ML20072F599
| ML20072F599 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 06/06/1983 |
| From: | LONG ISLAND LIGHTING CO. |
| To: | |
| Shared Package | |
| ML20072F587 | List: |
| References | |
| PROC-830606-01, NUDOCS 8306280046 | |
| Download: ML20072F599 (205) | |
Text
{{#Wiki_filter:. . - _ . EPC 4# EPIP 2-8 : O aPProvea: Plant Manager ._ r - Pase i or io Effective Date 6/6/83 l EPIP 2-8 POST-ACCIDENT GASEOUS EFFLUENT SAMPLE ANALYSIS 1.0 PURPOSE To provide guidance for anaiyzing particulate, radiofodine, and gaseous samples from the Post-Accident Station Ventilation Monitor (RE-126) and the Post-Accident Reactor Building Standby Ventilation Monitor (RE-134) during emergencies. 2.0 RESPONSIBILITY O A Radiochemistry Technician is resPonstaie for Performing this procedure.
. 3.0 PRECAUTIONS 3.1 If the radiation measurements on the sample container exceed 50,000 cpm on an RM-14 or equivalent set-up, the sample will be considered a "high activity sample."
3.2 If the sample is not a high activity sample (i.e., less , than 50,000 cpm), then observe standard precautions for handling radioactive materials. O 8306280046 830623 Rev. O hDR ADOCK 05000M pDR 5/3/83 ,
EPIP 2-8 , Page 2 of 10 . 3.3 If the sample is considered a high activity sample, the Radiation Protection Coordinator (RPC) will specify additional precautions (e.g., shielding, remote handling equipment, etc.) to be used. 4.0 PREREQUISITES 4.1 EPIP 2-7, Post-Accident Gaseous Effluent Sampling, has been completed. 4.2 For high activity samples: 1 4.2.1 The RPC will specify the protective clothing and equipment required for sample analysis. O 4.2.2 The Radiochemistry taboratory fume hood is ! functioning properly or a suitable alternative is
- approved by the RPC.
l 4.2.3 A shield area is set up for the s,torage of the [ sampl e(s). i r I l - i !O i ! Rev. 0 l 5/3/83
__.~....__.....-__..._.__m_._..___.-.m.. .-... _ .._-.... _ . . _ _. i ( . l l EPIP 2-8 Page 3 of 10 ,O , 5.0 ACTIONS l l l l l CAUTION l l .- I
- 1 PERFORM SECTION 5.1 ONLY IF THE TECHNICIM(S) l l l TRANSPORTING THE SMPLES ARE NOT THE SME TECH ~ l l 1 NICIAN(S) ANALYZING IT. l l l 1 5.1 Sample Turnover -
5.1.1 Upon arrival at the Rad / Chem Lab, the technician (s) transporting the sample will brief I the technician (s) perfoming the analysis.
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- 5.1. 2 Ensure the briefing covers the following items I
(as a minimum):
- a. Monitor sampled (Station Vent Monitor RE-126
- or RBSYS Monitor RE-134) l b. Survey results (high or low activity sample) 5.1. 3 Obtain completed Stack Sampling Data Sheet I
(Attachment 3) from EPIP 2-7 Post-Accident Gaseous Effluent Sampling. ~ Rev. O 5/3/83
_ _ _ _ _ ~ . . _ _ _ . . . . _ _ . . _ _ . . . _ . _ . _ . _ ._ _ _ . _ _ _ . . EPIP 2-8 Page 4 of 10 5.2 Sample Preparation i I I CAUTION I I I I WHEN HANDLING THE SAMPLE (S) (THE FILTER I l ELEMENT) OR A MARINELLI BEAKER, ENSURE , 1 I RADIOLOGICAL PRECAUTIONS ARE TAKEN (e.g., I l PROTECTIVE GLOVES MUST BE WORN). 1 I i 5.2.1 Prepare the sample (s) and Marinelli beaker for counting in accordance with Sections 8.2, 8.4, and 8.6 of SP72.006.01, Sample Preparation--General . O 5.3 Sample Counting and Analysis 5.3.1 Place the filter, cartridge, and beaker in a GeLi detector and count each using the following steps. 5.3.2 Prepare the Multichannel Analyzer (MCA) for operation in conjunction with the computer in accordance with Section 8.2, SP73.033.08, Gantt Spectrometer System Startup. 5.3.3 Prepare the computer to analyze and provide a printout of the isotopic analysis in accordance with Section 8.1.1, SP73.033.10, Gama Spectrometer System Operation. 5.3.4 Perform an isotopic analysis of the samp1e in accordance with Section 8.1.3, SP73.033.10, Gama Spectrometer System Operation. Rev. 0 5/3/83
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( . I - t EPIP 2-8 , Page 5 of 10 I
.O .
l . I I CAUTION I
- 1 I I IF, DUE TO HIGH ACTIVITIES.IN A I I SAMPLE, THE DEADTIME ON THE COUNTING l l SYSTEM IS EXCESSIVE, THE SAMPLE MUSTI l BE DILUTED, ALLOWED TO DECAY, OR l l CHANGED TO A DIFFERENT COUNTING l l GE0ETRY. I I I 5.3.5 If a dilution of the sample is necessary, perform the following steps:
1 I I CAUTION l 1 1 I
- l THIS DILUTION PROCESS, WHICH CAN BE lQ i
l DONE FOR THE MARINELLI BEAKER ONLY,
.I SHOULD BE DONE UNDER A FLME HOOD, I
I I 1 AND ALL PREVIOUSLY STATED PRECAUTIONSI L i SHOULD BE STRICTLY ADHERED TO. I ! I I L i
- a. Using a microsyringe, extract a known amount j of gas from the Marinelli beaker through a rubber-capped top. Note this volume on Attachment 1, Gaseous Effluent Sample Analysis Data Sheet.
l [
- b. This extracted gas will be injected into an evacuated Marinelli beaker again utilizing a
{ rubber-capped top beaker assembly. i !O
- Rev. O l 5/3/83
. ...: . _ . ._ ......__ _ . _ _ ._ _ _ _ _ _ _._ _. - . .. . _ . , . ..__ . . _._..-.. l l
EPIP 2-8 O "' ' I i l CAUTION I I I I IF THE COMPUTER MALFUNCTIONS, AN I I ANALYSIS CAN BE PERFORMED USING l l ONLY THE GELI AND THE MCA. I I I 5.3.6 In order to do an isotopic analysis using only the GeLi detector and MCA, perfom Section 8.2, SP73.033.10, Gamma Spectrometer System Operation. I , l l CAUTION I I I I THE ANALYSIS, USING ONLY THE GELI l O i OETECTOR AND TnE aCA. #UST sE STORED i l ON A CASSETTE AND PRINTED OUT WHEN l 1 THE COMPUTER IS RETURNED TO OPERA- l l TION. , I I I 5.3.7 Attach the computer printout onto Attachment 1. ~ 5.3.8 Contact the Chemistry Coordinator when the required analysis is completed and relay the infomation from Attachment 1.
- 6. 0 REFERENCES 6.1 EPIP 2-7, Post-Accident Gaseous Effluent Sampling 6.2 SP72.006. 01, Sample Preparation--General ,
6.3 SP73.033.08, Gamma Spectrometer System Startup . 6.4 SP73.033.10, Gama Spectrometer System Operation O Rev. 0 5/3/83
4 EPIP 2-8 Page 7 of 10 1 0 . 7.0 ATTACl#lENTS
- 1. Gaseous Effluent Sample Analysis Data Sheet I
4 i J a lO i . . i i-t l' r-r l: 4 n i "I i 4 E O Rev. 0 5/3/83
.._u.._._.__. _ _ _ . . .c _.
C t EPIP 2-8 Page 8 of 10 O ' Attachment 1 Page 1 of 3 GASE0US EFFLUENT SMFLE ANALYSIS DATA SHEET l - 1. Date: Time (taken/ analyzed): / (use { 24-hour clock) i j 2. Team Members: , 3
- 3. Sample Radiation Survey Results:
4 4. Sample ID (type of sample and location): l 5. Sample Preparation Results: Volume of gaseous sample transferred to clean Marinelli beaker
- (if necessary)
- al (from Step 5.3.5.a)
I , 6. Analytical Results: l l Counting Time: _ tO '- ' d***'"'"-' " Ea""*: l From the isotopic analysis of the particulate filter and ! charcoal cartridge ietemine the concentration of the i isotopes listed below, multipy them by the correction , factor, and add the dose equivalents in order to l determine the I-131 Dose Equivalent. For the particulate filter: [ ! I I (1)l (2) l Dose 1 - l l l Concentration l Correction IEquivalent I i IIsotope (ill (uti/cc) l Factor (uci/cc) I I I -1 31 1 l l I-132 9.60 x 10-3 l t i I-133 l 1.81 x 10-1 l l l I-134 l 2.50 x 10-3 i
, I I-135 ., .
3.76 x 10-2 l ! l . Dose Equivalent: 1 I I-1 31 1,, I (1) From computer printout (2) From Table B-1, Regulatory Guide 1.109 LO . Rev. 0 l 5/3/83
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EPIP 2-8 Page 9 of 10 O Attachment 1 Page 2 of 3 GASE0US EFFLUENT SMFLE AllALYSIS DATA SHEET (continued) For the chan:oal cartridge: l l (111 (2) I Dose l l l Concentration 1. Correction l Equivalent l l Isotope (ill (uCi/cc) l Factor I (uci/cc) l l I -1 31 1 1 l l I-132 I , 9.60 x 10-3 l l I-133 l : 1 81 x 10-1 l l I-134 I Z.50 x 10-3 l l I-135 1 3.76 x 10-2 l l l l Dose Equivalent: l I I-131 l l By adding the dose equivalents (I-131) together, obtain the " Total Dose Equivalent,: DEI -131 - i Convert the Dose Equivalent of I-131 to a " release rate" !O sy uitipirin,ny the stacu fiow rate: uti DEI -131 (in. uCi/cc) x 1.42 x 108 cc/sec = sec l or by the RBSYS flow rate: uCi DEI -131 (in. uCi/cc) x 4.50 x 105 cc/sec = ,,e depending on the monitor that was sampled / analyzed. e' (1) From computer printout (2) From Table B-1, Regulatory Guide 1.109 Rev. 0 5/3/83 1
I l 4 l . EPIP 2-8 i Page 10 of 10
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0 Page 3 of 3 GASEOUS EFFLUEllT SAMLE AllALYSIS DATA SHEET (continued) i ,
- b. To detemine Xe-133 Dose Equivalent:
From the isotopic analysis of the Marinelli beaker, detemine the concentration of the nuclides listed I
, below, multiply by the correction factor, and add the dose equivalent in order to determine the Xe-133 Dose Equivalent.
- I l - l l Dose I i l Concentration l Correction lEguivalent,1, IIsoto)e (i)I (uCi/cc) l Factor l (uci/cc) l
', i Kr-33m i :.2.57 x 10-4 l l Kr-85m l ..3.98 l l Kr-85 I i l5.48 x 10-2 l l l Kr-87 l ll17.1 l l Kr-88 I ' 50.0 l l Kr-89 I 55.9 l l Kr-90 1 53.1 l r l Xe-131 m l . 3.11 x 10-1 l
\ l Xe-133m l uB.5T x 10-1 l l Xe-133 I ill l l Xe-135m l ll10.6 l l Xe-135 l 1,6.16 l
, l Xe-137 l l 4.83 l l Xe-138 l 1 30.0 l l Ar-41 1 1 30.1 l 1 l Total Dose Equivalent: I l Xe-133 l l Convert the Dose Equivalent of Xe-133 to a " release rate" by multiplying by the stack flow rate:
uCi DEX e-133 (in. uCi/cc) x 1.42 x 108 cc/sec = sec or by the RBSYS flow rate: uCi DEXe-133 (in. uCi/cc) x 4.50 x 105 cc/sec = sec depending on the monitor that was sampled / analyzed. O Rev. 0 5/3/83
O L. iC EMERGENCY PREPAREDNESS SIGNOFF FORM To: Holder, Manual No. /pg Shoreham Nuclear Power Station Emergency Plans and Implementation Procedures From: Emergency Planning Coordinator
Subject:
REVISION TRANSMITTAL f Attached is Revision No. CP to the shoreham Nuclear Power Station Emergency Plan and Implementation Procedures for insertion into your Controlled Copy No. (0 4 . Two Copies of this memorandum have been provided. Once you have complied with the instructions presented below, sign one copy and place it in front of your controlled copy of the Manual. Sign the second copy of the memorandum and return it . within seven (7) working days to Emergency Preparedness Coordinator >., . I LILCO 175 East Old Country Road Hicksville, NY 11801 l' The undersigned acknowledges completion of the following actions:
- 1. The receipt and incorporation of Revision No. dated in controlled Copy No. of the shoreham Nuclear Power Station Emergency Plan and Implementation Procedures.
- 2. Destruction of all pages which have been superseded by the revision.
Signature Date Rev. 0
CAL O 'l EMERGENCY PREPAREDNESS SIGNOFF FORM To: Shoreham Nuclear Power Station Holder,ManualNo.J Emergency Plans an Implementation Procedures From: Emergency Planning Coordinator Subject ~ REVISION TRANSMITTAL () to the Shoreham Nuclear Power Station Attached is Revision No. Emergency Plan and Implementation Procedures for insertion into your Controlled Copy No. Two copies of this memorandum have been provided. Once you haveldidl . complied with the instructions presented below, sign one copy and place it in front of your controlled copy of the Manual. Sign the second copy of the memorandum and return it within seven (7) working days to:
. Emergency Preparedness Coordinator
- j LILCO 175 Eastiold Country Road Hicksville, NY 11801 The undersigned acknowledges completion of the following actions:
i
- 1. The receipt and incorporation of Revision No. dated in Controlled Copy No. of the Shoreham Huclear Power Station Emergency Plan and Implementation Procedures.
- 2. Destruction of all pages which have been superseded by the revision.
Signature Date Rev. 0
* - . ~ v_.- . . , . . - - - - . _ - -___. _- ._
' . .-.... - . . . . - . . . . . . . . . - _ . . . ~ - . . . . , . . . ~ . _ . . . . . . _ . . . . . . - . . . . . .- l EPC 4.,c_ EPIP l-0 ant a ager Effective Datef 6/6/83 I COUTilEi. LED CCFY O i04 EPIP l-0 CLASSIFICATION OF EMERGENCY ACTION LEVELS b 1.0 PURPOSE To describe the method for classifying /re-classifying events. 2.0 RESPONSIBILITY Implementation of this procedure shall be the responsibility of f the Watch Engineer, the Emergency Director (ED), or the f Response Manager (RM).
