ML20249C495

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Rev 1 to EA-GEJ-98-01, Palisades Cycle 14 Disposition of Events Review
ML20249C495
Person / Time
Site: Palisades Entergy icon.png
Issue date: 06/17/1998
From:
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18066A216 List:
References
EA-GEJ-98-01, EA-GEJ-98-01-R01, EA-GEJ-98-1, EA-GEJ-98-1-R1, NUDOCS 9806300166
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J l9 EA-GEJ-98-01 Rev.1 lr PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS COVER SHEET

(

Title:

Palisades Cvele 14 Disposition of Events Review l Initiation & Review l

Calculation status Preliminary Pending Final superseded Rev Description Technically

  1. Initiated init Reviewed $

Review Method Revr i 0 Onginallssue By Date Appd Alt Detad Qual By Date Appd By Calc Review Test By ApPd JW 1 k%)/ [])ck!, 0 Y V' 0(n\Q g l

< &t W6

~

! /

l PURPOSE:

The purpose of this EA is to document the results of a Disposition of Licensing Basis Events l

Review as a result of changes made to the fuel and plant for Cycle 14 operation. All Palisades Licensing Basis Events are reviewed and dispositioned in this analysis.

Table 2.1 of reference 2.5 lists the PCS flow assumptions used in the safety analyses. Note 2 to this table lists lower minimum flows to be used in future analyses. After revision 0 of this EA was completed, a decision was made to reanalyze all events (see reference 2.33) which would be l impacted by the use oflower minimum flows. Reference 2.33 of this EA provides dispositions

! or analyses for these lower minimum flows. Note that the control rod ejection (reference 2.31),

the main steam line break (reference 2.30), the large break LOCA (reference 2.15), and the small break LOCA (reference 2.18) used these lower minimum flows.

SUMMARY

OF RESULTS:

This EA reviewed all Palisades Licensing Basis Events (FSAR Chapter 14 [Ref. 2.2]) and determined whether the current analysis of record was bounding or required analysis as a result of changes being made to the fuel and plant for Cycle 14 operation. The effects on the Safety Analyses of the changes to the fuel and plant for Cycle 14 (see Background Section) meet all regulations and SRP requirements and are considered acceptable. Offsite doses were calculated for those events with radiological consequences that may not be bounded by past analyses and are also acceptable.

The effects on the Safety Analyses of the changes to the lower minirnum flows for Cycle 14 meet all regulations and SRP requirements and are considered acceptable.

Total Sheets 15 9006300166 900623 2 DR ADOCK O

t. EA-GEJ-98-01 Rev.1

.'< PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS CONTINUATION SHEET I

M OBJECTIVE:

The purpose of this EA is to document the results of a Disposition of Licensing Basis Events Review as a result of changes made to the fuel and plant for Cycle 14 operation. All Palisades Licensing Basis Events are to be reviewed and dispositioned in this analysis.

M

REFERENCES:

2.1 EMF-98-013, Revision 0, " Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 15 Events," Siemens Power Corporation, May 1998.

2.2 Palisades Final Safety Analysis Report (FSAR).

2.3 Palisades Technical Specifications.

2.4 NUREG 0800, USNRC Standard Review Plan, Rev.1 - July 1981 (Various Sections).

2.5 EMF-97-051, " Palisades Cycle 14 Principal Plant Parameters," January 1998, l 2.6 ANF-90-078, " Palisades Cycle 9: Analysis of Standard Review Plan Chapter 15 Events," September 1990; Cart / Frame C667/1283.

2.7 ANF-90-181, " Review and Analysis of SRP Chapter 15 Events for Palisades with a

]

15% Variable High Power Trip Reset," November 1990.

2.8 EMF-95-022, " Palisades Cycle 12: Disposition & Analysis of Standard Review Plan Chapter 15 Events," Siemens Power Corporation, April 1995.

2.9 EA TAM-96-04 Rev. O, "Offsite Radiological Dose Consequences of a Cask Drop in the Spent Fuel Pool," October 1996; Cart / Frame G781/2304.

2.10 EMF-93-086(P), " Palisades Loss of Load Analysis," Siemens Power Corporation -

Nuclear Division, April 1993.

2.11 EMF-87-150(NP), " Palisades Modified Reactor Protection System Report: Analysis of Chapter 15 Events," Volume 2, Advanced Nuclear Fuels Corporation, June 1988.

2.12 EA-TAM-95-03 Rev.1, " Cycle 12 Off-Site Radiological Dose Consequences for the Main Steam Line Break," August 1995; Cart / Frame: G352/2379.

2.13 deleted 2.14 EA-TAM-95-01 Rev. O, " Cycle 12 Off-Site Radiological Dose Consequences for the l Control Rod Ejection Accident," July 1995; Cart / Frame: G331/2475.

l 2.15 EMF-98-026, Revision 1, " Palisades Large Break LOCA/ECCS Analysis, April 1998.

