ML021340147
| ML021340147 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/30/2002 |
| From: | Abney T Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML021340147 (38) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 April 30, 2002 10 CFR Part 50, App E U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Gentleman:
In the Matter of )) Docket Nos. 50-259 Tennessee Valley Authority 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) - UNITS 1, 2, and 3 EMERGENCY PLAN IMPLEMENTING PROCEDURE (EPIP) REVISIONS TVA is submitting this notification in accordance with the requirements of 10 CFR Part 50, Appendix E, Section V, to provide NRC with Revision 32 to EPIP-1. The EPIP revision date for this change is April 7, 2002.
The enclosed information is being sent by certified mail.
The signed receipt signifies that you have received this information. If you have any questions, please telephone me at (256) 729-2636.
Manager of Licensin ryffiK Prnted on reeycled paper
U.S. Nuclear Regulatory Commission Page 2 April 30, 2002 cc (Enclosure)
NRC Resident Inspector (Enclosure provided by Browns Ferry Nuclear Plant BFN Document Control Unit)
P.O. Box 189 Athens, Alabama 35611 Mr. Paul E. Fredrickson (2 Enclosures)
U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street S.W., Suite 23T85 Atlanta, Georgia 30303-8931 Mr. Kahtan N. Jabbour, Senior Project Manager (w/o Enclosure)
U.S. Nuclear Regulatory Commission One White Flint, North (MS 08G9)
Office of Nuclear Reactor Regulation 11555 Rockville Pike Rockville, Maryland 20852-2738
ENCLOSURE TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 EMERGENCY PLAN IMPLEMENTING PROCEDURES (EPIP)
EPIP-1 SEE ATTACHED
GENERAL REVISIONS GENERIC FILING INSTRUCTIONS FILE DOCUMENTS AS FOLLOWS:
PAGES TO BE REMOVED PAGES TO BE INSERTED EPIP-1 REVISION 31 (AFFECTED EPIP-1 REVISION 32 (SUPPLIED PAGES) PAGES)
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE EPIP-1 EMERGENCY CLASSIFICATION PROCEDURE REVISION 32 PREPARED BY: T. W. CORNELIUS PHONE: 2038 RESPONSIBLE ORGANIZATION: EMERGENCY PREPAREDNESS APPROVED BY: GILBERT V. LITTLE DATE: 04/04/2002 EFFECTIVE DATE: 04/07/2002 LEVEL OF USE: REFERENCE USE QUALITY-RELATED
REVISION LOG Procedure Number: EPIP-1 Revision Number 32 Pages Affected: 1, 14. 15. 17, 84. 86. 89 Description of Change:
IC - 42 EPIP 1. rev. 31 revision is being conducted to change the Site Boundary Radiation Reading from a beta-gamma value to gamma only value. This change does not involve the numerical value. This revision is in compliance with the REP and doesn't affect the BFN EP standard emergency classification and action level scheme. This revision is being conducted to ensure consistency with NUMARC/NESP-007., Reg Guide 1.101. and NEI 99-01 (Rev. 4).
IC - 43 EPIP 1. rev. 32 is being conducted to modify information that support EAL 1.1 -G 1.
- 1. l-G2. and 1.2-G. The revision incorporates changes resulting from modifications to calculations that support Minimum Alternate RPV Flooding Pressures (MARFP) and Minimum Steam Cooling Reactor Water Level (MSCRWL). Revisions to these calculations were conducted in support of the EOI Program Manual Revision 21 (U3Cl 1).
EPIP-1 EMERGENCY CLASSIFICATION PROCEDURE TABLE OF CONTENTS PAGE NUMBER REVISION TA BLE O F CON TENT S ............................................................................ 1 32 SECTION I INTRODUCTION CLASSIFICATION INSTRUCTIONS ................................ ...... 3 29 GLOSSARY ......................................................... .......... .29 EVENT CLASSIFICATION INDEX ............................................ 11 29 SECTION II EVENT CLASSIFICATION MATRIX 1.0 REACTOR ....... ............... .............. 13 30 2.0 PRIMARY CONTAINMENT ................................................. 21 28 3.0 SECONDARY CONTAINMENT ........................................... 29 29 4.0 RADIOACTIVITY RELEASES .............................................. 33 30 5.0 LO SS OF POWER .................................................................. 39 29 6.0 H AZ A RDS .............................................................................. 45 29 7.0 NATURAL EVEN TS .............................................................. 61 28 8.0 EMERGENCY DIRECTOR JUDGEMENT ............................ 67 29 SECTION III BASIS 1.0 REA C TO R .............................................................................. 75 30 2.0 PRIMARY CONTAINMENT ................................................. 97 28 3.0 SECONDARY CONTAINMENT ......................................... 116 29 4.0 RADIOACTIVITY RELEASE .............................................. 126 30 5.0 LO SS O F POW ER ................................................................ 139 29 6.0 H AZ AR D S ............................................................................ 155 29 7.0 NATURA L EVENTS ............................................................ 183 28 8.0 EMERGENCY DIRECTOR JUDGEMENT .......................... 190 29 PAGE 1 OF 207 REVISION 32
SEPIP-1 EMERGENCY CLASSIFICATION PR*OCEDURE THIS PAGE INTENTIONALLY BLANK PAGE 2 OF 207 REVISION 32
EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASS....CAT. N MATRnIX f. T)E' 1 K REACTOR 1.0 1.0 REACTOR PAGE 13 OF 207 REVISION 30
SECTION II CLASSIFICATION 1.0 REACTOR EVENTCLASSIFICATION MATRIIX PROCEDURE NOTES:
1.1-Ul/1.1-A1 Applicable when the Reactor Head is removed and the Reactor Cavity is flooded.