- O 3.0 PRECAUTIONS l
j 3.1 When initiating conditions fall on the upper limit of an Emergency Action Level (EAL), the classification shall be made at the next higher level. 4.0 PREREQUISITES l l
- l. 4.1 The appropriate emergency operating, .larm response i and/or abnonnal occurrence procedures have been initiated.
l i l
- O Rev. 0 5/23/83
EPIP 1-0 Page 2 of 7 5.0 ACTIONS 5.1 Using an Event Classification Sheet (Attachment 1), place a check mark beside every applicable event category. The Emergency Guide Flowchart ( Attachment 2) is used for quick reference to the tabbed categories. I i l CAUTION I i l l MORE THAN 1 EVENT CATEGORY MAY APPLY l i I 5.2 For the event category checked, turn to tha corresponding ! Event Category tab. l O 5.3 Review the list of generic initiating conditions, select l the one(s) applicable, place a check mark next to the l corresponding classification and number (s) on the Event Classification Sheet. l l l CAUTION I I I I MORE THAN 1 INITIATING CONDITION MAY APPLY I I l 5.4 Turn to the tab of the highest classification for which an initiating condition was observed. 5.5 Select the applicable EAL(s) and enter a simple descrip-tion of the EAL on the Event Classification Sheet. !O l Rev. 0
, 5/23/83
r-EPIP 1-0 Page 3 of 7 l l l CAUTION I I I 4 l IF THE EVENT HAS BEEN CLASSIFIED A GENERAL l l EE RGENCY, PROCEED TO STEP 5.8. I I I 5.6 If more than one event category has been checked, repeat steps 5.2 through 5.5 for each category. 5.7 Select a final classification based on the category, initiating conditions, and EAL which results in the highest accident classification. 5.8 Record the final classification in the appropriate space O at the top of the Event Classification Sheet. 5.9 Implement one of the following procedures based on the final classification:
- a. Unusual Event EPIP 1-1
- b. Alert EPIP 1-2
- c. Site Area Emergency EPIP 1-3
- d. General Emergency EPIP 1-4
6.0 REFERENCES
6.1 EPIP 1-1, Unusual Event 6.2 EPIP 1-2, Alert EPIP l-3, Site Area Emergency 6.3 6.4 EPIP 1-4, General Emergency O Rev. 0 5/23/83
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EPIP 1-0 Page 4 of 7 O 7.0 ATTACHMENTS
- 1. Event Classification Sheet
- 2. Emergency Action Levels O
l O Rev. 0 5/23/83 t
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t. EPIP l-0 i Page 5 of 7 j. Attachment 1 i Page 1 of 3 r 1 EVENT CLASSIFICATION SHEET ! I i I Date Time i Perfomed by Final Classification I CLASSIFICATION APPLICABLE EVENT CATEGORY (IES) TAB AND NUMBER SPECIFIC EAL l Abnomal Primary Leak Rate 1. UE 5 i i Alert 5 I Site 1 ! GE 6c ,! t Abnomal Core Conditions and 2. UE 3 ! Fuel Damage UE 4 Alert 1 l Alert 9 ! Site 2 '. GE 2 - i GE 6b i I Steam Line Break or MS RV/SY 3. UE 6 ; Failure Alert 4 [ Site 4 : Other LCO's 4. UE 1 I UE 8 UE 9 l Abnormal Radiological 5. UE 2 Effluent or Radiation Levels Alert 6 , Alert 15 Site 13 GE 1 Rev. 1 g g 5/2g
EPIP l-0 Page 6 of 7 Attachment 1 Page 2 of 3 EVENT CLASSIFICATION SHEET (continued) CLASSIFICATION APPLICABLE EVENT CATEGORY (IES) TAB AND NUMBER SPECIFIC EAL Loss of Shutdown Functions: 6. Alert 10 Decay Heat or Reactivity Alert 11 Site 8 Site 9 GE 6a GE 6d Electrical or Power Failures 7. UE 7 Alert 7 Alert 8 Site 6 Site 7 Fire 8. UE 10 Alert 13 Site 11 GE 7 Control Room Evacuation 9. Alert 10 Site 18 Loss of Monitors, Alanns, 10. UE 11 etc. Alert 14 Site 12 Fuel Handling Accident 11. Alert 12 Site 10 Rev. 0 5/23/83 O O O
EPIP l-0 Page 7 of 7 Attachment 1
. Page 3 of 3 EVENT CLASSIFICATION SHEET (continued)
CLASSIFICATION APPLICABLE EVENT CATEGORY (IES) TAB AND NUMBER SPECIFIC EAL Hazards to Plant Operation 12. UE 4 Alert 8 Site 16 GE 7 Security Threats 13. UE 2 Alert 6 Site 14 GE 3 , Natural Events 14. UE 3 Alert 7 - Site 15 GE 7 Others 15. UE 15 UE 16 Alert 19 : Site 17 ' GE 4 Rev. 0 5/23
Also Avadabb On i Aperture Card l SHOREH AM NUCLE EMERGENCY CLASSIFIC
, EVENT CATEGORY UNUSUAL EVENT ALERT 13 A8 NORMAL PRINARY LEAR RATg EXCEEDING EITHER PR iuaR Y /S ECOND ARY - PRIMARY COOLANT LE AK R ATE LEAN R ATE TECHNICAL SPECIFICATION OR GREATER TMAN 50 CPM PRIW4RT SYSTEM LF AK RATE TECHNICAL SPECIFICATION 23 ABNORMAL CORE CONOITIONS AND FUEL DAMAGE INDICATION SEVERE LOSS OF FUEL CLACC FUEL DAhlAGE \ ABNORMAL COOLANT TEWP AND /04 COOLANT PUW SEIZURE LEA" PRESSURE OR AENORMAL FUEL TEWS. TO FUEL FAILbRE. l WHICM EXCEED TECH. SPEC. LIMITS.
- 3) STEAM LINE BREAK OR kS RV/SV FAILURE OF A S AFETY OR RELIEF STEAM LINE B ACAM wlTM WSi FA8 LUR E VALVE IN A SAFETY RELATED Sv3 FEM MALFLNCTION CAUSING LEAAD TO CLOSE FOLLOWING A REDUCTION OF APP *.lCA8tE PRESSURE
- 4) OTHIER LCO'S LOSS OF CONTAtNENT INTEGRITY N REQUIRING lhKDIATE SHUTDOWN SY TECH. SPEC.
LOSS OF ENGINEERED S AFETY FEAT'JRE OR
\ FIRE PROTECTION FUNCTION REQUIRING SHuTDoulN SY TECH. SPEC.
EERGENCY CCRE COOLING SYSIDI (ECCSI INITIATED AND DtSCHARGED TO VESSEL
$) ABNORMAL R ADIOLOGICAL EFFLUENT R A010 LOGICAL EFFLUENT TECHNICAL ' R ADi& TION LEYELS OR AIR 801 OR R ADI ATION LEFELS. SPECIFICATION LIMITS EXCEEDED. CONT AW8MA TION WHICH INDICA SEVERE DEGRADAfloM IN CON 1 CF RADICACTivE MATERI ALS. l RADIGLOOf CAL EFFLUENTS CREJ TNAN 10 fines TECH. SPECS.
INSTANTANEQUS L1riliS.
- 8) LOSS OF SduTDow4 FUNCTIONS: FAILURE OF REACTOR PPOTEC DECAY NEAT OR REACTIVITY STSTEW To INaTI ATE AND C001 A SCR AM WHICH BRINGS THE SU8 CRITICAL COMPLETE LOSS OF AP(Y FUM IEEDED FOR PLANT COLD SHU'
- 7) ELECTRICAL CR POER Loss CF CFFSITE PonER OR LOSS LOSS Or OFFSITE PowR ANoi FAsLURES Or ONSITE AC power CAP A8tLITY OF ALL CNSITE AC POWER.
LOSS Or ALL CMSITE DC pow 1 83 FIRE FIRE WITHIN TkE PLANT LASTING WORE FIRE POTENTI ALLY AFFECTIC TH AN e0 WINUTES. SYSTEWS. 91 CCNTROL R00W EVACUATION EVACU ATICN OF CONTROL Roci ANT!CIP ATED OR REQUIRE 3 wel CCNTROL OF SNUTDOWN SYSTJ ESTA8LtSHED FRold LOCAL ST tt) LOSS OF MONITORS, INosCATI0ltS OR ALARMS ON PROCESS OR - MOST OR ALL ALARWS (ANNUN ALARMS. ETC. F'FLbENT P AR A&ETERS NOT FUNCTION AL IN CONTROL ROOM TO AN EXTENT REQUIRING PL ANT SHuTD01pn OR CTNER StG8HFICANT LOSS OF ASSESSIENT OF Coim4ANtCAT10N CAP ABILIT F. Lil FUEL 14AMILING ACCIDENT 0 gy yy o yfg g M APOLING BUILDING.
- 12) MAZARDS TO PLANT CPERATIONS N AZARDS SEING EXPERIENCED CR SEVERE M A2 ARDS 6EING EXPE' PROJECTED TM AT AFFECT *LANT OR mROJECTED POTENTI ALLY CPERAf f 3NS. AFFECT;NG S AFETY SYSTEWS.
131 SECURITY THREATS SECURITw TuREAT. ATTEMP*ED ENTR Y ON00ime SECbRITY COMPR04d CR ATTEv'ED S A80TAGE
- 14) NATURAL EVENTS NATURAL PMENC E NA BEthG EXPEP'ENCED - SEVERE NATbM AL PwENCnENA OR P904f CTED SETON 0 USU AL LEVELS EXPERIENCED CR PROJECTED.
l ISI OTHERS OTHER PL &MT CONolTIONS exist W AT OTHER PLANT CONDFf!ONS Ex: WARR ANT INCRE ASED AWAREMESS CN *W E WAPR ANTING PRECAdTION ART R AMT OF A PLAMT OPER ATING ST AFF OR ACTivAT;0N OF TNE TSC. STATE arc /OR LOCAL OFFSITE AUTHORITIES OR RE0utRE PL ANT SNUTDOWN L*NDER TECNICAL SPEC 4FICATICN REQQlREhENTS 04
, INVOLVE OTkER THAN NORMAL CONTROLLED SMUTDo est.
( TR AlsSPORT& TION OF CONTAMINATED INJURED INDIVIDUAL FRott SITE f TO ,OFFSITE HOSPITAL.
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R
- AR POWER STATION '
eTION GUIDE FLOWCH ART SITE AREA EMERGENCY GENERAL EMERGENCY abOWN LOSS Or CCOL ANT ACCIOENT S"4LL OR L ARGE Breast LOCA OCCURS ILOCal GREATER TH AN MAKE%P CAP AC ATY. aNO CONTAINMENT PERFORMANCE l$ UNSUCCESSFUL AFFECTING LONGER TERM SUCCESS CF TNE ECCS COULD LE AD TO CCRE DEGR ADATION 04 MELT IN SEVERAL MouRS WITHOUT CONTAINENT BOUNDARY. Q j DECR ACED CORE wtTM POSS'BLE LCSS OF 2 OF S FISSION PR00uCT 'gg LOSS OF C00LABLE CECIE TRY B ARRIERS wlTM POTENTI AL LOSS OF THE TM4RD B ARRIER SMALL OR L ARGE LOCA'S WITH Fall.URE CF ECCS TO r PERFORM LEADING 70 CORE WLT CEGR ADATION OR STE AM LINE BRE Alt OUTSIDE WLT IN Mim'JTES TO NOURS. LOSS OF CONTAINMENT >E. CONTAINENT WITHOUT ISOLATION. INTEGRITY MAY BE IMMINENT, NE EFrLUENT MONITORS DETECT LEVELS
'I S EFFLUENT MONITORS DETECT LEvtLS CORRESPONDING 70 ORE ATER TH AN S0 MR/HR CCRRESPONDING To i REM /MR w.8. OR S REM /kR ROL FOR 1/2 hour OR GREATER THAN $00 MR/WR Twveoso AT TWE SITE BOUNDARY UNttR ACTUAL W S. FOR Two MINUTES (CR FIVE T6MES THESE METEOROLCGICAL CONolTIONS.
TER LEVELS 70 THE TYRolD) AT THE SITEBOUNDART gCR 40yEjgSg MgTgoNgt0Gy* TWESE DOSE RATES ARE PROJECTED B ASED ON THESE 00$E R ATES ARE PROJECTED E ASED ON OTHER PLANT P AR AMETERS OR ARE W ASURED OTHER PL ANT P AR AWTERS OR ARE ME ASURED IN THE ENYlRCNS. IN THE ENVIRONS. EPA PROTECTIVE ACTION QultfLINES ARE PROJECTED SITE 70 BE EXCEEDED OUTSIDE THE SOUpCART. Tion TR ANSIENT REQUIRING OPER ATION OF TRANSIENT (E.G. LOSS OF orFSITE power) PLUS FAILURE PLE TE SM'JTDOWN SYSTEMS wlTW F AILURE TO 4EACTQ SCN AM (CONT 18AJED POWER GENERATION OF REcutSITE CORE SHUT DOWN S YS TEMS (E e.. SCR AMI. wtTH NO CORE DAMAGE IMEDI ATELT Cout.D LE AD TO CORE MELT 54 SEVER AL McLRS wnTM EvlCENT). CONTAINMENT FAILURE ListELY. MORE SEVERE CONSEQUENCES IF Putr TRIPS DOES NOT FUNCTION. 710el LOSS OF ANY FUNCTION tlEIDED TDowN. SMUTDOWN OCCURS SUT REQUISITE DECAV HEAT REMOVAL FOR PLANT HQT SHUTDOWN. SYSTEM IE G . RMR) OR NON! AFETT SYSTEMS MEAT REMOVAL MEANS ARE REMOCTED UNAwalLA8LE. CORE DEGR ADATl0N OR LISS MELT COULD OCCUR IN ABOUT TEN HOURS WITH SUBSEQUENT LOSS OF 0FFSITE P0uER AND LOSS OF CONTalNENT FAILURE. ALL ONalTE AC poser FOR MORE THAN 13 MlpL [ R. LOSS OF ALL vlTAL ONSITE QC power FOR MORE TNAN l$ MIN.
