor Protection 1-8 Event is not applicable.

I 7

Table 15.4.9-8 Disposition of Events for the Spectrum of Rod Drop Accidents (BWR) b l

l Reactor Operating Conditions Disoosition l

1-8 Event is not applicable; no analysis required.

l l

[

I I

J 121 ANF-87-150(NP)

Volume 1 15.5 INCREASES IN REACTOR COOLANT SYSTEM INVENTORY f

15.5.1 Inadvertent Ooeration of the ECCS That Increases Reactor Coolant Inventory 15.5.1.1 Event Initiator - Inadvertent actuation of the ECCS system can result from an operator error or spurious electrical actuation signal.

15.5.1.2 Event Descriotion - Upon actuation of the ECC3, the charging pump suction is taken from the Safety Injection and Refueling Water Tank (SIRWT).

This results in the injection of highly concentrated borated water into the primary system. On an ECCS actuation signal, the safety injection pumps will also start; however, because the shutoff discharge pressure is approximately 1200 psig, these pumps 'will not deliver flow at power operating conditions.

Likewise, the passive ECCS accumulators will not deliver flow at power operating conditions.

When the borated water from the charging pumps reaches the reactor core, the reactor power will decrease. If the steam generator load is maintained, the primary system will begin to cool off as the power decreases. The primary system fluid will contract causing an outsurge from the pressurizer and a reduction in the primary system pressure. Boron dilution which might result from charging pump actuation is considered in Event 15.4.6.

I 15.5.1.3 Reactor Protection - Reactor protection is provided for by the high h pressurizer pressure, thermal margin / low pressure, and manual trips. Reactor protection for the Inadvertent Operation of the ECCS that Increases Reactor Coolant Inventory event is summarized in Table 15.5.1-A.

The Primary Coolant System Overpressurization Subsystem is designed to provide automatic pressure relief of the Primary Coolant System whenever the conditions of low temperature (1260'F) and high pressure (375 psia increasing) exist concurrently.

I i

122 ANF-87-150(NP) /

Volume 1

{

(

i Another potential scenario, but one which terminates the transient immediately, would be for a reactor trip signal to be genc.riited by the signal which actuated the safety injection signal. As a result of the reactor trip <

the plant would be brought to hot standby or cold shutdown c nditions. ,

15.5.1.4 Disoosition and Justification - The potential overpressurization I

consequences of this event will be mitigated by the primary safety valves. In addition to this, the relief capacity of one PORV is sufficient to mitigate the worst case overpressurization event the primary coolant system would be  !

(

subjected to when in a solid water condition. The volumetric saturated steam discharge capacity of the three safety valves is approximately 8.35 cfs. Even

{

though the volumetric discharge of water could be expected to be significantly <

1ess than this, it is much greater than the volumetric addition of liquid by i the three charging pumps at 2500 psia which is approximately 0.30 cfs.

Therefore, there is sufficient discharge capacity to prevent the primary l system from being pressurized to a pressure greater than the primary safety valve setpoint.

I I

This event does not pose a challenge to the fuel design limits and the primary l system pressure will not exceed 110% of the design valuo.

l Therefore, the potential consequences of the inadvertent operation of the ECCS that increases the RCS inventory are similar to or bounded by the results of j l other events (15.4.6 Boron Dilution and 15.2.1 Loss of External Load) which are to be analyzed. The disposition of events for the Inauertent Operation

{

of the ECCS that Increases the Reactor Coolant Inventory event is summarized in Table 15.5.1-B. j i

1 1

i 123 ANF-87-150(NP)

' Volume 1 15.5.2 CVCS Malfunction That Increases Reactor Coolant Inventory 15.5.2.1 Event Initiator - Due to operator error or an erroneous electrical signal, the charging pumps could provide fall makeup flow to the primary system with the letdown system inoperative.

15.5.2.2 Event Descriotion - This could result in an increase in the primary system inventory causing an increase in the primary system pressure and possib'ly, if unborated water is added, a dilution of the primary system boron concentration.

