ML041190358
| ML041190358 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/15/2004 |
| From: | Abney T Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| EPIP-1, Rev 36 | |
| Download: ML041190358 (38) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Decatur, Aabarna 35609-2000 April 15, 2004 10 CFR Part 50, App E U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Mail Stop:
OWFN, Pl-35 Washington, D.C. 20555-0001 Gentlemen:
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.
Specifically, EPIP-l, Table of Contents, Revision 36; and Sections II-1.0 and III-1.0, Revision 32 are being provided.
The effective date for these revisions is March 22, 2004.
If you have any questions, please telephone me at (256) 729-2636.
St cerely T. E.
y Manager of Licensin and I ry Affa s
cc:
See Page 2 p"ted n Lcya MM
U.S. Nuclear Regulatory Commission Page 2 April 15, 2004 cc (Enclosure):
NRC Resident Inspector (Enclosure provided by Browns Ferry Nuclear Plant BFN Document Control Unit) 10833 Shaw Road Athens, Alabama 35611-6970 Mr. Stephen J. Cahill, Branch Chief (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 Eva A. Brown, 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 (BFN)
UNITS 1, 2, AND 3 EMERGENCY PLAN IMPLEMENTING PROCEDURE (EPIP) REVISIONS EPIP-1 SEE ATTACHED
GENERAL REVISIONS FILING INSTRUCTIONS FILE DOCUMENTS AS FOLLOWS:
PAGES TO BE REMOVED PAGES TO BE INSERTED EPIP-1, TOC Revision 35 SECTION II -
1.0 Revision 31 SECTION III -
1.0 Revision 31 EPIP-1, TOC Revision 36 SECTION II -
1.0 Revision 32 SECTION III -
1.0 Revision 32
TENNESSEE VALLEY AUTHORITY BROWN'NS FERRY NUCLEAR PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE EPIP-1 EMERGENCY CLASSIFICATION PROCEDURE REVISION 36 PREPARED BY: RANDY WALDREP PHONE: 2038 RESPONSIBLE ORGANIZATION: EMERGENCY PREPAREDNESS APPROVED BY: PHILLIP CHADWELL EFFECTIVE DATE: 03/22/2004 DATE: 03/19/2004 LEVEL OF USE: REFERENCE USE QUALITY-RELATED
REVISION LOG Procedure Number:
EPIP-1 Revision Number 36 Pages Affected:
15,16,17,18,19,84,85,89,90,96 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 NUMARCINESP-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-GI,
- 1. 1 -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 (U3CI 1).
IC-44 EPIP 1, rev. 33 is being issued to modify EAL 6.7-U. The revision incorporates changes resulting from the letter from NEI to NRC (to Mr. Bruce A. Boger) dated December 18, 2001 requesting confirmation for EAL basis change to include response to a Site -Specific Credible Threat. This was developed in response to NRC's October 6, 2001 Safeguards Advisor. This is additional information and does not change existing criteria in the EAL Basis.
IC-45 EPIP 1, rev. 34 is being conducted to modify information that support EAL 1.1-GI, 1.1 -G2, and 3.1 -S. The revision incorporates changes resulting from modifications to calculations that support Minimum Alternate RPV Flooding Pressures (MARFP),
Minimum Steam Cooling Reactor Water Level (MSCRWL) and Maximum Safe Operating Area Temperature Limits. Revisions to these calculations were conducted in support of the EOI Program Manual Revision 22 (U2C 13).
IC-46 EPIP-I rev. 35 is being conducted to more effectively clarify the Emergency Action Level, 6.3-A. This change was initiated through the BFN Drill and Exercise Program as an Improvement Item.
IC-47 EPIP 1, rev. 36 is being conducted to modify information that supports EAL 1.I-GI, I.I-G2, 1.2-G, and 1.5-S. This revision incorporates changes resulting from engineering calculations that support Minimum Alternate RPV Flooding Pressures (MARFP), Minimum Steam Cooling Reactor Water Level (MSCRWL) and Heat Capacity Temperature Limits. Revisions to these calculations were conducted in support of the EOI Program Manual Revision 23 (U3C 12).
EPIP-1 EMERGENCY CLASSIFICATION PROCEDURE TABLE OF CONTENTS PAGE NUMBER REVISION TABLE OF CONTENTS........................................
1 36 SECTION I INTRODUCTION CLASSIFICATION INSTRUCTIONS........................................
3 29 GLOSSARY........................................
5 29 EVENT CLASSIFICATION INDEX.......................................
I 1 29 SECTION II EVENT CLASSIFICATION MATRIX 1.0 REACTOR.......................................
