Letter Sequence Other |
|---|
|
|
MONTHYEARML20080A1571983-07-12012 July 1983
Confirms 830627 Telcon Re Results of Heat Balance Calculations of Containment Atmosphere Temp Calculations. Post-accident Containment Sampling Procedure Will Be Modified to Include 10 Minute Sampling Period
Project stage: Other
ML20081L3101983-11-0404 November 1983
Monthly Operating Repts for Oct 1983
Project stage: Request
ML20082Q5401983-12-0606 December 1983
Advises That TMI Item II.B.3, Core Damage Assessment Methodology, Will Be Available for Ref & Use Immediately Upon Issuance by Westinghouse Owners Group in Feb 1984
Project stage: Other
ML20096G6931984-08-15015 August 1984
Rev 4 to Emergency Plan Implementing Procedure EPIP 1.7, Core Damage Estimation Procedure
Project stage: Other
ML20096G6781984-09-0404 September 1984
Forwards Rev 4 to Emergency Plan Implementing Procedure EPIP 1.7, Core Damage Estimation Procedure, in Response to Jr Miller 831103 Request.Submittal Fulfills Final Requirements in 831103 SER Re post-accident Sampling Sys
Project stage: Other
1983-07-12
[Table View] |
|
|---|
Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20217C0521999-10-0505 October 1999
Proposed Tech Specs 15.3.1.A,15.3.3.A & 15.3.3.C,eliminating Inconsistencies & Conflict Between TSs
ML20210M6171999-08-0404 August 1999
Proposed Tech Specs 15.3.5,removing Word Plasma from Discussion of Type of Display Monitor for sub-cooling Info in CR & TS 15.3.7,removing Discussion That Allowed Delay in Declaring EDG Inoperable Due to OOS Fuel Oil Transfer Sys
ML20210B5121999-07-15015 July 1999
Pbnp Simulator Four-Yr Rept
ML20209C1831999-07-0101 July 1999
Proposed Tech Specs Page Re Amend to Licenses DPR-24 & DPR-27,removing Ifba Enrichment Curve Methodology from TS
ML20205R6741999-04-12012 April 1999
Proposed Tech Specs Updating References to Reflect Relocation of Referenced Info in UFSAR
ML20203A2261999-02-0202 February 1999
Proposed Tech Specs Bases Page 15.4.4,correcting References to Pbnp FSAR Re Reactor Containment Design
ML20202F6521999-01-29029 January 1999
Proposed Tech Specs 15.6 & 15.7,reflecting Administrative Control Changes
ML20154G5601998-10-0707 October 1998
Proposed Tech Specs Ensuring That 4 Kv Bus Undervoltage Input to Reactor Trip Is Controlled IAW Design & Licensing Basis for Facility
ML20154H4231998-10-0505 October 1998
Proposed Tech Specs Section 15.4.1,removing Requirement to Check Environ Monitors on Monthly Basis
ML20154C0561998-09-30030 September 1998
Rev 5 to Pbnp Units 1 & 2 IST Program Third 10-Yr Interval
ML20154C1941998-09-28028 September 1998
Proposed Tech Specs Table 15.4.1-1 Re Min Frequencies for Checks,Calibrations & Tests of Instrument Channels
ML20153G7361998-09-23023 September 1998
Proposed Tech Specs Removing Explicit Testing Requirements of TS Section 15.4.13, Shock Suppressors (Snubbers)
ML20206E2731998-08-26026 August 1998
Rev 11 to Pbnps Odcm
ML20206E2681998-08-26026 August 1998
Rev 12 to EM Environ Manual for Wepc
ML20236W7861998-07-30030 July 1998
Proposed Tech Specs Change 206,revising to Incorporate Changes to TS to More Clearly Define Requirements for Service Water Sys Operability