2.16 EA-TAM-95-05 Rev. O, " Radiological Consequences for the Palisades Maximum Hypothetical Accident & Loss of Coolant Accident," November 1995; Cart / Frame

, G534/0032. i l 2.17 EA-TAM-96-02, Revision 1, " Palisades Control Room Habitability Following FSAR I Chapter 14 Accidents with Radiological Consequences", May,1998.

2.18 "SBLOCA Evaluation of Palisades Cycle 14 Fuel Design Change," A-PAL-FE--0001, Rev 00, February 1998. .

2.19 EA-TAM-96-07, Owners Review - Palisades Updated Steam Generator Tube Rupture Analysis, ABB CENO 001-ST96-C-007 Rev. 00," December 1996; Cart / Frame H032/0975.

2.20 001-ST96-C-007 Rev. 00, " Palisades Updated Steam Generator Tube Rupture Analy11s," November 1996, ABB CENO.

L ._ _. _ _ _ _ _ _ _ _ ______

4 t EA-GEJ-98-01 Rev.1

,'r PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS CONTINUATION SHEET 2.21 EA-A-NL-91-169-01 Rev. 2, "Offsite Dose Calculations of Fuel Handling Accident,"

March 1993; Cart / Frame F385/1446 2.22 deleted 2.23 deleted 2.24 deleted 2.25 FC-970, Palisades Nuclear Plant Facility Change, " Reload R," May 1998, 2.26 EA-LOCA-98-01, " Containment Response to an LOCA Using CONTEMPT-LT/28.",

May,1998.

l 2.27 EA-MSLB-98-01, " Containment Response to an MSLB Using CONTEMPT LT/28.",

May,1998. l 2.28 DA-001-NT90-014, Rev. 00, " Palisades Steam Generator Tube Rupture for the Replacement Steam Generator Project," MGolbabai,1IMAR91.

2.29 EMF-96-140, " Palisades Cycle 13: Disposition & Analysis of Standard Review Plan Chapter 15 Events," Siemens Power Corporation, November 1996, 2.30 EMF-98-012, Revision 0, '" Main Steam Line Break Analysis for Palisades", May 1998.

2.31 EMF-98-021, Revision 0, " Palisades Control Rod Ejection Analysis", April 1998.

2.32 ST98-C-326, " Assessment of Reduced PCS Flowrate on SGTR and Containment Analyses for Palisades Plant," June 151996, memo from Michael J. Gancarz (ABB) to GA Baustian.

2.33 EMF-98-042, Revision 0, " Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 15 Events fo'r Reduced Primary Coolnat Flow,"

Siemens Power Corporation, June 1998.

2.34 Memo RlW:98:190 Responses to Consumers' Comments on EMF-98-042 dated June 16,1998 10 BACKGROUND:

The changes for Cycle 14 that impact the disposition of the Chapter 15 events and the verification of the setpoints, MDNBR and fuel centerline melt (FCM) analyses consist of the following (See Reference 2.1):

3.1 Plant Operating Conditions Changes Implemented Maximum nominal T of 544 F was added.

The nominal steam generator flow decreased from 11.57 Mibm/hr to 10.982 Mibm/hr.

The nominal HZP steam generator pressure of 900 psia was added.

- The HFP and HZP steam generator masses were changed:

The HFP liquid mass increased from 120,385 lbm to 133,595 lbm and the vapor mass decreased from 8,779 lbm to 8,545 lbm.

The HZP liquid mass of 203,783 lbm and the HZP vapor mass of 6,976 lbm were added.

The maximum steam generator tube plugging was assumed to be 15% for analysis purposes.

The shutdown margin with less than four reactor pumps operating was reduced from 3.75% to 3.5%.

. 3

e

.* EA-GEJ-98-01 Rev.1 9 PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS CONTINUATION SHEET 4

The minimum flows listed in note 2 to table 2.1 of reference 2.5 were bounded for all events (see references 2.33 and 2.34). These minimum flows are:

PCS flow Rate Technical Specification Limit, gpm 352,000 PCS Flow Rate Analytical Limit, gpm 341,400 Core Flow Rate Analytical Limit, gpm 331,200 Per a telecon with Siemens (Bill Nutt) on June 16,1998, for all MDNBR events renanalyzed at lower flowrates, the core average mass flux in the existing XCOBRA input files was reduced by 5% to bound the above minimum flowrates.

3.2 Plant Control Systems Changes Implemented

' The delay time for the opening or closing of the normal feedwater regulating valves increased from 20.5 sec to 22 sec.

The letdown capacity increased to 120 gpm.

l The minimum charging pump capacity of 33 gpm.

l The maximum pressurizer spray flow rate increased from 280 ppm to 500 gpm.

3.3 Plant Safety Systems Changes Implemented l The pressurizer pressure for actuation of the Si decreased from an allowable l" value of 1,593. psig to an allowable value of 1,590 psig.

l The LPSI pump performance has generally degraded over the range of operating pressures.

A containment high-pressure trip setpoint of 5 3.7 psig was added.