13 32 2.0 PRIMARY CONTAINMENT.......................................
21 28 3.0 SECONDARY CONTAINMENT.......................................
29 30 4.0 RADIOACTIVITY RELEASES.......................................
33 30 5.0 LOSS OF POWER.......................................
39 29 6.0 HAZARDS.......................................
45 31 7.0 NATURAL EVENTS.......................................
61 28 8.0 EMERGENCY DIRECTOR JUDGEMENT.............................. 67 29 SECTION III BASIS 1.0 REACTOR.......................................
75 32 2.0 PRIMARY CONTAINMENT.......................................
97 28 3.0 SECONDARY CONTAINMENT.......................................
116 30 4.0 RADIOACTIVITY RELEASE.......................................
126 30 5.0 LOSS OF POWER.......................................
139 29 6.0 HAZARDS.......................................
155 31 7.0 NATURAL EVENTS.......................................
183 28 8.0 EMERGENCY DIRECTOR JUDGEMENT............................ 190 29 PAGE 1 OF 207 REVISION 36
EPIP-1 EMERGENCY CLASSEFCATNON PROCEDURE
-1 THIS PAGE INTENTIONALLY BLANK PAGE 2 OF 207 REVISION 36
EMERGENCY CLASSIFICATION PROCEDURE
- EPIP-1
-- - SECTION II EVENT CLASSIFICATION MATRIX 1.0 REACTOR m
-REALC-T OR 1.0 1.0 REACTOR PAGE 13 OF 207' RIO3 REVISION 32
EPIP-1 SECTION II EVENT CLASSIFICATION MATRIX EMERGENCYX CLASSIFICATION PROCEDURE-f 1.0 REACTOR 1
. adnj NOTES:
1-2,~
-I..
r
.;i :_,-,! ':,
1.1l-Ul/.1l-Al 1.1-Si 1.1-G2 Applicable when the Reactor Head is removed and the Reactor Cavity is flooded.
Applicable in Mode 5 when the Reactor Head is installed.
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:
-w!:n.~,~;n~iTABLE
.1-~G2 MINIMUM`ALTERNATE RPVFLODINGPRSS-O' NUMBER OF OPEN MSRVs MARFP (PSIG) 6 or More 190 5
230
- 4.
290 I
i i
I REVISfON 32 PAGE 14 OF 207 1.0 REACTOR
EMERGENCYE CCLASSIFICATION PROCEDURE EPIP-1 SECTION H EVENT CLASSIFICATION MATRIX 1.0 REACTOR 9 Di Iola N DkTJ Di
DESCRIPTION I
DESCRIPTION I
lt' I
.' 5-
.1
- 1. 1-1.1-ul KI Uncontrolled water level decrease in Reactor Cavity with irradiated fuel assemblies expected to remain covered by water.
Uncontrolled water level decrease in Spent Fuel Pool with irradiated fuel assemblies expected to remain covered by water.
.PERATING CONDITION:
-All z
OPERATING CONDmON:
- Mode 5 Uncontrolled water level decrease in Reactor Cavity Uncontrolled water level decrease in Spent Fuel Storag expected to result in irradiated Fuel assemblies being Pool expected to result in irradiated fuel assemblies
-1 uncovered.
. being uncovered.
~~~~~~-
. -_..e OPERATING CONDITION:
OPERATING CONDITION:
-Mode
-Mll 1.1-Sl 1.1-S2 C
Reactor water level CANNOT be maintained above Reactor water 1evel CANNOT be determined.
162 IN. (TAF)
OPERATING CONDmON:
OPERATING CONDITION:
X
-All
-Mode I
--Mode3
.- Mode2 1.1-Gi ll.1-G2 t-,
Reactor water level CANNOT be determinedT Reactor water level CANNOT be restored and AND maintained above:
EITHER of the following conditions exists:
The btor voill enuin subcrtical w/oban nrwx ais conditions find Unit 2
-185 IN.
l'ssthan4 MSRVscanbeopened, Unit3
-1801N.
OR Reactor pess CANNOT be restored and maintained at least
,.65 PSI forUnit 2, or70 PSI forUnit 3, above S ession
.amzberpessum
- It has NOT been detennined that the reactor will remain subcritical Wlo boron under all conditions and unable to OPERATING CONDITION:
restore and nintain MARFP in Table 1.14G2.