ML20249C3811998-06-22022 June 1998
Proposed Corrected Tech Specs Pages Re Radiological Effluents
ML20248C6651998-05-28028 May 1998
Proposed Tech Specs Providing Specific Numerical Setting for Rt,Rcp Trip & AFW Initiation on Loss of Power to 4 Kv Buses
ML20248G1671998-05-27027 May 1998
Proposed Edited Tech Specs Pages Adding CR & Condenser Air Ejector Radiation Monitor SR & Correcting Errors in Notes of Table 15.4.1-1
ML20203K9571998-02-26026 February 1998
Proposed Tech Specs,Revising Sections 15.3.6, Containment Sys bases,15.3.12 CREFS Including bases,15.4.4 Containment Tests bases,15.4.11 CREFS Including Bases, 15.6.8 Plant Operating Procedures & 15.6.12
ML20203K2981997-12-15015 December 1997
Proposed Tech Specs Re Administrative Controls
ML20203K3081997-12-0808 December 1997
Rev 0 to Draft Radiological Effluent & Matls Control & Accountability Program
ML20203K3231997-12-0202 December 1997
Rev 11 to Draft Odcm
ML20203K3121997-12-0101 December 1997
Rev 12 to Draft Environ Manual Wisconsin Electric
ML20203K3201997-11-25025 November 1997
Rev 0 to Draft Radiological Effluent Control Manual Wisconsin Electric
ML20247P2681997-09-23023 September 1997
Proposed Tech Specs Implementing Boron Concentration Changes Re Planned Conversion of Unit 2 to 18-month Fuel Cycles
ML20210M5151997-08-14014 August 1997
Proposed Tech Specs,Removing Requirement in Plant TS to Perform Pbnp Unit 2 Containment Integrated Leak Rate 60-months from Previous Test
ML20210J0331997-08-0707 August 1997
Proposed Tech Specs,Replacing Wording & Double Underlining of Revised Wording
ML20140C9111997-06-0303 June 1997
Proposed Tech Specs,Modifying TS Section 15.3.3, Eccs,Acs, Air Recirculation Fan Coolers & Containment Spray, to Incorporate AOT Similar to Ones Contained in NUREG-1431,Rev 1
ML20141C5931997-05-0909 May 1997
Proposed Tech Specs Section 15.3.3, Eccs,Acs,Air Recirculation Fan Coolers & Containment Spray to Incorporate Allowed Outage Times Similar to Those Contained in NUREG-1431,Rev 1
ML20140H0321997-05-0101 May 1997
Rev 5 to Training Programs Trpr 33.0, Licensed Operator Requalification Training Program
ML20140H0171997-05-0101 May 1997
Rev 12 to Training Courses Trcr 86.0, Administrative
ML20140G9811997-04-29029 April 1997
Assessment of Corrective Action Program Pbnp
ML20140D7251997-04-14014 April 1997
Proposed Tech Specs,Changing Title of Corporate Officer Responsible for Nuclear Operations from Vice President- Nuclear Power, to Chief Nuclear Officer Per TS Section 15.6.2.1.c
ML20140D7091997-04-14014 April 1997
Proposed Tech Specs,Eliminating Provisions for Operation of Units at Below 3.5% Rated Power W/Only One RCP
ML20140G9241997-04-11011 April 1997
Rev 6 to ISTs IT 536, Containment Sump B Suction Line Leak Test (Refueling Shutdown)
ML20140G9071997-04-0707 April 1997
Rev 8 to ISTs IT 525B, Leakage Reduction & Preventive Maint Program Test of 2SI-896A&B,SI Pump Suction Valves (Refueling)
ML20140G8741997-04-0707 April 1997
Rev 35 to ISTs IT 04, Low Head Safety Injection Pumps & Valves (Quarterly)
ML20217P2851997-04-0303 April 1997
