ML18057B351

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Cycle 10:Disposition & Analysis of SRP Chapter 15 Events.
ML18057B351
Person / Time
Site: Palisades Entergy icon.png
Issue date: 10/31/1991
From: Cole S, Lindquist T, Little R
SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML18057B349 List:
References
EMF-91-176, NUDOCS 9111080151
Download: ML18057B351 (106)
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Text

SIEMENS EMF-91-176 Issue Date: 10/01 /91 PALISADES CYCLE 1O: DISPOSITION AND ANALYSIS OF STANDARD REVIEW PLAN CHAPTER 15 EVENTS r

re I Prepared by:

T. R. Lindquist, Team ead' r PWR Fuel Engineering Fuel Engineering and Licensing Contributors: S. E. Cole R. B. Little G. B. Peeler (Consultant)

K. C. Segard E. L Tolman S.C.Yung October 1991

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Siemens Nuclear Power Corporation Engineering and Manufacturing Facility 2101 Horn Rapids Road, PO Box 130 Richland, WA 99352-0130 Tel: (509) 375-8100 Fax: (509) 375-8402

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.1 CUSTOMER DISCLAIMER I IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT I PLEASE READ CAREFULLY Siemens Nuclear Power Corporation's war:ranties and representations concerning t

  • the subject matter of this document are those set forth in the Agreement between Siemens Nuclear Power Corporation and the Customer pursuant to which this document is issued. Accordingly, except as otherwise expressly provided in such Agreement. neither Siemens Nuclear Power Corporation nor any person acting on its behalf makes any warranty or representation, expressed or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this document, or that the use of any information, apparatus, method or process disclosed in this document will not infringe privately owned rights; or assumes any liabilities with respect to the use of any information, apparatus, method or process disclosed in this document.

el The information contained herein is for the sole use of the Customer. I In order to avoid impairment of rights of Siemens Nuclear Power Corporation in patents or inventions which may be included in the information contained in this document, the recipient, by its acceptance of this document, agrees not to publish or make public use (in the patent use of the term) of such information until so authorized J

in writing by Siemens Nuclear Power Corporation or until after six*(6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless expressly provided in the Agreement. No rights or licenses in or to any patents are implied by the furnishing of this document

EMF-91-176 Page i TABLE OF CONTENTS Section 1.0 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.0 Summary and Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3.0 Disposition and Analysis of Plant Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 15.0 Accident Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 15.0.1 Classification of Plant Conditions . . . . . . . . . . . . . . . . . . . 12 15.0.2 Plant Characteristics and Initial Conditions . . . . . . . . . . . 18 15.0.3 Power Distribution . ; ............. : . . . . . . . . . . . . . . 22 15.0.4 Range of Plant Operating Parameters and States . . . . . . *25 15.0.5 Reactivity Coefficients Used in the Safety Analysis . . . . . . 27 15.0.6 Scram Insertion Characteristics . . . . . . . . . . . . . . . . . . . 29 15.0.7 Reactor Protection System Trip Setpoints and Time Delays....................................... 31 15.0.8 Component Capacities and Setpoints . . . . . . . . . . . . . . . 44 15.0.9 Plant Systems and Components Available for Mitigation of Accident Effects . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 15.0.10 Effects of Mixed Assembly Types and Fuel Rod Bowing . 52 15.0.11 Plant Licensing Basis and Single Failure Criteria . . . . . . . 53 15.1 Increase in Heat Removal by the Secondary System . . . . . . . . . . 55 15.1.1 Decrease in Feedwater Temperature . . . . . . . . . . . . . . . . 55 15.1.2 Increase in Feedwater Flow . . . . . . . . . . . . . . . . . . . . . . . 56 15.1.3 Increase in Steam Flow . . . . . . . . . . . . . . . . . . . . . . . . . . 56 15.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve ....................... *..... . . . . . . 59 15.1.5 Steam System Piping Failures Inside and Outside of Containment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59 15.2 Decrease in Heat Removal by the Secondary System . . . . . . . . . 61 15.2.1 Loss of External Load . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 15.2.2 Turbine Trip . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62 15.2.3 Loss of Condenser Vacuum . . . . . . . . . . . . . . . . . . . . . . 63 15.2.4 Closure of the Main Steam Isolation Valves (MSIV)

(BWR) ...................................... . 63 15.2.5 Steam Pressure Regulator Failure ................. . 64 15.2.6 Loss of Nonemergency A.C. Power to the Station Auxiliaries . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 64 15.2.7 Loss of Normal Feedwater Flow .................. . 65 15.2.8 Feedwater System Pipe Breaks Inside and Outside Containment ................................. . 65

t EMF-91-176 I Page ii

.* 1 TABLE OF CONTENTS Section 1 15.3 Decrease in Reactor Coolant System Flow ; . . . . . . . . . . . . . . . . 68 i

.15.3.1 Loss of Forced Reactor Coolant Flow . . . . . . . . . . . . . . . 68 15.3.2 15.3.3 Flow Controller Malfunction . . . . . . . . . . . . . . . . . . . . . . .

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

70 70 I

15.3.4 Reactor Coolant Pump Shaft Break . . . . . . . . . . . . . . . . . 72 15.4 15.4.1

. Reactivity and. Power Distribution Anomalies . . . . . . . . . . . . . . . .

Uncontrolled Control Rod Assembly Withdrawal from a 73 l

Subcritical or Low Power Startup Condition . . . . . . . . . . . 73 15.4.2 Uncontrolled Control Rod Bank Withdrawal at Power . . . . 74 ;I 15.4.3 Control Rod Misoperation . . . . . . . . . . . . . . . . . . . . . . . . 76 15.4.4 Startup of an Inactive Loop . . . . . . . . . . . . . . . . . . . . . . . 85 15.4.5 Flow Controller Malfunction . . . . . . . . . . . . . . . . . . . . . . . 85 15.4.6 CVCS Malfunction that Results in a Decrease in the 1 Boron Concentration in the Reactor Coolant . . . . . . . . . . 86 15.4.7 Inadvertent Loading and Operation of a Fuel Assembly 15.4.8 in an Improper Position . . . . . . . . . . . . . . . . . . . . . . . . . .

Spectrum of Control Rod Ejection Accidents . . . . . . . . . .

86 87 el 15.4.9 Spectrum of Rod Drop Accidents (BWR) . . . . . . . . . . . . . 87 15.5 Increases in Reactor Coolant System Inventory . . . . . . . . . . . . . 88 1 15.5.1 Inadvertent Operation of the. ECCS that Increases Reactor Coolant Inventory . . . . . . . . . . . . . . . . . . . . . . . . 88 15.5.2 CVCS Malfunction that Increases Reactor Coolant 'J Inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 88 15.6 Decreases in Reactor Coolant Inventory . . . . . . . . . . . . . . . . . . . 90 15.6.1

  • Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 90 1

15.6.2 Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside of Containment . 91 15.6.3 Radiological Consequences of Steam Generator Tube Failure .......................... ~ . . . . . . . . . . . . 92 15.6.4 Radiological Consequences of a Main Steam Line Failure Outside Containment (BWR) . . . . . . . . . . . . . . . . . 92 15.6.5 Loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary . . . . . . . . . . . . . . . . . . . . . . ; . . . . . . 93

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TABLE OF CONTENTS EMF-91-176 Page iii I Section l 15. 7 Radioactive Releases from a Subsystem or Component . . . . . . . 94 15.7.1 Waste Gas System Failure........................ 94 I 15.7.2 Radioactive Liquid Waste System Leak or Failure (Release to Atmosphere) . . . . . . . . . . . . . . . . . . . . . . . . . 94

15. 7.3 Postulated Radioactive Releases Due to Liquid-t 15.7.4 Containing Tank Failures . . . . . . . . . . . . . . . . . . . . . . . . .

Radiological Consequences of Fuel Handling Accident . .

94 94 15.7.5 Spent Fuel Cask Drop Accidents . . . . . . . . . . . . . . . . . . . 95 J

4.0 References ......................................... : .......... *. 96 re t

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LIST OF TABLES EMF-91-176 Page iv 2-1 DISPOSITION OF EVENTS

SUMMARY

FOR PALISADES CYCLE 1O . . . . . 4 2-2

SUMMARY

OF PLANT LICENSING BASIS . . . . . . . . . . . . . . . . . . . . . . . . 9 15.0.1-1 ACCIDENT CATEGORY USED FOR EACH ANALVZED EVENT . . . . . . . . 15 15.0.2-1 PLANT OPERATIONAL MODES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 15.0.2-2 NOMINAL PLANT OPERATING CONDITIONS . . . . . . . . . . . . . . . . . . . . . 20 15.0.2-3 NOMINAL RELOAD N FUEL DESIGN PARAMETERS . . . . . . . . . . . . . . . 21 15.0.3-1 CORE POWER DISTRIBUTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 15.0.4-1 RANGE OF KEY INITIAL CONDITION OPERATING PARAMETERS . . . . . 26 15.0.5-1 REACTIVITY PARAMETERS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 15.0.7-1 TRIP SETPOINTS FOR OPERATION AT 2530 MWT . . . . . . . . . . . . . . . . 39 15.0.7-2 TM/LP UNCERTAINTIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 15.0.8-1 COMPONENT CAPACITIES AND SETPOINTS . , . . . . . . . . . . . . . . . . . . 45 15.0.9-1 OVERVIEW OF PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR TRANSIENT AND ACCIDENT CONDITIONS . . . . . . . . . . . . . . . . . . . . . . 47

    • I 15.4.3-1

SUMMARY

OF MDNBRS FOR CONTROL ROD MISOPERATION EVENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 84

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LIST OF FIGURES EMF-91-176 Page v l Figure

~ 15.0.3-1 LIMITING AXIAL POWER SHAPE {ASI = -0.112) . . . . . . . . . . . . . . . . ... 24 15.0.6-1 INTEGRATED SCRAM WORTH WITH MOST REACTIVE ROD STUCK OUT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 r 15.0.7-1 TINLET LIMITING CONDITION OF OPERATION . . . . . . . . . . . . . . . . . . . . 41 15.0.7-2 AXIAL SHAPE FUNCTION FOR TM/LP TRIP . . . . . . . . . . . . . . . . . . . . . . 42 15.0.7-3 CORE PROTECTION LIMITS FOR TM/LP TRIP {ASI = +0.200) . . . . . . . 43

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EMF-91-176 Page 1 1.0 Introduction This report documents the results of a disposition and analysis of the FSAR Chapter 14 events in support of Palisades Cycle 1O operation with up to 15.0% steam generator tube plugging. The events were evaluated in accordance with Chapter 15 of the Standard Review Plan (SRP}(1) and Siemens Nuclear Power (SNP} Corporation methodologyC2>. The changes l that will be implemented for Cycle 1O include the following:

l (1} The insertion of the second full reload of fuel that utilize High Thermal Performance (HTP} grid spacers.

(2) An increase in assembly radial power peaking, FrA* to accommodate a low radial leakage loading pattern. The proposed Technical Specification radial peaking factor limit's for Cycle 1O are: (a} a fuel rod peaking factor limit for all

~-. fuer assembly types of 1.92, and (b) assembly peaking factor limits of 1.48 for 208 fuel rod assemblies, 1.57 for 216 fuel rod assemblies (Reloads L and M}

and 1.66 for Reload N. The proposed Technical Specification radial peaking limits for Cycle 1O are given in Table 15.0.3-1 .

(3)

I Inclusion of eight partial shielding assemblies (PSA} in low powered peripheral locations to reduce vessel fluence.

(4) Reactor Protection System modifications (FC-888)(~>.

I (5) Main Feedwater Controller upgrade (FC-920)(a).

The minimum departure from nucleate boiling ratio (MDNBR) calculations using the ANFP critical heat flux correlationC9*1°> were performed for Reload N fuel with HTP spacers and radial peaking factors of 1.66 and 1.92 for the assembly and peak rod, respectively. The Cycle 9 analysisC3> supports the proposed Technical Specification radial peaking limits for previous reloads and remains conservatively applicable for Cycle 10.

Section 2.0 presents a summary of results for this analysis. Section 3.0 presents the conditions employed in the event analyses and a discussion of the event disposition and

  • a CPCo is addressing this change.

EMF-91-176 Page 2 MDNBR results for the SRP Chapter 15 events. The events are numbered in accordance with the SRP to facilitate review.

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f EMF-91-176 Page 3 2.0 Summary and Conclusions A summary of the Disposition of Events for Cycle 1O is given in Table 2-1. This table lists each SRP Chapter 15 event, indicates whether that event is reanalyzed for this submittal, provides a reference to the bounding event or analysis of record for events not reanalyzed and a cross reference between SRP event numbers and the Palisades Updated FSAR(S).

The results of Anticipated Operational Occurrences and Postulated Accidents reanalyzed for this submittal are listed in Table 2-2 along with a complete summary of the plant licensing basis. The MDNBR analyses performed for Cycle 1O incorporated the ANFP critical heat flux correlation, the second full reload of HTP spacer fuel, Reload N specific fuel design (e.g., fuel rod enrichments, loading, etc.) and an axial shape characteristic of full power control rod position. The effect of the aforementioned items more than offsets the DNBR margin lost due to increasing assembly radial peaking. In general, the there was an increase in thermal te margin for Cycle 1O relative to Cycle 9. The MDNBR for one event, the dropped bank, decreased slightly relative to Cycle 9 due to a larger radial peaking augmentation factor. The I results reported herein confirm that event acceptance criteria (Section 15.0.1.1) are met for Cycle 1O operation as defined by the operating parameter ranges in Sections 15.0.1 to 15.0.8 of this report. These results support operation with up to 15.0% average steam generator tube plugging at a rated thermal power of 2530 MWt. Table 2-2 also contains a compilation of the current licensing basis for Palisades.

1 EMF-91-176 Page 4 TABLE 2-1 DISPOSITION OF EVENTS

SUMMARY

FOR PALISADES CYCLE 10 SRP Bounding Event Event or UFSAR Designation Event Name Disposition Reference Designation 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM I

15.1.1 15.1.2 Decrease in Feedwater Temperature Increase in Feedwater Flow Bounded 15.1.3 14.9.4 I

1) Power Bounded 15.1.3 14.9.6
2) Startup Bounded 15.1.3 14.9.5 15.1.3 Increase in Steam Flow Analyze(a) Ref. 3(b) t 14.10 15.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve
1) Power Bounded 15.1.3 1
2) Scram Shutdown Margin Boun.ded 15.1.3 15.1.5 Steam System Piping Failures Inside and Outside of Containment Bounded Ref. 3 14.14 et 15.2 15.2.1 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Loss of External Load Bounded Ref. 3 14.12 I

15.2.2 Turbine Trip Bounded 15.2.1 15.2.3 Loss of Condenser Vacuum Bounded 15.2.1 15.2.4 Closure of the Main Steam Isolation Valves Bounded 15.2.1 (MSIV) 15.2.5 Steam Pressure Regulator Failure Not Applicable; BWR Event 15.2.6 Loss of Nonemergency A.C. Power to the Short term: 15.3.1 Station Auxiliaries Bounded Long term: 15.2.7 Bounded a

MDNBR analysis will be performed for Cycle 1O.

b PTSPWR2 analysis is given reference for this event.