, The low steam generator level reactor. trip has been changed to an allowable value of 18.77% of the narrow range reading. -

The low steam generator level trip for AFW actuation has been changed to an l allowable value of 25.9% of the narrow range reading.

L Si actuation on high containment pressure was given an allowable range of 3.7 psig to 4.3 psig.

The MSSV capacities were made consistent with design capacities. The MSSV opening. pressure and flow data are now staged by bank.

The steam assisted MSIV closure time of 2 sec.

The single flow rate value (2,125 gpm) for the containment sprays was replaced with a table of values that are based on the failure assumptions.

The minimum, maximum and nominal flow rates and the startup delay times for

! the AFWs were added.

The SIT water volume was increased from 2,000 ft' to 2,011 ft . I l

3.4 Thermal-Hydraulic Characteristics Changes l l The lower tie plate design changed for Reload R. These tie plates have a FUELGUARD debris-resistant insert.

New pressure drop measurements were performed with the control blades fully inserted.

. 4

9 EA-GEJ-98-01 Rev.1 f PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS CONTINUATION SHEET 3.5 Fuel Rod Mechanical Design Changes A new fuel design will be introduced for Reload R, incorporating the following changes:

0.009-inch increase in the pellet OD.

0.152-inch increase in the pellet length.

0.4% reduction in the pellet dish volume.

0.009-inch increase in the cladding ID.

0.8-inch increase in the fuel rod active length.

0.111 inch increase in the effective plenum length.  !

Initial rod internal pressure of the gadolinia bearing rods has increased.

3.6 Neutronics Design Changes The enrichment pattern for the Reload R assemblies was modified and it produces different local power distributions.

The UO2enrichment in the gadolinia bearing fuel rods increased.

The maximum gadolinia concentration in the gadolinia-bearing fuel rods increased in Cycle 14 from 6 wt% to 8 wt%.

The U23ecapture-fission ratios for Cycle 14 increased slightly, resulting in an increase in the decay heat.

The 50% and 100% power differential rod worths at EOC conditions increased.

The moderator density and Doppler reactivity tables changed for Cycle 14.

The boron endpoint concentrations for Cycle 14 changed for all conditions.

3.7 Uncertainties and Response Times .

Cy;l; 14 de;; act ;;..;;;a ; S:l cere of freeh d;;;;ter..

Delay times for the HPSI and LPSI pumps were increased from 27 see to 40 sec.

RTD delay times were increased from 12 to 15 sec.

The trip delays for the low reactor coolant flow, high pressurizer pressure , low ,

pressurizer pressure, low steam generator pressure and TM/LP and VHP trips ,

j were increased by 0.2 sec.  !

!- )

i The Technical Specifications total and assembly radial peaking factor limits for Reload i

T A l l R fuel, Fr = 2.04 and Fr = 1.76, are unchanged from Cycle 13 Reload Q. Radial l peaking factors for fuel prior to Reload O are unchanged from those contained in (Core l Operating Limits Report) COLR Rev. 3. A copy of the COLR can be found attached to Reference 2.3.

The aforementioned changes necessitate a review of all Palisades Licensing Basis Events, which are contained in Ch.14 of the Palisades Final Safety Analysis Report

'(Reference 2.2). The majority of these events are dispositioned in Reference 2.1.

Events that are not dispositioned by Reference 2.1 are dispositioned in this EA. This l

,_ EA will review each Ch.14 Event and its associated Standard Review Plan (SRP) Ch.

. 5

EA-GEJ-98-01 Rev.1

." PALISADES NUCL EAR PLANT ENGINEERING ANALYSIS CONTINUATION SHEET 15 Section (Ref. 2.4}, and will either provide a disposition or reference the disposition in i Reference 2.1. It will be noted whether the event is bounded by an analysis of record or was analyzed for Cycle 14 operation.

M ASSUMPTIONS:

1.) Each event is dispositioned relative to the known changes made to the Palisades Plant and its fuel for Cycle 14 operation.

2.) The Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 15 Events, [Ref 2.1] has been technically reviewed.

M ANALYSIS INPUT:

The only inputs used in this analysis are the changes to be made to the fuel and plant for Cycle 14 operation, as described in References 2.1,2.5 and Ref. 2.25.

64 ANALYSIS:

Each FSAR Chapter 14 event is reviewed and dispositioned into one of the following categories: Unchanged from the previous analysis, Bounded by another analysis, Not applicable (outside the licensing basis, deleted from FSAR or SRP, or Not Credible), or analyzed for Cycle 14. Some events are a combination of these (i.e. the thermal hydraulic system response remains unchanged from previous analyses, but the MDNBR or LHGR may be calculated for the current cycle).

The following events are dispositioned in reference 2.33.

14.2 Uncontrolled Rod Withdrawal 14.4 Control Rod Drop 14.7 Decreased Reactor Coolant Flow 14.10 Increase in Steam Flow (Excess Load)

Each event is reviewed and dispositioned below in the order in which they appear in Ch.14 of the Palisades FSAR (Reference 2.2). Standard Review Plan events that are dispositioned in Reference 2.1, but bounded by another Ch.14 analysis, are also referenced.