-M-Mo1
-2Mode OPERATING CONDmON:
-M-Mode I
-Mode3 1.0 REACTOR PAGE 15 OF 207 REVISION 32
EPIP-1 SECTION II EVENT CLASSIFICATION MATRIX EMERGENCY, i CLASSIFICATION PROCEDURE,,,,-:
1.0 REACTOR NOTES:
1.2 Subcritical is defined as Reactor power below the heating range and not trending upward.
CURVES/TABLES:
}
I CURVE 1.2-G UNIT 2 HEAT CAPACITY TEMP LIMIT
- g8Sgi.et.. ;;Xdt S BELOW 65 PSIG gi 240 R
Pes 65 220 210 Owl RPV Press. 300 IL 200 P
500 n 190 RPVPresI.700 170 RPV Press. 11 35 11.5 12 12.5 13 13.5 14 14.5.15: 15.5 16 16.5 17 17.5 18 18.5 19 4SUPPRPLLV
'(F,.....
ACTION REQUIRED IF ABOVE CURVE FOR EXISTING RX I
II III I
t i
CURVE 1,2-G UNIT 3 HEAT CAPACITYiTEMP LIMIT
- 260, 720 presS 2SUP PL LV (FT) l3 3
t I0 ABOVE R
FOR EIN RX ESS l 1 ig,^,,o r~ii;;0 tts;b 3
E
~i L
SUPRPLLV0(T l DCINlQI=
BV UV O
XSIGR RS J
REVISION 32 PAGE 16 OF 207 1.RAT 4.0 REACTwOR
EMERGENCY CLASSIFICATION PROCEDURE-SECTION II EVENT CLASSIFICATION MATRIX 1.0 REACTOR DESCRIPTION DESCRIPTION Reactor coolant activity exceeds 26 RCi/gm dose eq wiival~ntI-'131 (Technicial Sp~eifi~ati6n LAi*i) as
- determined by chemistry sample.
a En a
1 7
.. I,
. I
. OPERATING CONDITION:
'. -,A LL, I..
1.2-A1.
Failure of automatic scram functic'rs tobring the.,
Reactor coolant activity exceeds 300 paCi/gm dose Reactor subcritical tEquivalent lodine..l 31 as determined by chemistry 4N D.,i pl.
i-Manual scram or ARt iWas Soccessful OPERATING CONDi-oN:
OPERATING CONDITION:
-Mode I Mode I~~~
-Md~ 3
-Mode 2 Moce2 1.2-S Failure of automatic scram, manual scram, and ARI to
- .I bigteReactor subcritical.
z OPERATING CONDITION:
-Mode I.
1.2-GC Failure of automaitic scrarn, manual scram, and ARL Reactor power ~>3%
.*AND EITHR of the following conditions exists:
- Suppression Poo! t-.rnp exceeds HCTL. Refer to Curve 1.2-G.-
- Reactor water level CANNOT be restored and maintained at or abov:
Unit 2 --185 IN.
Unit 3 -180OIN.
OPERATING CONDITION:
- Mode 2
-. M E C O RPoE 1dFe0 R2S I N 3 4.0 REACTOR PAGE 17 OF 207 --, -
RET*rISION32
EPIP-1 SECTION II EMERGENCY"r CLASSIFICATION 1.0 REACTOR EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:
CURVES/TABLES:
CURVE 1.5-S UNIT 2 HEAT CAPACITY TEMP LIMIT 260 250 s
'.e
>g' E. '-,Bt,2._i
.'SAFE WHEN RX PRESS *-,
240 X
=
Ef i;B; Lg
- ii 30sYI,,;
- . S BELOW; 65 PSIG.. -
'Z 230
£ 220
- 0. 200 R0V. re.
500 tL 190
- RPViress.700 w180 P~es90a 170 RPV Press. 1135 SAFE 150_
11.5'12 i25 13 "13.514 4t4.5 151, 15.5'-165"16.5' 17' 17.5 18'18.5 19 SUPPR PL LVL (FT)
B1 ACTION REQUIRED IF ABOVE CURVE FOR EXSTING RX
-,
4 -
-
I I
CURVE 1.5-S UNIT 3 HEAT CAPACITY TEMP LIMIT 260 BPV rIS BELOW 70PSIGs
- 150, 20S URP Press.
30 W (C
200 5:IR IF AB C
E FOr ISTI RX PRESS (L
-A 7o>:- ?gag e
180
- a 11.5 12 13 14 1 5 1 6 17 18 is SUPPR PL LVIL (Fr) 1..An RQUIRED IFABOVE CURVE FOR EXIS'nNG RX PRESS i
i REVISION 32 PAGE 18 OF 207 1.0 REACTOR
EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 !
I SECTION II EVENT CLASSIFICATION MATRIX:
1.0 REACTOR SAI 6
MI IM A
A1 tW'M me I DEN DESCRIPTiON I
DESCRIPTION.