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I TABLE 2-1 DISPOSITION OF EVENTS

SUMMARY

FOR PALISADES CYCLE 10, CONTINUED SRP Bounding Event Event or UFSAR Designation Event Name Dis12osition Reference Designation 15.2.7 Loss of Normal Feedwater Flow

1) Maximum PCS pressure Bounded Ref. 17 14.13
2) Maximum Primary-to-Secondary Bounded Ref. 17 14.13 pressure difference
3) Minimum steam generator inventory Bounded Ref. 17 15.2.8 Feedwater System Pipe Breaks Inside and Cool down: 15.1.5 Outside Containment Bounded Heatup: 15.2.1 Bounded 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW 15.3.1 Loss of Forced Reactor Coolant Flow Analyze(a) Ref. 3(b) 14.7 15.3.2 Flow Controller Malfunction Not Applicable 14.7 15.3.3 Reactor Coolant Pump Rotor Seizure Analyze(a) Ref. 3(b) 14.7 15.3.4 Reactor Coolant Pump Shaft Break
  • Bounded 15.3.3 14.7 15.4 REACTIVITY AND POWER DISTRIBUTION ANO MAU ES 15.4.1 Uncontrolled Control Rod Bank Withdrawal Bounded Ref. 3 14.2.1 from a Subcritical or Low Power Condition 15.4.2 Uncontrolled Control Rod Bank Withdrawal at Analyze(a) Ref. a(b) 14.2.2 Power Operation Conditions a MDNBR analysis will be performed for Cycle 10.

b PTSPWR2 analysis is given reference for this event.

1 EMF-91-176 Page 6 TABLE 2-1 DISPOSITION OF EVENTS

SUMMARY

FOR PALISADES CYCLE 10, CONTINUED SRP Bounding Event Event or UFSAR Designation Event Name DisQosition Reference Designation 15.4.3 Control Rod Misoperation

1) Dropped Control Bank/Rod Analyze(a) Ref. 3(b) 14.4
2) Dropped Part-Length Control Rod Bounded 15.4.3(1) 14.6
3) Malpositioning of the Part-Length Not Applicable Control Group
4) Statically Misaligned. Control Bounded 15.4.3(1) 14.6 Rod/Bank
5) Single Control Rod Withdrawal Analyze(a) Ref. 9(b) 14.2.3
6) Core Barrel Failure Bounded 15.4.8 14.5 15.4.4 Startup of an Inactive Loop Bounded by 14.8 rated power MDNBR 15.4.5 Flow Controller Malfunction Not Applicable: No Flow Controller et 15.4.6 CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant I
1) Rated and Power Operation Bounded Ref. 3 14.3 2)

Conditions Reactor Critical, Hot Standby and Hot Shutdown Bounded Ref. 3 14.3 I

3) Refueling Shutdown Condition, Cold Bounded Ref. 3 14.3 Shutdown Condition and Refueling Operation 15.4.7 Inadvertent Loading and Operation of a Fuel Administrative Assembly in an Improper Position procedures preclude this event a

MDNBR analysis will be performed for Cycle 1o.

b PTSPWR2 analysis is given reference for this event.

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  • EMF-91-176 Page 7 l TABLE 2-1 DISPOSITION OF EVENTS

SUMMARY

FOR PALISADES CYCLE 10, CONTINUED I SAP Bounding Event Event or UFSAR I Designation 15.4.8 Event Name Spectrum of Control Rod Ejection Accidents Disgosition Bounded Reference Ref. 3 Designation 14.16 I 15.4.9 Spectrum of Rod Drop Accidents (BWR) Not Applicable; BWR Event 15.5 INCREASES IN REACTOR COOLANT t* INVENTORY 15.5.1 Inadvertent Operation of the ECCS that Overpressure: 15.2.1 Increases Reactor Coolant Inventory Bounded Reactivity: 15.4.6 Bounded 15.5.2 CVCS Malfunction that Increases Reactor Overpressure: 15.2.1 te Coolant Inventory Bounded Reactivity:

Bounded 15.4.6 I 15.6 DECREASES IN REACTOR COOLANT INVENTORY 15.6.1 Inadvertent Opening of a PWR Pressurizer Analyze(a) Ref. 3(bJ, 6 Pressure Relief Valve 15.6.2 Radiological Consequences of the Failure of Bounded 15.6.5 14.23 Small Lines Carrying Primary Coolant Outside of Containment 15.6.3 Radiological Consequences of Steam Bounded Ref. 6 14.15 Generator Tube Failure 15.6.4 Radiological Consequences of a Main Steam Not Line Failure Outside Containment Applicable; BWR Event 15.6.5 Loss of Coolant Accidents Resulting from a Analyze Ref. 6, 14 14.17 Spectrum of Postulated Piping Breaks Within 14.18 the Reactor Coolant Pressure Boundary 14.22 a MDNBR analysis will be performed for Cycle 10.

  • b PTSPWR2 analysis is given reference for this event.

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EMF-91-176 Page 8

  • I TABLE 2-1 DISPOSITION OF EVENTS

SUMMARY

FOR PALISADES CYCLE 10, CONTINUED l SRP Bounding I

Event Event or UFSAR Designation Event Name Disposition Reference Designation 15.7 RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.1 Waste Gas System Failure Bounded Ref. 6 14.21 15.7.2 Radioactive Liquid Waste System Leak or Bounded Ref. 6 14.20 Failure (Release to Atmesphere) 15.7.3 Postulated Radioactive Releases due to Bounded Ref. 6 14.20 Liquid-Containing Tank Failures 15.7.4 Radiological Consequences of Fuel Handling Analyze(a) 14.19 Accidents 15.7.5 Spent Fuel Cask Drop Accidents Analyze(a) 14.11 el l

a CPCo will provide the reanalysis of this event.

  • TABLE 2-2

SUMMARY

OF PLANT LICENSING BASIS MAXIMUM CORE MAXIMUM MAXIMUM POWER AVERAGE HEAT PRESSURIZER PEAK EVENT(aJ LEVEL (MWt} FLUX (Btu[hr-tt2} PRESSURE (~sia} MDNBR LHR (kWfn} REFERENCE 15.1.3 Increase In Steam Flow 2735.4 175166 2033.12 1.81ibJ 15.00 3 15.1.5 Steam System Piping 318.0 20272 2060.0 Note (c) 19.00 3 Failures Inside and Outside of Containment 15.2.1 Loss of External Load 2668.4 165499 2625.36 Note (d) 3 15.2.7 Loss of Normal Feedwater 2580.6 167000 2271.9 Note (e) 17 15.3.1 Loss of Forced Reactor 2686.7 165473 2127.78 1.391(bJ 13.77 3 Coolant Flow 15.3.3 Reactor Coolant Pump 2743.1 165473 2145.19 1.341(bJ 14.11 3 Rotor Seizure 15.4.2 Uncontrolled Control Bank 2900.9 178867 2267.14 1.640(b) 16.98 8 Withdrawal at Power 15.4.3 Control Rod Mlsoperatlon

  • Dropped Rod 2580.6 165473 2010 1,400(b) 16.29 3
  • Dropped Bank 2340.7 146400 2060 1,553(bJ 16.30 3
  • Single Rod Withdrawal 2900.9 178867 2267.14 1,375(bJ 18.34 8 15.4.6 CVCS Malfunction Resulting Adequacy of the shutdown margin Is demonstrated 3 in Decreased Boron Concentration 15.4.8 Control Rod Ejection 3494.3 173528 2115.00 Note (f) 3 15.6.1 Inadvertent Opening of a 2690.4 168558 .. 2110.10 . 1.741(bJ 14.43 3 m PWR Pressurizer Pressure ,,s:

Relief Valve "ti ch SU _.

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TABLE 2-2

SUMMARY

OF PLANT LICENSING BASIS, CONTINUED MAXIMUM CORE MAXIMUM MAXIMUM POWER AVERAGE HEAT PRESSURIZER. PEAK EVENT(a) LEVEL (MWt) FLUX (Btu/hr-tt2l PRESSURE (pslal MDNBR LHR (kW/ftl REFERENCE 15.6.3 Radiological Consequences Radiological consequence acceptance criteria are met 6 of Steam Generator Tube Failure 15.6.5 Losa of Coolant Accidents 10 CFR 50.46(b) acceptance criteria are shown to be met 14 Resultlng from a Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary 15.7.4 Radlologlcal Radlologlcal consequence acceptance criteria are shown to be met 6 Consequences of Fuel Handling Accidents 15.7.5 Spent Fuel Cask Drop Radlologlcal consequence acceptance criteria are shown to be met 6 Accidents

a. The events that are reanalyzed as a result of the Cycle 1O changes are Indicated in bold type.
b. The MDNBRs are based on the radial peaking assumptions given In Table 15.0.3-1 and theANFP correlation (95/95 limit= 1.154). A2% mixed core penalty was Included In the MDNBR analyses such that the effective DNB correlation limit is 1.174.
c. Approximately 2% of the fuel fails as a result of penetrating DNB limits.
d. MDNBR for the loss of load event is bounded by other AOOs.
e. MDNBR Is bounded by the loss of flow event (15.3.1)
f. 14.7% of the fuel rods in the core are calculated to fail as a result of penetrating DNB limits. The offsite doses that result from this event are 38 rem (thyro~

and 0.4 rem (whole body). These doses are within the acceptance criteria for this event (75 rem for thyroid and 6.25 rem for whole body). "'U 11 P> <<>

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I EMF-91-176 Page 11 3.0 Disposition and Analysis of Plant Events This section provides the results of the disposition of events review and MDNBR analyses performed to support Cycle 1O operation. Event numbering and nomenclature are consistent with the SRP to facilitate review. Reference 3 contains information on the plant licensing basis as it affects the event analyses including classification of plant conditions and classification of accident events by category. This section provides information on the plant l operating mode, initial conditions, neutronics data, core and fuel design parameters. Listings of systems and components available for accident mitigation, trip setpoints, time delays and component capacities are also included. These data, together with the design parameters(20) t and the event specific input data, represent a comprehensive summary of analysis inputs.

A system transient analysis for non-LOCA events was previously performed for Cycle 9(3). The changes introduced in Cycle 1O (i.e., increase in assembly radial peaking and Cycle te 1O core loading) affect the only the event MDNBR and will not affect the system response to a non-LOCA transient event. Thus, the system thermal-hydraulic response for the transient I analysis performed for Cycle 9 remains applicable to Cycle 10. The effect of increased assembly radial peaking on MDNBR and LHR will be assessed for anticipated operational occurrences (AOO) for Cycle 10. The effect of the Cycle 1O changes on fuel failures and radiological consequences for the postulated accidents will be assessed. The large break LOCA is being reanalyzed for Cycle 1O because of the increase in assembly radial peaking factor.

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EMF-91-176 Page 12

  • I 15.0 Accident Analyses 15.0.1 Classification of Plant Conditions I

The classification of plant conditions is given below. For Cycle 1O, these classifications are unchanged from previous analyses beginning with Cycle a<1n. Plant operations are placed I

in one of four categories. These categories are those adopted by the American Nuclear Society (ANS). The categories are:

NORMAL OPERATION AND OPERATIONAL TRANSIENT

., Events which are expected to occur frequently in the course of power operation, refueling, maintenance, or plant maneuvering.

FAULTS OF MODERATE FREQUENCY

  • Events which are expected to occur on a frequency of once per year during plant operation.

et INFREQUENT FAULTS

  • Events which are expected to occur once during the lifetime of the I plant.

LIMITING FAULTS

  • Events which are not expected to ()Ccur but which are evaluated to demonstrate the adequacy of the design.

15.0.1.1 Acceptance Criteria Operational Events This condition describes the normal operational modes of the reactor. As such, occurrences in this category must maintain margin between operating conditions and the plant setpoints. The setpoints are established to assure maintenance of margin to design limits.

The set of operating conditions, together with conservative allowances for uncertainties establish the set of initial conditions for the other event categories.

I EMF-91-176 Page 13 Moderate Frequency Events

1. The pressures in reactor coolant and main steam systems should be less than 110%

of design values.

2. The fuel cladding integrity should be maintained by ensuring that fuel design limits are not exceeded. That is, the minimum calculated departure from .nucleate boilillQ ratio I (DNBR) is not less than the applicable limits of the DNBR correlation being used.
3. The radiological consequences should be less than 10 CFR 20 guidelines.
4. The event should not generate a more serious plant condition without other faults occurring independently.

Infrequent Events

1. *The pressures in reactor coolant and main steam systems should be less than 11 0%

of design values.

'*I 2.

3.

A small fraction of fuel failures may occur, but these failures should not hinder the core coolability.

The radiological consequences should be a small fraction of 10 CFR 100 guidelines (generally < 10%).

4. The event should not generate a limiting fault or result in the consequential loss of the reactor coolant or containment barriers.

Limiting Fault Events

1. Radiological consequences should not exceed 1o CFR 100 guidelines.
2. The event should not cause a consequential loss of the required functions of systems needed to cope with the reactor coolant and containment systems transients.
3. Additional criteria to be satisfied by specific events are:
a. LOCA - 10 CFR 50.46 and Appendix K.
b. Rod Ejection - Maximum deposited fuel enthalpy <280 cal/gm .

EMF-91-176 Page 14 15.0.1.2 Classification Of Accident Events By Category Table 15.0.1-1 lists the accident category used for each event considered in this report.

This classification is used in evaluating the acceptability of the results obtained from the analysis. These are unchanged from previous analyses beginning with Cycle aC17).

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EMF-91-176 Page 15 TABLE 15.0.1-1 ACCIDENT CATEGORY USED FOR EACH ANALVZED EVENT SRP Event Designation Event Name Category(a) 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.1 Decrease in Feedwater Temperature Moderate (AOO) 15.1.2 Increase in Feedwater Flow Moderate (AOO) 15.1.3 Increase in Steam Flow Moderate (AOO) 15.1.4 Inadvertent Opening of a Steam Generator Moderate (AOO)

Relief or Safety Valve 15.1.5 Steam System Piping Failures Inside and Limiting Fault (PA)

Outside of Containment 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.1 Loss of External Load Moderate (AOO)

~. 15.2.2 15.2.3 Turbine Trip Loss of Condenser Vacuum Moderate (AOO)

Moderate (AOO)

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

Moderate (AOO) 15.2.5 Steam Pressure Regulator Failure Moderate (AOO)

I 15.2.6 . Loss of Nonemergency AC. Power to the Station Auxiliaries Moderate (AOO) 15.2.7 Loss of Normal Feedwater Flow Moderate (AOO)

I 15.2.8 Feedwater System Pipe Breaks Inside and Outside Containment Limiting Fault (PA) 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW 15.3.1 Loss of Forced Reactor Coolant Flow Moderate (AOO) 15.3.2 Flow Controller Malfunction Moderate (AOO) 15.3.3 Reactor Coolant Pump Rotor Seizure Infrequent (PA) 15.3.4 Reactor Coolant Pump Shaft Break Limiting Fault (PA) a Anticipated Operational Occurrence (AOO)

  • Postulated Accident (PA)

I EMF-91-176 Page 16 TABLE 15.0.1-1 ACCIDENT CATEGORY USED FOR EACH ANALVZED EVENT, CONTINUED SAP Event Designation Event Name Category(a) 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1 Uncontrolled Control Rod Bank Withdrawal from a Subcritical or Low Power Condition Moderate (AOO) I 15.4.2 . UncontrolJed Control Rod Bank Withdrawal at Moderate (AOO)

Power Operation Conditions J Control Rod *Misoperation 1) 2)

3)

Dropped Control Bank/Rod Dropped Part-Length Control Rod Malpositioping of the Part-Length Moderate (AOO)

Moderate (AOO)

Moderate (AOO)

I Control Group

4) Statically Misaligned Control Rod/Bank Moderate (AOO)

I

5) Single Control Rod Withdrawal Infrequent (PA)
6) Core Barrel Failure Limiting Fault (PA) 15.4.4 Startup of an Inactive Loop Moderate (AOO) et 15.4.5 Flow Controller Malfunction Moderate (AOO) 15.4.6 eves Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Moderate (AOO) I Coolant 15.4.7 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position Infrequent (PA) I 15.4.8 Spectrum of Control Rod Ejection Accidents Limiting Fault (PA) 15.4.9 . Spectrum of Rod Drop Accidents (BWR) Not Applicable 15.5 INCREASES IN REACTOR COOLANT SYSTEM INVENTORY 15.5.1 Inadvertent Operation of the ECes that Moderate (AOO)

Increases Reactor Coolant Inventory 15.5.2 CVes Malfunction that Increases Reactor Moderate (AOO)

Coolant Inventory a Anticipated Operational Occurrence (AOO) 1 Postulated Accident (PA)

I

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I EMF-91-176 Page 17 1*

TABLE 15.0.1-1 ACCIDENT CATEGORY USED FOR EACH ANALVZED EVENT, I SRP CONTINUED I Event Designation 15.6 Event Name DECREASES IN REACTOR COOLANT Categoryfa)

INVENTORY I 15.6.1 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve Moderate (AOO}

I 15.6.2 Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside of Containment Infrequent (PA}

I 15.6.3 Radiological Consequences of Steam Generator Tube Failure Limiting Fault (PA}

15.6.4 Radiological Consequences of a Main Steam Limiting Fault (PA}

I 15.6.5 Line Failure Outside Containment Loss of Coolant Accidents Resulting from a Limiting Fault (PA}

Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary 15.7 RADIOACTIVE RELEASES FROM A SUBSYSTEM OR COMPONENT 15.7.1 Waste Gas System Failure Note (b}

15.7.2 Radioactive Liquid Waste System Leak or Note (b}

Failure (Release to Atmosphere) 15.7.3 Postulated Radioactive Releases Due to Infrequent (PA)

Liquid-Containing Tank Failures 15.7.4 Radiological Consequences of Fuel Handling Limiting Fault (PA}

Accident 15.7.5 Spent Fuel Cask Drop Accidents Infrequent (PA}

a Anticipated Operational Occurrence (AOO)

Postulated Accident (PA) b This event has been deleted .from the SRP but is part of the Palisades licensing basis.