14.2 Uncontrolled Rod Withdrawal 14.2.1 Uncontrolled Control Rod Bank Withdrawal from a Sub-Critical or Low Power Start-Up Condition.

1. SRP 15.4.1
2. Dispositioned in Reference 2.1
3. System response was analyzed for Cycle 13, [Ref. 2.29]. MDNBR and maximum fuel temperature were analyzed for Cycle 14

[Ref. 2.33]. l

.. 6

Q EA-GEJ-98-01 Rev.1 i . PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS CONTINUATION SHEET i

14.2.2 Uncontrolled Control Rod Bank Withdrawal at Power

1. SRP 15.4.2
2. Dispositioned in Reference 2.1
3. = System resoonse unchanged froin analyses in References 2.6 & 2.7.

MDNBR and maximum LHGR were analyzed for Cycle 14

[ Reference 2.33]. l

~

14.2.3 Single Control Rod Withdrawal

1. SRP 15.4.3
2. Dispositioned in Reference 2.1
3. System response unchanged from analyses in References 2.6 & 2.7. I MDNBR and maximum LHGR were analyzed for Cycle 14

[ Reference 2.33]. l 14.3 Boron Dilution Note: All Boron Dilution scenarios (except at Hot Standby and Reactor Critical) were i addressed for Cycle 12 [Ref. 2.8] due to changes in the initial and critical boron concentration for each operating mode. Dilution during Hot Standby and Reactor'- l Critical operating modes is taken into account in the analysis for SRP Sec 15.4.2 I (Uncontrolled Control Rod Bank Withdrawal at Power). The analysis for Cycle 12.

, (Ref. 2.8) demonstrated significant margins in the times to criticality. The analysis for l

Cycle 12 bounds Cycle 14.

14.3.1 Dilution During Refueling

l. SRP 15.4.6
2. Dispositioned in Reference 2.1
3. Shutdown Margin Adequacy is addressed in Reference 2.1.

14.3.2. Dilution During Start-Up .  !

1. SRP 15.4.6 l
2. Dispositioned in Reference 2.1
3. Shutdown Margin Adequacy is addressed in Reference 2.1.

14.3.3 Hot Standby or Reactor Critical L. 1. SRP 15.4.6

2. . Dispositioned in Reference 2.1 1 3. Event is bounded by SRP 15.4.2 (Uncontrolled l Control Rod Bank Withdrawal at Power) in Reference 2.1.

14.3.4 Dilution During Power Operation L . 7

l s EA-GEJ-98-01 Rev.1 f PALISADES NUCLEAR PLANT l ENGINEERING ANALYSIS CONTINUATION SHEET l

l'

l. SRP 15.4.6 .
2. Dispositioned in Reference 2.1
3. Event is bounded by the analysis for SRP 15.4.2 (Uncontrolled Control Rod Bank Withdrawal at Power) found in Reference 2.1.

14.3.5 Failure to Add Boron to Co:npensate for Reactivity Changes After Shutdown

1. SRP 15.4.6
2. Event is dispositioned in Reference 2.8.

14.4 Control Rod Drop 14.4.1 Dropped Rod Event

1. SRP 15.4.3
2. Dispositioned in Reference 2.33. l
3. System response unchanged from analysis in Reference 2.6.

l MDNBR analyzed in Reference 2.33. l 14.4.2 Rod Bank Drop Event

1. SRP 15.4.3 l 2. Dispositioned in Reference 2.33. l
3. System response unchanged from analysis in Reference 2.6.

MDNBR analyzed in Reference 2.33. l 14.5 ~ Core Barrel Failure

1. SRP 15.4.3
2. Dispositioned in Reference 2.1
3. Event is bounded by the Control Rod Ejection Event, (SRP Sec.15.4.8).

See Section 14.16 below.

l 14.6 Control Rod Misoperation

l. 14.6.1 Malposition of the Part-Length Control Rod Group
1. SRP 15.4.3
2. Dispositioned as not applicable to Palisades in Reference 2.1 Note: Current T.S. (Reference 2.3) do not allow the use of part length rods. These rods are j maintained in a fully withdrawn position and misalignment of the rods is not considered a credible event.

I 14.6.2 Statically Misaligned Control Rod / Bank  !

. 8

r EA-GEJ-98-01 Rev.1 J PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS CONTINUATION SHEET

1. SPJ 15.4.3
2. Dispositioned in Reference 3:+ 2.33.
3. System response unchanged from analysis in Reference 2,6.

MDNBR analyzed in Reference 2.33.

14.7 Decreased Reactor Coolant Flow 14.7.1 Loss of Forced Reactor Coolant Flow

1. SRP 15.3.1
2. Dispositioned in Reference 2.33.

l

3. System response unchanged from analysis in Reference 2.6.

MDNBR analyzed in Reference 2.33 14.7.2 Reactor Coolant Pump Rotor Seizure

1. SRP 15.3.3
2. Dispositioned in Reference 2.33
3. System response unchanged from analysis in Reference 2.6.

MDNBR analyzed in Reference 2.33 l 14.8 Start-up of an Inactive Loop

1. SRP 15.4.4
2. Dispositioned in Reference 2.1 -
3. Bounded by Rated Power MDNBR with four primary coolant pumps operating as stated in Reference 2.1.