Rev 10 to Point Beach Nuclear Plant,Units 1 & 2 Odcm
ML20137N6451997-04-0202 April 1997
Proposed Tech Specs 14.2.4, Steam Generator Tube Rupture
ML20140G9031997-03-21021 March 1997
Rev 8 to ISTs IT 325, CVCS Valves (Cold Shutdown)
ML20140G9741997-03-21021 March 1997
Rev 0 to Operations Refueling Tests Ort 10A, Recovery from Integrated Lrt W/Core Off-Loaded
ML20137C8331997-03-20020 March 1997
Proposed Tech Specs 15.2.2, Safety Limit,Reactor Coolant Sys Pressure
ML20140G9621997-03-19019 March 1997
Rev 0 to Operations Refueling Tests Ort 9A, Preparation for Integrated Lrt W/Core Off-Loaded
ML20140G9131997-03-17017 March 1997
Rev 4 to ISTs IT 535B, Leakage Reduction & Preventive Maint Program Test of Train B HHSI & RHR Sys (Refueling)
ML20140G9521997-03-17017 March 1997
Rev 16 to Operations Refueling Tests Ort 6, Containment Spray Sequence Test
ML20140G8821997-03-0404 March 1997
Rev 9 to ISTs IT 115, Instrument Air Valves (Quarterly)
ML20140H3401997-02-28028 February 1997
WO Work Plan WO 9612073, Removal/Replacement of Breakers 1Y-06-11
ML20140H3011997-02-28028 February 1997
WO Work Plan WO 9612056, Removal/Replacement of Breakers 1Y-06-05
ML20140H3331997-02-28028 February 1997
WO Work Plan WO 9612072, Removal/Replacement of Breakers 1Y-06-01
ML20140H3241997-02-27027 February 1997
WO Work Plan WO 9612057, Removal/Replacement of Breakers 1Y-06-03
1999-08-04
[Table view]
Category:TEST/INSPECTION/OPERATING PROCEDURES
MONTHYEARML20154C0561998-09-30030 September 1998
Rev 5 to Pbnp Units 1 & 2 IST Program Third 10-Yr Interval
ML20206E2731998-08-26026 August 1998
Rev 11 to Pbnps Odcm
ML20206E2681998-08-26026 August 1998
Rev 12 to EM Environ Manual for Wepc
ML20203K3081997-12-0808 December 1997
Rev 0 to Draft Radiological Effluent & Matls Control & Accountability Program
ML20203K3231997-12-0202 December 1997
Rev 11 to Draft Odcm
ML20203K3121997-12-0101 December 1997
Rev 12 to Draft Environ Manual Wisconsin Electric
ML20203K3201997-11-25025 November 1997
Rev 0 to Draft Radiological Effluent Control Manual Wisconsin Electric
ML20140H0171997-05-0101 May 1997
Rev 12 to Training Courses Trcr 86.0, Administrative
ML20140H0321997-05-0101 May 1997
Rev 5 to Training Programs Trpr 33.0, Licensed Operator Requalification Training Program
ML20140G9811997-04-29029 April 1997
Assessment of Corrective Action Program Pbnp
ML20140G9241997-04-11011 April 1997
Rev 6 to ISTs IT 536, Containment Sump B Suction Line Leak Test (Refueling Shutdown)
ML20140G8741997-04-0707 April 1997
Rev 35 to ISTs IT 04, Low Head Safety Injection Pumps & Valves (Quarterly)
ML20140G9071997-04-0707 April 1997
Rev 8 to ISTs IT 525B, Leakage Reduction & Preventive Maint Program Test of 2SI-896A&B,SI Pump Suction Valves (Refueling)
ML20217P2851997-04-0303 April 1997
Rev 10 to Point Beach Nuclear Plant,Units 1 & 2 Odcm
ML20140G9031997-03-21021 March 1997
Rev 8 to ISTs IT 325, CVCS Valves (Cold Shutdown)
ML20140G9741997-03-21021 March 1997
Rev 0 to Operations Refueling Tests Ort 10A, Recovery from Integrated Lrt W/Core Off-Loaded
ML20140G9621997-03-19019 March 1997
Rev 0 to Operations Refueling Tests Ort 9A, Preparation for Integrated Lrt W/Core Off-Loaded