EMF-91-176

  • Page 18 15.0.2 Plant Characteristics and Initial Conditions Eight operational modes have been considered in the disposition and analyses. These are given in Ta~le 15.0.2-1. These operational modes have been considered in establishing the subevents associated with each event initiator. A set of initial conditions is established for the events analyzed that is consistent with the conditions for each mode of operation. The modes of plant operation are unchanged for Cycle 1o.

The nominal plant rated operating conditions are presented in Table 15.0.2-2 and principal fuel design characteristics in Table 15.0.2-3. The following uncertainties were used in the Reference 3 accident analysis and are applicable to Cycle 1o:

Core Power +/-2%

Primary Coolant Temperature +/- 5 °F Primary Coolant Pressure +/- 50 psi Primary Coolant Flow +/-3%

The maximum PCS mass flow rate supported by this analysis is defined by that 1

I assumed in the main steam line break analysis (i.e., 145.9 Mlbm/hr at 554 °F and 2060 psia)C3).

This is equivalent to 150.3 Mlbm/hr at 532 °F and 2060 psia.

I I

EMF-91-176 Page 19 TABLE 15.0.2-1 PLANT OPERATIONAL MODES Average Mode(a) Reactivity Power(b) Core Temperature Rated Power (1) Critical 100% (2530 MWt) >525°F

  • Power Operation (2) Critical > 2% >525°F Reactor Critical (3) Critical >10-4% >525°F Hot Standby (4) Any* Withdrawn Rod 10-43 to 2% >525°F Hot Shutdown (5) Shutdown Margin <10-4% >525°F

>2%Ap

--~'

Refueling Shutdown Shutdown margin of at 0 <210°F Condition (6) least 5%A p with all control rods withdrawn Cold Shutdown kett _s.98 with all <210°F Condition (7) control rods in the core and the highest worth control rod fully withdrawn t Refueling Operation Any operation (8) involving movement of core components when the vessel head is unbolted or removed a

Mode numbers are given in parenthesis.

b Does not include decay heat.

II I

TABLE 15.0.2-2 EMF-91-176 NOMINAL PLANT OPERATING CONDITIONS Page 20

  • I I

Core Thermal Power (MWt)

Pump Thermal Power, total (MWt) 2530 15 I

System Pressure (psia)

Vessel Coolant Flow Rate(a)(Mlbm/hr) 2060 138.6 I

Core Coolant Flow Rate(b)(Mlbm/hr)

Core Inlet Coolant Temperature (°F) 134.4 543.65 I

Steam Generator Pressure (psia) 722 I Steam Flow Rate (Mlbm/hr) 10.97 Feedwater Temperature (°F)

Number of Active Tubes per Steam 435 6986 I

Generator(a) 1 a Reflects 15.0% average.steam generator tube plugging.

b Reflects a 3% bypass flow.

I I

TABLE 15.0.2-3 EMF-91-176 Page 21 NOMINAL RELOAD N FUEL DESIGN PARAMETERS I

I Total Number of Fuel Assemblies Fuel Assembly Design Type

  • 204 15 x 15 I Fueled Rods per Assembly Instrument Tubes per Assembly 216 1

I Guide Bars per Assembly Assembly Pitch (inches) 8 8.485 I Rod Pitch (inches)

Fuel Pellet Outside Diameter (inches) 0.550 0.3510 I Clad Inside Diameter (inches)

Clad Outside Diameter (inches) 0.358 0.417

'*I Active Fuel Length (inches)

Number of Spacers 131.8 10

EMF-91-176 Page 22 15.0.3 Power Distribution The radial and axial power peaking factors used in the analysis are presented in Table I

15.0.3-1. The analyses for the inlet temperature LCO and for the TM/LP trip utilize axial power distributions and associated ASls. The axial power distributions are generated from a three- I dimensional core physics model. Figure 15.0.3-1 shows the DNB limiting axial shape for 100%

power events. This axial shape has an axial shape index (ASI) of -0.112. This axial shape is I representative of a full power hot channel shape, an associated excore ASI and the control rod insertion limits at full power. In this context, ASI is defined as: I p LDWBI' - p UpfW I

p LDw* + p UpfW I

Plower corresponds to the power generated in the lower half of the core and Pupper corresponds to the power generated in the upper half of the core. *'I MDNBR occurs on an interior pin of a Reload N assembly. The Technical Specification(7) Limiting Conditions of Operation (LCO) ensure that the power distribution is maintained within these limits during normal operation. However, some events analyzed result in transient redistribution of the radial power peaking factors. Transient radial power redistribution is treated as described in Section 15.4.3.2 of Reference 2.

I I

TABLE 15.0.3-1 CORE POWER DISTRIBUTION EMF-91-176 Page 23 I

I Radial Peaking Factor: Assembly {F r':l Peak Rod {F r:1 208 fuel rods/assembly 1.48 1.92 I 216 fuel rods/assembly

  • Reload L(a) 1.66 2.03
  • Reload M(a) 1.57 1.92
  • Reload N 1.66 1.92 Axial Peaking Factor See Figure 15.0.3-1 Engineering Tolerance Uncertainty 1.03 Fraction of Power Deposited in the Fuel .974

'*I

  • a The Cycle 9 transient analysis (Reference 3) addresses the radial peaking limits for Reload L and M.

I EMF-91-176 Page 24 1.4

(..

1.2 Q)

s 0

o_

_, 1.0 0

x a:

"'O 0.8 Q)

N 0

e

(.. 0.6 z

0 I

0.4 I

0.2 o.o i......a........_..._...._........._._....._.................._.__._..........................._.._........_._....._..._.._._.__._..._._._._....._..._._._.__..._..._........

0.0 0.1 0.2 0.3 0.4 O.S 0.6 0.7 0.8 0.9 l.O f roclLon of AclLve fuel HeLghl FIGURE 15.0.3-1 LIMITING AXIAL POWER SHAPE (ASI = -0.112)

I I EMF-91-176 Page 25 1*

15.0.4 Range of Plant Operating Parameters and States I Table 15.0.4-1 presents the range of key plant operating parameters considered in the Reference 3 transient analysis. A broader range of power, core inlet temperature, and primary I pressure is considered in establishing the trip setpoints verified by the analysis results presented in this document. These are unchanged for Cycle 1O.

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I

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  • I EMF-91-176 Page 26 TABLE 15.0.4-1 RANGE OF KEY INITIAL CONDITION OPERATING PARAMETERS I Core thermal power Subcritical to 2580.6 MWt(a)

I Inlet coolant temperature (Power operation)

Primary coolant system pressure T(in) LCO +/- 5 °F 2060 psia +/- 50 psi I

Pressurizer water level Feedwater flow and temperature Programmed +/- 5% of level span Range consistent with power level I

I I

  • 'I a 1.02 x 2530 MWt

I EMF-91-176 Page 27 15.0.5 Reactivity Coefficients Used in the Safety Analysis Table 15.0.5-1 presents the reactivity coefficients used in the reference transient analy-sis(3). The analysis conservatively supports the Technical Specification moderator temperature coefficient {MTC) of < +0.5 x 104 Ji p/°F. The nominal Cycle 1O burnup is 11,090 MWd/MTU, however, the safety analysis is applicable to an end-of-Cycle 1Oexposure of 11,373 MWd/MTU.

le I

I EMF-91-176 Page 28

.1 I

TABLE 15.0.5-1 REACTIVITY PARAMETERS I

Item BOC Bounding EOC Bounding I Moderator Temperature Coefficient, 1o-4 ll. p/°F 0.5 -3.5 Doppler Temperature Coefficient, 1o-5 ll. p!°F -1.09 -1.76 I Moderator Pressure Coefficient, 1o-6 ll. p/psi -1.0 7.0 Delayed Neutron Fraction 000075 0.0045 I Effective Neutron Lifetime, 1o~ seconds 41.9 19.9 U238 Atoms Consumed per Total Atoms Fissioned 0.54 0.70 el I

  • 15.0.6 Scram Insertion Characteristics EMF-91-176 Page 29 The insertion worth of 2.0% IJ. p and a control rod drop time of 2.5 seconds (to 90%

insertion} have been supported by the transient analysis used for Cycle 1o. Figure 15.0.6-1 presents the negative insertion used in the reference transient analysis(3) for reactor trip. The insertion worth includes the most reactive rod stuck out.

I le I

EMF-91-176 Page 30

..,:n 0.8 0

0 Q) a::: 0.6 e

0

(..

0 (J1 "U

Q)

N

-> O.i 0

e

(..

0 1

z 0.2 O.Ol_..._......__._..i.:.._--=:::i::=:::i:::::i':===-_.__,___.___.,___.__JL..._..__....._....._...L.-_._.....__,__J l

I o.o o.s 1.0 l.5 2.0 2.5 3.0 TLme ofler Scram, seconds FIGURE 15.0.6-.1 INTEGRATED SCRAM WORTH WITH MOST REACTIVE ROD STUCK OUT J

EMF-91-176 Page 31 15.0.7 Reactor Protection System Trip Setpoints and Time Delays Table 15.0.7-1 presents the trip setpoints, uncertainties, and time delays used in the Reference 3 analysis. The Reference 3 setpoint and transient analysis is applicable to Cycle

10. The setpoints used are discussed the section describing each transient in Reference 3.

An inlet temperature limiting condition of operation, Tinlet LCO, and thermal margin/low pressure (TM/LP) trip were developed for Cycle 9. The analysis performed herein confirms their continued applicability to the Cycle 1O core. Their development, unchanged from Reference 3, and the results of the Cycle 1Overification analyses are presented in the following sections. The Tinlet LCO described in Section 15.0.7.1 was used to develop the initial conditions used in the reference transient analyses. The TM/LP trip described in Section 15.0.7.2 was included in the reference transient simulations.

15.0.7.1 Inlet Temperature Limiting Condition of Operation The Tinlet LCO provides protection against penetrating DNB during limiting AOO transients from full power operation. The most limiting AOO transient that does not produce a reactor trip is the inadvertent drop of a full length control assembly. Therefore, the Tinlet LCO must provide DNB protection for this transient assuming a return to full power with enhanced peaking due to the anomalous control assembly insertion pattern.

The Tinlet LCO was developed in Reference 3 using the XCOBRA-lllC computer code( 13>

with augmented radial peaking. The XCOBRA-lllC calculations were run to determine the inlet temperature which resulted in a DNB equal to the 95/95 safety limit for a range of pressurizer pressures, core power levels and primary coolant system flow rates. These calculations were performed over a range of ASls from -0.63 to +0.63.

The results of the above analysis correspond to plant measured values of pressurizer pressure, primary coolant system flow rate and the inlet temperature and includes a 2% power and a +.06 ASI measurement uncertainty. These results must, therefore, be biased to account

  • for both measurement uncertainty and variations due to the control assembly drop transient.

EMF-91-176 Page 32 The Cycle 1O verification analysis included the same uncertainties and transient allowances as in Reference 3, namely: (1) a +50 psia pressure measurement uncertainty, (2) a + 7 °F inlet temperature uncertainty (Reference 17, +5 °F tilt allowance + 2 °F measurement uncertainty),

(3) a +6 percent to the flow rate (+/-3% bypass flow+ 3% measurement uncertainty) and, (4) 1 a -20 psia transient pressure bias. Accounting for measurement uncertainties and transient biases, the Tinlet LCO equation is as follows:

T1n1et ~ 542.99 + :O._OS~Ox(P - 2060) + 1.0x1 o-5 x(P - 2060)2

+ 1.125x(W - 138) - 0.0205x(W - 138)2 From Reference 3, the above equation is applicable for pressurizer pressures (P) between 1800 and 2200 psia and primary coolant system (PCS) mass flow rates rt') between 100 and 150 Mlbm/hr. For primary loop flow rates greater than 150 Mlbm/hr, the inlet temperature is limited to the Tinlet LCO value at 150 Mlbm/hr. The Tinlet LCO is applicable for measured ASI in the range from +0.40 to -0.08 and can be compared to an average cold leg temperature for each of the four loops. The applicability of the Tinlet LCO equation was extended to a measured ASI of-0.30 at 70% rated power and -0.55 at 25% rated power. The applicable range of the Tinlet LCO is shown in Figure 15.0.7-1.

Calculations were performed for Cycle 1O using the XCOBRA-lllC computer _code and a conservative peaking augmentation factor to demonstrate the continued applicability of the above Tinlet LCO equation. The XCOBRA-lllC calculations were run to demonstrate that the inlet temperature allowed by the Tinlet LCO results in a DNBR greater than the 95/95 limit for the ANFP correlation over a range of pressurizer pressures, primary coolant system flow rates, axial shape indices and core power levels.

I I EMF-91-176 Page 33 r*

15.0.7.2 Thermal Margin/Low Pressure ITM/LP) Trip

~ The function of the TM/LP trip is to protect the core against slow heatup and depressurization transient events. In order to perform this function, the TM/LP trip must initiate J' a scram signal prior to exceeding the specified acceptable fuel design limits (SAFDLs) on departure from nucleate boiling (DNB) or before the average core exit temperature exceeds

~ the saturation temperature. The SAFDL ensures that there is no damage to the fuel rods and the limit on core exit saturation ensures meaningful thermal power measurements.

[i The NSSS is protected against penetrating DNB during rapid power, flow, and pressure l transient events by the variable high power (VHP) trip, the low flow trip and the high pressurizer pressure trip, respectively. For slow transient events, however, it is possible that either the SAFDL on DNB or hot leg saturation could be reached prior to activating these trip setpoints. These slow transients generally involve a slow heatup of the primary coolant system caused by: (1) a power mismatch between the primarY and secondary systems or, (2) a slow depressurization of the primary system with or without a slow power ramp. Transient events that exhibit these characteristics and must, therefore, be protected by the TM/LP include an uncontrolled rod withdrawal, an inadvertent boron dilution, an excess load, a loss of feedwater .1 or a PCS depressurization. The TM/LP trip works in conjunction with the other trips and the LCO on control rod group position, radial peaking and primary coolant flow.