- SAFETY FIRE COMPRONISIW THE Fue10TIONS OF S AFE TY S YS TEMS. ANY MAJOR tNTERNAL OR EXTERNAL EVENTS (EG FIRES. EARTHCUAltES SUBS'AN TI ALLY BEYOND DEslGN 84563) wNiCM COULD C AUSE MAS $lvE commons CAMAGE TO PLANT SYSTEMS u EVACUATION OF CONTROL ROOM AND T **
CONTROL OF SMUT 00ww SYSTEMN NOT IMS ESTA8LSMED FROM LOCAL STATIONS g.p . LTIONS. e IN 1$ bl N. 4-
~
ICf. TORS! LOST.- MOST OR ALL AL ARMS ( ANNUNCI ATORS) LCST - l I"_ AND Pt ANT TR ANSIENT INITI ATED OR IN PROGRFSS. AI i RELEASE CF MAJOR DAMAGE TO SDENT FUEL I4 4T OR FUEL CONTAINI(NT OR FUEL HANOLING BUILQlNG. [ ' "ENCED SEVEME wa7 AR05 BEING EXPER:ENCID OR PRO.ECTED Tu aT CowROM*SE TME ANY %IAJOR IN'ERNAL 09 EXTEoN AL Evna 'S (E3 FGES, E ARTHCU ARES SUBS T Ah %L* FUNCTIONS or S AFETY Sv3TEMS. GEYONO OESIGN 8ASl5) WCCM C OLLO CA.:!E REACTOR ae0T IN COLD SHUTDOWN. MASSIVE COMMON CAMAGE TO PLANT S f S TE '.13
.E INMiNENT LOSS OF PHYSICAL CONTROL OF PLANT. LOSS CF PHYSICAL CONTROL CF THE FAC:LdTY B EiPsG SEVE9E 'eATUR AL PMENouEN A BE'NG ANY MA,,0R rNvEa4At CR Ex' eon AL EvrNr$
E XPf R:ENCED CR PRosECTED 'b4T (EG F tRES. EARTHOU AuES SLBS T ANTI ALLY COMPROMISE THE FUNCTIONS OF $4FETv BEv0ND CESIGN B ASIS) AC Cm COLLO CA:.JSE SYS,TEMS. REACTOR NOT sN COL 3 SMUTCow4 MASS.vf COMMON CAMAGE To PL ANY SYSTEMS ET O THF R PL ANT CCNDITIONS Ez?$T OTHER PLANT CCNCITIONS Er:ST, FpCW W ATEbEm A&RR ANTING ACTlvatf 0N OF EOF. Sol., ACE. Tw AT NAAE RELEASE OF L A4GE recWS CF EW%ENCY CENTERS AND MolilTORiNG TE AIAS. CR ISSU ANCE R ACIC ACTivl TV im A SMCRT TIE PERIOD P MSIBLE. EG ANT CCPE WLT SITU A TiCN. C# A PRECAUTIONARY NO TIFICA*0N TO THE PUBL!C NE AR TME SITE. 1 EMERGENCY CLASSIFICATION OU!CE FLOWCHART i 11559.Def.23 slCL ,el FWI 9.58111
, _m . . . . . _ _ ._ ,
Postulated Acci:ients Analyzed ,j In Shoreham FSAR-Chapter 15 FSAR Comparison with Shoreham Section Accident Title EAL's 15.1.1 Generator Load Rejection ECCS initiated'- covered (With Failure of Bypass) by Unusual Event No. 1 15.1.2 Turbine Trip ECCS initiated - covered (With Failure of Bypass) by Unusual Event No. I 15.1.3 Turbine Trip with Failure Bounded by 15.1.2 of Generator Breakers to Open 15.1.4 MSIV Closure ECCS initiated - covered by Unusual Event No. 1 15.1.5 Pressure Regulator Failure No thermal margins are
- Open significantly affected -
if necessary, ECCS will be initiated 15.1.6 Pressure Regulator Failure Bounded by 15.1.2
- Closed 15.1.7 Feedwater Controller Failure Bounded by 15.1.2
- Maximum Demand
- C 15.1.8 Loss of Feedwater Heating Minimum Critical Power .
Ratio (MCPR is above 1.06 - Notification not required 15.1.9 Shutdown Cooling (RHR) No thermal margins are-Manlfunction - Decreasing significantly affected Temperature 15.1.10 Inadvertent HPCI Pump MCPR is above 1.06 - Start Notification not required / 15.1.11 Continuous Rod Withdrawal MCPR is above 1.17 - During Power Operation Notification not required 15.1.12 Continuous Rod Withdrawal Hot considered credible During Startup Operation 15.1.13 Control Rod Removal Error Criticality cannot occur During Refueling with only one rod removed
- interlocks prevent more rods being removed 15.1.14 Fuel Assembly Insertion No consequences Error During Refueling Page 1 of 3 l
I 4.
~ .
FSAR Comparison with Shoreham fs Section . Accident Title EAL's U 15.1.15/ Inadvertent Loading-and No significant 15.1.16 Operation of Fuel Assembly consequences in Improper Location 15.1.17 Inadvertent Opening of No significant Safety / Relief Valves consequences 15.1.18 Loss of Feedwater Flow ECCS initiated - covered by Unusual Event No. 1 15.1.19 Loss of A.C. Powen ECCS initiated - covered by Unusual Event No. 1 15.1.20 Recirculation Pump No fuel failures - Trip Bounded by Alert No. 9 15.1.21 Loss of Condenser Vacuum .ECCS initiated - covered by Unusual Event No. 1 15.1.22 Recirculation Pump No fuel failures - Seizure Eounded by Alert No. 9 15.1.23 Recirculation Flow Control Bounded by Alert No. 9 Failure - Decreasing Flow 15.1.24 Recirculation Flow Control MCPR is above 1.06 - Failure - Increasing Flow Notification not required 15.1.25 Abnormal Startup of No significant Idle Recirculation Pump consequences 15.1.26 Core Coolant Temperature Loss of RHR System - Increase covered by Alert No. 10 15.1.27 Anticipated Transients Not Design Basis - l Without Scrams.,(ATWS) covered by Alert No. 11 15.1.28 Cask Drop Accident Not considered credible - covered by Site Area Emergency No. 10 15.1.29 Miscellaneous Small Bounded by 15.1.24 and
-Releases outside Containment 15.1.37 15.1.30 Instrument Line No fuel failure - small Failure LOCA covered by Unusual Event No. 5 - Potential Radiological releases
- covered by Alert No. 15
- and Site Area Emergency No. 13 k') Page 2 of 3 l
FSAR Comparison with Shoreham Section Accident Title TAL's O 15.1.31 Main Condenser Gas Potential radiological Treatment System Failure releases are covered by Alert No. 15 15.1.32 Liquid Radwaste Tank Potential radiological Rupture releases are covered by Alert No. 6 and Alert No. 15 15.1.33 Control Rod Drop Potential radiological Accident releases are covered by Site Area Emergency No. 13 15.1.34 Loss of Coolant Accident Assumed fuel failure - Potential radiological releases are covered by Site Area Emergency No. 13 15.1.35 Steam Line Break No fuel failure - Accident . Potential radiological releases covered by Site Area Emergency No. 13 n' 15.1.36 Fuel Handling Potential radiological l Accident released covered by Site Area Emergency No. 13 15.1.37 Feedwater Line Break No fuel failure - Ejector Lines Potential radiological l releases are covered by Alert No. 6 and Alert No. 15 15.1.38 Failure of Air Ejector No fuel failure - Lines Potential radiological releases are cover'ed by l Site Area Emergency l No. 13 l Note: Potential radiological releases are based on accident analyses using conservative NRC analysis assumptions. G [ [~/ s Page 3 of 3 l N-
EVENT CATEGORY 1 () ABNORMAL PRIMARY LEAK RATE INITIATING CONDITIONS UNUSUAL EVENT NO. 5 Exceeding a primary system leak rate Technical Specification. ALERT NO. 5 Primary coolant leak rate greater than 50 gpm with the reactor at operating temperature and pressure. SITE AREA EMERGENCY NO. 1 Known loss of coolant accident greater than makeup pump capacity. GENERAL EMERGENCY No. 6c LOCA occurs and containment performance is unsuccessful affecting longer term success of the ECCS. Could lead to core degradation or melt in several hours without containment boundary. O 6
'x
\
- - rw r ~ 't -V
UNUSUAL EVENT !!O. 5 O Initiating Conditions Exceeding a primary system leak rate Technical Specification. Emergency Action Levels ,
- 1. Drywell Leakage
- a. Rad Waste Sump Trouble Alarm on 1H11*MCB-01 (ARP-0423 RW SUMPS TROUBLE) and a Drywell Equipment Drain Tank Pump Frequent Start or Excessive Run Alarm both on 1G11-PNL-047 (ARP 4828 DRYWELL EQPT DRAIN
' TANK PUMP FREQUENT START or ARP 4835 DRYWELL EQPT DRAIN TANK PUMP EXCESSIVE RUN) indicating that Drywell Equipment Drain Tank Pump 1G11-P-032A or B is starting more than every 12 minutes or after pump has
, stopped' running longer than 12 minutes each time. EE
- b. Rad Waste Sump Trouble Alarm on 1H11*MCB-01 (ARP-0423 RW SUMPS TROUBLE) and a Drywell Floor Drain Tank' Pump Frequent Start or Excessive Run Alarm both on 1G11-PNL-047 (ARP-4829 DRYWELL FLOOR DRAIN TANK PUMP FREQUENT START or ARP-4P36 DRYWELL FLOOR DRAIN TANK PUMP EXCESSIVE RUN) indicating that Drywell Floor Drain Tank Pump 1G11-P-161A or B is starting more than every 12 minutes or after pump has
_s stopped running longer than 12 minutes each time. _O_R
- c. High Drywell Air Cooler Condensate Leakoff Flow Alarm on panel 1H11
- PNL-VC2 (ARP 3053 DRYWELL UNIT CLR 17A FLOW HI and ARP-3056 DRYWELL UNIT CLR-17B FLOW HI) indicating flow in excess of 0.4 gpm SE
- d. Routine determination by the Watch Engineer based upon observation and calculation that either there has been a 2 gpm increase in unidentified leakage in any 4 hour period or the conditions of TS 3.4.3.2d are violated.
Logic Diagram See Separate Sheet I 4 UE 5 Page 1 of 2
- /<
t UNUSUAL EVENT NO. 5 O LOGIC DIAGRAM l l RAD WASTE I SUMP TROUBLE 2GPM LEAKAGE ALARM INCREASE IN 4 HOURS 1 OR T.S.3.4.3.2 d ! VIOLATED l DRYWELL EQUIPMENT ORYWELL FLOOR HIGH DRYWELL ; DRAIN TANK DRAIN TANK AIR COOLER FREQUENT START FREQUENT START CONDENSATE 1 OR EXCESSIVE OR EXCESSIVE LEAKOFF FLOW l RUN ALARM RUN ALARM ALARM O 1 l l ! UNUSUAL I EVENT NO. 5 ., I l l l l
)
O- FIG. UE5 -l l JE 5 = AGE 2 CF 2
)
. ALERT NO. 5 Initiating Conditions Primary coolant leak rate greater than 50 gpm with the reactor at operating temperature and pressure. .*
Emergency Action Levels 1.a. Drywell Equipment Drain Tank Hi Level Alarm on psiel 1H11*PNL-602. (ARP-0011 DW EQP DRN TK LEVEL HI) with indication on Drywell Floor and Equipment Drain Level Recorder 1G11-LR-505 on panel 1G11-PNL-047 indicating level in excess of 61 inches. . i AND
- b. Indication on panel 1G11-PNL-047 that Drywell Equipment Drain Tank Pump is running.
SE . 2.a. Rad Waste Sump Trouble Alarm on panel 1H11*MCB-01 (ARP-0423 RW SUMPS TROUBLE) and a Drywell Equipment Drain Tank Pump Excessive Run Alarm on panel 1G11-PNL-047 (ARP-4835 DRYWELL EQPT DRAIN TANK PUMP EXCESSIVE ( RUN) indicating that Drywell Equipment Drain Tank Pump 1G11-P-032A i or B is running longer than 12 minutes each time. . !O l
-o
- b. Level Recorder 10(1*LR-505 on panel 1G11-PNL-047 or Level Recorder 1G11*LR-505X on panel 1H11*PNL-602 indicates that the level in the Drywell Equipment Drain Tank does not decrease.
SE
?.a Drywell Floor Drain Tank Hi Level Alarm on panel 1H11*PNL-602 (ARP-0012 DW FLOOR DRN TK LEVEL HI) indicating a level in excess of 61 inches.
AND
- b. Indication on panel 1G11-PNL-047 that a Drywell Floor Drain Tank Pump is running.
OR
- 4.a Rad Waste Sump Trouble Alarm on 1H11*MCB-01 (ARP-0423 RW SUMPS j 1 TROUBLE) and a Drywell Floor Drain Tank Pump Excessive Run Alarm on l panel 1G11-PNL-047 (ARP-4836 DRYWELL FLOOR DRAIN TANK PUMP EXCESSIVE RUN) indicating that Drywell Floor Drain Tank Pump 1G11-P-161A or B is running longer than 12 minutes each time
() A 5 Page 1 of 3
AND
- b. Level Recorder 1G11-LR-505 on panel 1G11-PNL-047 or Level Recorder 1G11*LR-505X on panel 1H11*PNL-602 indicates that the level in the Drywell Floor Drain Tank Pump does not decrease.
Logic Diagram See Separate Sheet e 1 9 g O . i f i A 5 Page 2 of 3
- ,--.-, -----. ,-,..-.- - ~. -,-.
6
.O ALERT NO. 5 LOGIC DIAGRAM DRYWELL EQUIPMENT RAD WASTE DRYWELL FLOOR DRAIN TANK SUMP TROUBLE DRAIN TANK HIGH LEVEL ALARM ALARM HIGH LEVEL ALARM i
DRYWELL DRYWELL DRYWELL FLOOR EQUIPMENT DRAIN EQUIPMENT DRAIN DRAIN TANK TANK PUMP ' TANK EXCESSIVE EXCESSIVE RUN
^I" ^"
RUNNING RUN ALARM ALARM O .
~
LEVEL IN LEVEL IN DRYWELL DRYWELL EQUIPMENT DRAIN FLOOR DRAIN IS NOT DECREASING IS NOT DECREASING ALERT N O. 5 FIG A 5 -l A5 PAGE 3 OF 3 f
~
, SITE AREA EHERGENCY NO. 1
(~ L) - Initiating Conditions: Known loss of coolant accident greater than make up pump capacity.
. Emergency Action Levels .
- 1. Reactor water level instrument 1B21-L1-007 on panel 1H11*PNL-601 indicates that reactor water level cannot be maintained above top of active fuel zone
(-158 inches) E
- b. Drywell Containment Area Temp Hi Alarm on panel 1H11*PNL-VC2 (ARP-3096 DRYWELL AREA TEMP HI) when any of the following conditions are satisfied:
- 1. Supply air from the Drywell Unit Coolers into the drywell exceeds 1108F (as indicated by IT47-TRS010 recorder).
E 1 2. Zone 2 (annular space between reactor pressure' vessel and biological shield) or Zone 3 (CRD area) temperature exceeds 1708F i (as indicated by IT47-TRS010 recorder on panel 1H11*PNL-VC2). E
- 3. Return air from: primary containment; lone 2; drywell floor; or vicinity of return duct inlets exceeds 1558F (as indicated by IT47-TRS020 recorder on panel 1H11*PNL-VC2). ,
E
- 4. Return air from: Drywell Head Area; RPV head flange area; or vicinity of refueling bellows and bulkhead assembly exceeds 190eF (as indicated by IT47-TRS030 recorder on panel 1H11*PNL-VC2).
E '
- c. Radiation Monitoring System Div. A Common Alarm on panel 1H11-PNL-601 (ARP-1401 DIV I RAD MONITOR RAD HI) with indication on recorder 1D11*RR111 on panel 1H11*PNL-080A of high gaseous or particulate radiation in the drywell.