15.5.2.3 Reactor Protection - Reactor protection is provided by the variable overpower trip, thermal margin / low pressure trip, high pressurizer pressure trip, nonsafety grade high rate of-change of power trip (between 10'4,7and 15'/.

power), and manual trips. The potential overpressurization consequences of this event will be mitigated by the pri:tary safety valves. In addition to this, the relief capacity of one PORV is sufficient to mitigate the worst case overpressurization event the primary coolant system would be st.bjected to when j in a solid water condition. Finally, the Primary Coolant System Overpressurization Subsystem is design to provide automatic pressure relief of 4 the Primary Coolant System whenever the conditions of low temperature (1260*F) and high pressure (375 psia increasing) exist concurrently.

Reactor protection for the CVCS Halfunction that Increases the Reactor Coolast Inventory event is summarized in Table 15.5.2-A.

15.5.2.4 Discosi3 ion and Justification - The potential consequences of cverfilling the primary system are discussed above. The potential consequences of diluting the primary system boron concentration are addressed in Event 15.4.6, and the results of that analysis are applicable for this event. The disposition of events for the CVCS Malfunction that Increases the Reactor Coolant Inventory event is summarized in Table 15.5.2-B.

_ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - . _ _ _ _ _ _________________-_____A

124 ANF-87-150(NP)

Volume 1 l

Table 15.5.1-A Available Reactor Protection for the Intivertent .

Operation of the ECCS that Increases Reactor Coolant i

-Inventory Event-l l

Reactor Operating  :

Conditions Reactor Protection 1 1-5 Variable Overpower Trip Thermal Margin / Low Pressure Trip .

High Pressurizer Pressure Trip NonsafetyGradeHigh,$ ate-of-Change ,

of Power, between 10 % and 15% power, No credit taken 6-8 No significant consequences for these reactor operating conditions High pressure safety injectico pump shutoff head -1200 psia. Reactor operator has sufficient time to detect and terminate the increase in Reactor Coolant I:.ventory. ,

Primary Coolant System Overpres- '

surization Subsystem is available.

i t

1 1

L 125 ANF-87-150(NP) j' Volume 1 Table 15.5.1-B Disposition of Events for the Inadvertent Operation of the ECCS that Increases Reactor Coolant Inventory Event i

Reactor Operating

' Conditions Disoosition 1-5 No analysis required; event does not pose a challenge to fuel design limits and the primary pressure will

,! not exceed 110% of the design value.

Bounded 15.4.6 (Boron Dilution) and 15.2.1 (Loss of External Load)

J, .

6-8 No analysis required i

I-1

_J

126 ANF-87-150(NP)

Volume 1 Table 15.5.2-A Available Reactor Protection for the CVCS Malfunction that "

Increases Reactor Coolant Inventory Event Reactor Operating Conditions Reactor Protection 1-5 Variable Overpower Trip Thermal Margin / Low Pressure Trip High Pressurizer Pressure Trip Nonsafety Grade High Rate 4of-Change ,

of Power Trip between 10' % and 15%

• power, No credit taken .

Manual Trips ,

6-8 No significant consequences for these reactor operating conditions Primary Coolant System overpressuri-zation Subsystem is available.

Reactor operator has sufficient time to detect and terminate the increase in Reactor Coolant Inventory i

l l

l

127 ANF-87-150(NP) f Volume 1 Table 15.5.2-8 Disposition of Events for the CVCS Malfunction that Increases Reactor Coolant Inventory Event h

Reactor Operating Conditions Disoosition 1-5 Overpressurization potential bounded by 15.2.1 (Loss of External Load). Dilution potential bounded by Event 15.4.6, CVCS Malfunction that >

Results in a Decrease in the Boron Concentration in the Reactor Coolant.

No analysis required; event does not pose a challenge to fuel design limits and the primary pressure will not exceed 110% of the design value.

6-8 No analysis required l

a b

i 6

k 128 ANF-87-150(NP)

Volume 1 15.6 DECREASES IN REACTOR COOLANT INVENTORY 15.6.1 Inadvertent Ooenina of a PWR Pressurizer Pressure Relief Valve 15.6.1.1 Event Initiator - The event is postulated to occur as a result of the inadvertent opening of PWR pressurizer pressure relief or safety valve due to an electrical or mechanical failure.