The functional form for the Palisades TM/LP trip is:

where: Pvar is the low pressure trip limit; ex, p, and y are constants to be determined; Teal is the highest measured cold leg temperature adjusted for possible coolant stratification in the cold leg; and, QDNB is a function representing axial and radial power peaking effects. The adjusted cold leg temperature Teal is calculated from:

r-J EMF-91-176 .1 Page 34

  • 1 J

1 where B is the measured 4 T power, Kc is a flow stratification factor, and Tin is the highest measured cold leg temperature in each steam generator. For NSSS like Palisades, which have J the cold leg temperature sensors located downstream of the primary coolant pumps, there is sufficient mixing so that ~ = 0.0 and T cal equals Tin* J The OoNB function is represented as:

J QDNB = QAxQR1 where QA is a function representing the variation in power versus axial shape at constant DNB, and OR 1 is a function representing the variation in power with radial p_eaking and/or hot leg saturation.

15.0.7.2.1 TM/LP Uncertainties In setting the TM/LP trip it is necessary to conservatively account for the uncertainties in the measured parameters used to determine the trip. These uncertainties result not only from the inability of the instrumentation to exactly measure the value of a parameter, but also from the fact that the changes in the parameters being measured may actually lag behind the event of interest. Therefore, in any transient uncertainty analysis, both static and transient effects must be considered.

The input parameters for the TM/LP for which uncertainties must be determined and accounted for in the trip development are inlet temperature, core power, pressure and axial shape index. The uncertainty applied to the pressure in this analysis is 165 psi<18>. This uncertainty was developed to account for most of the uncertainties in the TM/LP. Included in

I I

f, EMF-91-176 r* Page 35 this 165 psi are: instrument drift in both power ~nd inlet temperature, calorimetric power (1

measurement, inlet temperature measurement and, primary pressure measurement. The uncertainties in these parameters will, consequently, not be treated separat~ly in this analysis.

~

An additional uncertainty, not accounted for in the 165 psi, is associated with the inlet l* temperature. This uncertainty accounts for the lag time in the RTDs {12 seconds) and primary coolant system transit times. These times were converted to temperature uncertainties through t the use of a typical temperature ramp for a slow rod withdrawal. The inlet temperature uncertainty for these time delays was found to be bounded by 1.5°F.

l Also, uncertainties used in this analysis include a 3% uncertainty to account for

,, manufacturing tolerances and a 6% uncertainty in PCS flow to account for 3% core bypass flow and 3% measurement uncertainty.

le The final uncertainty is in the measured ASI used in the TM/LP. This uncertainty was taken to be +0.06 consistent with the ASI uncertainty assumed for other Combustion Engineering plants. The uncertainties applied in the development of the TM/LP in Reference u 3 are summarized in Table 15.0.7-2.

15.0.7.2.2 TM/LP Development In the actual development of the TM/LP in Reference 3, a definite step-by-step procedure is followed. First, the axial shape function QA is developed. This is followed by the determination of the radial peaking function QR 1. F!nally, the coefficients a, p, and y are derived. Throughout this development the various uncertainties are applied to ensure that the final TM/LP function is conservative.

The first function derived, QA, is the axial shape function and corrects the TM/LP for variations in power with axial shape index (ASI) at a constant DNBR value. This function was generated by finding the limiting axial shapes from 756 axial shapes. The limiting axial shapes cover an ASI range from -0.63 to +0.63. The limiting axial shapes were used in the XCOBRA-

l EMF-91-176 Page 36 .i 1

lllC model to determine the power level required to reach a DNBR equivalent to the correlation 1) safety limit. These results are conservatively fitted as a function of ASI with three straight lines with maximized slopes. The QA function is derived by normalizing the straight line functions to the peak power and inverting the result. The derived QA function is plotted in Figure 15.0.7- 1

2. Note that this QA function has been developed to cover the full range of the possible ASl's.

J The radial peaking function, QR1, accounts for the changes in slope of the core protection. limit lines with radial peaking and hot leg saturation. The core protection limits are J parallel equally spaced lines representing the variation in the maximum allowed inlet temperature with power and pressure. These lines are composed of two limiting portions: the J first, which dominates at low powers, is the inlet temperature which produces saturation in the hot leg; and, the second is the inlet temperature which produces the minimum allowed DNBR.

The core protection limits for an ASI of +0.200 (QA = 1.00) are given in Figure 15.0.7-3.

l Adjustments were made to the trip coefficients to ensure that the trip conservatively el bounds the core protection limits. The TM/LP trip equation developed in Reference 3 is as follows:

[

l EMF-91-176 r* Page 37

(, Pv* = 2012x(QA)x(OR1) + 17.0x(T.J - 9493 t QR1 = 0.412x(Q) + 0.588 for Q ~ 1.0 QR1 = Q for Q > 1.0 I/

QA = -0.120x(AS~ + 1.028 for -0.628 ~AS/< -0.100 t'

= -0.333x(AS~

l . QA + 1.067 for.-0,100 .~ AS/< +0.200 QA = +0.375x(AS~ + 0.925 ~ ~

l for +0.200 AS/ +0.565 le Pvar is defined as the low pressurizer pressure trip limit, QA is the axial shape function, QR1 is the radial peaking function, Tin is the highest measured cold leg temperature, Q is the fraction of rated power and ASI is the axial shape index. From Reference 3, this TM/LP equation is applicable over a pressure range from 1700 psia to 2300 psia and to a minimum measured HZP primary coolant flow rate of 140.7 Mlbm/hr. Figure 15.0.7-2 shows the QA function. The core protection limits for an ASI of +0.200 are shown in Figure 15.0.7-3.

The TM/LP trip function was verified for Cycle 1O by first determining a set of limiting axial shapes. The limiting axial shapes covered the ASI range defined by the Tinlet LCO. The limiting axial shapes were used in the XCOBRA-lllC model to ensure that the MDNBR for conditions allowed by the TM/LP trip function is greater than the ANFP correlation 95/95 limit.

Thus, the TM/LP trip was verified to be applicable over the possible range of axial shapes for Cycle 10.

EMF-91-176 Page 38

  • 15.0.7.3 Variable High Power NHP) Trip Reference 3 (Section 15.0.7.3) discussed the VHP trip uncertainties employed in the reference transient analyses. The following discussion reiterates that in Reference 3.

From Table 15.0.7-1, the VHP trip uncertainty is given as +8.5%. A +3.0% allowance that accounts for potential differences between the AT and neutron flux power measurements is included. The reactor protection system (RPS) circuitry uses an auctioneered high value of AT and neutron flux power_to set the VHP trip setpoint. If the neutron flux power is 3.0% less than the ATpower, theVHP trip setpoint is based on the 4 T power. For fast' transients in which the AT power response lags behind the neutron power, a 3.0% AT-neutron flux power difference can delay the occurrence of a VHP trip. l)lus, for transient events that are protected by the VHP trip, an uncertainty of +8.5% is used.

-For "slow transients, however, the plant is protected by either the VHP or TM/LP trip.

During these events, the thermal lags associated with the calculation of AT power are less than the overall time of the event, i.e., AT power is able to track the neutron flux power. Thus, a 3.0% difference in AT-neutron flux power will not impact the time of trip since the trip setpoint

  • 1~

I

~

is based on the higher of AT or neutron power. The effective VHP trip uncertainty under these conditions.is +5.5%.

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EMF-91-176 Page 39 TABLE 15.0.7-1 TRIP SETPOINTS FOR OPERATION AT 2530 MWT Trip Setpoint Uncertain~ Delay Time Low Reactor Coolant Flow 95% of rated four + 2.0% 0.6 sec pump flow High Pressurizer Pressure 2255 psia + 22 psi 0.6 Low Pressurizer Pressure 1750 psia + 22 psi 0.6 Low Steam Generator Pressure 500 psia + 22 psi 0.6 Low Steam Generator Level(a) 6 feet + 10 inches 0.6 Thermal Margin/Low Pressure(b) P = f(TH Tc) 1 + 165 psi 0.6 Variable High Power 106.5% maximum + 8.5% 0.4 30.0% minimum 15.0% above thermal :i":

power a Below operating level.

b The thermal margin trip setpoint is a functional pressurizer pressure (P) setpoint, varying as function of the maximum cold leg temperature (Tc). the measured power, and the measured axial shape Index.

EMF-91-176

  • Page 40 TABLE 15.0.7-2 TM/LP UNCERTAINTIES Instrument Drift (Power, Tinlet)

Calorimetric Power Tinlet measurement 165 psi Pressure Measurement RTD Measurement

  • Engineering Tolerances 3%

Primary Coolant Flow 6%

Inlet Temperature Time Delay 1.5°F Axial Shape Index +.06 I

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  • l

l r EMF-91-176 Page 41 r/

t 1.15

~ Unocceploble Operoli..ons 5

3:

1.00

c (1 0 n

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J EMF-91-176 l

Page 44

  • 1 15.0.8 Component Capacities and Setpoints Table 15.0.8-1 presents the component setpoints and capacities supported by this

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analysis. These are unchanged from the Reference 3 analysis.

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  • TABLE 15.0.8-1 COMPONENT CAPACITIES AND SETPOINTS Response Nominal Setpoint Time Setgoint Uncertainty Total Capacity Pressurizer Safety Valves (3) 2500 psia 75 psia 191 lbm/sec 2540 psia 2580 psia Pressurizer Relief Valves (2) 2 sec 2400 psia 50 psia 319 lbm/sec Steam Line Relief Valves (24) Group A at 1000 psia 3% 3244 lbm/sec at Group B at 1020 psia 1000 psia Group C at 1040 psia Turbine Stop and Control Valves 0.1 sec Steam Dump Valves and Turbine 3.0 sec Turbine trip then Tavg 1173 lbm/sec at Bypass program 770 psia Pfessurizer Backup Heaters Always on 1350 kW Pressurizer Proportional Heaters Full on- 201 O psia 50 psia 150 kW Full off- 2060 psia 50 psia Pressurizer Sprays Full on- 211 O psia 50 psia 29.4 lbm/sec (1.5 Full off- 2060 psia 50 psia gpm continuous flow)

Letdown Orifice Valves Level controller 12.6 lbm/sec CVCS Makeup System Level controller 18.5 lbm/sec Normal Feedwater system 20.5 sec Feedwater controller 3321 .4 lbm/sec m

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  • 1 15.0.9 Plant Systems and Components Available for Mitigation of Accident Effects Table 15.0.9-1 is a summary of trip functions, engineered safety features, and other ,J equipment available for mitigation of accident effects. These are listed for all Chapter 15 SAP events and are unchanged from the Reference 3 analysis.

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TABLE 15.0.9-1 OVERVIEW OF PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR TRANSIENT AND ACCIDENT CONDITIONS, CONTINUED SAP Designation Event Reactor TriQ Functions Other Signals[EguiQment 15.2 Decrease In Heat Removal by the Secondary System 15.2.1 Loss of External Load/Turbine High Pressurizer Pressure Trip

"*'* Low Steam Generator Pressure Trip Auxiliary Feedwater System

" Pressurizer Sprays and level Control

  • TABLE 15.0.9-1 OVERVIEW OF PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR TRANSIENT AND ACCIDENT CONDITIONS, CONTINUED SAP Designation Event Reactor-TrlE! Functions Other Slgnals[EgulE!ment 15.3 Decrease In Reactor Coolant System Flow Rate 15.3.1 Loss of Forced Reactor Coolant Low Reactor Coolant Flow Trip Atmospheric Steam Dump Controller Flow TM/LP Trip Steam Bypass to Condenser Controller High Pressurizer Pressure Trip Steam Generator Safety Valves Pressurizer Safety Valves 15.3.3 Reactor Coolant Pump Rotor Low Reactor Coolant Flow Trip Atmospheric Steam Dump Controller 15.3.4 Seizure/Shaft Break High Pressurizer Pressure Trip Steam Bypass to Condenser Controller Steam Generator Safety Valves Pressurizer Safety Valves 15.4 Reactivity and Power Distribution Ano ma lies 15.4.1 Uncontrolled Control Rod Bank TM/LP Trip Non-safety Grade High Rate-of-Change Withdrawal from a Subcrltlcal or Low VHP trip of Power Trip Power Startup Condition High Pressurizer Pressure Trip High Rate-of-Change of Power Alarms, which Initiate Rod Withdrawal Prohibit Action 15.4.2 Uncontrolled Control Rod Bank VHP trip Pressurizer Safety Valves Withdrawal at Power Operation TM/LP Trip Steam Generator Safety Valves Conditions High Pressurizer Pressure Trip Pressurizer Spray and Level Control Control Rod and Bank Deviation Alarms m

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Plant Operation with less than all four primary coolant pumps Is not permitted by Technical Specifications except for very short periods of time and at reduced power levels (Tech Spec Table 2.3.1).

15.4.6 Chemical Volume and Control VHP trip Non-safety Grade High Rate-of-Change System (CVes) Malfunction that TM/LP Trip of Power Trip Results In a Decrease In the Boron High Pressurizer Pressure Trip Administrative Procedures Concentration In the Reactor Coolant Sufficient Operator Response Time 15.4.7 Inadvertent Loading and Operation Technical Specification measurement of a Fuel Assembly In an Improper requirements and administrative Position procedures preclude occurrence 15.4.8 Spectrum of Control Rod Ejection VHP trip Non-safety Grade High Rate-of-Change Accidents TM/LP Trip of Power Trip Long Term, Safety Injection Actuation eves Signal m

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EMF-91-176

  • Page 52 15.0.10 Effects of Mixed Assembly Types and Fuel Rod Bowing The Cycle 1O core contains only fuel assemblies manufactured by SNP.
  • The fuel assemblies prior to Reload M (Cycle 9) used bi-metallic spacers. Reloads M and N, however, utilize High Thermal Performance (HTP) spacers. Due to the inclusion of these assemblies, a 2% mixed core DNBR penalty is applied(S). The assemblies with bi-metallic spacers are highly burnt and are, therefore, not limiting with respect to DNBR. For assemblies with HTP spacers, DNBRs are calculated with the ANFP critical heat flux correlation(9 *10) with a 95/95 safety limit of 1.t54. Adjusting for .the 2% mixed core penalty, the effective DNB correlation limit used in this analysis for the ANFP correlation is 1.174.

The effects of rod bow on limiting DNB and heat flux peaking were considered.

Reference 11 concludes that due to the short distances between spacers, the 15 x 15 design does not exhibit fuel rod bow of any significance to plant operating margins. Therefore, no penalty is applied due to rod bow effects. * *

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  • 15.0.11 Plant Licensing Basis and Single Failure Criteria EMF-91-176 Page 53 The licensing basis for Palisades is as stated in the Final Safety Analysis Report(6). The event scenarios depend on single failure criteria established by the plant licensing basis.

Examination of the Palisades licensing basis yields the following single failure criteria:

(1) The RPS is designed with redundancy and independence to assure that no single failure or removal from service of any component or channel of a system will result in the loss of the protection function.

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

(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 SAFDls riot be exceeded during Anticipated Operational Occurrences (AOOs). As per the definition of AOC in 1o 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

EMF.:91-176

  • Page 54 of the plant and. include but are not limited to loss of pow.er to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power".

The SAFDLs are that: (1) the fuel shall not experience centerline melt, i.e., linear heat rate, LHR, to be less than 21 kW/ft, and (2) the DNBR shall have a minimum allowable limit such that there is a 95% probability with a 95% confidence interval that DNB has not occurred, i.e.,

DNBR to be greater than the effective correlation limit for ANFP.