AND -
- 2. a. High Drywell Pressure Alarm on panel 1H11
- PNL - 601 (System A, ARP-1128 DRYWELL SYS A PRESS HI or System B, ARP-1129 DRYWELL SYS B PRESS HI) indicating drywell pressure in excess of 1.69 psig l
SAE 1 Page 1 of 3 4 l l t k
Logic Diagram See Separate Sheet t l I ( lO
~
SAE 1 Page 2 of 3 O i l..-...____.._. ..__ -,_
I O SITE AREA EMERGENCY NO. I LOGIC DIAGRAM HIGH DRYWELL HIGH DRYWELL RADIATION PRESSURE CONTAINTMENT MONITORING TEMPERATURE SYSTEM DIV. A ALAR M ALARM COMMON ALARM REACTOR WATER CANNOT BE MAINTAINED ABOVE TOP OF ACTIVE FUEL ZONE (-158") l SITE AREA EMERGENCY NO. l l 1 FIG. SAE f-l SAE I PAGE 3 OF 3
6 GENERAL EMERGENCY NO. 6c O Initiating Conditions
~
LOCA occurs and containment performance is unsuccessful affecting longer term success of the ECCS. Could lead to core degradation or melt in several hours without containment boundary.
- Emergency Action Levels
- 1. On-duty Watch Engineer has determined that a LOCA has occurred.
AND
- 2. Failure of containment cooling indicated by:
- a. Process Computer, Pt L993 and L994 indicating drywell temperature above 2968F and rising SE
- b. Containment cooling has become inadequate and remains inadequate for 1/2 hour as indicated by: s
- 1) Containment spray valves 1E11*MOV-038A or 039A and.1E11*MOV-038B or 039B have remained closed as shown by position indicating lights on panel 1H11*PUL-601 and cannot be opened.
O SE
- 2) On-duty Watch Engineer concludes from other indications that-containment cooling has become and remains inadequate.
l Logic Diagram ? See Separate Sheet l [ i l GE 6c Page 1 of 2
GENERAL EMERGENCY NO. 6C O LOGIC DI AGRAM ON SHIFT WATCH ENGINEER HAS DETERMINED THAT A SM ALL OR LARGE LOCA HAS OCCURRED PROCESS COMPUTER, ON SHIFT WATCH CONTAINMENT PT L 993 AND PT L 994 ENGINEER JUDGEMENT SPRAY VALVES THAT CONTAINMENT INDICATE DRYWELL HAVE REMAINED TEM PERATURE COOLING HAS BEEN CLOSED FOR INADEQUATE FOR ABOVE 296* F AND I/2 HOUR STILL RISING 1/2 HOUR GENERAL EMERGENCY NO. 6C SE-So-' 'O - GE GC PAGE 2 OF 2 l l l l
EVENT CATEGORY 2 ABNORMAL CORE (~' CONDITIONS AND FUEL DAMAGE INITIATING CONDITIONS
, UNUSUAL EVENT NO. 3 Fuel damage indication. Examples:
- 1. High offgas at BWR air ejector monitor (greater than 500,000 uci/sec; corresponding to 16 isotopes decayed to 30 minutes; or an increase of 100,000
, uti/see within a 30 minute time period)
- 2. High coolant activity sample (e.g.,
exceeding coolant technical specifica-tions for iodine spike)
. UNUSUAL EVENT NO. 4 Abnormal coolant temperature and/or pressure or abnormal fuel temperatures outside of technical specification limits.
ALERT NO. 1 Severe loss of fuel cladding:
- 1. High offgas at BWR air ejector monitor (greater than 5 Ci/sec; corresponding to 16 isotopes decayed 30 minutes)
- 2. Very high coolant activity sample (e.g.,
300 uCi/cc equivalent of I-131) ALERT No. 9 Coolant pump seizure leading to fuel failure. SITE AREA EMERGENCY NO. 2 Degraded core with possible loss of coolable geometry (indicators should include instrumentation to detect inadequate core cooling, coolant activity and/or containment radioactivity levels). GENERAL EMERGENCY No. 2 Loss of 2 of 3 fission product barriers with a potential loss of 3rd barrier (e.g., loss of primary coolant boundary, clad failure, and high potential for loss of containment). GENERAL EMERGENCY NO. 6b LOCA's with failure of ECCS to perform leading to core degradation in minutes to hours. Loss of containment integrity may be imminent.
U!! USUAL EVEllT !!O. 3 O Initiating Conditions Fuel damage indication. Examples: ,
- 1. High offgas at BWR air ejector monitor (greater than 500,000 uCi/sec, corresponding to 16. isotopes decayed 30 minutes; or an increase of 100,000 uCi/see within a 30 minute time period)
- 2. High coolant activity sample (e.g., exceeding coolant technical specifications for iodine spike)
Emergency Action Levels
- 1. Cat II Common Radiation Monitoring Alert Alarm on panel 1H11*PNL-601 (ARP-1408 CAT II RAD M0!!ITOR RAD ALERT) and indication on the Radiation
- Monitoring System CRT display of the associated BWR Air. Ejector Offgas Monitors PM12A, PM12B, or PM14.
O_,R
- 2. Analysis of reactor coolant sample indicates activity in excess of 0.2 uCi/gm dose equivalent I-131 for Steady State Operation or 4 uCi/gm dose equivalent I-131 during transient changes in reactor power.
Logic Diagram See separate sheet UE 3 Page 1 of 2
^
O UNUSUAL EVENT NO. 3 LOGIC DI AGRAM CATEGORY H COMMON RADIATION MONITORING REACTOR COOLANT ALERT ALARM ACTIVITY GREATER THAN O.2 uCi/gm (STEADY STATE) OR 4uCi/gm (POWER CHANGES) DOSE EQUlVALENT INDICATION ON RMS OF 1-131 ASSOCIATED AIR EJECTOR OFFGAS MONITORS . PMl2A , PM128, PMl4 O l l UNUSUAL EVENT NO. 3
.i O
FIG UE 3- 1 UE 3 PAGE 2 OF 2
- - -- --- .,,_,,e
UNUSUAL EVENT NO. 4 O Initiating Conditions Abnormal coolant temperature and/or pressure or abnormal fuel temperatures outside of technical specification limits. Emergency Action Levels When any of the following plant Safety Limits are exceeded: l I .
- 1. Recorders 1C51*XR-802, IC51*XR-803A,B,C,D, on panel 1H11*PNL-603 indicates 1
thermal power greater than 25 percent of rated power. 1 l l AND EITHER i l a. Pressure Indicator 1C32-PI-003 on panel 1H11*PNL-603 indicates reactor 1 l vessel steam dome pressure less than 785 psig. i OR 4
)
- b. Recorder 1B21-XR-014 on panel 1H11*PUL-603 indicates reactor core flow less than 10 percent of rated flow.
O QR ; I
- 2. Minimum Critical Power Ratio (MCPR) is less than 1.06. This will be determined by the Periodic Core Performance Program, P1, indicating a Maximum Fraction of the Limiting Critical Power Ratio (MFLCPR) equal to or ,
- greater than one and a subsequent reactor trip due to a power or flo~w transient.
AND BOTH l
- a. Pressure Indicator IC32-PI-003 on panel 1H11*PNL-603 indicates reactor vessel steam dome pressure greater than 785 psig.
A__N_D ,
- b. Recorder 1B21-XR-014 on panel 1H11*PNL-603 indicates reactor core flow I
( is greater than 10 percent of rated flow. . l I i l l ER
- 3. Pressure Indication 1 (32-PI-003) on panel 1H11*PNL-603 indicates reactor l vessel steam dome pressure is greater than 1325 psig. ;
l UE 4 Page 1 of 3 l l
, l
__ . _ . - . . ._... - .-.__ -= _ .. - ._ _ . - _ d l t OR
- 4. Reactor water level instrument 1B21-LI-007 on panel 1H11*PNL-601 indicates
- reactor vessel water level cannot be maintained above top of active fuel
, zone (-158 inches). f Logic Diagram See separate sheet i 4 1 i i O l
^
UE 4 Page 2 of 3 O-
-,wge,.w,e*+y9,p=_g._.%7, - - - - - - - - - m -
gv.-- wr., *ym y3m
'B' w*-et- eem,- we t we*,- ,w ea +w w _ www m- w 4ew ygy-e.wg-._r, g. w -.--sw4,we,w.g- ,
9 -9
l UNUSUAL EVENT NO. 4 L GI DIAGRAM O l IN RA oR E EL RA O E GREATER HA CRITI LP g 25% CF R ATIO PRESSU RE WATER LEVEL IS BELOW THETOP R ATED POWER ( l.06 >!325 psig OF ACTIVE FUEL ( -158 ") l l I REACTOR REACTOR STEAM DOME PRESSURE < 10 % OF (785 poig .RARD W 1
)
REACTOR VESSEL STEAM DOME ~ PRESSURE
> T85 psig I
I REACTOR CORE FLOW > 10% OF R ATED FLO W UNUSUAL EVENT o NO. 4 O F I G. U E4- I UE4 PAGE 3OF3
ALERT NO. 1 O Initiating Conditions Severe loss of fuel cladding:
- 1. High offgas at BWR air ejector monitor (greater than 5 Ci/sec; corresponding to 16 isotopes decayed 30 minutes)
- 2. Very high coolant activity sample (e.g., 300 uci/cc equivalent of I-131)
Emergency Action Levels
- 1. a. Cat II Common Radiation Monitoring High Alarm on panel 1H11*PNL-601 (ARP-1407 CAT II RAD MONITOR RAD HI) and indication on the Radiation Monitoring System CRT display of the associated BWR Air Ejector Offgas Monitors PM12A, PM12B, or PK14.
SE
- b. Main Steam Line High Radiation Trip Alarm on panel 1H11*PNL-603 (ARP-1192 MAIN STM LINE HI RAD TRIP B for Main Steam Lines B and D; or ARP-1208 MAIN -STM LINE HI RAD TRIP A for Main Steam Lines A and C) indicating radiation levels exceeding 3x normal.
E O
- 2. Analysis of a reactor coolant sample indicates activity in excess of 300 uCi/gm dose equivalent I-131.
Logic Diagram See separate sheet l l l l A 1 Page 1 of 2 i {
O ALERT NO.I . LOGIC DI AGR AM CATEGORYH COMMON RADIATION MONITORING HIG H ALARM INDICATION ON RMS RE ACTOR COOL ANT MAIN STEAM OF A SSOCI ATED ACTIVITY GREATER LINE HIGH THA N 360 uCi/gm AIR EJECTOR OFF GAS MONITORS PMl2 A, RADI^T N T IP DOSE EOUIVALENT AL RM PMl2B , PM 14 I -131 O i i ALERT NO.I Fl6 A l-l Al PAGE 2 OF 2
-y rr,- y-pm--- wvry ,,-we --w- ,, ,e-, ,--,+--m- --
ALERT NO. 9 Initiating Conditions coolant pump seizure leading to fuel failure. Emergency Action Levels The combination of the following indications will define this condition:
- 1. Recirculation pump or motor high temperature indicated by a common high temperature alarm on panel 1H11*PNL-602 (ARP-1281 RECIRC PUMP A/B MTR TEMP HI) indicating any of the following:
- a. Pump motor thrust or guide bearing temperature in excess of 200ey,
- b. Pump seal cavity temperature in excess of 1608F.
- c. Pump motor air cooler circulating water discharge temperature in excess of 110sp,
- d. Pump motor bearing or seal cavity cooler circulating water discharge temperature in excess of 1208F.
gR
- 2. MG Set Generator Lockout Alarms on panel 1H11*PNL-602 (ARP-1254 MG SET O. A GEN LOCKOUT or ARP-1255 MG SET B GEN LOCKOUT) indicating a pump motor trip due to instantaneous differential protection or generator overcurrent protection activation.
AND
- 3. Turbine trip alarm together with Reactor Vessel High Level Trip Alarm l on panel IH11*PNL-603 (ARP-0140 MAIN TURB TRIPPED and ARP-1248 RX VESSEL HI LEVEL TRIP) indicating that reactor vessel high water level (Level 8-greater than 54.5 inches) has tripped the turbine resulting in a reactor scram.
l AND l
- 4. Fuel damage indication according to the Emergency Action Levels of l
Unusual Event No. 3. Logic Diagram See Separate Sheet A 9 Page 1 of 2 i
O ALERT NO.9 LOGIC DIAGRAM TEMP RA R ALARM LOC OUT l INDICATING PUMP MOTOR ALARMS HIGH TEMPERATURES TURBINE TRIP ALARMS INDICATING REACTOR VESSEL ! HIGH WATER LEVEL (>54.5 INCHES) l l l EAL FOR UE NO.3 EXCEEDED INDICATING FUEL DAMAGE ALERT NO. 9 i O FIG. A 9 - 1 A9 PAGE 2 OF 2
SITE AREA EMERGENCY NO. 2 () Initiating Conditions Degraded core with possible loss of coolable geometry (indicators should include instrumentation to detect inadequate core cooling, coolant activity and/or containment radioactivity levels). Emergency Action Levels l This condition will be defined by the following indications:
- 1. Severe loss of fuel cladding according to the Emergency Action Levels of Alert No. 1.
AND EITHER
- a. With recirculation pumps operating, recorder IB21-XR-014 on panel 1H11*PNL-603 indicates a 25 to 50 percent increase in core plate differential pressure for the existing jet pump flow and power level, compared to previous operating experience.
9E l b. With recirculation pumps not operating, reactor water level instrument IB21-LI-007 on panel 1H11*PNL-601 indicates reactor vessel water level cannot be maintained above top of active fuel (
-See separate sheet i
l
~
SAE 2 Page 1 of 2 I L-
. _- . , - - . . - . . - , - - . - . ._- .- __a__ __
O i SITE AREA EMERGENCY NO.2 l LOGIC DIAGRAM EAL'S FOR ALERT NO.I INDICATING SEVERE LOSS OF FUEL CLADDING FUEL ZONE INSTRUMENT 25 TO 50 PERCENTINCREASE INDIC ATES REACTOR IN CORE SUPPORT PLATE AP WATER LEVEL BELOW (R ECIRC PUMPS OPERATING) THE TOP OF ACTIVE FUEL (RECIRC PUMPS NOTOFER ATING) (-158 INCHES) i SITE AREA EMERGENCY NO.2 l FIG.S AE 2-1 S AE 2 PAGE 2 OF 2
GEllERAL E!!ERGEllCY !!O. 2 fl V Initiating conditions Loss of 2 of 3 fission product barriers with a potential loss of 3rd barrier (e.g., loss of primary coolant boundary, clad failure, and high potential for loss of containment). Emergency Action Levels Loss of any 2 of the 3 fission product barriers with a potential loss of the 3rd, as follows:
- 1. Clad failure and failure of primary coolant boundary with high potential for loss of primary containment.
- a. Clad failure indicated by:
- 1) EAL's for Site Area Emergency No. 2 exceeded.
EE
- 2) EAL's for Alert No.1 exceeded.
AND
- b. Primary coolant boundary failure indicated by:
O 1) High Drywell Pressure Alarms on panel 1H11*PNL-601 (ARP-1128 DRYWELL SYS A PRESS HI and ARP-1129 DRYWELL SYS B PRESS HI) indicating drywell pressure in excess of 1.69 psig as indicated by pressure instruments on panel 1H11*PNL-60, j 1H11*PIS-165 B, D and 1H22*PIS-165 A,C. EE
- 2) Drywell Containment Area High Temperature Alarm on panel 1H11*PUL-VC2 (ARP-3096 DRYWELL AREA TEMP HI). See Site Area Emergency No. 1 EAL for conditions which will cause the alarm.
gR
- 3) Radiation Monitoring System Div. I Common Alarm on panel 1H11-PNL-601 (ARP-1401 DIV I RAD MONITOR RAD HI) with indication on recorder 1DI1*RR111 on panel 1H11*PNL-080A of high gaseous or particulate radiation in the dryvell.