15.6.1.2 Event Descriotion - Two power operated relief valves are located in parallel pipes connected to the pressurizer 4" relief nozzle. In the event of the inadvertent opening of a power operated relief valve, a motor actuated isolation (block) valve upstream of each power operated relief valve provides isolation capabilities, it should be noted that the isolation (block) valves are closed during power operations (pressure 400 psi) so that no discharge to atmosphere would occur if the power operated relief valves were inadvertently opened. If it is determined that a safety valve has stuck open following a transient, then the emergency procedures for a loss of coolant accident are initiated.

15.6.1.3 Reactor Protection - The thermal margin / low pressure trip and the low pressurizer pressure trip provides initial protection against loss of thermal margin and possible fuel damage. Reactor protection for the Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve event is summarized in Table 15.6.1-A.

i 15.6.1.4 Disoosition and Justification - The overall long term consequences of a stuck open pressurizer safety valve, following a transient, would be l bounded by the small break loss of coolant accident analysis. The disposition of events for the Inadvertent Opening of a PWP. Pressurizer Pressure Relief Valve event is summarized in Table 15.6.1-8.

I

R 129 ANF-87-150(NP) ,

Volume 1 l 15.6.2 Radioloaical Conseauences of the Failure of Small Lines Carrvina Primary Coolant Outside of Containment 15.6.2.1 Event Initiator - The event is postulated to occur as a result of the failure outside the containment of a small line carrying primary coolant. '

1 15.6.2.2 Event Descriotion - The rupture of a small line will result in the depletion of the primary coolant inventory. Charging and HPSI flow will provide sufficient makeup capacity to avoid core unccvery. As such, no core f

damage should occur.

15.6.2.3 Reactcr Protection - Reactor protection is provided by the low ,

- pressurizer pressure and thermal margin / low pressure trips, and a Safety Injection Actuation Signal. Offsite dose consequences are limited by the containment isolation system.

Reactor protection for the Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside of Containment event is summarized in Table 15.6.2-A.

15.6.2.4 Disoosition and Justification - The transient response and the offsite dose consequences of a failure of a small line carrying primary coolant outside of containment are bounded by the small break loss of coolant analyses and, as such, will not be analyzed. The disposition of event: for the Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside of Containment event is summarized in Table 15.6.2-8. i 15.6.3 Radioloaical Consecuences of Steam Generator Tube Failure l 15.6.3.1 Event Initiator - This event is initiated by the complete severance l

of a single generator tube.

15. 6. 3 . ?. Event Description -

Experience with nuclear steam generators

130 ANF-87-150(NP)

Volume 1 indicates that the probability of complete severance of a tube is small. The more probable modes of failure are those involving the occurrence of pinholes or small cracks in the tubes, and of cracks in the seal welds- between the f_

tubes and tube sheet.

i A' leaking steam generator tube would allow transport of primary coolant into the main steam system. Radioactivity contained in the primary coolant would mix with chell side water in the affected steam generator. Some of this radioactivity would be transported by steam to the turbine and then to the condenser. Noncondensible radioactive materials would then be passed to atmosphere through the condenser air ejector discharge via the Plant stack.

The radioactive products would be sensed by the condenser a.i r ejector radiation monitor or the stack radiation monitor. These monitors have audible alarms that will be annunciated in the control room to alert th'e operator to abnormal activity levels so that corrective action could be taken.

The behavior of the systems will vary depending upon the size of the steam generator tube failure. For small leaks the chemical and volume control l charging pumps will be able to maintain the necessary primary coolant inventory and an automatic reactor trip will not occur. The gaseous fission products will be released from the main steam system at the air ejector i

discharge and will be discharged via the Plant stack. Nonvolatile fission products will tend to concentrate in the water of the steam generators, i

For leaks larger than the capacity of the charging pumps, the pressurizer

( water level and pressure will decrease and a reactor trip will occur. Upori reactor trip, the turbine will trip and the steam system atmospheric dump valves and the turbine bypass valve will open. In this case it is possible

! that in addition to the noble fission gases a substantial amount of the radioiodines contained in the secondary system may also be released through f

! the steam dump valves.

l l

I

131 ANF-87-150(NP)

Volume 1 The amount of radioactivity released increases with break size. For this analysis, a double-ended break of one tube was assumed. The selection of one double-ended break as an upper limit is conservatively based upon the experience obtained with other steam generators. No double-ended failure has ever occurred in such units.