1.1-Si Applicable in Mode 5 when the Reactor Head is installed.
1.l-G2 The reactor will remain subcritical under all conditions without boron when:
- All control rods are inserted to or beyond position 02
- All control rods except one are inserted to or beyond position 00
- Determined by reactor engineering CURVES/TABLES:
NUMBER OF OPEN MSRVs MARFP (PSIG)
UNIT2 UNIT 3 6 or More 180 190 5 720 230 4 280 290 REVISION 30 PAGE 14 OF 207 1.0 REACTOR
LLAbSLk ILAL LION SECTION II PRO t)6ikLA'1E 1 11 d*'k "11*11714. a EVENT CLASSIFICATION
.A..*T ID CTOR I IIJ1' DESCRIPTION I
1.l-U1 i 1.1-U2 Uncontrolled water level decrease in Reactor Cavity Uncontrolled water level decrease in Spent Fuel Pool V with irradiated fuel assemblies expected to remain I with irradiated fuel assemblies expected to remain covered by water. covered by water.
OPERATING CONDITION: OPERATING CONDITION:
- Mode 5 - All 1.-A.-A Uncontrolled water level decrease in Reactor Cavity Uncontrolled water level decrease in Spent Fuel expected to result in irradiated Fuel assemblies being Storage Pool expected to result in irradiated fuel uncovered. assemblies being uncovered.
OPERATING CONDITION: OPERATING CONDITION:
- Mode 5 All 1.1-Si 1.1-$2 Reactor water level CANNOT be maintained above Reactor water level CANNOT be determined.M
-162 IN. (TAF)
M OPERATING CONDITION: OPERATING CONDmON:
-All -Model -Mode 3
- Mode 2 1.1-G1 1.1-G2 FIN Reactor water level CANNOT be restored and Reactor water level CANNOT be determinedM AND maintained above: EITHER of the following conditions exists:
"* UNIT 2 -190 IN. M
- Therewowill mmamaitical w/o bxka UNIT 3 -185 IN. Ucmxham and I=#=n 4 MSRVscan beopied4, orReadctrplse CANNOT be restredandmairaedkat least 65 PSI aboveSu"pesicn Chanber
- It has NOT been determined that the reactor will remain subcritical w/o boron under all conditions and unable to OPERATING CONDITION: reameanrmilainMARP in Table 1.1-G2.
- Mode I - Mode 3 OPERATING CONDITION:
-Mode2 -Model -Mode3
-Mode2 1.0 REACTOR PAGE 15 OF 207 REVISION 30
SECT[ION H CLASSIFICATLON 1.0 REACTOR EVENT CLAP'IFICATION MAIRJIX PROCEDURE NOTES:
1.2 Subcritical is defined as Reactor pOwer below the heating range and not trending upward.
CURVES/TABLES:
CURVE 1.2-G HEAT CAPACITY TEMP LIMIT 260 250 SAFE WHEN RXPRS
............ . ....... EI OW 65 PSIG 240 230~~................IP rs.6 23230 220' AM .I.0 .....
- "iR V Press. ....
Lu 210 *
-0J S200 'RPV Press. 500 I 190, PV Press. 700 co 180. iRPV Press" 900 160 RPV Press. 1135 15 15.5 *16 16.5 17 18.5 19 REVISION 11.5 30 .. 12 12.5 13 13.5 14 14.5 PAG 16.F.07..0.EATO 17.5 18 SUPPR PL LVL (FT) .