AND
- 4) Reactor Water Level Alarms on panel 1H11*PNL-601 (ARP-1144 RX SYS A LEVEL LO and ARP-1145 RX SYS B LEVEL LO) with
() GE 2 Page 1 of 8 4
indication on level instrument 1B21-LIS-029 on ' panel 1H11*PNL-601 of reactor water level below -132.5 inches and
) rapidly decreasing.
E
- c. Potential loss of primary containment indicated by:
- 1) Process Computer PT F655 and F656 indicate primary containment pressure in excess of 43.2 psig.
3 -
- 2) Process Computer, PT L993. and PT L994 indicate drywell temperature in excess of 296'F and rising.
.E
- 3) On-duty Watch Engineer's opinion based on visual observation of structural cracks or penetration failure, or other independently obtained information.
3
- 2. Failure of primary coolant boundary and primary containment with high potential for clad failure.
- a. Primary coolant boundary failure indicated by:
O E
- 1) High Drywell Pressure Alarms on panel 1H11*PNL-601'(ARP-1128 DRYWELL SYS A PRESS H1 and ARP-1129 DRYWELL SYS B PRESS HI) indicating drywell pressure in excess of 1.69 psig as indicated by pressure instruments 1H11*PIS-165 B, D and 1H22*PIS-165 A, C on panel 1H11*PNL-601.
E
- 2) Drywell Containment Area High Temperature Alarm on panel 1H11*PNL-VC2 (ARP-3096 DRYWELL AREA TEMP HI). See Site Area Emergency No. 1 EAL for conditions which will cause this alarm.
E
- 3) Radiation Monitoring System Div. I Common Alarm on panel 1H11-PNL-601 (ARP-1401 DIV I RAD MONITOR RAD HI) with indication on recorder 1D11*RR111 on panel 1H11*PNL-080A of high gaseous or particulate radiation in the drywell.
M GE 2 Page 2 of 8
- 4) Reactor Water Level Alarms on panel 1H11*PNL-601 (ARP-1144 RX SYS A LEVEL LO and ARP-1145 RX SYS B LEVEL LO) with
[3 indication on level instrum?nt IB21-LIS-029 on panel V 1H11*PNL-601 of reactor water level below -132.5 inches and rapidly decreasing. AI!E
- b. Primary containment failure indicated by:
- 1) On-duty Watch Engineer observes from valve status indicators thaE all containment penetrations required for isolation are not valved off or closed, noting that during an accident closure by a single valve rather than two valves in series will suffice.
OR
- 2) On-duty Watch Engineer's opinion, based on visual observation or other independently obtained information, such as unexplained decrease in Suppression Pool level or
, unexplained increase in secondary containment activity.
, AND
- c. Potential clad failure indicated by:
O 1) Reactor water level indicator 1B21-LI-007 on panel 1H11*PNL-601 shows reactor vessel water level below top of active fuel zone (-158 inches) and still falling. 3
- 2) On-duty Watch Engineer's opinion, based on other indications in absence of falling water level, that ECCS has failed
- 3. Clad failure and primary containment failure with high potential for primary coolant boundary failure.
- a. Clad failure indicated by:
- 1) EAL's for Site Area Emergency No. 2 execeeded.
E
- 2) EAL's for Alert No. 1 exceeded.
l A,,N_,g l
- b. Primary containment failure indicated by:
I -
\ GE 2 Page 3 of 8 i
{J
. . ~ ,
4
. 1) On-duty Watch Engineer observes valve status indicators that 7s all containment penetrations required for isolation are not ! ( valved off or closed, noting that during an accident closure by a single valve rather than two valves in series will suffice. .
EE
- 2) on-duty Watch Engineer's opinion, based on visual observation or other independently obtained information, such as unexplained decrease in Suppression Pool level or unexplained increase in secondary containment activity AND
- c. Potential failure of primary coolant boundary indicated by:
- 1) Pressure Indicator IC32-PI-003 on panel 1H11*PNL-603 indicates reactor pressure in excess of 1250 psig and increasing.
EE
- 2) On-duty Watch Engineer's opinion, based on other indications in absence of rising reactor pressure, that ECCS has failed.
Logic D,iagtgm () See Separate Sheets 4 J l l f , ,-s GE 2 Page 4 of 8 k-o l l l
GENERAL EMERGENCY No.2 L GlC DIAGRAM iO . PATH A (FIG.GE 2-2) P ATH B (FIG.GE 2-3) P ATH C (FIG. GE2-4) PRIMARY CLAD COOLANT CLA D FAILURE BOUNDARY FAILURE FAILtRE I PRIMARY
' ~
PRIMARY PRIMARY COOLANT , BOUNOAR Y CONTAINMENT CONTAINMENT FALLURE FAILURE FAILURE
~
lO POTENTIAL POTENTIAL TENTIE PRIMARY COOLANT PRIMARY O ' CONTAINMENT BOUNDARY FAILURE MME FAILURE
+
t GEN ER AL EMERGENCY No.2 l
.O FIG. GE2-1 GE 2 PAGE 5 0F 8
GENERAL EMERGENCY NO. 2 O LOGIC DI AGR AM (CONT'D) i PATH A l ! EAL FOR EAL FOR SITE AREAEMERGENCY ALERT NO.2 EXCEEDED NO.1 EXCEEDED MYWE COMM ' HIGH DRYWELL COMMON MW PRESSURE AREA HICH INDICATION Or HIGH T MPERA M E RADIATION 'N s ALARMS ALARM ,THE DRYWELL RE ACTOR WATER LEVEL AL ARMS WITH IN DICATION ON LEVEL INSTRU MENT OF WATER LEVEL LESS THAN
-132.5 INCHES AND R APIDLY DECREASING PROCESS COMPUTER PROCESS COMPUTER W ATCH ENGINEER'S INDICATES INDICATES JUCGEMENTOF POTENTIAL PRIMARY CONTAIN MENT DRYWELLTEMPERATURE LOSS OF PRIM ARY PRESSURE IN IN EXCESS OF 296*F CONTAIN M ENT BASED EXCESS OF 43.2 psig AND STILL RISING hfNDPEDE INFORMATION GENERAL EMERGENCY
#2 FI G. G E2 -2 GE 2 PAGE 6 OF 8
4 'l GENERAL EMERGENCY NO. 2 LOGIC DIAGRAM (CONT D) l PATH B DRYWELL . RMS DIVISION 1 HIGH DRYWELL CONTAINMENT Cole 40N ALARM WITH PRESSURE AREA HIGH INDICATION OF HIGH ALARMS TEMPERATURE RADIATION IN ALARM THE DRYWELL REACTOR WATER LEVEL ALARMS WITH INDICATION ON LEVEL INSTRUMENT OF WATER LEVEL LESS THAN
-132.5 INCHES AND RAPIDLY DECREASING '
O WATCH ENGINEER WATCH ENGINEER'S OBSERVES THAT ALL OPINION OF PRIMARY CONTAINMENT PENETRATIONS CONTAINMENT FAILURE REQUIRED FOR ISOLATION BASED ON OBSERVATION ARE NOT VALVED OFF OR INDEPENDENT OR CLOSED INFORMATION I LEVEL INDICATOR WATCH ENGINEER'S SHOWS REACTOR VESSEL WATER LEVEL BELOW OTHE INDI A IONS N TOP OF ACTIVE FUEL ABSENCE OF FALLING AND STILL FALLING WATER LEVEL THAT ECCS HAS FAILED NO. 2 FIG GE2-3 GE 2 PAGE 7 OF 8
GENERAL EMERGENCY NO. 2 0- LOGIC DI AG RAM (CONT'D) PATH C EAL'S FOR EAL'S FOR SITE AREA EMERGD.'CY ALERT NO. 2 EXCEEDED NO.1 EXCEEDED WATCH ENGINEER WATCH ENGINEER'S OBSERVES TH AT ALL OPINION OF PetlMARY CONTAINMENT PENETRATIONS CONTAINMENT FAILURE REQUIRED FORISOLATION BASED ON OBSERVATION ARE NOTVALVED 0FF OR INDEPEPENDENT i O OR CLOSED INFORMATION WATCH ENGINEER'S REACTOR OPINION BASED ON PRESSUREIN EXCESS OTHERINDICATIONS IN OF 1250 PSIG AND ABSENCE OF RISING IN CRE ASIN G REACTOR PRESSURE THAT ECCS HAS FAILED GENERAL EMERGENCY
#2 O rio.ee 2 -4 GE 2 PAGE 8 OF 8
GENERAL EMERGE!!CY NO. 6b Initiating Conditions ! LOCA'S~ with failure of ECCS to perform leading to core degradation in minutes to hours. Loss of containment integrity may be imminent. Emergency Action Levels
- 1. Loss of primary coolant boundary indicated by:
- a. High Drywell Pressure Alarms on panel 1H11*PNL-601. (ARP-1128 DRYWELL SYS A PRESS HI and ARP-1129 DRYWELL SYS B PRESS HI) indicating drywell pressure in excess of 1.69 psig as indicated by pressure instruments i
1H11*PIS-165 B, D and 1H22*PIS-165 A, C on panel 1H11*PNL-601. l OR
- b. Drywell Containment Area Temp Hi Alarm on panel 1H11*PNL-VC2 (ARP-3096 DRYWELL AREA TEMP HI) when any of the following conditions are satisfied:
> 1. Supply air from the Drywell Unit Coolers into the drywell exceeds i 1108F (as indicated by IT47-TRS010 recorder on panel 1H11*PNL-VC2). . ! E
- 2. Zone 2 (annular space between reactor pressure vessel and
'. biological shield) or Zone 3 (CRD area) temperature exceeds 1708F (as indicated by IT47-TRS010 recorder on panel 1H11*PNL-VC2) l E
- 3. Return air from: primary containment; Zone 2; drywell floor; or vicinity of return duct inlets exceeds 155er (as indicated by IT47-TRS020 recorder on panel 1H11*PNL-VC2) l l OR 4.
Return air from: Drywell Head Area; RPV head flange area; or l vicinity of refueling bellows and bulkhead assembly exceeds 1908F
- (as indicated by IT47-TRS030 recorder on panel 1H11*PNL-VC2).
l 3 I
- c. Radiation Monitoring System Div. A Conon Alarm on panel 1H11-PNL-601 (ARP-1401 DIV I RAD HONITOR RAD HI) with indication on Recorder i
1D11*RR111 on panel 1D11*PNL-080 of high gaseous or high particulate radiation in the drywell. AND
- d. Reactor Water Level Alarms on panel 1H11*PNL-601 (APP-1144 RX SYS A LEVEL LO and ARP-1145 RX SYS B LEVEL LO) with indication on level GE 6b Page 1 of 3
instrument IB21-LIS-029 on panel 1H11*PNL-601 of reactor water level below -132.5 inches and rapidly decreasing. AND
- 2. Clad failure indicated by:
- a. Radiation Honitoring System Div. A common alarm on panel 1H11-PNL-601 (ARP-1401 DIV I RAD HONITOR RAD HI) with indication on Recorder 1D11*RR111 on panel 1D11*PNL-080 of high gaseous or high particulate radiation in the Drywell.
gR
- b. Analysis of a reactor coolant sample indicates activity in excess of 300 uCi/gm dose equivalent I-131.
AND
- 3. ECCS malfunctien, indicated by:
- a. Recctor water level instrument 1B21-LI-007 on panel 1H11*PNL-601 indicates water level cannot be maintained above top of active fuel zone (-158 inches) and still falling.
PND
- b. Core Spray System A Inoperative Alarm on panel 1H11*PNL-601 (ARP-1115 CS SYS A INOF)
AND
- c. Core Spray System B Inoperative Alarm on panel 1H11*PNL-601 (ARP-1113 CS SYS B INOP)
AND
- d. Residual Heat Removal System A Inoperative Alarm on panel 1H11*PNL-601 (ARP-1147 RHR SYS A INOP).
AND r e. Residual Heat Removal System B Inoperative Alarm on panel 1H11*PNL-601 ! (ARP-1155 RHR SYS B INOP). 1 Logic Diagram See Separate Sheet O GE 6b Page 2 of 3
GENERAL EMERGENCY N O. 6 B LOGIC DIAGRAM O MIGH DRYWELL DRYWELL CONTAINM!DT OM RMS DIV. A A AR P R ESSURE AREAHIGH hi ON ALARMS TE M PER ATURE OF HIGH RADIATION ALARM IN THE DRYWELL ' REACTOR WATER LEVEL ! ALARMS AND 1 ! INDICATION ON LEVEL ! INSTRUMENT OF WATER LEVEL LESS )
, TH AN-!J2.51N CHES
, AND RAPIDLY DCCREASING i l
l l REACTOR COOLANT 'RWS D:V. A COMMON ACTIVITY G REATER ALARM WITH INDLCATION
- TH AN SCO u Cf/gm OF HIGH RADIATION I DOSE EQUIVALENTI-ISI L
IN THE DRYWELL i .O REACTOR WATER ! LEVEL INDICATOR I SHOWS WATER LEVEL BELOW TOP OF l ACTIVE FUEL ZONE (158 INCHES) RESIDUAL HEAT REMOVAL SYSTEM A INOPERATIVE ALARM RESIDUAL HEAT REMOVAL SYSTEM B INOPERATIVE ALARM , CORE SPRAY SYSTEM A INOPERATIVE ALARM l CORE SPRAY SYSTEMB , INOPERATIVE ALARM FIG, GE68-1 EMERGE Y NO.6b
EVENT CATEGORY 3 STEAM LINE BREAK OR MS RV/SV FAILURE INITIATING CONDITIONS , UNUSUAL EVENT No. 6 Failure of a safety or relief valve in a safety related system to close following reduction of applicable pressure. ALERT NO. 4 Steam line break outside containment with MSIV malfunction causing leakage. SITE AREA EMERGENCY NO. 4 Steam line break outside containment without isolation. t e 5 O
~ < 1
UNUSUAL EVENT NO. 6
- ( )- -Initiating conditions Failure of a safety or relief valve in a safety related system to close following reduction of applicable pressure.
t Emergency Action Levels Safety or relief valve in Reactor Coolant System lifts on high pressure and remains open after pressure returns to normal and is indicated by an,y of the following: -
- 1. SRV leaking alarm on panel 1H11*PNL-602 (ARP-1337 SRV LEAKING) indicating SRV discharge pressure in excess of 25 psig (indicated by pressure indicating switch IE21*P15-153 A through L on 1B21-PMU-501 on panel 1H11*PNL-601) or discharga temperature in excess of 2GO'F, (indicated by temperature re: order IB21-TR-100 on panel 1H11*PNL-614).
f .
- 2. Any relief valve open (red) indicator on panel 1H11*PNL-602.
EE
- 3. Suppression Pool Water Temperature and/or Level increasing as indicated on IE41-LI-013 on panel 1H11*PNL-601.
O a
- 4. High Pressure on tailpipe pressure indicator IB21-PLS-153.
SE
- 5. Temporary increase in Reactor Water Level and/or Steam to feed flow mismatch.