EMF:.91-176 Page 55 The following sections, numbered according to the SRP, provide a discussion of the disposition of events review and MDNBR analyses performed to support Cycle 1O operation.

r 15.1 Increase in Heat Removal by the Secondary System The magnitude of the decrease in feedwater temperature, increase in feedwater flow r rate, increase in steam flow and inadvertent opening of a steam generator relief/safety valve for Events 15.1.1 , 15.1.2, 15.1 .3, and 15.1 .4, respectively, is not affected by the changes for t Cycle 1o in Section 1.0. Therefore, the relative PCS cooldown rate and severity of each of the above events remains unchanged from previous event dispositions beginning with Cycle a<19>.

l For this category of events, the Increase in Steam Flow event (15.1.3) remains bounding of events 15.1.1, 15.1 .2 and 15.1.4. The MDNBR for Event 15.1 .3 will be reanalyzed for the *Cycle 1O changes given in Section 1.0.

r te 15.1.1 Decrease in Feedwater Temperature 15.1.1.1 Event Description A decrease in feedwater temperature event may result from the loss of one of several r of the feedwater heaters. This loss may be due to the loss of extraction steam flow from the turbine generator or due to an accidental opening of a feedwater heater bypass line.

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

15.1.1 .2 Event Disposition and Justification

~I As noted in Section 15.1, this event is bounded by the Increase in Steam Flow event (Event 15.1.3).

EMF-91-176 Page 56

  • 15.1 .2 Increase in Feedwater Flow 15.1.2.1 Event Description The Increase in Feedwater Flow event is initiated by a failure in the feedwater system.

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

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

15.1 .2.2 Event Disposition and Justification As noted in Section 15.1 , this event is bounded by the Increase in Steam Flow event (Event 15.1.3).

15.1 .3 Increase in Steam Flow 15.1.3.1 Event Description The increase in steam flow event is initiated by i;ln increase in steam demand. The increased steam demand may be initiated by the operator or by regulating valve malfunction.

The step increase in steam flow results from a rapid opening of the turbine control valves, atmospheric dump valves or the turbine bypass valve to condenser.

The event initiator is a step increase in steam flow. The feedwater regulating valves open to increase the feedwater flow in an attempt to match the increased steam demand and maintain steam generator water level. In response to the increased steam flow, the secondary system pressure decreases, resulting in an increase in the primary-to-secondary heat transfer rate. The primary side steam generator outlet temperature decreases due to the enhanced

EMF-91-176 Page 57 heat removal. As a consequence, the primary system core average temperature decreases and the primary system fluid contracts, resulting in an outsurge of fluid from the pressurizer.

The pressurizer level and pressure decrease as fluid is expelled from the pressurizer. If the I MTC is negative, the reactor core power increases as the moderator temperature decreases due to the mismatch between the power being removed by the steam generators and the t power being generated in the core.

t TM/LP and VHP trips are available to prevent the violation of the acceptance criteria.

Depending on the magnitude of the increase in steam demand, a reactor trip may not be l activated. Instead, the reactor system will reach a new steady-state condition at a power level greater than the initial power level which is consistent with the increased heat removal rate.

The final steady-state conditions which are achieved will depend upon the magnitude of the r MTC. If the MTC is positive, the reactor power would decrease as the core average coolant temperature decreased, and this event would not produce a challenge to the acceptance criteria.

This event is a moderate frequency event (Table 15.0.1-1). The acceptance criteria for this event are listed in 15.0.1.1. Single failure criteria for Palisades are given in 15.0.11. For this analysis, the systems challenged in this. event are redundant; no single active failure in the

  • RPS or ESF will adversely affect the consequences of the event.

15.1.3.2 Event Disposition and Justification The parameters controlling the severity of the transient are: (1) the secondary steam flow (load), (2) the primary-to-secondary heat transfer, (3) the moderator reactivity coefficient, (4) the Doppler reactivity coefficient, (5) the reactor safety system setpoints, and (6) the scram reactivity. The reference analysis(3) evaluated the severity of the transient at full power. As shown in analyses for previous cycles, the transient initiated from hot full power bounds all other modes of operation. Because of the increase in FrA to 1.66 for Reload N, a DNBR analysis is required.

EMF-91.-176 15.1.3.3 Definition of Events Analyzed Page 58 This event is predominantly a depressurization event and is evaluated at full power conditions. At full power, the margin to limits is the smallest and, therefore, bounds operation at lower power levels. The end-of-cycle moderator and Doppler feedback coefficients were selected to maximize the challenge to the SAFDLs. The time in the cycle will determine the value of the MTC. If the MTC is negative, there will be a positive reactivity insertion, the magnitude of which is dependent upon the magnitude of the MTC. If the MTC is positive, then negative reactivity will be inserted as the coolant temperature decreases, causing the power to decrease with less challenge. The reactor control rod system at Palisades is disabled so

~

that the control rods will not withdraw automatically in response to the decrease in core average temperature. Therefore, the consequences of this event ~re bounded at end-of-cycle conditions when the MTC is at its maximum negative value.

.1 15.1 .3.4 Analysis Results The transient response of the reactor system was calculated using the PTSPWR2(12) computer program in Reference 3, including plots of key system variables and a sequence of events. The core thermal hydraulic boundary conditions from the PTSPWR2 calculation are used as input to the XCOBRA-lllC code(13) to predict the MDNBR. Based on the peaking factors given in Table 15,0.3-1 for Reload N fuel, the MDNBR for this event is 1.812. The peak pellet LHR is 15.00 kW/ft.

1 15.1.3.5 Conclusion The results of the analysis demonstrate that the event acceptance criteria are met since .)

the predicted MDNBR is greater than the safety limit The critical heat flux correlation limit ensures that, with 95% probability and 95% confidence, DNB is not expected to occur; therefore, no fuel is expected to fail. The fuel centerline melt threshold of 21 kW/ft is not penetrated during this event.

I I

1. EMF-91-176 Page 59 I

15.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve l 15.1.4.1 Event Description This event is initiated by an increase in steam flow caused by the inadvertent opening of a steam generator relief or safety valve. The increase in steam flow rate causes a mismatch between the heat generation rate on* the primary side and the heat removal rate on the

  • secondary side.

15.1.4.2 Event Disposition and Justification The increase in steam flow due to opening a steam generator valve is less than that considered in the Increase in Steam Flow event (Event 15.1.3), and therefore is bounded by Event 15.1 .3. -***

15.1.5 Steam System Piping Failures Inside and Outside of Containment 15.1.5.1 Event Description The steam line break event is initiated by a double ended guillotine break of the main .

steam line which leads to an uncontrolled steam release from the secondary system. The *

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increase in energy removal through the secondary system results in a severe overcooling of the primary system. In the presence of a negative MTC, this cooldown results in a large decrease in the shutdown margin and a return to power. This return to power is exacerbated because of the high power peaking factors which exist as result of the most reactive control rod being assumed to be stuck in a fully withdrawn position.

15.1 .5.2 Event Disposition and Justification

- The consequences of this event are controlled by the following: (1) steam flow rate out of the ruptured steam line, (2) primary-to-secondary heat transfer, (3) the primary pump assumptions (i.e., with or without offsite power), (4) moderator reactivity coefficient, (5) Doppler reactivity coefficient, (6) core power, and (7) Technical Specification shutdown margin. The consequences of a main steam line break are dependent on the competing effects of shutdown reactivity as a result of a reactor trip and the positive reactivity insertion due to PCS overcooling. The system response for this event remains unchanged from Cycle 9. In

EMF-91-176

  • Page 60 addition, the Cycle 1O changes do not affect the release path to the environment. The changes incorporated in Cycle 1O will not adversely impact MDNBR or peak LHR (e.g., the radial peaking at the time of MDNBR used in Reference 3 bounds that for Cycle 10).

Therefore, the amount of fuel failure calculated in the reference analysis<3> remains applicable for Cycle 1O.

The fuel failure analysis in Reference 6 is based on a steam line break inside of containment. Because of the outlet flow restrictors, the thermal-hydraulic response for a break inside containment will be identical to that for a break outside of containment. Further, the amount of fuel failure will be identical for the two break locations. Since the changes introduced by Cycle 1O will not adversely impact the amount of fuel failure relative to the Reference 3 analysis, the bounding radiological consequence analysis given in Reference 6 remains applicable to Cycle 1o.

J

EMF-91-176 I

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  • 15.2 Decrease in Heat Removal by the Secondarv System Page 61 J The initiating mechanisms for the loss of external load event (15.2.1 ), turbine trip event I (15.2.2), loss of condenser vacuum event (15.2.3), closure of the main steam isolation valves l

I (15.2.4) and the heatup phase of the feedwater system pipe break event (15.2.8) are not affected by any of the Cycle 1O changes. Therefore, the relative severity of these events as established in previous dispositions beginning with Cycle a(19) remains valid. The loss of load t:

event, 15.2.1, as biased in previous analyses(3,4) was found to be bounding of Events 15.2.2, l 15.2,3, 15.2.4 and the heatup period of Event 15.2.8.

15.2.1 Loss of External Load 15.2.1.1 Event Description A Loss of External Load event is initiated by either a loss of external electrical. load or a turbine trip. Upon either of these tWo conditions, the turbine stop valve is assumed to rapidly close (0.1 second). The plant response to this event is negligibly affected by assuming a faster valve closure time. Normally a reactor trip would occur on a turbine trip, however, to calculate a conservative system response, the reactor trip on turbine trip is disabled. 1:he steam dump system (atmospheric dump valves - ADVs) is assumed to be unavailable. These assumptions allow the Loss of External Load event to bound the* consequences of Event 15.2.2 (Turbine Trip - steam dump system unavailable) and Event 15.2.4 (Closure of both MSIVs - valve closure time is comparable to the turbine stop valve).

The Loss of External Load event primarily challenges the acceptance criteria for both primary and secondary system pressurization and DNBR. The event results in an increase in the primary system temperatures due to an increase in the secondary side temperature. As the primary system temperatures increase, the coolant expands into the pressurizer causing an increase in the pressurizer pressure. The primary system is protected against overpressurization by the pressurizer safety and relief valves. Pressure relief on the secondary side is afforded by the steam line safety/relief valves. Actuation of the primary and secondary system safety valves limits the magnitude of the primary system temperature and pressure increase.

i EMF-91-176 Page 62 1

  • 1 With a positive BOC.MTC, increasing primary system temperatures result in an increase in core power. The increasing primary side temperatures and power reduces the margin to thermal limits (i.e*., DNBR limits) and challenges the DNBR acceptance criteria.

1 15.2.1.2 Event Disposition and Justification I

The parameters influencing the severity of the transient include: (1) PCS high pressure .j trip setpoint, (2) PCS over pressure relief capacity, (3) Primary-to-secondary heat transfer, (4)

Secondary side press~re relief capability, (5) Moderator reactivity coefficients, and (6) Doppler reactivity coefficient. This event initiated from full power bounds all other operating modes.

I The reference analysis(3*17) evaluated the following two cases: maximum primary system overpressure and MDNBR. From Reference 17, the MDNBR case for this event is non-limiting

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relative to other AOOs. Since this is a pressurization event (increasing pressure increases DNBR) and other more limiting MDNBR events are being reanalyzed for Cycle 1O (e.g., the l

uncontrolled rod withdrawal), a MDNBR analysis for this event will not be performed for Cycle

10. The increase in FrA for Reload N will not adversely impact the evaluation of the transient eJ re~nalysis pressurizer response in Reference 3; thus, this subevent will not require 10.

for Cycle j

. 15.2.2 Turbine Trip I 15.2.2.1 Event Description This event is initiated by a turbine trip which results in the rapid closure of the turbine l stop valves. A reactor trip would occur on a turbine trip and the steam dump system would operate to mitigate the consequences of this event. The primary system is protected against l overpressurization by the pressurizer safety and relief valves. Pressure relief on the secondary side is afforded by the steam line safety/relief valves. l 15.2.2.2 Event Disposition and Justification J The following reactor conditions assumed for the Loss of External Load event (Event 15.2.1) are conservative relative the Turbine Trip event: (1) a conservatively fast turbine stop valve closure time, (2) no reactor trip coincident with the turbine trip, and (3) the atmosphe~ic

EMF-91-176 Page 63 dump valves are assumed to be unavailable. Therefore, this event is bounded by the Loss of External Load event.

l 15.2.3 Loss of Condenser Vacuum 15.2.3.1 Event. Description t This event is initiated by a reduction in the circulating water flow or an increase in the circulating water temperature which can impact the condenser back pressure. This condition l can result in a turbine trip without the availability of steam bypass to the condenser. The primary system is protected against overpressurization by the pressurizer safety and relief valves. Pressure relief on the secondary side is afforded by the steam line safety/relief valves.

15.2.3.2 Event Disposition and Justification At rated power and power operating modes, the assumptions made for the Loss of re External Load event (Event 15.2.1) are conservative relative to the Loss of Condenser Vacuum.

From operating modes other than rated power, the operator has sufficient time to control the l primary and secondary system temperatures. These dispositions will not change for Cycle 10..

15.2.4 Closure of the Main Steam Isolation Valves (MSIVl (BWR)

I 15.2.4.1 Event Description .

Closure of the Main Steam Isolation Valves event is initiated by the loss of control air l to the MSIV operator. The valves are swinging check valves designed to fail in the closed position. The inadvertent closure of the MSIVs is primarily a BWR event, however, the closure

[ of these valves in a PWR can drastically reduce the steam load.

15.2.4.2 Event Disposition and Justification Closure of the MSIVs is less than 5 seconds, but comparable to the value used in Event 15.2.1 (0.1 seconds). A MSIV closure event will progress in a similar fashion as a Loss of External Load (Event 15.2.1 ), but at a slower rate. The consequences of Event 15.2.1 will bound those for Event 15.2.4 because of the more rapid valve closure time. This disposition will not change for Cycle 1O.

I EMF-91-176 Page 64 .t I

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

J 15.2.6 Loss of Nonemergency A.C. Power to the Station Auxiliaries I 15.2.6.1 Event Description A Loss of Nonemergency A.C. Power to the Station Auxiliaries event may .be caused 1

by a complete loss of the offsite grid together with a turbine generator trip or by a failure in the J

onsite A.C. power distribution system.

t The loss of A.C. power may result in the loss of power to the primary coolant pumps I and condensate pumps which, in turn, results in the loss of the main feedwater pumps. The

. combination of the decrease in primary coolant flow rate, the cessation of main feedwater flow

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and trip of the turbine generator compounds the event consequences. The decrease of both primary coolant flow and main feedwater decreases the primary-to-secondary system heat e1 transfer rate resulting in the heatup of the primary system coolant. The increase in primary system coolant temperature increases the overpressurization potential and increases the threat J of penetrating DNB.

15.2.6.2 Event Disposition and Justification The event is mo~t limiting when initiated from full power conditions. During this mode .~

of operation the stored heat in the fuel rods is maximized and the margin to DNB is minimized.

This event can be separated into two distinct phases: the near-term and the long-term. The 1 near-term phase is characterized by the loss of power resulting in the coastdown of the primary coolant pumps, the coastdown of the main feedwater pumps and the trip of the turbine I generator. The coastdown of the primary coolant pumps causes an immediate reduction in thermal margin. The trip of the reactor and the subsequent insertion of control rods terminates J the challenge to DNB limits.

I t.