N .ACTION RIEQUIRIED IF ABOVE CURVE FOR EXISTING RX REVISION 30 PAGE 16 OF 207 1.0REACTOR
CLASSIFI(ATION' SECTION I1 MR A'n"vr
"'i ,, ' EVENT~~~I AI.R,
';I*.**
igl I .*
CTOR SCRAM FAILURE S TRECACO D COOLANTA
- .--.--- ACTIVITYJ LA DESCRIPTION DESCRIPTION
'1.3-U Reactor coolant activity exceeds 26 pCi/gm dose equivalent 1-131 (Technical Specification Limit) as determined by chemistry sample.
OPERATING CONDITION:
ALL 1.2-A 1.3-A Failure of automatic scram functions to bring the Reactor coolant activity exceeds 300 pCi/gm dose Reactor subcritical equivalent Iodine-I 31 as determined by chemistry ANDsape Manual scram or ARI was successful. Msample OPERATING CONDITION: OPERATING CONDITION:
- Mode 1 -Model -Mode 3
-Mode 2 -Mode 2 1.2-S W
Failure of automatic scram, manual scrani, Adl io M.,
bring the Reactor subcritical. M OPERATING CONDITION: .
- Mode 1.2-G Failure of automatic scram;: manual soram, and ART
.L, ..... '
Reactor power >3%
AND EITHER of the following conditio s exists:
- Suppression Pool temp exceeds HCTh.. .
Refer to Curve 1.2,G.-*
- Reactor water level CANNOT bý restored'."
and maintained at or above:.... .
SuNrr2. -190 IN.
- UNIT3 -1851N .
OPERATING CONDITION:
- Mode l
- Mode 2 1.0 REACTOR PAGE 17 OF 207 REVISION 30
SkECTION It CLASSIF TAN, TO 1.0 REACTOR EVENTCLASSIFICATION MATRIX PROCEDURE
- 4 CURVES/TABLES:
CURVE 1.5-S HEAT CAPACITY TEMP LIMIT.
260 ..
250*
240° m*VPes.
65 230,*
Lu 210RPV Press. 300
- -200, RPV Press. 500 La 190, RV Press. 700 S180, RPV Press. 0 150 .........i 11.5 12V 12. 131.54 345 155.
6" 16.5 17 17.5 18 18.5 19 SUPPR PL LVL (FT)
SACTION REQUIRED IF ABOVE CURVE FOR EXISTING RX REVISION 30 PAGE 18 OF 207 1.0 REACTOR
A-RO1k(C-A11tN SECTION 1-PRO tEI)VRE "WISLOFFAS RADATINT LOSS OFCA~
MTDECY EA 1.4-U[
Valid MAIN STEAM LINE RADIATION HIGH-HIGH alarm, RA-90-135C OR Valid OG PRETREATMENT RADIATION HIGH alarm, RA-90-157A.
OPERATING CONDITION:
-Model -Mode3
-Mode2 1.5A EV .M.A..'
N
.L S IFReactor V T moderator
.i.r maintained V temperature
.: ...iCANNOT
, be below 2120 F whenever Technical Specifications require Mode 4 conditions or during operations in Mode 5.
OPERATING CONDITION:
-Mode
.- S.
..... Suppression Pool temperature, level and
.... .RPV
.in thepressure safe areaCANNOT of Curve be maintained 1.5-S. M S
- f OPERATING CONDmON:,
": -Model
,-.... -Moe.2.
1.0 REACTOR PAGE 19 OF 207 REVISION 30
SECTION 11 CLASSIFjIjAT Io.N 1.0 REACTOR EVENT CLASSIFICATION MATRIX PROCEDURE THIS PAGE INTENTIONALLY BLANK REVISION 30 PAGE 20 OF 207 1.0 REACTOR
EMERGENCY EPIP-1 CLASSJ't[CATION SECTION lII PROCEDURE TECHNICAL BASIS 1.0 REACTOR TECHNICAL BASIS 1.0 REACTOR REACTOR 1.0 1.0 REACTOR PAGE 75 OF 207 REVISION 30 1
EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR UNUSUAL EVENT Uncontrolled water level decrease in Reactor Cavity with irradiated fuel assemblies expected to remain covered by water.
OPERATING - Mode 5 CONDITION BASIS This event classification only applies during Mode 5 when the Reactor Head is removed. For the purposes of this event classification the Reactor Cavity includes the cavity and the Reactor Vessel.
This event classification is anticipatory to 1.1-Al and should only be considered if, in the opinion of the Site Emergency Director, the water level decrease is substantial enough to ultimately result in increased dose rates in the area of the Reactor Cavity due to loss of shielding by water covering irradiated fuel.