15.6.3.3 Reactor Protection - The leak rate through the double-ended rupture of one tube is greater than the maximum flow available from the charging pumps; therefore, the Primary Coolant system pressure will decrease and a low pressurizer pressure trip or thermal margin / low-piessure trip will occur. The thermal margin trip has a low-pressure floor, set at 1,750 psia, below which ,

trip will always occur. Following the reactor trip the Primary Coolant System is cooled down by exhausting steam through the atmospheric steam dump valves and turbine bypass valve. The radioactivity exhausted through the steam dump valves passes directly to atmosphere. The radioactivity exhausted through the bypass valve flows to the condenser where the gaseous products remaining are vented to the atmosphere through the condenser air ejector and Plant stack.

The atmospheric steam dump valves close when 'the a';erage primary coolant temperature drops below 535'F. Once the dump valves are closed, operation of the bypass valve will allow steam to pass to the main condenser. (If the operator does not manually open the bypass valve, the steam bypass pressure controller will function to maintain pressure at 900 psia.) The Plant can continue to be cooled using the main condenser until the primary coolant temperature is about 400*F. To continue cooldown at the maximum cooldown rate ,

the atmospheric dump valves must also be manually opened. When the primary coolant temperature is below 300*F, the operator can place the shutdown heat removal syttem into operation and isolate both steam generators to terminate the incident.

Reactor protection for the Radiological Consequences of Steam Generator Tube

132 ANF-87-150(NP)

Volume 1 Failure event is summarized in Table 15.6.3-A.

15.6.3.4 Disoosition and Justification - The consequences of a steam generator tube rupture incident are maximized at rated power operation due to the decay heat which must be removed by the intact steam generator in order to l

bring the primary and secondary systems into pressure equilibrium terminating the primary to secondary leak. The results of a postulated steam generator l

tube ' rupture are presented in the Updated Palisades FSAR, Reference 4, with and without a loss of Offsite Power. The loss of offsite power case assumed that offsite power is lost at the time of reactor trip. In conjunction with the existing low primary pressure and rated power operation, the loss of forced flow at reactor trip time maximizes the challenge to thermal margin limits. Also, the analyses presented in the Updated FSAR assumed an initial operating power of 2650 MWt, which is higher than the current rated power operating level of 2530 MWt. As such the consequences of a postulated steam generator tube rupture are bounded by the results presented in the Palisades Updated FSAR.

l The disposition of events for the Radiological Consequences of Steam Generator l Tube Failure event is summar' J in Table 15.6.3-B.

15.6.4 Radioloaical Consecuences of a Main Steamline Failure Outside Containment (BWR)

This event is only applicable to Boiling Water Reactors (BWRs). As such, this l

event is not applicable to the Palisades plant.

15.6.5 Loss of Coolant Accidents Resultina from a Saectrum of Postulated Elpina Breaks Within the Reactor Coolant Pressure 8oundarY I

15.6.5.1 Event Initiator - This event is initiated by a breach in the Primary Coolant System pressure boundary. Basically, a range of break sizes from small leaks up to a complete double-ended severance of a Primary Coolant

133 ANF-87-150(NP)

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System pipe must be considered. Typically, these breaks are classified as Large Breaks or Small Breaks. I 15.6.5.2 Event Descriotion 1

_(1) Larae Breaks. The large Break LOCA events are characterized by four-sequential phases: 1) blowdown, 2) refill, 3) reflood, and 4) long term cooling.

The blowdown phase immediately follows the initiation of a large break.

Primary system water is discharged through the break into containment. The system pressure decreases rapidly during the initial subcooled blowdown. As the saturation pressure is approached, local boiling and flashing takes place in the core and the reactor goes subcritical via the negative moderator reactivity feedback. The blowdown flow becomes a water-vapor mixture. The depressurization rate is reduced when core pressure falls below the saturation pressure. The water level continues to decrease until a large amount of water from the ECCS passive accumulators reaches the lower plenum.

The refill chase starts when the accumulator water begins to fill the lower plenum. At this time, the core is uncovered by water and the fuel rods are cooled primarily by thermal radiation.

The reflood chase begins when the water level reaches the bottom of the core.

The lona term coolina chase starts after the core has quenched to the point where the zircaloy-water reaction is suppressed, or the water level covers the active fuel. During this phase, the water inventory is controlled by the safety injection pumps. The continuous operation of these pumps ensures the long term dissipation of the decay heat.