J EMF-91-176 Page 65 The near-term phase of the event is similar to that of a Loss of Forced Reactor Coolant t Flow transient (Event 15.3.1 ). The near-term consequences of this event are addressed in the analysis of Event 15.3.1.

l The long-term consequences of a Loss of A.C. Power event are determined by the heat i removal capacity of the auxiliary feedwater system. The long-term portion is similar to the Loss of Normal Feedwater Flow transient (Event 15.2.7). The long-term effects are, therefore, addressed by the disposition of the Loss of Normal Feedwater Flow event. The changes for r

Cycle 1O will not alter this disposition.

~i 15.2.7 Loss of Normal Feedwater Flow r 15.2. 7.1 Event Description A Loss of Normal Feedwater Flow transient is initiated by the trip of the main feedwater le pumps or a malfunction in the feedwater control valves. The loss of main feedwater flow decreases the amount of subcooling in the secondary-side downcomer which diminishes the l primary-to-secondary system heat transfer and leads to an increase in the primary system coolant temperature. As the primary system temperatures increase, the coolant expands 'into the pressurizer which increases the pressure by compressing the steam volume.

I

,, The opening of the secondary-side safety valves controls the heatup of the primary-side. The long-term cooling of the primary system is governed by the heat removal capacity of the auxiliary feedwater flow. The auxiliary feedwater pumps are automatically started upon r a steam generator low liquid level signal.

15.2.7.2 Event Disposition and Justification A Loss of Normal Feedwater Flow event is only credible for rated power and power operating conditions. The worst consequences occur when the feedwater is lost during rated power operation since more stored heat is contained in the fuel than in other modes of operation.

EMF-91-176 Page 66

  • For the initial PCS heatup phase of the transient, both the DNB and the primary system overpressurization acceptance criteria are challenged. The DNB challenge is maximized when it is assumed that offsite power is lost* causing the prir:nary coolant pumps to coastdown.

After the reactor trip system is activated, the core power is drastically reduced alleviating the challenge to DNB. The loss of forced reactor coolant flow event (Event 15.3.1) bounds the short term DNB consequences of a loss of normal feedwater transient.

For the longer term phase of the transient, the slow PCS heatup threatens both the overpressurization of the primary system by filling the pressurizer with liquid, and the dryout of the steam generators. The limiting cases in the reference analysis assumed an auxiliary feedwater flow rate of 300 gpm distributed to both steam generators.

. The parameters influencing the severity of the long term phase of the transient include:

(1) decay heat generation, (2) secondary safety/relief valve settings, (3) primary coolant pump operation, (4) auxiliary feedwater flow rate, and (5) steam generator secondary side mass at the time of reactor trip. The reference an~lysisC 1 7) evaluated the following three cases: (1)

Nominal PCS and secondary conditions, (2) Minimum steam generator inventory, and (3)

Maximized PCS pressure. The increase in FrA for Cycle 1owill not adversely impact this event.

Thus, the reference analysis remains bounding.

15.2.8 Feedwater System Pipe Breaks Inside and Outside Containment 15.2.8.1 Event Description A Feedwater System Pipe Break event occurs when a main feedwater system pipe is ruptured. The ruptured pipe will cause a blowdown of the affected steam generator if the break occurs upstream of the feedline check valve. If the rupture occurs downstream of the check valve, the event would behave much like the Loss of Normal Feedwater Flow transient.

Since the auxiliary feedwater flow is injected into the steam generators via a separate piping network than the main feedwater, the delivery of auxiliary feedwater will not be interrupted by the pipe rupture.

EMF-91-176 Page 67 The event results in both a primary system cooldown and a heatup. Initially, the event results in a cooldown of the primary-side coolant due to the energy removal during the blowdown stage of the event. The eventual depletion of secondary-side inventory and lack of main feedwater will cause the primary system to heatup much like a Loss of Normal Feedwater Flow event.

15.2.8.2 Event Disposition and Justification l The parameters controlling this event are the same* as those controlling the main steam line break event described in Section 15.1 .5. The cooldown consequences of this event are t bounded by the main steam line break event.

[ As noted in Section 15.2, the Cycle 1O changes will not alter previous dispositions for the heatup consequences of the event which was classified as bounded by the Loss of Load le Event (15.2.1). Therefore, this event remains bounded by Events 15.2.1 and 15.1.5.

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EMF-91-176 Page 68 15.3 Decrease in Reactor Coolant System Flow I

15.3.1 Loss of Forced Reactor Coolant Flow 15.3.1.1 Event Description t

This event is characterized by a total loss of forced reactor coolant flow which is J

caused by the simultaneous loss of electric power to all of the reactor primary coolant pumps.

Following the loss of electrical power, the primary coolant pumps begin to coast down.

1 If the reactor is at power when the event occurs, the loss of forced coolant flow causes the primary coolant temperatures to rise rapidly. This results in a rapid reduction in DNB J

~

margin, and could result in DNB if the reactor is not tripped promptly. Also, as the primary coolant temperatures rise the primary coolant expands, which causes an insurge into the pressurizer, a compression of the pressurizer steam space and a rapid increase in primary coolant system pressure. The primary system overpressurization will be mitigated by the J action of the primary system safety valves and the reduction in core power following reactor trip. Reactor trip signals are provided from low primary coolant system flow.

The MDNBR is controlled by the interaction of the primary coolant flow decay and the core power decrease following reactor trip. The power to flow ratio initially increases, peaks, and then declines as the challenge to the SAFDLs is mitigated by the decline in core power due to the reactor trip. Ha reactor trip can be obtained promptly, the power to flow ratio will first peak and then decrease during the transient such that the SAFDLs will be no longer challenged.

The pump coastdown characteristics and the timing of the reactor trip, trip delays and scram rod insertion characteristics are key parameters. Natural circulation flow is developed in the primary system and the steam generators are available to remove the decay power.

Therefore, long term cooling of the core can be achieved.

EMF-91-176

  • Page 69 The primary concern with this event is the challenge to the SAFDLs. The event is analyzed to verify that the reactor protection system can respond fast enough to prevent penetration of the DNBR SAFDL.

This event is classified as a moderate frequency event (Table 15.0.1-1). The acceptance criteria are as described in 15.0.1.1. For this analysis, the systems challenged in this event are redundant; no single active failure in the RPS or ESF will adversely affect the consequences of the event. Long term recovery is provided by the auxiliary feedwater system, as demonstrated in the analysis of Event 15.2.7 in Reference 17.

15.3.1.2 Event Disposition The most limiting loss of flow transient is initiated from rated power. The transient is initiated by tripping all four primary coolant pumps. As the pumps coast down, the core flow is reduced, causing a reactor scram on low flow. As the flow coasts down, p~imary temperatures increase. This increase in temperature causes a subsequent power rise due to moderator reactivity feedback. The primary challenge to DNB is from the decreasing flow rate

  • and resulting increase in coolant temperatures. Because of the increased assembly" radial peaking, a DNBR analysis will be required for Cycle 10.

l I

15.3.1.3 Definition of Events Analyzed The event is initiated by simultaneously tripping of all of the primary coolant pumps.

The pump coastdown is governed by a conservative estimate of the pump flywheel inertia, the r

I homologous pump curves and the loop hydraulics. Reactor trip is delayed until the low primary coolant system loop flow signal is obtained. This trip setpoint is conservatively

~ reduced to account for uncertainties in flow measurement.

I

~ This event is analyzed from full power initial conditions. The core thermal margins are minimized at full power conditions resulting in this being the bounding mode of operation for

~ this event. One case is analyzed for this event to assess the challenge to the DNB SAFDL.

The event analysis is biased to minimize DNBR. The steam line bypass and the atmospheric

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EMF-91-176 Page 70 dump valves are both assumed not to operate, which again most challenges the DNB SAFDL 15.3.1.4 Analvsis Results The transient response of the reactor system was calculated using the PTSPWR2 computer program in Reference 3, including plots of key system variables and a sequence of events. The core thermal hydraulic boundary conditions from the PTSPWR2 calculation are used as input to the XCOBRA-lllC code to predict the MDNBR for the event. Based on the peaking factors given in Table 15.0.3-1 for Reload N fuel, the bounding MDNBR for this event is 1.391 using. The peak pellet LHR is 13.n kW/ft 15.3.1.5 Conclusion The 95/95 DNB correlation safety limit is not penetrated by this event. Maximum peak pellet LHR for this event is below the incipient fuel centerline melt criterion of 21 kW/ft.. Thus, all applicable acceptance criteria are. met.

15.3.2 Flow Controller Malfunction .

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

15.3.3 Reactor Coolant Pump Rotor Seizure 15.3.3.1 Event Description The locked rotor event is caused by an instantaneous seizure of a primary coolant pump rotor. Flow through the affected loop is rapidly reduced, causing a reactor trip due to a low primary loop flow signal. Following the reactor trip, the heat stored in the fuel rods continues to be transferred to the. primary coolant. Because of the reduced core flow, the coolant temperatures will begin to rise, causing expansion of the primary coolant and consequent pressurizer insurge flow and PCS pressurization. As the pressure increases, pressurizer sprays and safety valves would act to mitigate the pressure transient.

EMF-91-176 Page 71 The rapid reduction in core flow and the increase in coolant temperature may seriously challenge or penetrate the DNBR SAFDL. The event is thus evaluated to assess the DNBR challenge. The fuel centerline melt SAFDL is not seriously challenged by the small power increase typical of this event. PCS pressurization criteria have not been. approached in SNP analyses of this event; no case addressing pressurization is therefore performed.

The event as simulated is structured to provide a bounding determination of MDNBR for both the locked rotor and broken shaft (15.3.4) events.

~

I The reactor pump rotor seizure is an infrequent event (Table 15.0.1-1 ). The acceptance criteria for this event are presented in Section 15.0.1.1. For this analysis, the systems I

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challenged in this event are redundant; no single active failure in the RPS or ESF will adversely affect the consequences of the event. The auxiliary feedwater pumps will provide cooling capability after scram, as demonstrated in Section 15.2.7 of Reference 17.

le 15.3.3.2 Event Disposition l The controlling parameters for the pump seizure event are identical to those for the loss of flow event. Therefore, the same arguments will hold for this event as noted in Section I 15.3.1.2, requiring a DNBR analysis for Cycle 1o.

l 15.3.3.3 Definition of Events Analyzed

  • One case is analyzed for this event to maximize the challenge to the DNB limit. The

' bounding operating mode for this event is full power initial conditions.

15.3.3.4 Analysis Results The transient response of the reactor system was calculated using the PTSPWR2 computer program in Reference 3, including plots of key system variables and a sequence of events. The core thermal hydraulic boundary conditions from the PTSPWR2 calculation are used as input to the XCOBRA-lllC code to predict the MDNBR for the event. The MDNBR for this event is 1.341. Since the MDNBR for Reload N is above the 95/95 limit for the ANFP

EMF-91-176 Page 72 correlation, this infrequent event results no fuel failure due to penetrating DNB limits. The peak pellet LHR is 14.11 kW/ft.

15.3.3.5 Conclusion No radiological consequences occur due to fuel failure that results from penetrating DNB limits. The peak LHR is less than the 21 kW/ft limit to centerline melt. Thus,- the event acceptance criteria are met.

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

impeller. The impact of a coolant pump shaft break is a loss of pumping power from the affected pump and a reduction in the PCS flow rate. The flow reduction due to the seizure of 1 a pump rotor is more severe than that for a shaft break; however, the potential for flow reversal is greater for the shaft break event. The event is terminated by the low reactor coolant flow el trip.

1 15.3.4.2 Event Disposition The event is most limiting at rated power conditions because of a minimum margin to I DNBR limits. The initial flow reduction for this event is bounded by that for the Reactor Coolant Pump Rotor Seizure event (Event 15.3.3). The potential for greater reverse flow due

  • I to a shaft break is accounted for in the seized rotor analysis by decreasing the rotor inertia to zero at the time of predicted reversed flow. The severity of the pump shaft break event is l bounded by Event 15.3.3. Since the Cycle 1O changes summarized in Section 1.0 will not adversely impact the relative PCS flow reduction for the loss of flow Events (15.3), this event is bounded by Event 15.3.3 as in previous dispositions.

EMF-91-176 Page 73 15.4 Reactivity and Power Distribution Anomalies l 15.4.1 Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition I 15.4.1.1 Event Description This event is initiated by the uncontrolled withdrawal of a control rod bank, which results in the insertion of positive reactivity and consequently a power excursion. It could be f 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)

I operating conditions are considered in this event category; the consequences at power operating conditions are considered in Event 15.4.2.

t The control rods are wired together into preselected bank configurations. These r 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 l* at the same time and in their proper withdrawal sequence.

I The reactivity insertion rate is rapid enough that very high neutron powers are

  • calculated, but of short enough duration that excessive energy deposition does not occur.

~ Rod surface heat flux lags the neutron power but still approaches a significant fraction of full power. Because the event is very rapid, primary coolant temperature lags behind power. The I reactivity insertion rate is initially countered by the fuel temperature reactivity (Doppler) coefficient followed by trip and rod insertion.

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

(1) Non-safety grade high rate-of-change of power trip, .0001 % to 15% power (no credit taken)

(2) VHP trip (3) TM/LP trip

EMF-91-176 Page 74 *

(4) High pressurizer pressure trip (5) . High rate-of-change of power alarms, which initiate Rod Withdrawal Prohibit Action (no credit taken).

(6) The reactor is tripped (circuit breakers 42-01 and 42-02 open) and the plant is in hot shutdown or below when a pump out of service for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Reference 7).

Further protection is provided by the Doppler reactivity feedback in the fuel and by available DNBR margin betWeen the initial operating condition and the DNB thermal limit 15.4.1.2 Event Disposition and Justification Based on the Reference 3 analysis, this event is non-limiting relative to other AOO events that are being reanalyzed for Cycle 1O. Changes introduced in Cycle 1O will not alter this conclusion. Thus, this event will not be reanalyzed for Cycle 1O.

15.4.2 Uncontrolled Control Rod Bank Withdrawal at Power 15.4.2.1 Event Description As with Event 15.4.1, this event is initiated by an uncontrolled withdrawal of a control rod bank. This withdrawal adds positive reactivity to the core which leads to potential power and temperature excursions. Event 15.4.2 consid.ers the consequences of control bank withdrawals at rated and operating initial power levels.

The reactor protection trip system is designed and set to preclude penetration of the SAFDLs. Because of the design of this analysis, the TM/LP and VHP trips are principally challenged.

The TM/LP trip function is designed and set to protect against DNB. Principal DNB parameters such as power (the highest auctioned value of either calorimetric or neutronic power), core inlet temperature and core power distribution are measured. The function decreases margin to trip setpoint when process variables indicate a decrease in operating

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I I** EMF-91-176 Page 75 1* margin. This function is based on the core protection boundaries. Operation within these l boundaries assures protection of the SAFDLs.

I A broad range of reactivity insertion rates and initial operating conditions are possible.

The range of reactivity insertion is from very slow, as would be associated with a gradual boron dilution, and bounded on the fast end of the range by a bank withdrawal.

f l The objective of the analysis is to demonstrate the adequacy of the trip setpoints to assure meeting the acceptance criteria. To assure this objective, the analysis considers a spectrum of reactivity insertion rates and initial power levels. Since neutronic feedback as a J function of cycle exposure and design also influences the results, these effects are also included in the analysis.

r This everit is classified as a moderate frequency event (Table 15.0.1-1}. The le acceptance criteria are as described in 15.0.1.1. The single failure criteria are given in 15.0.11.

The safety systems challenged in this event are redundant and no single active failure will l adversely affect the consequences of the event.

t 15.4.2.2 Event Disposition and Justification The reference analysis<3*8>evaluated the severity of the transient over a range of values I for the reactivity insertion rate, moderator and Doppler reactivity coefficients and initial power.

The increase in assembly radial peaking will require a DNBR analysis for Cycle 10.