14.9 Excessive Feedwater Incident (Deleted from FSAR) l

l. 1. SRP 15.1.2 l 2. Dispositioned in Reference 2.1
3. Event is bounded by the Increase in Steam Flow Event (SRP Sec.15.1.3).

See Section 14.10 below.

l 14.10. Increase in Steam Flow (Excess Load) -

1. SRP 15.1.3
2. Dispositioned in Reference 2.33
3. System response unchanged from analysis in Reference 2.6.

l MDNBR analyzed in Reference 2.33 l 14.11 Posmlated Cask Drop Accident I I

1. SRP 15.7.5
2. Dispositioned in Reference 2.1 l i

. 9 l

- -_ a

.- EA-GEJ-98-01 Rev.1 i

PAIJSADES NUCLEAR PLANT i*

ENGINEERING ANALYSIS CONTINUATION SHEET l

3. Radiological Consequences of this event were recalculated during Cycle 12 due j to the discovery of unanalyzed leak paths from the fuel handling building. See EA-TAM-96-04 Rev. 0 [Ref. 2.9].

14.12 Loss of Extemal Load

1. SRP 15.2.1
2. Dispositioned in Reference 2.1

! 3. System response is bounded by EMF-93-086(P), " Palisades Loss of Load Analysis," [Ref. 2.10]. MDNBR is bounded by analysis in Reference 2.11.

14.13 Loss of Normal Feedwater

1. SRP 15.2.7
2. Dispositioned in Reference 2.1
3. Event is bounded by analysis in ANF-87-150(NP) Vol. 2, " Palisades Modified Reactor Protection System Report: Analysis of Chapter 15 l Events," Reference 2.11.

14.14 Steam Line Rupture Incident

1. SRP 15.1.5
2. Dispositioned in Reference 2.1
3. Analyzed in Reference 2.30. Radiological Consequences of a Main Steam Line Break Outside Containment are bounded by EA-TAM-95-03 Rev.1,

[Ref. 2.12] which assumed 2% fuel failures. The Cycle 14 MSLB analysis (reference 2.30) determined no fuel failures.

I 14.15 Steam Generator Tube Rupture With a Loss of Offsite Power I

1. SRP 15.6.3
2. Dispositioned in Reference 2.1
3. Event is bounded by analysis in FSAR. See [Ref. 2.28].

In Reference 2.32, ABB provides a disposition for the lower assumed minimum flowrates. The l current FSAR analysis [Ref. 2.28] remains the bounding SGTR analysis of record for Palisades.

. Radiological Consequences were calculated as part of the analysis in Reference 2.17.

Note: A Updated SGTR analysis [Ref. 2.19] was performed by ABB/CE for Cycle 13 to i examine the feasibility of using the HPSI system to feed the auxiliary pressurizer sprays as opposed to the charging pumps. The results of this analysis were found to be acceptable and are detailed in Reference 2.20.

14.16 Control Rod Ejection

1. SRP 15.4.8 10

t EA-GEJ-98-01 Rev.1 PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS CONTINUATION SHEET

2. Dispositioned in Reference 2.1
3. Analyzed in Reference 2.31. Radiological consequences of this event are bounded by EA-TAM-95-01, [Ref. 2.14), which assumed 14.7% fuel failures.

The Cycle 14 CRE analysis (reference 2.31) determined no fuel failures.

14.17 Loss of Coolant Accident The results of the Cycle 14 reanalysis for both the SBLOCA and LBLOCA ECCS performance calculations show that the calculated values for peak clad temperature, cladding oxidation, hydrogen generation, and core geometry are within the allowed limits of 10 CFR 50.46 (see Reference 2.25).

14.17,1 Large Break LOCA

1. SRP 15.6.5'
2. Dispositioned in Reference 2.1
3. - A LBLOCA analysis was performed by Siemens for Cycle 14 (Reference 2.15). Radiological Consequences are bounded by the current MHA analysis EA-TAM-95-05 Rev. O, Reference 2.16.

14.17.2 Small Break LOCA

1. SRP 15.6.5 l
2. Dispositioned in Reference 2.17
3. A SBLOCA analysis was performed by ABB/CE for Cycle 14. The results of this analysis were found to be acceptable and are detailed in Reference 2.18. Radiological Consequences are bounded by the current MHA analysis EA-TAM-95-05 Rev. O, Reference 2.16.