(0) Small Breaks. The small break LOCA, as generally defined, includes any

134 ANF-87-150(NP)

Volume 1 break in the pressure boundary that has an area of 0.5 ft 2 or less. -The principal PWR design feature for mitigating the consequences of a small break LOCA is the ECCS which maintains the water inventory. Its major subsystems for restoring water inventory are the high pressure safety injection (HPSI) system, and the low pressure safety injection (LPSI) system and the safety injection (accumulator) tanks.

A small break LOCA is characterized by slow RCS depressurization rates and mass transfer rates within the RCS relative to similar parameters calculated for large break LOCA. If the break area is large enough that the HPSI pumps cannot maintain the reactor coolant inventory and allow RCS pressure control, the RCS will depressurize. The depressurization produces a low pressurizer pressure or thermal margin / low pressure reactor trip and a safety injection actuation signal (SIAS). The rate of RCS depressurization following SIAS depends on the break area and the HPSI shutoff head. With a combination of a very small break and a sufficiently high HPSI shutoff head, the depressurization may be arrested.

If the break area is sufficiently large to allow continued depressurization I and net loss of coolant inventory even with the HPSI pumps in operation, the coolant level in th~e reactor vessel may recede below the top of the reactor core. If sufficient steam is produced in the RCS, natural circulation (the reac+,or coolant pumps will have been tripped by this time to reduce coolant loss out of the break) around the RCS loops will cease. Eventually, loss of reactor coolant inventory is arrested by ECCS flow exceeding the flow out the break. In either case, the coolant level within the reactor vessel will rise, and the RCS will eventually be refilled (although leaking).

15.6.5.3 Reactor Protection (1) Larae Breaks. Basically no credit is taken for a reactor trip by the f

l Reactor Protection System (RPS) due to the rapid depletion of the moderator I

L

135 ANF-87-150(NP)

Volume 1 which shuts down the reactor core almost immediately, followed by ECCS injection which contains sufficient baron to maintain the reactor core in a subcritical configuration. Technical Specification limits on hot rod power serve to limit the peak cladding temperature.

(2) .Small Breaks. Primary reactor protection for this event is provided by j the low pressurizer pressure and low pressure / thermal margin trips and the j Safety Injection Actuation Signal (SIAS) on a low pressurizer pressure signal.

1 Reactor protection for the Loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary event is summarized in Table 15.6.5-A.

15.6.5.4 Disoosition and Justification (1) Larae Breaks. The current analytical basis for the Palisades Loss of Coolant Accident consists of References 11 through 16 and is applicable to Palisades for the reference cycle.

Additionally, this licensing action does not impact the Containment Pressure Analysis (Section 14.18 of the Updated FSAR) and Maximum Hypothetical Accident (Section 14.22 of the Updated FSAR) consequences since these analyses were -

performed at an assumed operating level of 2650 MWt, well above the current rated power. ,

(2) Small Breaks. The event is of greatest concern for the rated power j operating condition. The peak clad temperature which occurs during the transient is very dependent on the initial hot rod power. At intermediate, low, and zero pcxer operating conditions, the hot rod power is substantially reduced from that which characterizes rated power operating conditions.

Consequently, the peak cladding temperatures for these initial conditions will '

be less than those predicted for the rated power operating condition.

=

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The effects of additional steam generator tube plugging on the event consequences are expected to be quite minimal. These effects are three-fold:

1) the volume of primary water resident in the steam generators at event initiation is reduced, 2) the resistance to steam flow is increased in the steam generator tubes, and 3) the effective heat transfer area of the steam generators is reduced.

A smaller volume of water in the tubes reduces the amount of coolant which i must be discharged prior to the break being uncovered. This results in increased addition of safety injection water; however, the increased l

resistance in the U-tubes may slightly increase the time until the break uncovers which tends to decrease the addition of safety injection water. Thus, the smaller volume of water and increased resistance in the U-tubes produce offsetting effects.

During the forward heat transfer period, a reduction of effective heat transfer area in the steam generators can increase the temperature difference i

between the primary and secondary. This would result in a slight reduction in safety injection flow.

I During the reverse heat transfer period, the reduction in heat transfer surface area will reduce the amount of heat transferred from secondary to primary. The effect is to reduce primary system energy, and therefore aids primary system depressurization and leads to increased safety injection flow.