Uncontrolled water level decrease during Mode 5 is indicative of valve manipulation error or failure of equipment in such a manner as to cause uncontrolled drainage of the Reactor Cavity. Uncontrolled water level decrease may be detected by the presence of the low level alarm in the spent fuel storage pool, Visual observation, increased radiation levels or various other symptoms that the Site Emergency Director considers valid indicators of the event.
The degraded status of safety systems designed to makeup water to the Reactor Vessel is of particular concern during Mode 5 although plant Technical Specifications require minimum makeup systems be operable except with the spent fuel storage gates removed and water level > 22 feet over the top of the reactor pressure vessel flange. These events tend to have long lead times relative to potential for release outside the site boundary, thus impact to public health and safety is very low. Classification as Unusual Event is warranted as a precursor to a more serious event.
Escalation to Alert is by actual uncovery of irradiated fuel assemblies.
1.0 REACTOR PAGE 76 OF 207 REVISION 301
EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR k= iii WATER LEVEL 1.1-Ul UNUSUAL EVENT (CONTINUED)
REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-AU2 example -1)
Technical Specifications 3.5.2 NOTES NOTE 1.1-U1/1.1-Al Applicable when the Reactor Head is removed and the Reactor Cavity is flooded.
1.0 REACTOR PAGE 77 OF 207 REVISION 30 1
EMERGENCY EPIP-1 CLASSIFICATION SECTION IH PROCEDURE TECHNICAL BASIS 1.0 REACTOR UNUSUAL EVENT Uncontrolled water level decrease in Spent Fuel Storage Pool with irradiated fuel assemblies expected to remain covered by water.
OPERATING All CONDITION BASIS This event classification is anticipatory to 1.1-A2 and should only be considered if, in the opinion of the Site Emergency Director, the water level decrease is substantial enough to ultimately result in increased dose rates in the area of the Spent Fuel Storage Pool due to loss of shielding by water covering irradiated fuel.
Uncontrolled water level decrease may be detected by the presence of the low level alarm in the spent fuel storage pool, visual observation, increased radiation levels or various other symptoms that the Site Emergency Director considers valid indicators of the event.
Uncontrolled water level decrease in Spent Fuel Storage Pools is indicative of failure of equipment in such a manner as to cause uncontrolled drainage. These events tend to have long lead times relative to potential for release outside the site boundary, thus impact to public health and safety is very low. Classification as Unusual Event is warranted as a precursor to a more serious event.
Escalation to Alert is by actual uncovery of irradiated fuel assemblies.
REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-AU2 example-2) 1.0 REACTOR PAGE 78 OF 207 REVISION 30 1
EMERGENCI EPIP-1 CLASSIFICATION SECTION IH1 PROCEDURE TECHNICAL BASIS 1.0 REACTOR ALERT Uncontrolled water level decrease in Reactor Cavity expected to result in irradiated fuel assemblies being uncovered.
OPERATING - Mode 5 CONDITION BASIS This event classification only applies during Mode 5 when the Reactor Head is removed. For the purposes of this event classification the Reactor Cavity includes the cavity and the Reactor Vessel.
Uncontrolled water level decrease during Mode 5 is indicative of valve manipulation error or failure of equipment in such a manner as to cause uncontrolled drainage of the Reactor Cavity. The degraded status of safety systems designed to makeup water to the Reactor Vessel is of particular concern during Mode 5 although plant Technical Specifications require minimum makeup systems be operable except with the spent fuel storage gates removed and water level >.22 feet over the top of the reactor pressure vessel flange.
Uncontrolled water level decrease may be detected by visual observation, increased radiation levels or various other symptoms that the Site Emergency Director considers valid indicators of the event. Expected fuel uncovery may be detected by increased radiation levels, Visual observation, RPV level instrumentation expected to drop below - 162 inches, or best judgement of the Site Emergency Director based on present and past events and trends.
Due to the long lead times associated with these events there is time available to take corrective actions, and there is little potential for substantial fuel damage.
Significant exposures to onsite personnel is likely during these events and it is probable that additional personnel will be needed onsite; therefore the Alert classification is warranted.
Escalation is by Radiological Release event classifications.
1.0 REACTOR PAGE 79 OF 207 REVISION 30 1
EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR WATER LEVEL 1.1-Al ALERT (CONTINUED)
REFERENCES Reg Guide 1.101 Rev. 3, (NUMARC-AA2 example-3)
Technical Specifications 3.5.2 NOTES NOTE 1.1-U1/1.1-A1 Applicable when the Reactor Head is removed and the Reactor Cavity is flooded.
J
+.
1.0 REACTOR PAGE 80 OF 207 REVISION 30 1
EMERGENCY EPIP-1 CLASSIFICATION SECTION 1I1 PROCEDURE TECHNICAL BASIS 1.0 REACTOR ALERT Uncontrolled water level decrease in Spent Fuel Storage Pool expected to result in irradiated fuel assemblies being uncovered.