14.18 Containment Pressure and Temperature Analysis 14.18.1 LOCA Analysis

1. SRP 6.2.1
2. Dispositioned in Reference 2.26

, 3. In Reference 2.32, ABB provides a disposition for the lower assumed i minimum flowrates on the mass and energy release which is used as an I g input to reference 2.26. Event will be reanalyzed for Cycle 14 (see

reference 2.26).

i 14.8.2 MSLB Inside Containment

1. . SRP 6.2.1
2. Dispositioned in Reference 2.27.
3. - In Reference 2.32, ABB provides a disposition for the lower assumed minimum flo.wrates on the mass and energy release which is used as an

. 11

l EA-GEJ-98-01 Rev.1 -

l .,' PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS CONTINUATION SHEET input to reference 2.27. Event will be reanalyzed for Cycle 14 (see l reference 2.27).

14.19 Fuel Handling Incident

1. SRP 15.7.4
2. Dispositioned in Reference 2.1
3. Radiological Consequences of this event are bounded by EA-A-NL-91-169-01 Rev. 2, (Ref. 2.21].

14.20 Liquid Waste Incident

! 1. SRP 15.7.3 i

2. Dispositioned in Reference 2.1
3. Esent is bounded by the current FSAR analysis [Ref 2.2] as  ;

discussed in Reference 2.1.

l 14.21 Waste Gas Incident 14.21.1 Gas Decay Tank Rupture l

1. SRP 15.7.2 (Deleted) l
2. Dispositioned in Attachment I to this analysis
3. Event is bounded by the current FSAR analysis [Ref 2.2] as discussed in Attachment I to this analysis.

14.21.2 Volume Control Tank Rupture l 1. SRP 15.7.2 (Deleted)

! 2. Dispositioned in [Ref 2.17]

3. Event is bounded by the current FSAR analysis [Ref 2.2]. VCT rupture
dose consequence analysis was expanded to include an event Generated l Iodine Spike (GIS). The results of this analysis showed that the Previous

, Iodine Spike (PIS) case remains bounding. Results are documented in EA-TAM-96-02 Rev i [Ref. 2.17] for both control room and off site doses for the VCTR.

l 14.22 Maximum Hypothetical Accident

1. SRP 15.6.5 Appendix A
2. Dispositioned in Attachment I to this analysis.
3. Event is bounded by EA-TAM-95-05 Rev. 0 Reference 2.16, as discussed f in Attachment I to this analysis.

14.23 ' Radiological Consequences of Failure of Small Lines Carrying Primary I

Coolant Outside Containment (SMLBOC)

. 12

EA-GEJ-98-01 Rev.1

. PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS CONTINUATION SHEET

1. SRP 15.6.2
2. Dispositioned in Reference 2.1
3. Event is bounded by the current FSAR analysis [Ref 2.2] as discussed in Reference 2.1. SMBLOC dose consequences were expanded to include an event Generated Iodine Spike (GIS). The results show that the SMLBOC for off site doses meet the requirements of 10 CFR 100 and are within the requirements of Standard Review Plan section 15.6.2. The control room doses are within the requirements of General Design Criteria 19 and Standard Review Plan 6.4. Results are documented in EA-TAM-96-02 Rev i (Ref. 2.17].

14.24 Control Room Radiological Habitability

1. SRP 6.4 *
2. Dispositioned in Attachment I to this analysis.
3. Event is bounded by EA-TAM-96-02 Rev.1, Reference 2.17, as discussed in Attachment I to this analysis.

. ~ 13

-/ EA-GEJ-98-01 Rev.1,

, PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS CONTINUATION SHEET i

2A - CONCLUSIONS:

This EA reviewed all Palisades Licensing Basis Events (FSAR Chapter 14 [Ref. 2.2]) and determined whether the current analysis of record was bounding or required analysis as a result of changes being made to the fuel and plant for Cycle 14 operation. The effects on the Safety Analyses of the changes to the fuel and plant for Cycle 14 (see Background Section) meet all regulations and SRP requirements and are considered acceptable. Offsite doses for those events l with radiological consequences are also acceptable.

. The effects on the Safety Analyses of the changes to the lower minimum flows for Cycle 14 meet all regulations and SRP requirements and are considered acceptable.

8J LIST OF ATTACHMENTS:

1.) Event Disposition and Justification for FSAR Sections: 14.21,14.22, & 14.24 2.) Admin Procedure 9.11 Rev. 9, Attachment 4, " Engineering Analysis Checklist" 3.) Admin Procedure 9.11 Rev. 9, Attachment 5, "T :hnical Review Checklist" l 4.) Admin Procedure 9.11 Rev. 9, Attachment 6, " Engineering Analysis Review Sheet" 4

1 l'

I c..

14 ~~ *

  • Ni'l'.:'? ~ .

O

~

EA-GEJ-98-01 Rev.1 .

, PALISADES NUCLEAR PLANT l ENGINEERING ANALYSIS CONTINUATION SHEET j Attachment 1 l

l Event Disposition and Justification for FSAR Sections 14.21, 14.22, & 14.24 i-14.21 Waste Gas Incident l

Event Disposition and Justification 1 The primary inputs to the analysis are the unount of fission product gases contained in the tanks. A conservative source term was assumed for this analysis based on a conservative percentage of the core assumed to contain leaking rods. This event has been deleted from the SRP and no longer needs to be dispositioned. However, the Cycle 14 fuel changes will l not impact the results of the current FSAR analysis.

l 14.22 Maximum Hypothetical Accident

, Event Disposition and Justification i

This hypothesized event is designed to bound the radiological consequences of all i postulated and anticipated accidents. The entire core is assumed to fail due to fuel melting.