EE
- 6. Decreasing turbine generator load.
Logic Diagram See Separate Sheet UE 6 Page 1 of 2
UNUSUAL EVENT No.6 LOGIC DI AGRAM 1 SRV LE AKIN G I ALARM l ' l l l i HIGH PRESSURE i T AILPIPE PRESSURE INDICATOR l l SUPRESSION POOL ANY RELIEF TEMPORARY INCREASE DECRE ASING TEMPERATURE VALVE RED IN R EACTOR WATER URBINE GENERATOR AND/OR LIGHT OPEN LEVEL AND/OR STEAM TOFEED FLOW MISMATCH LOAD LEVEL INCREASING IN DICATION l l UNUSUAL EVENT No.6 O FIG.UE6-l UE 6 PAGE 2 CF 2
~
ALERT NO. 4 k) Initiating Conditions i Steam line break outside containment with' MSIV malfunction causing leakage
. Emergency Action Levels A- steam line break with Main Steam Isolation Valve (MSIV) malfunction causing leakage will be defined by the following indications:
- 1. Automatic MSIV closure indicated by any of the following:
- a. Main Steam Line (MSL) High Flow Alarms on panel 1H11*PNL-603 (ARP 1314 for MSL A/B FLOW HIGN and AAP 1315 for MSL C/D FLOW HIGH)
! indicating high flow when pressure differential is in 'xcess of e l 106 psi. i . CR < l- b. MSL Tunnel Ambient Higi Terperature Alarms on panel 1H11*PNL-601 (ARP-1324 STM LINE DIV I DIFF TEMP HI-CH Al and ARP-1374 STM LINE DIV II DIFF TEMP HI-CH A2) indicating temperature in excess of i 1658 F. c
. _OR t c. MSL Tunnel Ventilation High Differential Temperature alarms on
~
panel 1H11*PNL-601 (ARP-132S STM LINE DIV II DIFF TEMP HI-CH 32
. and ARP-1373 STM LINE DIV I DIFF TEMP HI-CH B1) indicating differential temperature in excess of 50s y, EE -
- d. Main Steam Line Low Pressure Alarms on panel 1H11*PNL-603 (ARP-1318 for MSL A/B PRESS LO and ARP-1319 for MSL C/D PRESS LO) indicating pressure decreasing below 825 psig.
AND
- 2. Watch Engineer judgement that the MSIV's have malfunctioned based on evidence of continuing main steam leakage, e.g. :
- a. Main steam line flow
- b. Elevated space temperatures
- c. Increased area radiation levels
- d. Valve Position Switches Logic Diagram See Separate Sheet A 4 Page 1 of 2 l
- - . . v. y -%<- y .-----.v---,-,y- . , ,y- . - - . y.-,,........c, , , , . . - . , . . . . . . _ . . . . . . , - _ . . _ , _ __ _ _ . _, ___ _ _ _ - - - - . _
O , ALERT No. 4 LOGIC DI AGRAM l '- MAIN STEAM LINE M AIN STEAM LINE M AIN STEAM LOW PRESSUR"e MAIN STEAM LilJE TUNNEL AMBIENT LINE TUNNEL AtARMS HIGH FLOW ALARMS hlOH TEMPERTURE HIGH DIFFERENTIAL l ALArtME TEMPERATURE ALARMS ( i _- t I - l WAT,CH ENGINEER JUDGEMENT THAT MSiv S HAVE MALFUNCTIONED BASED ON EVIDENCE OF CONTINUING MAIN l STEAM LEAKAGE , E.G. : e MAIN STEAM LINE FLOW
- ELEVATED SPACE TEMPERATURES e INCREASED AREA RADIATION LEVELS s DIRECT OBSERVATION i
ALERT No.4 O FIG. A4-1 A4 PAGE 2 OF 2
SITE AREA EMERGENCY NO. 4 Initiating Conditions Steam line break outside containment without isolation. Emergency Action Levels
. 1.a. Main Steam Line (MSL) break outside containment indicated by any of the following: ,
- 1) MSL High Steam Flow Alarms on panel 1H11*P!;L-603 (ARP 1314 for MSL A/B FLOW HIGH and ARP-1315 for MSL C/D FLOW HIGH) indicating high flow when pressure differential is in excess of 106 psi.
3
- 2) MSL Tunrel Anbient High Temperature Alarras en panel 1H11*PNL-601 (ARP-1324 S*lM LINE DIV I DIFF !EMP HI-CH Al and ARP-1314 STM LINE DIV II DIFF TENP HI-CH A2) indicning temperature ir. excess of IE5s y, g
- 3) MSL Tunt.el Ver.tilation High Lifferencial Tempercture Als s en psne]
,1H11*PNL-601 (ARP-13I.5 S!M LINE DIV II DIFF TEMP HI-CH B2 and ARP-1 ",'T 3 STM LINE DIV I DIFF TEMP HI-CH E1) indicating differer.tial temperature in excess of 508F.
O e
- 4) Main Steam Line Low Pressure Alarms on panel 1H11*PNL-603 (ARP-1318 for MSL A/B PRESS LO and ARP-1319 for MSL C/D PRESS LO) indicating pressure decreasing below 825 psig.
AND {
- b. No isolation of the break as indicated by:
Failure of two MSL isolation valves (A0V-081A-D and A0V-082A-D) in
, series to close as indicated by the position in'dication lights on panels 1H11*PNL-601 and 602.
l 3 l 2.a. High Pressure Coolant Injection (HPCI) steam supply break outside containment indicated by any of the following i
- 1) HPCI High Steam Flow Alarm on panel 1H11*PNL-601 (ARP-1015 for HPCI
.STM LINE DIFF PRESS A HI or ARP-1016 for HPCI STM LINE DIFF PRESS B l HI) initiated by pressure differential switches IE41*PDS 022A/B at l pressure differentials in excess of 212 inches of water when flow is high.
i SAE 4 Page 1 of 4 l u
EE
-( ) 2) HPCI Equipment Area High Temperature Alarm on panel 1H11*PNL-601 (ARP-1322 STM LEAK DET DIV I AMBIENT TEMP HI or ARP-1323 STM LEAK DET DIV II AMBIENT TEMP HI) indicating temperature in excess of 1340F.
Associated temperature indications ~can be obtained from Steam Leak i Det Panel A or B Temperature Monitors on panel 1H11*PNL-614. i , O_R
- 3) HPCI Steam Supply Low Pressure Alarm on panel 1H11*PNL-601 (ARP-1041 HPCI ISOL SIGNAL INIT-Logic A, setpoint 130 psig; and ARP-1042 HPCI i . ISOL SIGNAL INIT-Logic B, setpoint 128 psig). Indication of steam l pressure is shown on pressure indicator 1E41-PI-004 on panel
! 1H11"PNL-601. l i ANE ( .
- b. No isolatisn of the break as indicated by:
i Failure cf any two HPCI isolation valves (E41F002 MOV-41 ar.d E41F003 MOV-
- 42) and bypass valves (E41F080 MOV-48 and E41F097 MOV-47) in series to close as indicated by the position indication lights on panel 1H11*PNL-601. :
f CR i i 3.a. Reactor Core Isolation Cooling (RCIC) steam supply break outside
-( ) containment indicated by any of the following:
- 1) RCIC High Steam Flow Alarm on panel 1H11*PNL-602 (ARP-1063 for RCIC STM LINE DIFF PRES A HI and ARP-1076 for RCIC STM LINE DIFF PRES B
, Ill) at pressure differentials in excess of 105 inches as indicated on
'1E51-PDS-022A/B of water when flow is high.
EE
- 2) RCIC Equipment Area High Temperature Alarm on panel 1H11*PNL-601 (ARP-1322 STM LEAK DET DIV I AMBIENT TEMP HI or ARP-1323' STEAM LEAK DET DIV II AMBIENT TEMP HIGH) indicating temperature in excess of 134s F. Associated temperature indication can be obtained from Steam Leak Detection Panel A or B Temperature Monitors on panel 1H11*PNL-614.
] EE , 3) RCIC Steam Supply Low Pressure Alarm on panel 1H11*PNL-602 (ARP-1067 i RCIC ISOL SIGNAL INIT-Logic A, setpoint 82 psig; and ARP-1078 RCIC ISOL SIGNAL INIT-Logic B, setpoint 76 psig). Indication of steam pressure is shown on pressure indicator 1E51-PI-004 on panel 1H11*PNL-602. i AND SAE 4 Page 2 of 4 l()
. ..- . . . _ . - --. . . .- - -=
2 . 1 ) b. No isolation of the break as indicated by: Failure of any two RCIC isolation valves (E51-F007 MOV-041 and E51-F008 i MOV-042) and bypass valves (E51-F085 MOV-047 and E51-F075 MOV-048) in series to close as indicated by the position indication lights on panel i 1H11*PNL-602. i Logic Diagram i See Separate Sheet i a I f i : i , O l j b a 4 h + ' SAE 4 Page 3 of 4 I
, w -o- -.- _ - - ,,--- ----.n,-.,_n , - _ , - , , _ . ,,~.,-,.-m , - - ,, u..-.-.-.-__,y_, , - . . - - , , . , , - , - . , - - . - - , , - - . - - - - , - , . - . - - -
O O O SITE AREA EMERGENCY No. 4 i . LOGIC DI AGRAM _ M AIN MSL HIGH MS L TUNNEL MSL TUNNEL RC!C HIGet RCIC EOPT RCIC STEAM STEAM LINE STEAM AMBIENT VENT ' LOW PRESSURE FLOW STEAM FLOW AREA HIGH SUPPLY LOW HIGH TEMP HIGH AT ALARM TEMP ALARM PRESSURE ALARM ALARMS ALARMS ALARMS ALARM i HPCI EQPT. HPCI STEAM HPCI HIGH SUPFLY LOW STEAM AREA H:GH PRESSURE FLOW ALARM TEMP ALARM ALARt.1 AUTOMATIC AUTO ESCLATION AUTO ISOLATION OF HPCl STEAM OF RCIC STEAM MSIV CLOSURE SUPPLY VALVES- SUPPLY VALVES LACK OF CLOSED VALVE l'JDiCATION FROM TWO VALVES i IN SERIES SITE AREA FIG. SAE 4-1 EMERGENCY Nc.4
, SAE 4 PAGE 4 OF 4 I
t EVENT CATEGORY 4 OTHER LCO'S INITIATING CONDITIONS UNUSUAL EVENT NO. 1 Emergency Core Cooling System (ECCS) initiated and discharge to vessel. UNUSUAL EVENT NO. 8 Loss of containment integrity requiring shutdown by technical specifications. UNUSUAL EVENT NO. 9 Loss of engineered safety feature or fire protection system function requiring shutdown by technical specifications (e.g., because of
- malfunction, pers,onnel error or procedural ,
inadequicy). t t t O i l l l I l
UNUSUAL EVENT NO. 1 j'51 Initiating Conditions Emergency Core Cooling System (ECCS) initiated and discharge to vessel Emergency Action Levels
'l. Hi'gh Pressure Coolant Injection flow to Reactor Vessel
- a. HPCI system initiation indicated by white indicating light for HPCI initiation signal sealed in on panel 1H11*PNL-601 AND
- b. HPCI flow indication IE41*FI-003 on panel 1H11*PUL-601 exceeds 300
- UP3. ;
AND 4
- c. HFCI _gump discharge valves iE41*MOV-031 and IE41*MOV-035 to reactor i
vescel are open, indicated by red lig.it on par.el 1H11*PNL-601 U].E
- d. The on shift Watch regineer determines that an Unusual Ever.t is occurring.
O 2R
- 2. Core Spray flow to Reactor Vessel
( a. ADS Logic A&C or B&D Initiation Alarms on panel 1H11*PNL-602 (ARPs-1352, 1353, 1354 and 1355 for ADS INIT LOGICS A,C,B and D respectively) indicatil.g ADS activation. AND
- b. CSS flow indication on IE21*FIS-002 A(B) on panel 1H11*PNL-601 exceeds 420 gpm AND
- c. CSS Injection valves 1E21*HOV-033 A(B) to Reactor Vessel are open, indicated by red light on panel 1H11*PNL-601 E
- 3. Low Preasure Coclant Injection flow to Reactor Vessel
() UE 1 Page 1 of 3
- a. . ADS Logic A&C or B&D Initiation Alarms on panel 1H11*PNL-602 (ARPs- '
, 1352, 1353, 1354 and 1355 for ADS INIT LOGIC A, C, B and D s-) respectively) indicating ADS activation.
AND
- b. LPCI flow indication on 1E11*FI-001 A(B) on panel 1H11*PNL-601 exceeds 400 gpm AND
- c. LPCI isolation valves 1E11*HOV-037 A(B) and 1E11*HOV-037 A(B) are open, indicated by red light on panel 1H11*PNL-601
$ Logic Diagram See Separate Sheet
?
F r , f a i l . l i l l l i
~
UE 1 Page 2 of 3
UNUSUAL EVENT No.1 O LOGIC DIAGRAM "PC ADS ACTUATED
- SYS INITIATED
~
5 ! { h 'r'C l F LOW l 9P* CSS FLOW LPCI FLOW
-- - > 420 gom > 400 gpm H P Cl PUMP O' D!SCHARGE VALV ES O PEN CSS INJECTION LPCI ISOLATION VALVES OPEN VALVES OPEN ON SHIFT l WATCH
! ENGINEER l OPINION UNUSUAL EVENT No,I f O rio. usi-i UEI P AGE 3 OF 3
UUUSUAL EVENT 8
- (~)
\s / Initiating Conditions Loss of containment integrity requiring shutdown by technical specifications Emergency Action Levels
- 1. Failing to meet any one of the following Limiting Conditions for Operation (LCO) for primary containment and requiring shutdown:
- a. TS 3.6.1.3 Primary Containment Air Locks Leakage rate of less than or equal to 0.20 La at Pa, 46.4 psig with at least one air lock docr closed.
. O_R.
- b. TS 3.6.1.4 Mair. Staas Isolation Valva.s_(MSI:/) . eakage Control Syste,m i Both MSIV subsystees opereble 2E
- c. OS 3.5.1.6 Primary Containraent Irjterp_a1 Tressura Drywell and suppression chamber internal pressura shall be
, - between - 1.0 and +1.6 psig SE
- d. TS 3.6.1.7 Drywell Average Air Temperature Drywell average air temperature shall be less than or equal to 1458F SE
- e. TS 3.6.1.8 Primary Containment Purge System The 6" purge supply and exhaust isol. valves shall be operable and each 18" purge valve sealed closed. Each 6" purge valve may be open for inerting, de-inverting, pressure control and vacuum breaker test only.
EE
- f. TS 3.6.2.1 Suppression Chamber Depressurization System
- 1. Pool Water volume equivalent to a level between -6" gauge and +6" gauge or;
,-~
( ,/. UE 8 Page 1 of 4
- 2. Pool Water average temperature maximum: 908F during r"'g operation or startup; 1058F during testing; 1108F with (m,/ thermal power less than or equal to 1% of rated thermal
; power; 1208F with the main steam isolation valves closed. following a scram or;
- 3. Drywell-to-suppression chamber bypass leakage less than or equal to 3% of the acceptable A4rx design value of 0.16 ft2 or;
- 4. Drywell floor perimeter nitrogen-pressurized seal pressurized to greater than or equal to 53 psig.