ML20140G9521997-03-17017 March 1997
Rev 16 to Operations Refueling Tests Ort 6, Containment Spray Sequence Test
ML20140G9131997-03-17017 March 1997
Rev 4 to ISTs IT 535B, Leakage Reduction & Preventive Maint Program Test of Train B HHSI & RHR Sys (Refueling)
ML20140G8821997-03-0404 March 1997
Rev 9 to ISTs IT 115, Instrument Air Valves (Quarterly)
ML20140H3401997-02-28028 February 1997
WO Work Plan WO 9612073, Removal/Replacement of Breakers 1Y-06-11
ML20140H3011997-02-28028 February 1997
WO Work Plan WO 9612056, Removal/Replacement of Breakers 1Y-06-05
ML20140H3331997-02-28028 February 1997
WO Work Plan WO 9612072, Removal/Replacement of Breakers 1Y-06-01
ML20140H3241997-02-27027 February 1997
WO Work Plan WO 9612057, Removal/Replacement of Breakers 1Y-06-03
ML20137P8441997-02-14014 February 1997
Rev 5 to Authorization of Changes,Test & Experiments (10CFR50.59 & 72.48 Reviews)
ML20140G9351997-02-14014 February 1997
Rev 13 to ISTs IT 585, Leakage Reduction & Preventive Maint Program Test of Post-Accident Containment Atmospheric Sampling Sys (Refueling)
ML20140G8921997-02-14014 February 1997
Rev 7 to ISTs IT 215, SI Valves (Cold Shutdown)
ML20140G9401997-02-0707 February 1997
Rev 21 to TS Tests TS 10A, Containment Airlock Door Seal Testing
ML20138K2991997-01-31031 January 1997
Rev 46 to OP 4D,Part 1, Draining RCS to CVCS Hut W/O Entering Reduced Inventory
ML20138K2031997-01-31031 January 1997
Rev 6 to IT 380, Purge Valve Air Sys Check Valve (Quarterly) for Unit 1
ML20138K2551997-01-31031 January 1997
Rev 2 to IT 530B, Leakage Reduction & Preventive Maint Program Test of Train B HHSI & RHR Sys (Refueling) for Unit 1
ML20138K2411997-01-31031 January 1997
Rev 12 to IT 530, Leakage Reduction & Preventive Maint Program Test of RHR Sys (Refueling) for Unit 1
ML20138K3231997-01-31031 January 1997
Rev 9 to RP 6A, SG Crevice Flush (Vacuum Mode)
ML20137P8551997-01-31031 January 1997
Brine Tank Standpipe Mod
ML20138K3071997-01-31031 January 1997
Rev 38 to OP 9C, Containment Venting & Purging
ML20138K2901997-01-31031 January 1997
Rev 2 to IT 535A, Leakage Reduction & Preventive Maint Program Test of Train a HHSI & RHR Sys (Refueling) Unit 2
ML20138K2641997-01-31031 January 1997
Rev 12 to IT 535, Leakage Reduction & Preventive Maint Program Test of RHR Sys (Refueling) for Unit 2
ML20138K4991997-01-31031 January 1997
Rev 9 to Ort 4, Main Turbine Mechanical Overspeed Trip Device for Unit 2
ML20138K3181997-01-31031 January 1997
Rev 33 to RP 1B, Recovery from Refueling
ML20138K3031997-01-31031 January 1997
Rev 1 to OP 4D,Part 5, Draining RCS to CVCS Hut W/O Entering Reduced Inventory & W/O SG Tubes
ML20138K2251997-01-31031 January 1997
Rev 7 to IT 385, Purge Valve Air Sys Check Valve (Quarterly) for Unit 2
ML20138K4791997-01-31031 January 1997
Rev 15 to Ort 67, Component Cooling Water to & from Excess Letdown Heat Exchanger-Refueling Shutdown for Unit 1
ML20138K4701997-01-31031 January 1997
Rev 17 to Ort 59, Train Spray Sys CIV Leakage Test for Unit 1
ML20138K5041997-01-31031 January 1997
Rev 7 to Ort 50, PRT to Gas Analyzer Sample Line for Unit 2