Due to this assumption, fuel cycle changes do not affect the source terms used in the l l ~ reference analysis EA-TAM-95-05 Rev. 0 [Ref. 2.16]. Modifications to the plant for Cycle l 14 operation will not alter the results of this analysis. Therefore EA-TAM-95-05 Rev. O I

[Ref. 2.16] remains bounding for Cycle 14 operation. '

l l 14.24 Control Room Radiological Habitability Event Disposition and Justification ,

General Design Criterion 19 of 10CFR50, Appendix A requires that the Control room HVAC be designed to provide protection of the Control Room personnel from radiation

. doses. The whole body radiation dose limit of 5 rem or its equivalent to any other part of L the body (30 rem thyroid) for control room personnel is also specified in GDC-19. The l results of this analysis do not change for modifications being performed for Cycle 14 L operation. Therefore EA-TAM 96-02 Rev.1 [Ref. 2.17] remains bounding for Cycle 14 operation..

, . 15 l

PALISADES NUCLEAR PLANT Proc No 3.07 10CFR50.59 SAFETY REVIEW Attachment 1 Revision 8 Page 1 of 1 PS&L Log No D '/Nf item Identification: No EA-GEJ-98-01 Rev 1 Title Palisades Cvele 14 Disposition of Events SE Review REV Desenbe issue / Change: Reduced core and vessel flow assumptions in the safety analyses.

Reason for issue / Change: Analysis needed for proposed toch spec chance.

1. Does the item involve a change to procedures as desenbed in the FSAR? Yes No FSAR Sections affected None FSAR Sections reviewed See #2 X
2. Does the item involve a change to the facility as described or implied in the FSAR7 FSAR Sections affected Tables 14.1-2.14.15-1.14.18.2-2 14.17.2-1.14.17.1 1 References 14.1.14.15 FSAR Sections reviewed Chapters 1. 4. and 14 X DBD Sections reviewed 2.01 - 2.06 1
3. Does the item involve a test or experiment not described in the FSAR? l FSAR Sections affected None FSAR Sections reviewed See #2 X DBD Sections reviewed 2.01 - 2.06
4. Should the Technical Specifications or any of their Bases be changed in conjunction 1 with this item?

TS Sections affected none X TS Sections reviewed 1. 2. 3. COLR. and Standina Orders Justify "NO" answers below if logic is not obvious:

Searched Zyindex for. flow j If any Safety Review question listed above is answered "YES " perform a written USQ Evaluation according to Section 5.3.

i If all Safety Review questions listed above are answered "NO," written USQ Evaluation is not required.

However, this attachment shall accompany other review materials for the item to document that a Safety Evaluation was not required.  ;

1 1 l

/b I Al fg AllYb IY/75 PrKaredBy '6dte fteviewed By Date l a

l

[

16

,' Proc No 3.07

, Attachment 3 PALISADES NUCLEAR PLANT Revision 8 SAFETY ANALYSIS Page1 of1 I

Item identification: No EA-GEJ-98-01 Rev LTitle.Ealisades Cvele 14 Disposition of Events Review SECTION I Yes No

1. Will the probabikty of an accident previously evaluated in the FSAR be increased? X
2. Will the consequences of an accident previously evaluated in the FSAR be increased? X
3. Will the probability of malfunctions of equipment important to safety previously evaluated in the FSAR be increased? X
4. Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR be increased? X
5. Will the possibihty of an accident of a different type than any previously evaluated in the FSAR be created? X
6. Will the possibility of a malfunction of a different type than any previously evaluated in the FSAR be created? X
7. Will the margin of safety as defined by Plant Licensing Bases be reduced?

X A written Safety Analysis which documents the bases for these answers shall accompany this form. If any of the above questions are answered "YES," an unreviewed safety question is involved and the item shall not be implemented without prior NRC concurrence.

SECTION il Yes No

1. Should this be included in an FSAR update? X
2. Is an application for amendment to the Palisades Operating License required? X
3. Is NRC approval required prior to implementation of this item?

l SECTION 111 f

Y f /A I

$h I Yll$

eparVd 'By Date ' Reviewed By Date 17 u------------_

PALISADES NUCLEAR PLANT SAFETY ANALYSIS Item identification: No EA-GEJ-98-01 Rev 1 Palisades Cycle 14 Disposition of Events I Review DESCRIPTION OF ITEM Analyses and dispositions were accepta aility_of lower PCS flowrates in the safety analyses. performed to document JUSTIFICATION l 1. No. This is an analysis or disposition of a previously analyzed accidents and does not l change any plant equipment, procedure, or process as described in the FSAR. Therefore, the  !

probability of an accident previously evaluated in the FSAR will not be increased.