\

The consequences of the small break LOCA are therefore only slightly affected by the increased steam generator tube plugging. As such, the analysis of record (References 17-21) remains applicable to Palisades for the reference cycle.

Disposition of events for the loss of Coolant Accidents Resulting from a

[

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Spectrum of Postulated Piping Breaks Within- the Reactc' Coolant Pressure Boundary. event is sumarized in Table 15.6.5-B. i i

1 1

4

)

4 I

1 t

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Table 15.6.1-A Available Reactor Protection for the Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve Event i

l L Reactor Operating l Conditions Reactor Protection 1-5 Low Pressurizer Pressure Trip l Thermal Margin / Low Pressure Trip Safety Injection Actuation Signal 6-7 No significant consequences for these reactor operating conditions 8 Not a credible event; the reactor vessel head is unbolted or removed l

l l

l l

t l

l. '

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Volume-1 l

Table 15.6.1-B Disposition of Events for the Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve Event

]

Reactor Operating Conditions Disposition 1-5 Bounded by the Small Break LOCA analyses presented in References 17-21 6-7 No analysis required

~

8 No analysis required i

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Volume 1

, Table 15.6.2-A Available Reactor Protection for the Radiological

! Consequences of the Failure of Small Lines Carrying Primary Coolant Outside of Containment Event Reactor Operating i- Conditions Reactor Protection 1-5 Low Pressurizer Pressure Trip l

Thermal Margin / Low Pressure Trip i

Safety Injection Actuation Signal 6-8 No significant consequences for i these reactor operating conditions t

i 1

141 ANF-87-150(NP)

Volume 1 Table 15.6.2 B Disposition of Events for the Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside of Containment Event t

Reactor Operating Conditions Disoosition 1-5 Bounded by Small Break LOCA analyses l presented in References 17-21 6-8 No analysis required t

142 ANF-87-150(NP)

Volume 1 Table 15.6.3-A Available Reactor Protection for the Radiological Consequences of Steam Generator Tube Rupture Event 1

Reactor Operating Conditions Reactor Protection I

! 1 Thermal Margin / Low Pressure Trip Low Pressurizer Pressure Trip Safety Injection Actuation Signal 2-5 Same as Above 6-8 No significant consequences for these reactor operating conditions i

I l

l l

143 ANF-87-150(NP)

Volume 1 Table 15.6.3-8 Disposition _of Events for the Radiological Consequences of Steam Generator Tube Rupture Event l

Reactor Operating i 1

Conditions Disoosition 1 Bcunded by the analyses presented in the Updated Palisades FSAR, Section 14.15, Reference 4 2-5 Bounded by the above 6-8 No analysis required I

I l

l f.

l

f-144 ANF-87-150(NP)

Volume 1 Table 15.6.4 A Available Reactor Protection for the Radiological Consequences of a Main Steamline Failure Outside Containment (BWR) Event Reactor Operating Conditions Reactor Protection 4

1-8 Not applicable;.a Boiling Water Reactor (BWR) event Table 15.6.4-B Disposition of Events for the Radiological Consequences of a Main Steamline Failure Outside Containment (BWR) Event Reactor Operating Conditions Disoosition l

1-8 Not applicable; a Boiling Water Reactor (BWR) event a

i

\

(

145 ANF-87-150(NP) <

Volume 1 Table 15.6.5-A Available Reactor Protection for the Loss of Coolant  !

Accidents Resulting from a Spectrum of Postulated (

Piping Breaks Within the Reactor Coolant Pressure Boundary Reactor Operating i Conditions Reactor protection 1 Larae Brgiki -

No credit taken for reactor trip by the Reactor Protection System (RPS)

ECCS - short and long term cooling 3 I

1 Small Breaks -

Low Pressurizer Pressure Trip l Thermal Margin / Low Pressure Trip Low Reactor Coolant Flow Trip Safety injection Actuation Signal 2 Larae Breaks -

No credit taken for reactor trip by the Reactor Protection System (RPS) i ECCS - short and long term cooling l 2 Small Breaks -

Low Pressurizer Pressure Trip Thermal Margin / Low Pressure Trip Low Reactor Coolant Flow Trip ,