ML20138K5161997-01-31031 January 1997
Rev 18 to Ort 60, Train B Spray Sys CIV Leakage Test for Unit 2
ML20138K4751997-01-31031 January 1997
Rev 19 to Ort 60, Train B Spray Sys CIV Leakage Test for Unit 1
ML20138K5231997-01-31031 January 1997
Rev 7 to Ort 71, Electrical Penetration Leak Test for Unit 2
ML20138K7311997-01-31031 January 1997
Rev 9 to TS 80, Sampling of Emergency Fuel Oil Tank (T-72), Fuel Oil Storage Tanks (T-175A,B) & EDG Day Tanks (T-31A,B & T-176A,B)(Quarterly)
ML20138K7171997-01-31031 January 1997
Rev 6 to TS 37, Containment Spray Nozzles Check (Frequency of Less than or Equal to Five Yrs) for Unit 1
ML20138K3121997-01-31031 January 1997
Rev 43 to RP 1A, Preparation for Refueling
1998-09-30
[Table view] |
Text
-- . . _- - ~
f~__ m
. . . . . .y T ' EPIP 1.7 MINOP.
Revision 4 l 3 08-15-84
~ CORE DAMAGE ESTIMATION PROCEDURE
[1.'Oll PURPOSE-This procedure.wil'1 be used to implement the Westinghouse owner's. Group U Core Damage' Assessment Methodology. This. methodology provides the primary means of estimating core iamage through the measurement of fission product-concentrations in the reactor coolant system and containment. Also used-
. to . verify the extent 'of core damage are the auxiliary indicators of core exit thermocouple temperatures, reactor. vessel water level, containment-J radiation monitors, and containment hydrogen concentrations. .
l2.0 REFERENCE' Westinghouse Owner's Group Post-Accident Core Damage Assessment
~ -
'tethodology i
3 .'O RESPONSIBILITY This procedure is normally to be carried out by the Radwaste Technical p- } ; )SupportCoordinator. _However, the Core Physics Coordinator may be 1 1 1 directed by the Emergency Support Manager to complete this procedure.
? There is no~specified time frame in which this methodology should be
' carried out; rather, it'will be performed whenever needed and-appropriate.
~
Implementation'of the methodology-shall not? interfere with-the overall-response to the: emergency. +
- ' 14.0L PRECAUTION 5'.
3s .
~
~
i; 74.1 If. fuel damage or loss of: reactor coolant system integrity has-occurred, some'or all of the' following would be present:
~
4 .1.1- 1The letdown radiation monitor.(1RE-109) or.the failed
-fuel radiation monitor (2RE-109) may'be unusually high-
[ 'or offscale. f '
~
.. ; > - .V
. 4.1. 2 - -The containment radiation monitors (1RE-211 & 1RE-212 -
, or~2RE-211 & 2RE-212) may,be unusually high or offscale.-
, , 3
' ~
-4.1.3 The containment area monitors (1RE-102 & 1RE-107 or
^
2RE-102 & 2RE-107) may be unusually high or offscale.
r
.4.2- ' Health Physics procedures and requirements must be followed'when applicable, i.e., when entering a high radiation area.
1 c.
8409110098 840904
. 1 PDR ADOCK 05000266
, F. T PDR _
b r,
-_ 4.c.. . - . ..._,~.._,-,,...-;--_.u-._.--._._._.---.~u-.._.__,-_.-.~..---.;.,
b EPIP 1.7 Page 2 4.3 The user should use as many indicators as possible to differentiate between the various core damage states. Because of overlapping values of release and potential simultaneous conditions of core damage, overtemperature, and/or core melt, considerable judgement needs to be applied.