L 1.4-U.
Valid MAIN STEAM IINE RADIATION HIGH-HIGI alarm, RA-90-135C,.
OR Valid OG PRETREATMENT RADIATION HIGH alarm, RA-90-157A.
U zr c
OPERATING CONDITION:
- Mode I
- Mode 2
-Mode 3
., 1 A..,
Reactor moderator temperature CANNOT be
' mrtained below 2120 F whenever Technical iications require Mode 4 conditions or during
- ^ -... >._.* J.'.-
. OPERATING CONDITION:
.:Mode4.
.4
-Mode5 1.5-S
,SlN Suppression Pool temperature, level and RPV pressure CANNOT be maintained in the safe area of Curve 1.5-S.
M OPERATING CONDmON:
-Mode I
-Mode 3
-Mode 2
~~...-..r 1.0 RE A^CTOR-PAGE 19 OF 207 RE-
.ISON3 REVISION 32
EPIP-1 SECTION HII EVENT CLASSIFICATION MATRIX EMERGENCY..
CLASSIFICATION PROCEDURE.,,
1.0 REACTOR i
i 5
s
,3
'u':9
t
'I pi j
-, -a THIS fiAGE ENTIONALT Y BiANK S
's
.x
.r:
a z
REVISION 32 PAGE 20 OF 207 1.0 REACTOR
EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 SECTION III TECHNICAL BASIS 1.0 REACTOR
-4. T..
si, REACTOR1 4,
10 1.0 REACTOR PAGE 75 OF 207 REVISION 32
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 32
EMIERGENCY CLASSIFICATION PROCEDURE EPIP-1 SECTION III TECHNICAL BASIS 1.0 REACTOR A
UNUSUAL EVENT (CONTINUED)
REFERENCES Reg Guide 1.101 Rev. 3, (NUMARC-AU2 example -1)
Technical Specifications 3.5.2 NOTES NOTE 1.1-Ul/1.1-AI Applicable when the Reactor Head is removed and the Reactor Cavity is flooded.
1.0 REACTOR PAGE 77 OF 207 REVISION 32
EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR AI 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 32
EMERGENCY
.EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR AaA 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 32
EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 SECTION III TECHNICAL BASIS 1.0 REACTOR I
ALERT (CONTINUED)
REFERENCES Reg Guide 1.101 Rev. 3, (NUMARC-AA2 examnple-3)
Technical Specifications 3.5.2 NOTES NOTE 1.1-Ul/1.1-Al Applicable when the Reactor Head is removed and the Reactor Cavity is flooded.
1.0 REACTOR PAGE 80 OF 207 REVISION 32
EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR A
A; 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 REVSION 32
EMERGENCY EPIP CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR A
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 1.101 Rev. 3, (NUMARC-FS, SS5, SS4, example-i)
EOI Program Manual Section VI-J NOTES NOTE 1.1-S 1 Applicable in Mode 5 when the Reactor Head is installed.
1.0 REACTOR PAGE 82 OF 207 REVISION 32
EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 SECTION II TECHNICAL BASIS 1.0 REACTOR I
itaNWV SITE AREA EMERGENCY Reactor water level cannot be determined.
OPERATING CONDITION Mode l 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 32
EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 SECTION III TECHNICAL BASIS 1.0 REACTOR S
GENERAL EMERGENCY Reactor water level CANNOT be restored and maintained above:
- Unit 2 eUnit 3
-185 IN.
-180 IN.
OPERATING CONDITION Mode l 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 32
EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 SECTION III TECHNICAL BASIS 1.0 REACTOR K>
GENERAL EMERGENCY Reactor water level CANNOT be determined AND EITHER of the following conditions exists:
- The reactor will remain subcritical v/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 for Unit 2 or 70 PSI for Unit 3 above Suppression Chamber pressure.
- It has NOT been determined that the reactor will remain subcritical Nv/o boron under all conditions and unable to restore and maintain MARFP in Table 1.1-G2.
OPERATING Mode 1 f
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. If 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 for Unit 2 or 70 PSI for Unit 3 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 1.0 REACTOR PAGE 85 OF 207 REVISION 32
EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 SECTION III TECHNICAL BASIS 1.0 REACTOR
'A r..
-t4 I
L.;
411".
A GENERAL EMERGENCY (CONTINUED)
CURVES/TABLES
'eATABLE 11-G2 MINIMUM ALTERNATE RPV FLOODING PRESS NUMBER OF OPEN MSRVs MARFP (PSIG) 6 or More 190 5
230 4
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 32
EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 SECTION III TECHNICAL BASIS 1.0 REACTOR lEHIA AI ECO I
A 9
.. ~l.