I Safety Injection Actuation Signal NonsafetyGrag%to15%

of Power, 10' Power,Noe High Rate-of-Cha

)

eredit taken 38 No significant consequences for these reactor operating conditions

i-146 ANF-87-150(NP)-

Volume 1 Table 15.6.5-B Disposition of Events for the Loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary Reactor Operating Conditions Disoosition 1 Larae Break -

Bounded by analyses presented in References 11-16 1 Small Break -

Bounded by analyses presented in References 17 21 2 Larae Break -

Bounded by analyses presented in References 11-16 2 Small Break -

Bounded by analyses presented in References 17-21 3-8 No analysis required i

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147 ANF-87-150(NP)

Volume 1 15.7 RADIOACTIVE RELEASES FROM A SUBSYSTEM OR COMP 0NENT 15.7.1 Waste Gas System Failure 15.7.2 Radioactive Liouid Waste System Leak or Failure (Release to Atmosohere) 15.7.3 Postulated Radioactive Releases Due to Liouid-Containina Tank Failures The results of the three events above are not dependent on either fuel type, steam generator tube plugging, reactor coolant flow rate, reactor coolant inlet temperature, or reactor protection system modifications. The reference analysis is therefore not affected by the current licensing action and remains the bounding analysis for this event. The reference analysis is provided in

+ the Updated Palisades FSAR, Reference 4.

15.7.4 Radioloaical Consecuences of Fuel Handlina Accident Calculations for the worst case fuel handling accident are provided in Section 14.19 of the Updated Palisades FSAR, Reference 4. Consequences were evaluated for an accident in the spent fuel area of the auxiliary building and for an accident occurring inside of containment. These calculations were performed in such a manner as to be bounding for the reference cycle. The fission product activity in the fuel rod gap was determined for the average fuel rod having a residence time of three full power years at 2650 MWt. The results were then multiplied by 1.65 to accommodate maximum potential radial peaking for the highest power fuel rod. in both cases, the accident was conservatively assumed to occur 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following a reactor shutdown.

4 _

1 1

l 148 ANF 87-150(NP)

Volume 1 15.7.5 Spent Fuel Cask Droo Accidents Calculations for ti;e worst case Spent Fuel Cask Drop Accident are provided-in Section 14.11 of the Updated Palisades FSAR, Reference 4. These calculations were performed in such c manner as to be bounding for the reference cycle. The accident was conservatively assumed to occur 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following a reactor shutdown.

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149 ANF 87-150(NP)

Volume 1

, Table 15.7.3-A Available Reactor Protection for the Postulated Radioactive Releases Due to liquid Containing Tank Failures Event

}

l Reactor Operating Conditions Reactor Protection 1-8 Reactor Protection System (RPS) action not required Table 15.7.3-B Disposition of Events for the Postulated Radioactive Releases Due to Liquid-Containing Tank Failures Event Reactor Operating Conditions Disposition 18 Bounding analysis presented in Section 14.20 of the Updated Palisades FSAR, Reference 4

1 i

\

(

150 ANF-87-150(NP) l Volume 1 i i

l Table 15.7.4-A Available Reactor Protection for the Radiological i Consequences of Fuel Handling Accidents  ;

l l

Reactor Operating Conditions Reactor Protection 1-8 Reactor Protection System (RPS) <

action not required

(

I l

Table 15.7.4-B Disposition of Events for the Radiological Consequences of Fuel Handling Accidents Reactor Operating Conditions Di ..sition ,

1-8 Bounding analysis presented in '

Sectior. 14.19 of the Updated Palisades FSAR, Reference 4 I

1

I 1

i

, 151 ANF-87-150(NP)

I Volume 1 [

Table 15.7.5-A Available Reactor Protection for the Spent Fuel i Cask Drop Accidents

• 1 Reactor Operating Conditions Reactor Protection 1-8 Reactor Protection System (RPS) action not required ,

r I

)

. Table 15.7.5-B Disposition of Events for the Spent Fuel 2

Cask Drop Accidents

• Reactor Operating -

Conditions Disoosition 1-8 Bounding analysis presented in f Section 14.11 of the Updated  :

Palisades FSAR, Reference 4 I

1

152 ANF-87-150(NP)

Volume 1

4.0 REFERENCES

1. "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," NUREG-0800, U.S. Nuclear Regulatory Commission, July 1981. ,

2. "Exxon Methodology for Pressurized Water Reactors - Analysis of Chapter 15 Events," XN-NF-84-73(P), Exxon Nuclear Comoany, August 1984.