9E
- g. TS 3.6.2.2 Suppression Chamber S.cray Both independent RHR system spray icops are optrable. Each i loop consists Of one opurable EHR pump ar.d an cperable flow
( path capable of recirculating water from the s2ppression chamber spray spargers. i SE
- h. TS 3.'s.2.3 Suppression Pool cooling ,
Both independent RHF. system suppression ycol cooling mode loops are operable. Each loop consists of one operable RHR s~ pump and an operable flow path capable of recirculating water from the suppression chamber through an RHR exchanger. ( 9E l i. TS 3.6.3 Primary Containment Isolation Valves i. l Primary containment isolation valves and reactor instrumentation line excess flow check valve shall be operable within the time allocated to isolate as shown in Table 3.6.3-1 of Technical Specification 3.6.3. EE
- j. TS 3.6.4 Suppression Chamber-Drywell Vacuum Breakers Each pair of suppression chamber-drywell vacuum breakers shall be operable and closed l ([]) uE a Page 2 of 4 l
l
- 2. Failing to meet any one of the following Limiting Conditions for Operation (LCO) for secondary containment and requiring shutdown:
- a. TS 3.6.5.1 Secondary Containment Integrity
- 1. Secondary containment pressure shall be maintained less than or equal to 0.5" of vacuum water gauge.
- 2. All equipment hatches closed and sealed
- 3. At least one door in each access to the secondary containment is closed except for routine entry and exit.
- b. TS 3.6.5.2 Reactor Building Automatic Isolation Valves Reactor Buildir.g ventilation systen automatic isolation volves 1T46*AGV35A, B and IT46*AOV37A, E shall be operable and isolate secondary containment within 10 te:cnds.
EE , c, TS 3.6.5.3 Rea-tor Building Standby vr.n-ilation g Systec ; Both RB5V subrystems th311 be operable Logic Diagram () See Separate Sheet i l 1 ~ UE 8 Page 3 of 4
UNUSUAL EVENT No. 8 LOGIC DIAGRAM PRIMARY CONTAINMENT Al R LOCKS (T.S. 3.6.l.3) MSIV LEAKAGE CONTROL SYSTEM (T.S. 3.6. l.4) PRIMARY CONTAINMENT INTERNAL PRESSURE (T S. 3.6.1.6 ) PRIMARY CONTAINMENT AIR TEMPERATURE (T.S. 3.6.l.7) _ r .. l PRIMARY PRIMARY CONTAINMENT __ . ,- CONTAINMENT PURGE JYSTEM f T.S.3.6.1.8 ) l
~ ~ ~
l TECH SPECS buPPRESSION CHAMBER l ! OEPPESSURIZATION S YSTEM (T.S. 3 6.2.1) I SUPPRESSION l CHAMBER SPRAY (T.S. 3.62.2) SUPPRESSION POOL COOLING (T.S. 3.6.2.3) UNUSUAL PRIMARY CONTAINMENT EVENT ISOL ATION VALVES No.8 (T.S. 3.6.3) t SUPPRESSION CHAMBER
-DRYWELL VACUUM BREAKERS (T S. 3.6.4 )
SECONDARY CONTAINMENT INTEGRITY (T.S. 3.6.5.1) SECONDARY j CONTAINMENT REACTOR BUILDING AUTOMATIC ISOLATION TECH VALVES (T.S. 3.6.5.2) SPECS REACTOR BUILDING STANDBY VENTILATION SYSTEM (T.S. 3.6.5.3) FIG. UE 8-l UE 8 PAGE 4 OF 4
UI! USUAL EVEliT No. 9 O Initiating Conditions Loss of engineered safety feature or fire protection system function requiring shutdown by technical specifications (e.g., because of malfunction, personr.el error or procedural inadequacy) Emergency Action Levels
- 1. Failing to meet any one of the following Limiting Conditions for operation (LCO) for engineered safety features and requiring shutdown:
- a. TS 3.1.3.8 Control rod drive housing support Control rod drive housing support is in place 3
- b. TS 3 1.5 Standby liquid centrol system Standby liquid centrol syntam is operable ,
?
c, TS 3.4.2, Safety /Eelief valtres ' yO) Eight or more of the following reactor coolant system safety / relief valves are operable with the valve lift settings at nominal operating temperatures and pressures: 4 Safety-relief valves @ 1115 psig i 1% 4 Safety-relief valves @ 1125 psig i 1% (See Note) 3 Safety-relief valves @ 1135 psig i 1% (See Note) Note: Up to two inoperable valves may be replaced with spare operable valves with lower setpoints until the next refueling cutage.. E
- d. TS 3.4.7, Main Steam Line Isolation Valves Two MSIV's for each main steam line isolation valve shall be operable within 3 to 5 sec. to close and isolate.
O Uz e eaee 1 of 5
- e. TS 3.5.1, Emergency Core Cooling Systems (ECCS)
() 1 Automatic Depressurization System (ADS) The ADS system shall have at least seven operable ADS valves unless the reactor steam dome pressure is less than or equal to 113 psig as indicated by pressure indicator 1C32-PI-003 on panel 1H11*PNL-603. I SE
- f. 'TS 3.5.1, Emergency Core Cooling Systems (ECCS)
High Pressure Cooling Injection System (HPCI) The HPCI system consisting of cne operable HPCI pump and an operable ficw path from the suppression pool or condensate tank to the reactor pressure vessel shall be op rable, urless the' reactor steam dome pressure is less than or equal to 150 psig as indicated by pressure indicator IC32-PI-003 on panel 1H11*PNL*S03. 3 SE 4 g. TS 3.5.1, Emy gery Cere Ct,oling Systens (ECCS) g3re spj37_Sg stem Mss) (*') Bath independent CSS system loops shall be operable. Each loop consists of one operable CSS pump and an operable flow path from suppression pool to the spray sparger of the reactor pressure vessel. 93 -
- h. TS 3.5.1, Emergency Core Cooling Systems (ECCS)
Low Pressure Coolant Injection System (LPCI) i Both independent LPCI system loops shall be operable. Each loop consists of two operable RHR pumps and an operable flow path from the suppression pool to the reactor vessel. , SE
- i. TS 3.5.3, Suppression Pool Water Level (for ECCS use)
Suppression pool water volume shall be equivalent to a level of at least -6" gauge during operation or startup. 1 OR UE 9 Page 2 of 5 1
i
- j. TS'3.6.6.1, Drywell and Suppression Chamber Hydrogen Recombiner Systems Both independent drywell and suppression chamber hydrogen recombiner system shall be operable.
EE
- k. T.S.:3.7.1.1 Plant Service Water System Both independent plant service water system loops shall be operable. Each loop consists of two operable plant service water pumps and an operable flow path from the screenwell intake to a Reactor Building Service Water heat exchanger and associated safety related equipment of the RBSW system and to a Residual Heat Removal (RHR)
- heat exchanger.
EE
- 1. T.S. 3.7.1.4, Ultin. ate Heat $ ink The intake rtructure shall be flooded. Sediment depth shall not exceed an average of 1 ft above the intake '
structure minimum bottom depth of at least -11.0 ft Me9n Low Watar. O m. TS 3.7.2, Control Room Air Conditioning System (Emergency Filter System) Both independent control room air conditioning emergency filter systems shall be operable. EE
- n. TS 3.7.4, Reactor Core Isolation Cooling System (RCIC)
The RCIC system consisting of one operable RCIC pump j and an operable flow path from the suppression pool or condensate tank to the reactor pressure vessel shall be operable. SE
- o. TS 3.8.1.1 AC Sources A minimum of the following AC electrical power sources shall be operable: .
- 1. Two physically independent circuits between the 4
offsite transmission network and the onsite class IE distribution system, and;. UE 9 Page 3 of 5 I e-- e
- 2. Three separate and independent diesel generators g-~g each with a separate day fuel tank of 275 gallons
's,,/ of fuel minimum and with a fuel storage system containing at least 40,600 gallons of fuel and at ,
least one fuel transfer pump. SE
- p. T.S.-3.8.3.1 Onsite Power Distribution Systems'(D.C. Power Distribution)
The following three independent 125V DC distribution buses shall be energized: 125V D.C. distribution BUS A - for Div i 125V D.C. distribution' BUS B - for Div 2 125V D.C. distribution BUS C - for Div 3 SE c Logic Diagram See separate sheet . l l O 1 I f UE 9 Page 4 of 5 O.
UNUSUAL EVENT No.9 . . , LOGIC DIAGRAM i CCNTROL ROD DRIVE HOUSING SUPPORT (I S. ' 3.1. 3. 8) l l STAND 8Y LIQUID CONTROL , ! SYSTEM (T.S. 3.13) l SAFETY / RELIEF VALVES (T S. 3. 4. 2 ) -
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MAIN STEAM ISOLATION VALVES (T.S. 3.4.7) , AUTOMATIC DEPRESSURIZATION #' ' l SYSTEM (T.S. 3.5.1) , HIGH PRESSURE COOLING INJECTION SYSTEM
~
(T. S. 3. 5.1 ) ,
- l. ,.
" J CORE SPRAY SYSTEM (T. S. 3.5. I)
LOW PRESSURE COOLANT , FAILING TO MEET INJECTION SYSTEM ANY OF THESE (T.S. 3.5.1 ) UNUSUAL LCO'S FOR ' '
' EVENT SA RES SUPPRESSION POOL WATER N o. 9
' LEVEL (T.S. 3.5.3)
DRYWELL AND SUPPRESSION , - CHAMBER HYDROGEN RECOMBINER SYSTEM - (T.S. 3.6.6.1) _ PL ANT SERVICE ,-- WATER SYSTEM (T.S. 3.7.1.1)
~ ~
ULTIMATE HEAT SINK <
' ' (T.S. 3.7. l .3 ) ,
CONTROL ROOM AIR .E . CONDITIONING SYSTEM 4 (T.S. 3.7. 2) REACTOR CORE ISOLATION
, COOLING SYSTEM (T.S.2k7 4)
A.C. SOURCES ( T.S 3.8.1.1) , ,
. FIG. UE9-l 1
UE 9 P AGE 5 OF 5 - ONSITE POWER DISTRIBUTION SYSTEM (TS. 3.8.3.1)
._. ._ _ _.._ .'__. d
s. . ?!i EVENT CATEGORY 5 ABHORMAL RADIOLOGICAL EFFLUENT [ OR RADIATION LEVELS i INITIATING CONDITIONS UNUSUAL EVENT NO. 2 Radiological effluent technical specification limits exceeded. ALERT NO. 6 Radiation levels or airborne contamination which indicate a_ severe degradation in the i control ,of radioactive materials (e .g. , increase of factor of 1000 in direct radiation l readings within facility). ALERT NO. 15 Radiological effluents greater than 10 times technical specification instantaneous limits (an instantaneous rate which, if continued over 2 hours, would result in about 1 mr at i the site boundary under average meterological conditions). SITE AREA EMERGENCY No. 13 1. Effluent monitors detect levels corres-ponding to greater than 50 mr/hr for 1/2 hour or greater than 500 mr/hr W.B. for O two minutes (or five times these levels to the thyroid) at the site boundary for adverse meteorology (no meteorological data required for determination).
- 2. These dose rates are projected based on other plant parameters (e.g., radiation level in containment with leak rate appropriate for existing containment
- pressure) or are measured in the environs.
3 .' EPA Protective Action Guidelines are projected to be exceeded outside the site boundary. GENERAL EMERGENCY NO. 1 1. Effluent monitors detect levels corres-ponding to i rem /hr wholebody or 5 rem /hr thyroid at the site boundary under actual j . meteorological conditions (meteorological ' data required for determination).
- 2. These dose rates are projected based on other plant parameters (e.g., radiation levels - in containment with leak rate appropriate for existing containment 0 1 1
4
. . ~ . . - , - , , , , , , . - - - - . , . , , - - . .
__ _ = - - _ . .- - pressure with some confirmation from effluent monitors) or are measured in the environs. O O
F UllUSUAL EVENT NO. 2 Initiating Conditions Radiologi'al c effluent technical specification limits exceeded Emergency Action Levels The following table lists the radiological effluent monitors in the gaseous and liquid streams and their corresponding technical specification alarm setpoints Logic Diagram Not applicable . l p I I i O l
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f f i 4 i I UE 2 Page 1 of 2 1
. O 4 +. .,~,-.--,,,,w,-,,, ,,,,.,_,,.,,.,.~,...n,,,.,--,,,.,-- ,._._,,,,.,,,__.,.nn,, .,n.,---,-._,,,,,,-.,, - - , - , . . . . - , _ _ - . - ,. , , , . . . . , .