I 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 32
EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 SECTION III TECH]NICAL BASIS 1.0 REACTOR TECHNICAL BASIS 1.0 REACTOR
'A A
SITE AREA EMERGENCY Failure of automatic scram, manual scram, and ARI to bring the Reactor subcritical.
OPERATING CONDITION Mode I 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 REN'ISION 32
EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 SECTION III TECHNICAL BASIS 1.0 REACTOR
'A A
GENERAL EMERGENCY Failure of automatic scram, manual scram, and ARI. 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 Unit 3
-185 IN.
-180 IN.
OPERATING CONDITION Mode l 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 ION 32 1.0 REACTOR PAGE 89 OF 207
EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 SECTION III TECHNICAL BASIS 1.0 REACTOR GENERAL EMERGENCY (CONTINUED)
CURVES/TABLES
-t CURVE 1.2-G UNIT 2 HEAT CAPACITY TEMP LIMIT 260.
250-SAFE WHEN RX PRESS f-66tf e2sz i.X> wvX = -* > _I,.
S P BELOW 65 PSIG.
220 W
210 u < &
RPV Press. 30
- a. 200
.PVPr 500 _
aL 190 kRPVPres.700 L
W 180 k^t RPVPre s s9t-4it *7 90 170.
RPV Press.1135 1560 -
l
-65l S5>..SS!
-f e
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)
I771 AcTION RtOCUIRED IF ABOVE CURVE FOR EXISTING RX I
1:
I.0 REACTOR PAGE 90 OF 207 RMISION 32
EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 SECTION III TECHNICAL BASIS 1.0 REACTOR UNUSUAL EVENT Reactor coolant activity exceeds 26 pCi/gm dose equivalent 1-131 (Technical Specification limit) as determined by chemistry sample.
OPERATIN(
CONDITION BASIS I-All 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 32
EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 SECTION III TECHNICAL BASIS 1.0 REACTOR AI ALERT Reactor coolant activity exceeds 300 pCi/gm dose equivalent Iodine-131 as determined by Chemistry sample.
OPERATING CONDITION Mode 1 Mode 2 Mode 3 BASIS Reactor coolant activity samples exceeding 300 tCi/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 32
EMERGENCY EPIP-1.
CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR A
3 A3 A
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 tLCi/gm or Drywell radiation levels indicative of loss of the fuel cladding barrier.
REFERENCES Reg Guide 1.101 Rev. 3, (NUMARC-SU4 example-1) 1.0 REACTOR PAGE 93 OF 207 REVISION 32
EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 SECTION III TECHNICAL BASIS 1.0 REACTOR MOW5 AE IV A11 ALERT Reactor moderator temperature CANNOT be maintained below 2120F 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 REV'ISION 32
EMERGENCY EPIP-1 '*-
CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR A..
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, continued 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)
E0I Program Manual Sections VI-C and VI-F 1.0 REACTOR PAGE 95 OF 207 REVISION 32
EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 SECTION III TECHNICAL BASIS 1.0 REACTOR SITE AREA EMERGENCY (CONTINUED)
CURVES/TABLES CURVE 1.5-S UNIT 2 HEAT CAPACITY TEMP LIMIT 260 250 SAFE WHEN RXPRESS 24 Rres.K 7'
ISBELOW6SPSIG 6 5_
240 u
r^ <
Rv A
_230 t> &-sl t
G r
~220 91L 4,
~2 210-
..;RPV Press. 300 IL 200 _
- =RPV Press.500 l 190 RP__res.0 "180 "iP.es.0 170 RPV Press.1135 150 -
~
7 SAFE1 150 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) i ACTION KEOUIRED IFABOVE CURVE FOR EXISTING RX CURVE 1.5-S UNIT 3 HEAT CAPACITY TEMP LIMIT 260
-v2 22:L 250 3 3 SAFE WHEN RX PRESS 240 IS BELOW 70 PSIG
[
230 C
230 f
.f3l;tb S 3'
r4
.Z~
4~
,sSN3' c
a-w C-C-
a-If C7 I,,
t n',to-..%
A
^
200 Fpp 190
-Prss 1~n L ? ~ i SAFE
- V
-I t i 6 I
II 11.5 12 13 14 15 I
SUPPR PL LVL (FP)
F ABOVE CURVE FOR EXSTING RX PRESS 17 18 19 j7M #CTON w-awm II 1.0 REACTOR PAGE 96 OF 207 REVISION 32