OPERATING - All CONDITION BASIS Uncontrolled water level decrease in Spent Fuel Storage Pools is indicative of failure of equipment in such a manner as to cause uncontrolled drainage. These events tend to have long lead times relative to potential for release outside the site boundary, thus impact to public health and safety is very low.
Uncontrolled water level decrease may be detected by visual observation, increased radiation levels or various other symptoms that the Site Emergency Director considers valid indicators of the event. Expected fuel uncovery may be detected by increased radiation levels, Visual observation, or best judgement of the Site Emergency Director based on present and past events and trends.
There is time available to take corrective actions, and there is little potential for substantial fuel damage. Offsite exposures are expected to remain below the Environmental Protection Agency's Protective Action Guidelines; however, exposures to onsite personnel is of particular concern during this event; therefore the Alert classification is warranted.
Escalation is by Radiological Release event classifications.
REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-AA2 example-4) 1.0 REACTOR PAGE 81 OF 207 REVISION 30 1
EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR SITE AREA EMERGENCY Reactor water level cannot be maintained above -162 in. (TAF).
OPERATING - ALL BASIS If Reactor water level cannot be maintained above TAF the potential exist for fuel cladding damage. Events most likely to result in coolant inventory loss to this extent are RCS boundary degradation events or station blackout events. For this event to be declared, RPV water level must have decreased or be trending to a value that, in the opinion of the Site Emergency Director, has resulted in or will result in some actual core uncovery. Additionally, the Site Emergency Director must have evidence that Reactor level has been or can be recovered to above TAF.
This event classification also applies in Mode 5 when the Reactor Vessel head is installed. Inadvertent draining of the Reactor Vessel is possible under these conditions due to valving errors associated with the RHR system or failures associated with isolation valves during alignment changes of systems connected to the Reactor Vessel below the normal water level.
The fact that the transient was severe enough to result in inability to maintain RPV level coupled with the anticipatory nature of this event classification as a precursor to more serious event warrants the Site Area Emergency event classification.
For events that occur during operation, escalation to General Emergency is based on inability to assure adequate core cooling by restoring and maintaining RPV water level following transients that have resulted in extreme RPV water level decrease. For events that occur during shutdown or Mode 5, escalation is by radioactive release event classifications.
REFERENCES - Reg Guide 1.10 1 Rev. 3, (NUMARC-FS, SS5, SS4, example-I)
- EOI Program Manual Section VI-J NOTES NOTE 1.1-S I Applicable in Mode 5 when the Reactor Head is installed.
1.0 REACTOR PAGE 82 OF 207 REVISION 30 1
EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR WATER LEVEL U-S2 SITE AREA EMERGENCY Reactor water level cannot be determined.
OPERATING - Mode l CONDITION - Mode 2
- Mode 3 BASIS Inability to determine Reactor water level during operation may be due to boiling in the reference or variable instrument legs, instrument power failures, or conflicting information from uncontrolled indicator oscillations.
This condition requires Reactor flooding following emergency depressurization.
Adequate core cooling is assured by these measures. Due to the severity of these actions and the uncertainty of Reactor status it is appropriate to treat this as a potential loss for Reactor Coolant System and Fuel Cladding integrity; therefore, this event is appropriate for the Site Area Emergency classification.
Escalation to General Emergency is based on inability to assure adequate core cooling in this mode.
REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-FS)
- EOI Program Manual Section VI-J 1.0 REACTOR PAGE 83 OF 207 REVISION 30 1
EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR k = A _______
GENERAL EMERGENCY Reactor water level CANNOT be restored and maintained above:
UNIT 2 -190 IN.
UNIT 3 -185 IN.
OPERATING - Mode l CONDITION - Mode 2
- Mode 3 BASIS If Reactor water level cannot be restored and maintained above the Minimum Steam Cooling Reactor Water Level (MSCRWL), core damage is possible due to inadequate steam generation, by the covered portion of the Reactor core, to remove decay heat and prevent cladding heatup to a point that results in clad failure.
For either of the above conditions to be met, the control room operators should have progressed in the execution of the EOIs to the point that all high pressure and all low-pressure systems that are available within a reasonable time frame have been attempted and are unsuccessful in reversing the adverse RPV water level trend.
Events most likely to result in coolant inventory loss or loss of makeup capability to this extent are RCS boundary degradation events or events resulting from loss of multiple systems such as station blackout. During such transients or accidents the potential for Primary Containment failure increases substantially; therefore, the General Emergency classification is appropriate.
REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-FG)
- EOI Program Manual Section VI-J 1.0 REACTOR PAGE 84 OF 207 REVISION 30 1
EMERGENNCI EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR GENERAL EMERGENCY Reactor water level CANNOT be determined AND EITHER of the following conditions exists:
o The reactor will remain subcritical w/o boron under all conditions.
and Less than 4 MSRVs can be opened, or Reactor pressure CANNOT be restored and maintained at least 65 PSI above Suppression Chamber pressure.
o It has NOT been determined that the reactor will remain subcritical w/o boron under all conditions and unable to restore and maintain MARFP in Table 1.1-G2.
OPERATING - Mode 1 CONDITION - Mode 2
- Mode 3 BASIS Inability to determine Reactor water level during operation may be due to boiling in the reference or variable instrument legs, instrument power failures, or conflicting information from uncontrolled indicator oscillations. This condition requires Reactor flooding following emergency depressurization. It the reactor will remain subcritical without (w/o) boron under all conditions, adequate core cooling is assured only if at least 4 MSRVs are opened and Minimum Reactor Flooding Pressure (MRFP) is maintained with Reactor pressure at least 65 PSI above Suppression Chamber pressure. If it has not been determined that the reactor will remain subcritical without (w/o) boron under all conditions, adequate core cooling can only be assured when the Minimum Alternate Reactor Flooding Pressure (MARFP) is restored and maintained. If adequate core cooling is not assured core damage is probable under this scenario due to the extreme nature of the plant conditions that resulted in the inability to determine Reactor level (i.e.,
high containment temperatures, loss of multiple power supplies, etc.). Primary Containment integrity cannot be assured under all these conditions; therefore, the General Emergency classification is appropriate.
REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-FG) !--
- EOI Program Manual Section VI-J -- h 1.0 REACTOR PAGE 85 OF 207 REVISION 30
EMERGENCY EPIP-1 CLASSIFICATION SECTION I PROCEDURE TECHNICAL BASIS 1.0 REACTOR WATER LEVEL 1.1-G2 GENERAL EMERGENCY (CONTINUED)
CURVES/TABLES TABLE 1.1 - G2 NUMBER OF OPEN MSRVs MARFP (PSIG)
UNIT 2 UNIT 3 6 or More 180 190 5 220 230 4 280 290 NOTES NOTE 1. 1-G2 The reactor will remain subcritical under all conditions w/o boron when:
"* All control rods are inserted to or beyond position 02
"* All control rods except one are inserted to or beyond position 00
"* Determined by reactor engineering 1.0 REACTOR PAGE 86 OF 207 REVISION 30 1
EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE TECHNICAL BASIS 1.0 REACTOR SCRAM FAILURE 1.2-A ALERT Failure of automatic scram functions to bring the Reactor subcritical AND Manual scram or Alternate Rod Insertion (ARI) was successful.
OPERATING Mode 1 CONDITION Mode 2 BASIS A manual scram is any set of actions by the Reactor Operator(s) at the Reactor Control Console which causes control rods to be rapidly inserted into the core and brings the Reactor subcritical.
This event classification indicates failure of the RPS to automatically scram the Reactor. This condition is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient and thus plant safety has been compromised, and design limits of the fuel may have been exceeded.
An Alert is indicated because conditions exist that lead to potential loss of fuel clad or RCS barrier. Any set of actions by the Reactor Operator at Panel 9-5 that cause control rods to rapidly insert into the core and bring the Reactor subcritical is considered a manual scram.
Escalation to Site Area Emergency is based on fuel clad barrier or RCS barrier event classifications.
REFERENCE Reg Guide 1.101 Rev. 3, (NUMARC-SA2)
NOTES NOTE 1.2 Subcritical is defined as Reactor power below the heating range and not trending upward.
1.0 REACTOR PAGE 87 OF 207 REVISION 30 1
EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR SCRAM FAUUIL~URUEU 1.2-SITE AREA EMERGENCY Failure of automatic scram, manual scram, and ARI to bring the Reactor subcritical.
OPERATING - Mode l CONDITION BASIS Manual scram, and ARI are not considered successful if action away from the Reactor Control Console (Panel 9-5) was required to scram the Reactor.
A failure of the automatic and manual scram systems may result in the Reactor producing more heat than the maximum decay heat load for which the safety systems are designed. A Site Area Emergency classification is appropriate because conditions exist that lead to potential loss of both fuel clad and Reactor Coolant System (RCS) barriers. Therefore, this event classification ensures timely emergency response to the event before actual barriers loss has taken place.
Escalation to General Emergency is based upon inability to bring Reactor power within decay heat removal capability before Suppression Pool temperature reaches the Heat Capacity Temperature Limit (HCTL).
REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-SS2, SS4 example -1)
NOTES NOTE 1.2 Subcritical is defined as Reactor power below the heating range and not trending upward.
1.0 REACTOR PAGE 88 OF 207 REVISION 30 I
EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR SCRA FAILE!WURIEE 1.2 GENERAL EMERGENCY Failure of automatic scram, manual scram, and ARL Reactor power > 3%.
AND EITHER of the following conditions exists:
"* Suppression Pool temperature exceeds HCTL.
Refer to curve 1.2-G.
"* Reactor water level CANNOT be restored and maintained at or above:
UNIT 2 -190 IN.
UNIT 3 -185 IN.
OPERATING - Mode 1 CONDITION - Mode 2 BASIS Automatic scram, manual scram, and ARI are not considered successful if action away from the Reactor Control Console was required to scram the Reactor.
Under these conditions all efforts, including boron injection, have been unsuccessful in bringing Reactor power within the decay heat removal capability of the Emergency Core Cooling Systems (ECCS). Additionally, an extreme challenge to the ability to cool the Reactor Core exist if Reactor Pressure Vessel (RPV) water level cannot be maintained sufficient to ensure adequate core cooling.
Another consideration is the inability to remove heat using the Main Condenser or Suppression Pool. In the event that neither heat sink is effective and Reactor power remains above this level, then a core melt sequence exists. In this situation, core degradation can occur rapidly; therefore, a General Emergency classification is appropriate in anticipation of degradation of multiple fission product barriers.
REFERENCES - Reg Guide 1.101 Rev. 3,(NUMARC-SG2)
- EOI Program Manual Section V-K and Section V-D 1.0 REACTOR PAGE 89 OF 207 REVISION 30
EMERGENCY EPIP-1 CLASSIFICATION SECTION II1 PROCEDURE TECHNICAL BASIS 1.0 REACTOR I GENERAL EMERGENCY (CONTINUED)
[
CURVIES/TABLES CURVE 1.2-G HEAT CAPACITY TEMP LIMIT 260 250, i * ."SAFEs EWHEN65RXpsIPRESS
......... .*I ISP'rsSI6*
BELOW65PI
- "230 .. . ....................
... ~ ~ ~ *!*!**!::::::!*!!!!.........
- !:: ...... *i..........s .::: .... .. i : : .;-~-!( .:i.'....:::: ii:i S1 9 2 2,.0 :::"*"**. ............. p ....P.ress iiiii .::
I~i
- .... 70 0 .... .. :::, :: :. :: : :- ::i*::**;:ii i "..."",....
o*=~~
- ii:* :: :
..... 9 0...
'"l.r..;;::l::::::~::~~::l:!ii: : : - : *
=.......................
============= "'
- j10 2 ...... .....RPV Press. 003 F ....... ...
...A .....
,:SUPP PrLs (FT) ...
RE.UIRE.
i:ii:i:i: ..:... IFAOV.UREFO.XITNGR O..... .... .......... ..
1.0 REACTOR PAGE 90 OF 207 REVISION 30 I
EMERGENCI EPILP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR REACTORKS COSOLAN ACTwruIViITY' 1.3U UNUSUAL EVENT Reactor coolant activity exceeds 26 ItCi/gm dose equivalent 1-131 (Technical Specification limit) as determined by chemistry sample.
OPERATING - All CONDITION BASIS Reactor coolant activity samples exceeding Technical Specification limits for Iodine spikes are representative of fuel clad degradation. An Unusual Event is declared because of potential degradation in the level of safety of the plant. Iodine levels exceeding Technical Specification limits are a potential precursor of more serious problems.
Escalation to Alert would be based on higher Reactor coolant activity values indicative of significant fuel failure.
REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-SU4 example-2)
- Technical Specification 3.4.6 1.0 REACTOR PAGE 91 OF 207 REVISION 30
EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR
- ~ wRuEACTORSS COOLAN WACTU IVT 1.3-A ALERT Reactor coolant activity exceeds 300 #Ci/gm dose equivalent Iodine-131 as determined by Chemistry sample.
OPERATING - Mode l CONDITION - Mode 2
- Mode 3 BASIS Reactor coolant activity samples exceeding 300 pCi/gm dose equivalent Iodine 131 are well above those expected for Iodine spikes and represent a significant loss of the fuel clad barrier. Any loss or potential loss of the fuel clad barrier warrants the declaration of an Alert.
Escalation to Site Area Emergency would be based on the conditions given above coupled with a loss or potential loss of either the Primary Containment or Reactor Coolant System barrier or Radiological Releases.