5.0 INITIAL CONDITIONS 5.1 Applicable portions of EPIP 1.2, " Plant Status," are completed.
5.2 A reactor coolant sample has been.taken and analyzed by implementing EPIP 7.3.2, " Post-Accident Sampling & Aralysis of Potentially High Level Reactor Coolant."
5.3 A containment atmosphere sample has been taken and analyzed by implementing EPIP 7.3.3, " Post-Accident Sampling of Containment Atmosphere."
5.4 The Westinghouse Owner's Group (WOG) Post-Accident Core Damage Assessment Methodology is available for use. Copies of the WOG methodology are located in Nuclear Engineering office, Room 381, Public Service Building in Milwaukee and in the technical support center.
6.0 PLANT-SPECIFIC DIFFERENCES FROM THE WOG METHODOLOGY 6.1 Pressure & Temperature Adjustments The specific activities of the containment air sample recorded on EPIP-31 are referenced to actual containment pressure and temperature conditions. Therefore, the optional adjustment in the WOG methodology is not required for the PENP containment sample.
The reactor coolant system activities recorded on EPIP-30 are in units of pCi/cc. Therefore, a temperature adjustment is required for the reactor coolant system volume and estimated safety injection volume only if the reactor coolant system temperature or containment sump temperature, respectively, are greater than 200*F. Use Steps 6.1.1 through 6.1.5 on Pages 30 and 31 from the WOG methodology to calculate the total liquid volume.
' .' .GL/J~ TG..'y*;
i ' ' . O : . ..
m . ;. ( . : , . - . - . . * - ~ ;_' ; L. ; ; ; + : - Q. ' .
- g. g.i_. m ;> >
, o_ .
y;(, a.{ , '
SE EPIP 1.7 Page 3
~
.,A
'.r.
l '
n
'y. _
- Volume' Estimation
[ s
.'6'111.
. Conta'inment volume 51'x los fts = 2.83 x 1010 cc 6.1.2- iEstimated safety injection volume (ESIV) 1
- Available safety'injectionfdilution' sources are: l h, ',
iAccumulators: 2 at 8,000 gallons each
.~
~
. RWST's: -275,000 gallons
-BAST's: 1 of 3 at'5,000 gallons each +-
Spray Additive Tanks: 2,574 gallons Total ESIV: gal.
.ConveEt ESIV units to cc's:
-ESIV(cc) = ESIV_(gal)_X 0.15,56.ft 3 / gal X 28.3 X 103 cc/fts 6.1.3 ~ Assume RCS volume = 6,'040' ft8
~
'= 1.71\X los cc - -
)
^
[6.1.4 - If sump. temperature.or RCS temperature are greater than
'~
200*F, temperature adjust the volumes using' Page 31 of
.WOG methodology. Ensure;the! volumes,are in units-of cubic centimeters (cc).
- ESIV(adjusted) or RdS(adjusted)l= (ESIV or <RCS)' x STP p
l.
nsity' ratio (WOG_Page 31)
Where:
3
' = water P- ' _ 611.5: - Total liquid volume = ESIV(a'djustied), + RCS(adjusted) _
= . cc + cc-y, .
- cc
. -Total liquid volume =
.Y .
i ' 7.0.. CORE DAMAGE ASSESSMENT h_ - / Perform an assessment of core damage by implementing the WOG Core-Damage Assessment Methodology incorporating ~the plant _ specific differences listed f5 .
in Section 6 of this procedure.. Implementation of the WOG methodology will~
- - provide an estimation of core damage in several categories
- No clad damage,
~ ~
- 0-50% clad damage, 50-100% clad damage, 0-50% fuel pellet overtemperature, 50-100% fuel pellet overtemperature,.0-50% fuel melt, and 50-100% fuel melt.
[,
~
- . _ . - - - . - . - . - - . - - . - . -