15.4.2.3 Definition of Events Analyzed The References 3 and 8 evaluated the consequences of this event for an uncontrolled control rod bank withdrawal. A spectrum of reactivity insertion rates were evaluated in order to bound events ranging from a slow dilution of the primary system boron concentration to the fastest allowed control bank withdrawals. Specifically, the analysis encompasses reactivity insertion rates from 1. x 1o-6 to 5. x 1o-5 !J.. p/sec. For Cycle 1O, the MDNBR will be reevaluated for the case with the lowest MDNBR from the previous analyses<3 .a).

EMF-91-176 Page 76 15.4.2.4 Analysis Results The transient response of the reactor system was calculated using the PTSPWR2 computer program in Reference 8, including plots of key system variables and a sequence of events. The core thermal hydraulic boundary conditions from the PTSPWR2 calculation are used as input to the XCOBRA-lllC code to predict the MDNBR for the event. Based on the peaking factors given in Table 15.0.3-1 for Reload N fuel, the bounding MDNBR for this event 1

is 1.640 using. The peak pellet LHR is 16.98 kW/ft.

15.4.2.5 Conclusion l

Reactivity insertion transient calculations demonstrate that the DNB correlation limit will

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not be penetrated during any credible reactivity insertion transient at full power. The maximum peak peliet linear heat generation rate for this event is less than the fuel centerline melt criterion of 21 kw/ft. Applicable acceptance criteria are therefore met and the adequate I

functioning of the TM/LP trip is demonstrated.

15.4.3 Control Rod Misoperation The control rod misoperation events encompass transient and steady state configurations resulting from different event initiators. The specific events analyzed under this event category are:

  • Core Barrel Failure 15.4.3.1 Event Description Dropped Control Rod or Bank The dropped rod and dropped bank events are initiated by a de-energized control rod drive mechanism or by a malfunction associated with a control rod bank. The dropped rod I

events are classified as Moderate Frequency events. Acceptance criteria are given in 15.0.1.1.

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l EMF-91-176

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In the dropped rod or dropped bank events, the reactor power initially drops in response to the insertion of negative reactivity. This results in reduction of the moderator temperature due to a mismatch between core power being generated and secondary system I load demand. The core power redistributes due to the local power effect of the dropped assembly or bank. The reactor power will return to the initial level due to the combined effects I of a negative MTC and reduced moderator temperature. The moderator temperature will not decrease below the temperature necessary to return the core to initial power. The rod and l bank drop events challenge the DNBR SAFDL because of the increased radial peaking and the potential return to initial power. Depending on the worth of the dropped rod or bank, the reactor will trip on a VHP or a TM/LP trip.

t Statically Misaligned Control Rod or Bank

  • ~1...

r The static misalignment events occur when a malfunction of the Control Rod Drive

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mechanism causes a control rod to be out of alignment with its bank. Misalignment occurs l* when the rod is either higher or lower than any of the other control rods in the same bank or when. a bank(s) is out of alignment with the Power Dependent Insertion Limit (POil). During I this event, the reactor is at steady-state rated full power (Mode 1) or part-power (Mode 2) conditions with enhanced power peaking. This event is classified as a Moderate Frequency .$

t occurrence.. Acceptance criteria are given in 15.0.1.1.

I In the static rod misalignment event, a control bank is inserted but one of the rods remains in a withdrawn state. This results in a local increase of the radial power peaking factor and a corresponding reduction in the DNB margin. The most severe misalignment occurs at full power operation, with one bank inserted beyond its control rod insertion limit and one of the bank control assemblies fully withdrawn. The radial power redistribution consequences of a reverse misalignment, i.e., one rod is inserted while the bank remains withdrawn, are essentially the same as the dropped rod event. The bank misalignment event occurs when one bank is inserted or withdrawn beyond the POil. The situation of concern is the power interval between 35% to 65% where control rod banks 3 and 4 are used.

EMF-91-176 Page 78 Single Control Rod Withdrawal The rod withdrawal event is initiated by an electrical or mechanical failure in the Rod Control System that causes the inadvertent withdrawal of a single control rod. A rod is l

withdrawn from the reactor core causing an insertion of positive reactivity which results in a power excursion transient. The m~vement of a single rod out of sequence from the rest of the I

bank results in a local increase in the radial power peaking factor. The combination of these two factors results in a challenge to DNB margin. The system response is essentially the same 1

as that occurring in the Uncontrolled Bank Withdrawal event at power (15.4.2).

1 Core Barrel Failure This event is initiated by the circumferential rupture of the core support barrel. The t

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 1 approximately one-half inch at operating temperatures. The worst possible axial location of the barrel rupture is at the mid plane of the vessel nozzle penetrations. This forms a direct flow el path between the inlet and exit nozzles in parallel with the path that goes through the core.

The core sustains a small reactivity transient induced by the motion of the core relative to the I inserted rod bank(s).

t Reactor protection for the Core Barrel Failure event during hot shutdown, refueling shutdown, cold shutdown, and refueling operating conditions is provided by Technical I Specification Shutdown Margin requirements. For the reactor critical and hot standby operating conditions, reactor protection is provided by the VHP trip and a non-safety grade high rate-of-change of power trip. For the rated power and power operating conditions, reactor protection is afforded for the VHP and TM/LP trips.

The probability of a circumferential rupture of the core support barrel has the same low probability of occurrence as ~ major rupture of the primary system piping. Therefore, this l

event is classified as a Limiting Fault event.

l EMF-91-176 Page 79 l 15.4.3.2 Event Disposition and Justification Dropped Control Rod/Bank Because of the increased radial peaking factor, the dropped control rod/bank events will require a DNBR analysis for Cycle 1o. Limiting conditions will be used to bound both rod I and low worth bank drop events (i.e., bank drop events that do not initiate a reactor scram).

For larger worth banks (i.e., bank drop events that initiate a scram), the Reference 3 PTSPWR2 I transient analysis remains bounding for Cycle 1O since a conservative bank worth was used.

Radial peaking augmentation factors representative of Cycle 1O will be used for both the I dropped rod and dropped bank analyses.

t Dropped Part-Length Control Rod -,\

A dropped part-length control rod will not be as severe as a dropped fuU-length control l rod and is, therefore, bounded.

le Malpositioning of the Part-Length Control Rod Group Use of part-length control rods is not allowed during power operation. The part-length.

I control rods are maintained in a fully withdrawn state; therefore, this event is not credible.

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t Statically Misaligned Control Rod/Bank Because of a smaller radial peaking augmentation factor, a statically misaligned control t rod or bank will not be as severe as a dropped full-length control rod and is, therefore, bounded.

Single Control Rod Withdrawal The disposition of this event is controlled by the same parameters as Event 15.4.2. A DNBR analysis will be performed using the core thermal hydraulic conditions from Event 15.4.2 and an appropriate radial peaking augmentation factor for Cycle 1o.

The consequences of a single rod withdrawal from Modes 3, 4, and 5 are either bounded or the event does not challenge the acceptance criteria. Mode 3 operation (Reactor

I EMF-91-176 Page 80 Critical) is defined as having a power greater than 10-43 and Tave greater than 525°F. Since the peak power obtained during a low power reactivity insertion increases with increasing insertion rate, the results for a single rod withdrawal are bounded by the results for a bank withdrawal (Event 15.4.1 where the insertion rate is much larger). Mode 4 operation (Hot Standby) applies when the power is between 10-43 and 23 and any of the control rods are I

withdrawn. The peak heat flux following a rod withdrawal decreases with increasing initial power level. Since Mode 3 includes 10-43 power, Mode 4 is bounded*by the results of Mode 1

3. Finally, Mode 5 operation (Hot Shutdown) applies when the power is less than 10-43 and Tave is greater than 525°F. The most reactive rod worth is less than the required shutdown I

margin; therefore, the reactor could not become critical by the withdrawal of any single rod.

t Core Barrel Failure A core barrel failure is initiated by a circumferential rupture of the core barrel support.

1 During this event, the core experiences a small reactivity insertion due to motion of the core relative to the control rods. The event is established to be incredible during hot shutdown, el refueling shutdown, cold shutdown and refueling operation due to the Technical Specification shutdown margin requirements. The event initiated from rated power bounds the power 1 operating, reactor critical and hot standby operating modes.

The core barrel failure event is bounded by the consequences of the control rod ejection event (15.4.8). Both of these events are classified as Limiting Faults. Specifically, the reactivity insertion rate and radial power redistribution for the control rod ejection are worse than what occurs during a core barrel failure. In addition, the control rod ejection analysis assumed a coincident loss of offsite power which leads to a coastdown of the primary coolant pumps. The flow decrease due to PCS pump coastdown is more severe than the flow reduction due to increased core bypass from the core barrel failure.

15.4.3.3 Definition of Events Analyzed Acceptable outcomes for these control rod misoperation events rely only on the RPS or on the Technical Specifications limiting conditions of operation. The elements of the RPS

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I EMF-91-176 Page 81 challenged are redundant and have been designed to provide their function in the event of a I single failure in the RPS. Single failures in the RPS or Engineered Safety Features thus do not affect event results. Single failure criteria for the Palisades plant are given in 15.0.11 I Dropped Control Rod or Bank The *rod/bank drop events challenge the acceptance criteria only in Mode 1 (Chapter r 15.0.2) operation. This event is analyzed at the rated power condition with conservative l allowances applied in a direction to minimize DNBR.

Single Control Rod Withdrawal t The single rod withdrawal events challenge the acceptance criteria in operating Modes 1 through 5. The single rod withdrawal event is analyzed at conditions that exist at the time r of MDNBR as calculated by PTSPWR2 during the most limiting uncontrolled rod withdrawal event. The most limiting single rod withdraw~! occurs during Mode 1 conditions.

  • le 15.4.3.4 Analysis Results r The events considered have in common the radial redistribution of power in th~ core, and can result in radial peaking factors in excess of Technical Specification limits. The t analyses are performed by coupling a conservative power peak tp transient response and DNB calculations. The power peak associated with each event is characterized through an t augmentation factor which relates the maximum power peak to the steady- state power peak.

The steady-state power distributions and augmentation factors are calculated with the XTGPWR<16> reactor simulator.

Table 15.4.3-1 summarizes the results of the analysis of the control rod misoperation events.

J Dropped Control Rod or Bank The analysis of rod drop events is performed using XTGPWR and XCOBRA-lllC. The XTGPWR code is used to calculate neutronic param~ters such as rod worth and power f.

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EMF-91-176 peaking augmentation factors. A power peaking augmentation factor is included in the XCOBRA-lllC MDNBR calculation to account for radial power redistribution effects typical of Page 82 the event. Simulation of the system transient for rod drops is not performed. Because the secondary system load demand remains constant through the event, the moderator will continue to cool down until moderator feedback is sufficient to restore the initial power level.

At that point, the moderator temperature stabilizes because no mismatch between core power production and secondary system load demand exists. The transient thus results in a new steady state condition characterized by a power level equal to the initial power, a reduced primary system pressure and a . reduced core coolant temperature. The DNBR is conservatively evaluated with an XCOBRA-lllC calculation using the initial condition power, coolant temperature and flow at a reduced pressure. The redistribution of the radial peaking t I

I factor is incorporated as noted above. A conservative radial peaking augmentation factor of 1.15 was applied for this event. The MDNBR for this event is 1.400 for Reload N. The peak I LHR is 16.29 kW/ft.

el The dropped bank event is distinguished from the dropped rod eve"iit by the greater magnitude of augmentation factors. PTSPWR2 and XCOBRA-lllC were used to assess the 1 transient response and MDNBR for a dropped control bank. The PTSPWR2 performed in Reference 3 remains applicable to Cycle 1O since a bounding control bank worth was used. t A radial peaking augmentation factor of 1.30 was conservatively used for Cycle 1o. The MDNBR for this event is 1.553 for Reload N and the peak LHR is 16.30 kW/ft. 1 Single Control Rod Withdrawal 1

In the analysis of the single rod withdrawal event, the core boundary conditions of average heat flux, temperature, pressure and flow are selected to conservatively bound the consequences of this event at rated power. The bank withdrawal analysis (15.4.2) considers I

reactivity insertion rates down to 1. x 1o-6 !J. p/sec which is representative of a single rod. The boundary conditions used in the calculation of MDNBR are obtain~d from the limiting transient response from Reference 8 (Event 15.4.2) initiated from 91.5% of rated power. Those conservatively biased core boundary conditions are then combined in an XCOBRA-lllC

EMF-91-176 Page 83 calculation with a radial augmentation peaking factor calculated to bound the possible single rod withdrawal radial power redistribution. A radial peaking augmentation factor of 1.08 was used. The MDNBR for Reload N is 1.375 and the peak LHR is 18.34 kW/ft.

15.4.3.5 Conclusion For the control rod/bank drop, the 95/95 DNB correlation safety limit is not penetrated by this event. Maximum peak pellet LHR for this event is below the incipient fuel centerline melt criterion of 21 kW/ft. Thus, all applicable acceptance criteria are met for these moderate frequency events.

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

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I EMF-91-176 Page 84 .1 TABLE 15.4.3-1

SUMMARY

OF MDNBRS MISOPERATION EVENTS FOR CONTROL Maximum ROD I

Event (Power)

Dropped Control Rod (100%)

Mode(a}

1 MDNBR 1.400 LHR (kW/ft) 16.29 I

Dropped Control Bank (100%)

1 1.553 16.30 1

Statically Misaligned Control 1 Bounded Rod (100%) (Dropped Control Rod- 100%)

1 Statically Misaligned Bank (50%)

2 Bounded (Dropped Control I Rod-100%)

Statically Misaligned Bank (65%)

2 Bounded (Dropped Control 1

Rod-100%)

Rod Withdrawal (91.5%) 1 1.375 18.34 el Rod Withdrawal (50%) 2 Bounded (Single Rod Withdrawal-Mode 1) 1 Rod Withdrawal (104 %) 3 Bounded (15.4.1)

Rod Withdrawal (104 %) 4 Bounded (15.4.1) t Rod Withdrawal (S 104 %) 5 Subcritical Core Barrel Failure (100%) 1 Bounded (15.4.8) t a These operating modes are defined in Section 15.0.2. .,!

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15.4.4 Startup of an Inactive Loop 15.4.4.1 Event Description This event is initiated by the startup of an inactive primary coolant pump. The startup I of an inactive pump can lead to an introduction of colder primary coolant into the reactor core.

The lower coolant temperature, together with a negative MTC, can cause an increase in core power and a degradation of DNB margin. Sufficient protection is available to reduce the consequences of this event.

15.4.4.2' Event Disposition and Justification Continuous power operation with less than four primary coolant pumps is not allowed by the Technical Specifications. Additionally, startup with less than four primary coolant pumps above hot shutdown is not allowed. Thus, this event is most limiting for an initial

. condition of three operating primary coolant pumps with the corresponding reduced power le level and VHP trip setpoint.

For operation with one inoperative punip, the low flow trip setpoint and the VHP trip t setpoint are changed to the allowable values for the selected pump condition.

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  • Under this arrangement, the VHP trip will terminate any transient resulting from the activation of an idle pump before any significant decrease in thermal margin. Although a slight t temperature drop due to the startup of the inactive pump is experienced, the effect on system pressure and hot channel MDNBR is covered by the large power margin to full power conditions. Therefore, the consequences of this event are bounded by the nominal full power MDNBR with four primary coolant pump flow.

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

EMF-91-176

  • Page 86 15.4.6 CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant 15.4.6.1 Event Description A boron dilution event can occur when primary grade water is added to the primary coolant system via the Chemical Volume and Control System (CVCS) or the accidental transfer of the contents of the iodine removal system during cold shutdown or refueling shutdown conditions.