3. "Pa:isades, Cycle 7 Safety Analysis Report," XN-NF-85 34. Rev. 2, Exxon Nuclear Company, January 1986.

4. Palisades, Updated Final Safety Analysis Report.

5. "Plant Transient Analysis of the Palisades Reactor for Operation at 2530 MWt," XN-NF-77-18, Exxon Nuclear Company, July 1977.

6. "Systematic Evaluation Program Design Basis Transient Reanalysis for the Palisades Reactor," XN-NF-81-25, Exxon Nuclear Company, May 1981.

7. "Palisades Cycle 6 Safety Analysis Report," XN-NF-83-03, Exxon Nuclear Company, January 1983.

8. "Rod Withdrawal Transient Reanalysis for the Palisades Reactor," XN-NF-83-57, Exxon Nuclear Company, August 1983.

v. "Plant Transient Analysis for Palisades Nuclear Power Plant with 50%

Steam Generator Plugging," XN NF-84-18, Exwn Nuclear Company, March 1984.

10. Palisades Plant Final Safety Analysis Report, Consumers Power Company.

11. "LOCA Analysis for Palisades at 2530 MWt Using the ENC WREM-II PWR ECCS Evaluation Model," XN-NF-77-24, Exxon Nuclear Company, July 1977.

12. "Analysis of Axial Power Distribution Limits for the Palisades Nuclear Reactor at 2530 MWt," XN-NF-78-16, Exxon Nuclear Company, June 1978,

13. "Analysis of Axial Power Distribution Limits for the Palisades Nuclear Reactor at 2530 MWt: Sensitivity Studies," XN-NF-78-16. Suco. 1, Exxon Nuclear Company, April 1984.

14. "Palisades Cycle 5 Reload Fuel Safety Analysis Report," XN-NF-81-34(P),

Exxon Nuclear Company, May 1981.

15. "ECCS and Thermal-Hydraulic Analysis for the Palisades Reload H Design,"

)(N-NF-80-18, Exxon Nuclear Company, April 1980.

J l

153 ANF-87-150(NP) -

Volume 1

16. Combustio7i Engineering Report, CENPD-132P, "Calculative Methods for the CE Large Break LOCA Evaluaticn Model."

17. Combustion Engineering Report, CENPD-137P, ' Calculative Methods for the .

l CE Small Break LOCA Evaluation Model."

18. Combustion Engineering Report, CEN-ll4-P, Amendment 1-P, "Review of Small Break Transients in Combustion Engineering Nuclear Steam Supply Systems."

19. Combustion Engineering Report [fN-203(P). Rev. 1, "Response to NRC Action Plan Item II.K.3.30 - Justification of Small Break LOCA Methods.'

! 20. Letter, D.J. VandeWalle (CPCo) to Director, Nuclear Reactor Regulation, dated June 20, 1985, Docket S0-255, License DPR-20, Palisades Plant, I "NVREG-0737 Item II.K.3.30 Small Break LOCA Models, and Item II.K.3.31 l Plant Specific Analyses."

21. Letter, John A. Zwolinski (NRC) to Mr. VandeWalle (CPCo), dated July 3, 1985; Docket 50-255, License DRP-20, Palisades Plant, "THI Action Plan Item II.K.3.30, Small Break LOCA Analysis and II.K.3.31, Plant Analysis."

22. "Palisades Cycle 5 Reload Fuel Safety Analysis Report," XN-NF-81-34(P). q Suco.1, Exxon Nuclear Company, December 1981.

._ _ _ _ _ _ _ l

ANF-87-150(NP)

Volume 1 Issue Date: 6/13/88 PALISADES MODIFIED REACTOR PROTECTION SYSTEM REPORT -

DISPOSITION OF STANDARD REVIEW PLAN CHAPTER 15 EVENTS i

i-Djstribution I R. C. Gottula J. S. Holm J. W. Hulsman i

! J. D. Kahn l

T. R. Lindquist J. N. Morgan l L. A. Niel.

! L. D. C'0911 F. B. :;kogen G. N. Ward i H. E. Williamson CPCo/HG Shaw (20) l- Document Control (5) .

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