2. No. The results of the analyses or dispositions have demonstrated that all acceptance criteria are met. Therefore, the consequences of an accident previously evaluated in the FSAR will not be increased.

l l 3. No. The analyses or dispositions do not change any plant equipment, process, or l procedure. Therefore, the probability of malfunctions of equipment important to safety will not i

be increased.

4. No. The analyses or dispositions account for any required equipment failures. Therefore, the consequences of a malfunction of equipment important to safety will not be increased.
5. No. This is an accident analysis. It does not change any plant equipment, process, or procedure. Therefore, the possibility of an accident of a different type than previously evaluated in the FSAR will not be created.
6. No. This does not change any plant equipment, process, or procedure. Therefore, the possibility of a malfunction of a different type than previously evaluated in the FSAR will not be l created

, 7. No. Tne results of these analyses or dispositions demonstrate that the acceptcuce criteria l are met for all events. The margin of safety as defined by the plant licensing basis will not be reduced.

i t

l 18

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Proc No 9.11 PALISADES NUCLEAR PLANT Attachment 4 ENGINEERING ANALYSIS CHECKLIST Revision 10 Page 1 of 1 EA ~~ 66N N- b REV /

SECTION I - Affected ' Revision items Affected By This EA Yes No Required Identify' Closeout 1.0 Other EAs O Iff 2.0 Design Documents Electrical E-38 through E-49 O I 3.0 Design Documents Mechanical M239 M246, M249, M257-M261, M660, M664 M666 O ' Ef' 4.0 LICENSING DOCUMENTS 4.1 Final Safety Analysis Report (FSAR) O /,,, fe/ - 74 N 9"

4.2 . Technical Specifications - 58 e A,\

4.3 :- Operating Requirements Manual- O .G'-

5.0 PROCEDURES 5.1 Administrative Procedures lO B

5.2 - Operating Procedures (SOP,' EOP, ONP, etc) O (

5.3 - Working Procedures - 0 (

5.4 Tech Spec Surveillance Test Procedures &M . Rig 6.0-- OTHER DOCUMENTS 6.1 : Q List O 6.2 Plant Drawings O 6.3 Equipment Data Base 0 -

, 6.4 Spare Parts (Stock / MMS) 0 0' Fire Protection Program Report (FPPR) . O #'

w.d Design Basis Documents c O 6.7 Operating Checklists O 6.8 ' SPCC/PIPP Oil and Hazardous Material Spill Prevention Plan O Y 6.9 EQ Documents O Ef'

  • 6.10 MOV/AOV Program Documents (Voltage, ,

thrust, weak link, etc) 0 -['

6.11 Work Instructions 6.12 Other O

0 ((

SECTION ll Da any of the following documents need to be generated as a result of the conclusions reached in this EA:

Yes No

1. Corrective Action Document? O O' Reference

!' 2. EQ Evaluation Sheet? O O' Reference

3. Safety Evaluation? - #0 Reference T n e /vded
4. Design Basis Document Change Request? - O [ Reference l S. FSAR Change Request? (-O Reference /a f es'
6. Verification Test Procedure (for changes. O B' Reference to the Design Basis)?

Completed By N -

Data ddk

(-

r'

-f hnical Reviewed By - bb Date 4[/ Mf/

'IdIntify Section, No, Drawing, Document, etc.

-- -- - _ . - - - - -------_.---------------___----------_w

Proc No 9.11 TECHNICAL REVIEW CHECKLIST Attachment 5 Revision 10 EA - @fN@Ol REV [ Page 1 of 1 This checklist provides guidance for the review of engineering analyses. Answer questions Yes or No, or N/A if they do not apply. Document all comments on a EA Review Sheet. Satisfactory resolution of comments and completion of this checklist is noted by the Technical Review signature at the bottom of this sheet.

(Y, N, N/A)

1. Have the proper input codes, standards and design principles (/

been specified? /

2. Have the input codes, standards and design principles been V properly applied? /
3. Are allinputs and assumptions valid and the basis for their use \/

documented? /

~4. Is Vendor information used as input addressed correctly in the V analysis? /

5. If the analysis argument departs from Vendor Information/ /A Recommendations, is the departure justification documented?
6. Are assumptions accurately described and reasonable? V

/

7. Are the design basis changes permitted by this EA bounded by 'V the appficable Safety Review / Evaluation? /
8. Are all constants, variables and formulas correct and properly _

y applied? /

9. Have all comments been documented on an EA Review Sheet V and resolved, or have any minor (insignificant) errors been identified and their insignificance justified? (Indicate "No Comments," if none were made.)
10. If the analysis involves welding, is the following information tJ /k accurately represented on the analysis drawing (Output document)?

e Type of Weld e Size of Weld e Material Being Joined

e. Thickness of Material Being Joined e Location of Weld (s) e Appropriate Weld Symbology
11. Has.the objective of the analysis been met? V
12. Have administrative requirements such as numbering, format, v and indexing been satisfied? /

Technical Reviewer O'8 Date /708 J

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