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

15.4.6.2 Event Disposition and Justification The parameters affecting the boron dilution time-to-criticality include: (1) the mass of PCS coolant, (2) the PCS charging flow rate, (3) the PCS charging boron concentration, (4) the PCS boron concentration at event initiation versus operating mode, and (5) the PCS critical boron concentration versus operating mode. The changes introduced in Cycle 1o do not impact any of these factors. The Cycle 1O critical and shutdown boron concentrations are bounded by the values used in the Reference 3 analysis. Thus, this event will not be reanalyzed for Cycle 10.

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

15.4.7.2 Event Disposition and Justification The event is precluded due to the administrative controls and procedures, including startup testing, that ensure a properly loaded core. The Cycle 1O changes will not alter this disposition.

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15.4.8 Spectrum of Control Rod Ejection Accidents 15.4.8.1 Event Description This event is initiated by a failure in the control rod drive mechanism (CROM) pressure housing causing a rapid ejection of the affected control rod. The ejection of the control rod inserts positive reactivity causing an increase in core power. The resultant core thermal power excursion is limited primarily by the Doppler reactivity effect of the increased fuel temperatures t and is terminated by reactor trip of all remaining control rods, activated by neutron flux signals.

Because of the increase in core power this event challenges deposited enthalpy, radiological consequences and pressurization acceptance criteria. The criterion concerning the average enthalpy is addressed in Reference 15. The evaluation presented herein pertains to the radiological consequence criterion. The primary system overpressurization is bounded'by that for the loss of load event (15.2.1 ).

15.4.8.2 Event Disposition and Justification l* The reference analysisC3> contains conservative or applicable assumptions such that it bounds Cycle 10. The reference analysis was performed with a FrA of 1.66 and a FrT of 1.92.

t Conservative assumptions that were employed in the reference analysis include: (1) the reactivify of the ejected control rod, (2) the reactivity insertion rate, (3) the VHP trip setpoint, (4) the method of assessing fuel failure (i.e., the relationship between radial peaking and DNBR) and (5) a coincident four PCS pump coastdown at the event initiation. Thus, the amount of fuel failure predicted in the reference analysis (i.e., 14.7%) bounds Cycle 1O and reanalysis is not required. In addition, the changes for Cycle 1O will not affect the overpressurization event. As in previous dispositions, the overpressurization event will be bounded by the Loss of Load event (15.2.1 ) .

.15.4.9 Spectrum. of Rod Drop Accidents (BWR)

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

EMF-91-176 15.5 Increases in Reactor Coolant System Inventory Page 88 15.5.1 Inadvertent Operation of the ECCS that Increases Reactor Coolant Inventory 15.5.1 .1 Event Description This event is caused by an inadvertent actuation of the ECCS that results in an increase in the primary system inventory. The primary challenge is to the primary system overpressurization criteria.

15.5.1.2 Event Disposition and Justification The PCS over-pressurization for this- event is controlled by the charging system flow rate capacity and the relief capacity of the primary safety valves. The mass flow (steam discharge) capacity of the three safety valves is significantly greater than the inlet mass flow of the three charging pumps. Therefore, there is sufficient discharge capacity to prevent the primary system from being over-pressurized. No flow is initiated from the HPSI and LPSI due to high primary coolant system pressure.

The previous disposition and analysis concluded that the PCS pressurization for this event is bounded by the Loss of Load event (15.2.1 ). Since the Cycle 10 changes of Section 1.0 will not affect the pressurization rate from the charging pumps or the relief capacity of the PCS safety valves, the loss of load event remains bounding for this event. The potential boron dilution consequence of this event is bounded by the boron dilution event (15.4.6).

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

I EMF-91-176 Page 89 I

15.5.2.2 Event Disposition and Justification I Sufficient relief capacity from the pressurizer safety valves exists to limit the overpressurization potential to less than the 11 0% design value of 2750 psia. The potential I for dilution of the primary system boron is addressed in Event 15.4.6.

i Reference 3 disposed this event as being bounded by Events 15.4.6 and 15.2.1 . The Cycle 1O changes summarized in Section 1.0 will not impact this disposition.

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I EMF-91-176 Page 90 15.6 Decreases in Reactor Coolant Inventory 15.6.1 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve 15.6.1.1 Event Description This event is initiated by the inadvertent opening of a pressurizer pressure relief valve or safety valve, which results in the blowdown of primary coolant as steam. The primary system pressure decreases rapidly until the pressurizer liquid is depleted, and then to a pressure determined by the hot leg saturation temperature. Reactor scram occurs on TM/LP 1

well before the pressurizer liquid is depleted during the full power case, thus terminating the challenge.to SAFDLs. I This event is primarily considered a depressurization event, but with a negative I moderator pressure coefficient and a positive MTC, the thermal margin will be eroded with increased power, increased coolant inlet temperatures and decreased pressures. In addition, 1 the event can also uncover the core with a decrease in the primary coolant inventory.

el This accident is classified as a moderate frequency event (Table 15.0.1-1 ). The TM/LP trip affords protection against violation of the acceptance criteria for this event as described 1 in Section 15.0.1.1. The systems challenged are redundant; no single active failure in the RPS or ESF will adversely affect the consequences of the event. I 15.6.1.2 Event Disposition and Justification The event is principally of concern in the short term because of the DNBR challenge due to depressurization before scram_. The depressurization has little effect on core power or primary temperatures.

For non-power operating modes, the stored primary energy is less than that for the rated power case. Reactor power is limited to levels low enough that no challenge to DNB exists. At full power, this event is a depressurization event in which power, inlet temperature and flow remain essentially the same. The parameters controlling the severity of this transient

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are the rated PORV flow rate and the PCS low pressure trip setpoint. Because of the increase in FrA* a DNBR analysis is required for Cycle 1O.

' 15.6.1.3 Definition of Events Analyzed This event was analyzed for MDNBR at full power conditions since margin to thermal limits is minimized. The system response for the full power case was evaluated by PTSPWR2 in Reference 3. The event MDNBR was calculated using XCOBRA-lllC. The assumptions on equipment availability are such that primary system pressure is minimized. If sustained release of primary system coolant occurs, the event is bounded by the small break LOCA analysis CS).

15.6.1.4 Analysis Results The transient response of the reactor system was calculated using the PTSPWR2 computer program in Reference 3, including plots of key system variables and a sequence of events. The core thermal hydraulic boundary conditions from the PTSPWR2 calculation are l* used as input to the XCOBRA-lllC code to predict the MDNBR for the. event. Based on the peaking factors given in Table 15.0.3-1 for Reload N fuel, the bounding MDNBR for this event t is 1.741. The peak pellet LHR is 14.43 kW/ft.

15.6.1.5 Conclusion The 95/95 DNB correlation safety limit is not penetrated by this event. Maximum peak pellet LHR for this event is below the incipient fuel centerline melt criterion of 21 kW/ft. Thus, all applicable acceptance criteria are met. For sustained mass release from the primary coolant system, the small break LOCA analysis(S) is bounding.

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

I 1* The charging and HPSI systems provide sufficient coolant to replenish that which is lost.

I EMF-91-176 Page 92 .1 Consequently, no fuel failures would be predicted assuming a reactor trip on low pressurizer I

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

no reactor trip is assumed to occur.

I 15.6.2.2 Event Disposition and Justification The Cycle 1o changes will not impact initiating faults leading to the pipe break or the 1

primary coolant activity level, this event will remain bounded by the Reference 6 analysis.

I 15.6.3 Radiological Consequences of Steam Generator Tube Failure 15.6.3.1 Event Description I

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

t 15.6.3.2 Event Disposition and Justification This event is controlled by the NSSS system response. Since none of the changes I introduced in Cycle 1Owill affect the system response, the conclusions drawn in the Reference 6 analysis remain applicable for Cycle 1O. That is the radiological releases for this event are I well below the 10 CFR 100 limits for offsite doses. Thus, no further analysis is required for Cycle 10.

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

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

I 15.6.5 Loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks EMF-91-176 Page 93 I 15.6.5.1 Within the Reactor Coolant Pressure Boundary Event Description I boundary.

A loss of coolant accident is initiated by a breach in the primary system pressure The event initiators vary from relatively small break loss of coolant accidents i (SBLOCA) to complete ruptures of the PCS piping for large break LOCAs (LBLOCA). The limiting features of LBLOCA and SBLOCA analyses are the peak clad temperature (PCT) and, the time at elevated temperature that influence the extent of localized and core-wide zircaloy oxidation reaction.

15.6.5.2 Event Disposition and Justification The controlling parameters for the transient are: (1) the initial fuel storeq energy, (2) the ,'!

decay heat, {3) the radial and axial power profiles, (4) the fuel rod-to-PCS coolant heat transfer versus time, and (5) the operating conditions for the ECCS systems. The large break LOCA l* will be reanalyzed for Cycle 1O with increased radial peaking and reduced ECCS flow. Also, an increase in the fuel pellet diameter (reduction in pellet-to-clad gap) that was introduced in *~

t Reload M will be incorporated into the large break LOCA analysis. Reference 14 documents the results of this analysis.

The changes for Cycle 1O will not affect the relative severity between the large break and small break LOCAs. A review of the significant parameters listed in Reference 6 (Table 14.17.2-1) for the small break LOCA indicates that the parameters assumed in the reference small break LOCA analysis bound the corresponding values for Cycle 1o, including axial shape. Thus, the small break LOCA does not require reanalysis for Cycle 10.

EMF-91-176 Page 94 .1 I

15.7 Radioactive Releases from a Subsystem or Component 15.7.1 Waste Gas System Failure I 15.7.2 Radioactive Liquid Waste System Leak or Failure {Release to Atmosphere) 15.7.3 Postulated Radioactive Releases Due to Liquid-Containing Tank Failures I The results of ~he these events are not dependent on the changes being made for Cycle 10. Therefore, the reference analyses<6) remain bounding for the conditions of this disposition.

I 15;7.4 Radiological Consequences of Fuel Handling Accident 15.7.4.1 Event Description A fuel handling accident occurs when a fuel assembly is damaged during refueling operations such that fuel rods are ruptured, resulting in a release of radioactivity. The radiological dose is determined by the inventory of radioactive fission products in the affected fuel rods and the amount of release from the fuel pool and surrounding facility.

  • 1 15.7.4.2 Event Disposition and Justification The inventory of radioactive fission products is determined by the exposure and t

power level of the assemblies or fuel rods. The analysis presented in FSAR (Reference 6) assumes that the affected assembly was resident in the core for three full power years with a power of 2650 MWt and a radial peaking factor of 1.65. A fuel pool decontamination factor of 100 was assumed.

For Cycle 10, the core power is 2530 MWt and the peak assembly radial peaking factor is 1.66 for a 216 rod assembly. These parameters are bounded by those assumed in the reference FSAR analysis. Because of increased exposure for Cycle 10, the fission gas inventory and the internal gas pressure at the time of release could exceed the values used in reference analysis. The release gas pressure could affect the fuel pool decontamination factor. Because of this, the fission gas inventory and the fuel pool decontamination factors should be reevaluated for Cycle 10. Consumers Power Company will provide the results of this evaluation.

EMF-91-176 Page 95 I

15.7.5 Spent Fuel Cask Drop Accidents 15.7.5.1 Event Description A spent fuel cask drop accident can result in the damage of an irradiated fuel assembly

' and the subsequent release of radioactivity. The inventory of fission products is determined by the core power and power peaking (exposure).

I 15.7.5.2 Event Disposition and Justification Reference 6 contains an analysis of the radiological consequences of this event. The FSAR analysis conservatively assumes that the assembly with the maximum exposure is damaged. A radial peaking factor of 2.0 and a core power level of 2650 MWt are assumed in the Reference 6 analysis. For Cycle 10, the core power is 2530 MWt and the peak assembly radial peaking factor is 1.66 for a 216 rod assembly. These parameters are bounded by those assumed in the reference FSAR analysis. Like the Fuel Handling Accident (Event 15.7.4), the fission gas inventory and the fuel pool decontamination factors should be reevaluated for le Cycle 1o because of increased exposure. Consumers Power Company will provide this evaluation.

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4.0 References I 1. Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, NUREG-0800, LWR Edition, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, July 1981.

I 2. Advanced Nuclear Fuels Corporation Methodology for Pressurized Water Reactors:

Analysis of Chapter 15 Events, ANF-84-73(P)(A), Revision 5, Appendix B and I Supplements 1 and 2, Advanced Nuclear Fuels Corporation, October 1990.

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

041, Revision 2, Advanced Nuclear Fuels Corporation, September 1990.

Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed I Core Configurations, XN-NF-82-21 (A), Revision 1, Exxon Nuclear Company, September 1983.

6. Palisades Final Safety Analysis Report, Updated Version (through Revision 12),

t 7.

Consumers Power Company.

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

8. Review and Analysis of SRP Chapter 15 Events for Palisades with a 15% Variable High I Power Trip Reset, ANF-90-181, Advanced Nuclear Fuels Corporation, November 1990.
9. Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel, ANF-I 1224(A) and Supplement 1, Advanced Nuclear Fuels Corporation, April 1990.
10. Justification of the ANFP DNB Correlation for High Thermal Performance Fuel in the !I Palisades Reactor, ANF-89-192(P), Advanced Nuclear Fuels Corporation, January 1990.
11. Computational Procedure for Evaluating Fuel Rod Bowing, XN-NF-75-32(A) and Supplements 1-4, Exxon Nuclear Company, October 1983.
12. Description of the Exxon Nuclear Plant Transient Simulation Model for Pressurized Water Reactors (PTS-PWR}, XN-NF-74-5(A), Revision 2, Exxon Nuclear Company, October 1986, and Supplements 3-6.

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

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EMF-91-176 Page 97 .1 I

14. Palisades Large Break LOCNECCS Analysis with Increased Radial Peaking, EMF 177, Siemens Nuclear Power Corporation, October 1991. I
15. Palisades Cycle 1O Safety Analysis Report, EMF-91-093, Siemens Nuclear Power Corporation, September 1991.

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16. XTG: A Two Group Three-Dimensional Reactor Simulation Utilizing Coarse Mesh Spacing {PWR Version), XN-CC-28(A), Revision 3, Exxon Nuclear Company, January 1975. I
17. Palisades Modified Reactor Protection System Report: Analysis of Chapter 15 Events, ANF-87-150(NP), Volume 2, Advanced Nuclear Fuels Corporation, June 1988.

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18. "Determination of Palisades Thermal Margin/Low Pressure Trip Coefficients",

Combustion Engineering, Inc., September 1971.

19. Palisades Modified Reactor Protection System Report: Disposition of Standard Review Plan Chapter 15 Events, ANF-87-1 SO(NP), Volume 1, Advanced Nuclear Fuels Corporation, June 1988.
20. Palisades Principal Plant Parameters, EMF-90-084(P), Revision 2, Siemens Nuclear Power Corporation.

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EMF-91-176 Issue Date: 1O/Ol 191 PALISADES CYCLE 1O: DISPOSITION AND ANALYSIS OF STANDARD REVIEW PLAN CHAPTER 15 EVENTS Distribution SE Cole RA Copeland RC Gottula JC Hibbard JS Holm JW Hulsman

  • TR Lindquist RB Little JN Morgan KC Segard EL Tolman CJ Volmer RT Welzbacker SC Yung CPCo/HG Shaw (20)

Document Control (3)

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ATTACHMENT 4 Consumers Power Company Pali sades Pl ant Docket 50-255 CYCLE 10 TECHNICAL SPECIFICATIONS CHANGE REQUEST SIEMENS NUCLEAR POWER CORPORATION REPORT PALISADES LARGE BREAK LOCA/ECCS ANALYSIS WITH INCREASED RADIAL PEAKING AND REDUCED ECCS FLOW (EMF-91-177)

November 1, 1991

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