REFERENCE - Reg Guide 1.101 Rev. 3, (NUMARC-FA)
- RIMS L36 921201 806 1.0 REACTOR PAGE 92 OF 207 REVISION 30 1
EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR UNUSUAL EVENT Valid MAIN STEAM LINE RADIATION HIGH-HIGH alarm, RA-90-135C OR Valid OG PRETREATMENT RADIATION HIGH alarm, RA-90-157A.
OPERATING - Mode 1 CONDITION - Mode 2
- Mode 3 BASIS Main Steam Line radiation high high or offgas radiation high is indicative of fuel cladding leakage.
The Main Steam Line radiation high high alarm setpoint is normally set at 3 times normal full power background. 3 times normal full power background is in excess of any spikes expected from operational transients that do not result in cladding failure. This alarm setpoint is substantially above that which would be indicative of fuel cladding damage above Technical Specification allowable limits; however, the presence of a valid alarm warrants declaration of an Unusual Event and consideration of other symptoms and event classifications for possible upgrade of the event based on fission product barrier loss.
The offgas pretreatment radiation high alarm setpoint is set at a value that is indicative of the ODCM allowable limits for radiation release.
Either of these conditions is considered a potential degradation in the level of safety of the plant and a potential precursor of a more serious problem.
Escalation to the Alert is based on either Reactor coolant samples exceeding 300 ptCi/gm or Drywell radiation levels indicative of loss of the fuel cladding barrier.
REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-SU4 example-I) 1.0 REACTOR PAGE 93 OF 207 "RVISION 30 1
EMERGENCY EPLP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR ALERT Reactor moderator temperature CANNOT be maintained below 212'F whenever Technical Specifications require Mode 4 conditions or during operations in Mode 5.
OPERATING - Mode 4 CONDITION - Mode 5 BASIS This event classification addresses loss of decay heat removal functions when Mode 4 is required or during Mode 5. Loss of decay heat removal capability can result in more serious consequences depending upon whether Primary Containment is in tact and Emergency Core Cooling System (ECCS) equipment status. In any condition where Mode 4 is required, loss of decay heat removal capability represents a significant degradation in plant conditions that can lead to fuel cladding damage or RCS degradation. In order to maintain anticipatory philosophy the Alert classification is appropriate for this event.
Escalation to Site Area Emergency or General Emergency is by loss of Reactor water level that has or will uncover the fuel or Radiological Release Event classification.
REFERENCES Reg Guide 1.101 Rev. 3, (NUMARC-SA3) 1.0 REACTOR PAGE 94 OF 207 REVISION 30
EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR SITE AREA EMERGENCY Suppression Pool temperature, level and RPV pressure CANNOT be maintained in the safe area of Curve 1.5-S (Heat Capacity Temperature Limit)
OPERATING - Mode I CONDITION - Mode 2
- Mode 3 BASIS Suppression Pool temperature is limited by Curve 1.5-S as a function of suppression pool level and reactor pressure in order to preclude failure of Primary Containment or equipment necessary for the safe shutdown of the plant following emergency depressurization. When Suppression Pool temperature cannot be maintained below the limits of the curve corresponding to existing suppression pool level and reactor pressure, emergency depressurization is required and continued decay heat removal at operating temperature and pressure is no longer permissible.
Suppression Pool level is limited by Curve 1.5-S to the range of 11.5 feet to 19 feet in order to preclude failure of Primary Containment or equipment necessary for the safe shutdown of the plant and preserve the pressure suppression function of the containment for possible future emergency depressurization. When Suppression Pool level cannot be maintained within the limits of the curve, continied decay heat removal at operating pressures and temperatures is no longer permissible and emergency depressurization is required.
Exceeding the limits of Curve 1.5-S represents a loss of heat sink for decay heat removal and inability to maintain Mode 3. Under these conditions there is an actual failure of systems intended for protection of the public; therefore, Site Area Emergency is warranted. Escalation to General Emergency is by Abnormal Rad levels, Radiological Release or Primary Containment failure events.
REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-SS4)
- EOI Program Manual Sections VI-C and VI-F 1.0 REACTOR PAGE 95 OF 207 REVISION 30 1
EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR SITE AREA EMERGENCY (CONTINUED)
CURVES/TABLES CURVE 1.5-S HEAT CAPACITY TEMP LIMIT 260 250 240 2301 u
220 Lu a_ 210
-J 200 a
0m 190 180 170 160 I 1 1 150, I 'I I I I I I:Ii i I11 11.5 12 12.5 13 13.5 14 14.5 15 15.5 16 16.5 17 17.5 18 18.5 19 SUPPR PL LVL (FT)
ACTION REQUIRED IF ABOVE CURVE FOR EXISTING RX 1.0 REACTOR PAGE 96 OF 207 REVISION 30