O O O SHOREHAM NUCLEAR POWER STATION - RADIOLOGICAL EFFLUENTS MONITORS AND TECHNICAL-SPECIFICATION SETPOINTS DESCR I PT ION MONITOR LOCAL PANEL C.R. PANEL SET POI NT ALARM Station Ventilation Exhaust RE-041 ( Pa rt ic) PNL-041 PNL-059 RE-042 (Gas) , RE-043 (lodine) ! RBSYS Exhaust *RE-021A/B *PNL-021/022 *PNL-080
. *RE-022A/B Reactor Building Vent Exhaust RE-029 (Cas) PNL-029 RE-030 ( Pa rtic)
Turbine Building Vent Exhaust RE-057 (Cas) PNL-057 RE-058 (Partic) Radwaste Bldg Vent Exhaust RE-055 (Gas) PNL-055 - RE-056 ( Partic) Orr-gas (SJAE) Discha rge RE-012A/B (log) 1H11-PNL-604 4 1H11*PNL-6041 RE-014 ( Linear) IH11-PNL-604 l Orr-gas HEPA ri tter discha rge RE-065A/B 1H11-PNL-604 1H11*PNL-6011 Radwasto Bldg Tank Vent F ilter Dis. RE-077 (Cas) PNL-077 Sa l t-wa te r Dra in Tank Pump Di sch. RE-079 PNL-079 Radwaste System Liquid Erripent RE-013 PNL-013 PNL-059 PNL-058 i Annunciates only Radiation level indication and alarm functions are provided through the central processing units in all cases l l i I l t UE 2 Page 2 or 2
ALERT !!O. 6 O Initiating Conditions Radiation levels or airborne contamination which indicate a severe degradation in the control of radioactive materials (e.g., increase of factor of 1000 in direct radiation readings within facility). Emergency Action Levels Radiation alarm with indication on recorder or printer of 1000-fold increase in the radiation level under normal conditions for the following monitors: CATEGORY
. Containment drywell Monitors *RE-061 (particulate)
*RE-062 (gaseous)
Panels *PNL-061 (local)
*PNL-080 (in control room)
. Refueling level exhaust Monitors *RE-017A/B Panels 1H11*PNL-635 (radiation level indicator) 1H11*PNL-636 (radiation level indicator) 1H11*PNL-601 (annuncator)
CATEGORY II
. Radiation building ventilation exhaust Monitors RE-055 and RE-056 Panel PNL-055 (local) (also through CPU)
I
. Turbine building ventilation exhaust Monitors RE-057 and RE-058 Panel PNL-057 (local) (also through CPU)
. Reactor building ventilation exhaust Monitors RE-029 and RE-030 Panel PNL-029 (local) (also through CPU)
A 6 Page 1 of 2
4
- Area radiation monitors Monitors 1D21-RE-001 through ID21-RE-030 (GE) l Panel 1D21-PNL-600 (and through CPU)
Monitors ID21-RE-032 through 1D21-RE-042 (NMC) (through CPU only) i Annunciators on' panel 1H11*PNL-601 get activated on any Category II area or i process radiation monitoring in high radiation, alert or instrument failure. Logic Diagram Not applicable l 1 l l O l \ t P 1 i l 9 4
^" '
O
ALERT No.15 O Initiating Conditions Radiological effluents greater than 10 times technical specification instantaneous limits (an instantaneous rate which, if continued over 2 hours, would result in about 1 mr at the site boundary under average meteorological conditions). Emergency Action Levels The following table lists the radiological effluent monitors in the gaseous and liquid streams with their corresponding technical specification alarm setpoints. Ten times,these limits correspond to the required EAL's. Logic Diagram Not applicable O A 15 Page 1 of 2 O
r)J w] SHOREHAM NUCLEAR POWER STATION - RADIOLOGICAL EFFLUENT MONITORS AND TECHNICAL-SPECIFICATION SETPOINTS DESCR1 PT ION MONITOR LOCAL PANEL C.R. PANEL SETPOINT Al. ARM Station Ventilation Exhaust RE-041 ( Pa rt ic) PNL-041 PNL-059 RE-042 (Gas) RE-043 (todine) RBSYS Exhaust *RE-021A/B *PNL-021/022 *PNL-080
*RE-022A/B Reactor Building Vent Exhaust RE-029 (Cas) *PNL-029 RE-030 ( Pa rtic)
Turbine Building Vent Exhaust RE-057 (Gas) PNL-057 RE-058 (Pertic) Radwaste Bldg Vent Exhaust RE-055 (Cas) PNL-055 RE-056 ( Pa rt ic ) Orr-gas ( SJAE) Discha rge RE-012A/B (log) IH11-PNL-604 Ill11
- PNL-60 l t RE-014 (Linear) 1H11-PNL-604 Off-gas HEPA ri t ter discharge RE-065A/8 Ill11-PNL-604 1H11*PNL-6011 Radwaste Bldg Tank Vent Filter Dis. RE-077 (Gas) PNL-077 Sa l t-water Dra in Tank Pump Disch. RE-079 PNL-079 Radwaste System Liquid Errluent RE-013 PNL-013 PNL-059 PNL-058 i Annunciates only Radiation level indication and alarm functions are provided through the central processing units in all cases A 15 Page 2 of 2 l
I
SITE AREA EMERGENCY NO. 13 O Initiating Conditions
- 1. Effluent monitors detect levels corresponding to greater than 50 mr/hr for 1/2 hour or greater than 500 mr/hr W.B. for two minutes (or five times these levels to the thyroid) at the site boundary for adverse meteorology (no meteorological data required for_ determination).
- 2. These dose rates are projected based on other plant parameters (e.g.,
radiation levels in containment with leak rate appropriate for existing containment pressure) or are measured in the environs. 3.' EPA Protective Action Guidelines are projected to be exceeded outside the site boundary. Emergency Actions Levels
- 1. RBSVS intermediate range effluent monitor 1D11*RE-021B or 022B on panel 1H11*PNL-080 reads above the limits given in Figure 1 for the corresponding time after shutdown and RBSVS flow.
EE
- 2. a. Drywell high range monitor 1D11*RE-085A or B on panel ID11*PNL117A/B reads above the limits given in Figure 2 for the corresponding time after shutdown and drywell leak rate.
~~
EE
- b. Survey teams measure any of the following radiation levels at the site boundary: -
- 1. 500 mr/hr wholebody for a period of two minutes
- 2. 50 mr/hr wholebody for a period of thirty minutes
- 3. Procedure EPIP 2-24, Thyroid Dose Commitment Using TCS Air Sampler indicates one hour thyroid dose commitment values of:
- a. 2500 mrem for period of two minutes
- b. 250 mrem for a period of thirty minutes EE
- 3. a. Radiation Monitoring System (RMS) dose projection indicates that a wholebody dose of I rem or a thyroid dose of 5 rem is exceeded EE SAE 13 Page 1 of 5
({]) 1 i f
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- b. With RMS inoperative, procedure EPIP 2-2, Dose Projections, indicates that a wholebody dose of 1 rem or a thyroid dose of 5 rem is exceeded t
O for the projected duration of release. Logic Diagram See separate sheet J 4 4 4 i i. I J.
- O l
l SAE 13 Page 2 of 5 O
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O O O
- SITE AREA EMERGENCY No.13
- LOGIC DI AGRAM i
l SITE BOUNDARY RADIATION WITH RMS DRYWELL HIGH RANGE LEVELS EXCEED ANY OF RMS DOSE i RBSVS INTERMEDIATE I NOPERATIVE , RANGE MONITOR READS MONITOR READS ABOVE PROJECTION THE FOLLOWING: PROCEDURE i ABOVE THE CURVE SHOWN THE CURVE SHOWN IN EPIP 2-2 FIGURE 2 FOR THE
- 500 MR/HR WB FOR 2 MIN. IND ATES PA INDICATES EPA IN FIGURE I FOR THE e 50 MR/HR WB FOR 30 MIN. PAG'S EXCEEDED CORRESPONDING TIME CORRESPONDING TIME , ONE HR. THYROID DOSE
- AFTER SHUTDOWN AFTER SHUTDOWN AND COMMITMENT OF 2500 MREM AND RBSVS FLOW DRYWELL LEAK RATE FOR A PERIOD OF 2 MIN.
, e ONE HR. THYROID DOSE i COMMITMENT OF 250 MREM FOR A PERIOD OF 30 MIN l i I l SITE AREA l EMERGENCY No.13 l l l 4 FIG. S AE 13-1 SAE 13 P GE 3 OF 5
t,
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N wm e M M >g D eI % 2 >4 q%, f* o b,
) t % Q n *llf N k FIGURE 1 SHOREHAM NUCLEAR POWER STATION 7
RBSYS INTERMEDIATE RANGE MONITOR RESPONSE
- 10 --,, , . , .. .,, ,,
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. RBSYS MONITOR READING CORRESPONDING
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7 TO A DOSE RATE OF 500 MREM
/HRRWHOLEBODY
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q OR 2500 MREM /HR THYROID ATTHE SITE :r'-- +- - - -
$ 4 .
Ni ;h! ,$j,i.. _ _ .. ._ !' h.' , .' ' ..:.. T- rif'3' O..,$ i;
%[ d.
- , BOUNDARY FOR+yTWO MINUTES 4 Ig UNDER ADV
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'ii !4 . METEOROLOGICAL CONDITIONS -
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o '.U '. . 2 A...iu IIT S@A4hq i . U '
'it! i [ iTO OR 250 A DOSE MREM /HR RATE THYROID OF 50 ATTHE MREM SITE / HR WHOLEBODY Mie p.g" -
',' "+ 0 4 -
5 ' iN; ,' 'l' "-- - "" 2' 4 - 2 Q BOUNDARY FORTHIRTY MINUTES UNDER , 4[, ' t ADVERSE METEOROLOGICAL CONDITIONS
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rr[ q: rrr m .U a MULTIPLY CURVE VALUES BY: ._' iijl i-*4, tih he t ri F f 3a 4 :g. i e e rN Ut 4
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# FIG S AE 13-2 SAE 13 PAGE 4 OF 5
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c> p, FIGURE 2 bC m O SHOREHAM NUCLEAR POWER STATION c4 IO e DRYWELL MONITOR RESPONSE *
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- BASED ON LEAK RATE OF 0.5 % PER DAY. I' l'
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' FOR DIFFERENT LEAK RATES MULTIPLY !'
CURVE VALUES BY- -
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' - LEAK RATE (% PER DAY) ' _
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. DRYWELL MONITOR READING CORRESPONDING 7"4 ~ ~ ~ ~
i+
# TO A DOSE RATE OF 500 MREM /HR WHOLEBODY e 1-1-
i /j OR 2500 MREM /HR THYROID AT THE SITE 3 -- BOUNDARY FOR TWO MINUTES UNDER ADVERSE -
+- ---
5 I METEOROLOGICAL CONDITIONS a w 5 10 _ _ . N. . c-m '*
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THYROID ATTHE SITE BOUNDARY '
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,.n 1000 s : TIME AFTER REACTOR SCRAM (HRS.)
F .
~O O [ T-l g dm ,. r-]
I 1
' ' O' '
FIG. S AE 13-3 Id, i SAE 13 PAGE 5 OF 5 t 4 6
GENERAL EMERGEllCY No.1 O Initiating Conditions
- 1. Effluent monitors detect levels corresponding to 1 rem /hr wholebody or 5 rem /hr thyroid at the site boundary under actual meteorological conditions (meteorological data required for determination).
- 2. These dose rates are projected based on other plant parameters (e.g.,
radiation levels in containment with leak rate appropriate for existing containment pressure with some confirmation from effluent monitors) or are measured in the environs. Emergency Action Levels la. Radiation Monitoring System (RMS) dose projection indicates dose rates of 1 rem /hr wholebody or 5 rem /hr thyroid at the site boundary under actual meteorological conditons.
- b. With RMS inoperative, RBSVS intermediate range effluent monitor 1D11*RE-021B or 022B on panel 1H11*PNL-080 reads above the curve given in Figures 1-7 for the corresponding time after shutdown, stability class, effective plume height, wind speed, and RBSVS flow.
NOTE: Before using any of the curves, the following values must be known (from Health Physics Technician using EPIP 2-2) r ( a. RBSVS flow
- b. Stability class
- c. Effective plume height j d. Wind speed (ground or elevated)
PROCEDURE 1. Using stability class (A-G) locate correct figure. FOR USING CURVES 2. Pick correct curve using effective plume height. Do not interpolate (h r 0, 35, 70, 105 or 140 meters ONLY).
- 3. Read off value for monitor reading for appropriate time after scram.
- 4. Correct monitor reading using RBSVS flow and windspeed (use ground or elevated wind speed for ground or elevated releases respectively. Ground level release for effective plume height of zero.)
2a. Drywell high range monitor 1D11*RE-085A or B on panel ID11*PNL117A/B reads above the curve given in Figures 8-14 for the corresponding time after shutdown, stability class, effective plume height, wind speed and drywell l leak rate. GE 1 Pag" 1 of 17 i i
Y NOTE: Before using any of the curves, the following values must be known (from Health Physics Technician using EPIP 2-2)
'O a. Drywell leak rate 4
- b. Stability class
- c. Effective plume height
- d. Wind speed (ground or elevated) i I
PROCEDURE 1. Using stability class (A-G) locate correct figure. i FOR USING CURVES 2. Pick correct curve using effective plume height. Do not interpolate (h = 0, 35, 70, 105 or 140 meters ONLY).
- 3. Read off value for monitor reading for appropriate time after scram.
) 4. Correct monitor reading using drywell leak rate and windspeed (Use ground or elevated wind speed for ground or elevated releases respectively. Ground level j release for effective plume height of zero.) gR . l
- b. Survey teams measure any of the following radiation levels a't the site I
boundary: lO l 1. I rem /hr wholebody l
- 2. Procedure EPIP 2-24, Thyroid Dose Commitment Using TCS Air Sampler, indicates one hour thyroid dose commitment value of 5 rem Logic Diagram See separate sheet O GE 1 Page 2 of 17
O O O i j GENERAL EMERGENCY No. I LOGIC DI AGRAM i i l J WITH RMS INOPERATIVE' RBSVS INTERMEDI ATE DRYWELL HIGH RANGE RMS DOSE PROJECTION READS ABOVE ONE OF THE RANGE MONITOR READS SITE BOUNDARY RADI ATION INDICATES DOSE RATES LEVELS EXCEED ANY OF ABOVE ONE OF THE CURVES CURVES (FIGURES 8-14) FOR EXCEEDING IREM/HR THE CORRESPONDING TIME THE FOLLOWING:
" (FlGS l-7) FOR THE CORRESPONDING TIME ABOVE SHUTDOWN STABILITY HYROI T TH S C SS EFF T LUM eIREM/HR WB 1 BOUNDARY UNDER ACTUAL AFT SHUTDOWN TAB ITY METEOROLOGICAL CONDITIONS COMMITMENT OF 5 REM HEIGH'T, WIND SPEED, AND DRYWELL LEAK RATE RBSVS FLOW I
GENERAL . EMERGENCY No. I FIG. GE l-1 GE I PAGE 30F 17
g% W e e
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- STABILITY CLASS A lo.io... . ._ ;
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%$ 1 i h "~; j i* CURVES FOR RBSYS FLOW RATE OF 1160 CFM ii Uli : -.
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l { "E L .L i! AND WINDSPEED OF 1 METER /SEC. -., . . -
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; RBSVS INTERMEDIATE RANGE
~T-~
I WINDSPEEDS MULTIPLY CURVE VALUE BY: [ ['l ll[ {,7 T.? ~
}.l MONITOR READINGS CORRESPONDING 1160 x WINDSPEED (M/Sec). F ~7' ~' TO A DOSE RATE OF 1 REM /HR WHOLE ,
2 ,.
.. Jg .i ; BODY OR 5 REM /HR THYROID ATTHE j H L ' SITE BOUNDARY UNDER ACTUAL
- t ::,j-!i.:-
!! RBSYS FLOW RATE (CFM) ----
E~" METEOROLIGICAL CONDITIONS 10 6 j!Ml5IUlN!:l!-lil.fl: ;b...lE. lSlElIb!dUl@bl$ !!h " "
, Li i , i + .I .. l .. l +l.. ++ l-l I l . . l . . . l .+ 1 -! l . 1 O.i i io 10 0 3000 TIME AFTER REACTOR SCRAM (HRS)
I u O [Tl O
>w
.AJ l .AJ OC FIG. GE 1-2 (T] GE 1 PAGE 4 OF 17
-e 4 2 4
p~-~
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C > W e m3 o c 2 >< e. a C o :e= t o>a!C o woa M FIGURE 2 SHOREHAM NUCLEAR POWER STATION RBSVS INTERMEDIATE RANGE MONITOR RESPONSE
- u 10 STABILITY CLASS B 3
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1,... .... . . .... . . ..._ . , . . . , .
.,.. 4:.
.1, i- 1 AND WINDSPEED OF 1 METER /SEC. H n .. .,-,H - . ".n.,. ..
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'~ - :.n 3> 3 4
O FTl T 3> W W i
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/ ,0 1 m A,j FIG. GE 1-3 GE PAGE 5 OF 17
1. ann an*** *' D oO d > o >r T 'g >* W oc - R 5.,.: o - e
- c. )(r
- n. o Er e n
- FIGURE 3 ,
ts- ! l SHOREHAM NUCLEAR POWER STATION RBSVS INTERMEDIATE RANGE MONITOR RESPONSE
- lo H
STABILITY CLASS C
,a. &.J _
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