ML20133N786
| ML20133N786 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 01/03/1997 |
| From: | Dean Curtland, Hite R, Tirella C IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT |
| To: | |
| Shared Package | |
| ML20133N784 | List: |
| References | |
| NUDOCS 9701230408 | |
| Download: ML20133N786 (100) | |
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i DUANE ARNOLD ENERGY CENTER i
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EMERGENCY ACTION LEVEL (EAL) i TECHNICAL BASIS DOCUMENT l'!
Revision 2 (For NRC Revien)
December 1996 O
9701230408 970116 PDR ADOCK 05000331 F
PDR i
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Duane Arnold Energy Center EMERGENCY ACTION LEVEL SASES DOCUMENT Rev.2 (forNRCreview) i PAGE 1 of 1 i
APPROVAL SHEET EFFECTIVE DATE: TBD l
f l
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Prepared by:
John S. Fuoto, Ogden Environmental and Energy Services l
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Reviewed by:
__ IM.' Y2h Date:
/e-es-g Charles P. Tirella, Jr., Senior Edrgssky Planner
' Reviewed by:
. ' ML Date:
12'l$ /$
Dean Curtland, Operations Manager l
l Reviewed by:
1/
Date:
/7-M'$
Robert Hite, Radiological Protection Manager Approved by:
[M Date:
/ 77 Kenneth Peveler, Manager, Emergency Planning & Licensing i
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REOUEST FOR ADDITIONAL INFORMATION REGARDING DUANE ARNOLD ENERGY CENTER EAL REVISION TO NUMARC/NESP-007 METHODOI OGY The NRC has completed its initial review of the proposed Emergency Action Levels (EALs) contained in the September 15,1995, Duane Amold Energy Center submittal.
The submittal consisted of the proposed EAL procedure, the Duane Amold EAL Technical Basis Document, letters of agreement from State and local authorities, and I
copies of applicable Emergency Operating and Abnormal Operating procedures. The EAL procedure contained the EAL statements, the corresponding emergency classifications, a unique designator number for each EAL, the plant Operating Condition Applicability, and any tables or other data necessary for interpretation of the EAL. The Technical Basis Document gave further details on the EAL, providedjustification for any deviations from the NUMARC example EALs and cited specific Duane Arnold procedure numbers and other related references.
The proposed EALs were reviewed against the guidance in NUMARC/NESP-007,
" Methodology for Development of Emergency Action Levels," Revision 2. This document has been endorsed by the NRC in Regulatory Guide 1.101," Emergency Planning and Preparedness for Nuclear Power Reactors," Revision 3, as an attemative means by which licensees can meet the requirements in 10 CFR 50.47 (b) (4) and Appendix E to 10 CFR Part 50. Since the staff has previously endorsed the guidance in NUMARC/NESP-007, the review focused on those EALs that deviated from the guidance and those EALs that required the development of site-specific thresholds. As a result of the initial review, a number of EALs were identified which required additional informanon h order to determine whether the EALs conform to NUMARC/NESP-007.
Please providt this additional information as discussed below.
GENERAL Issued No.1 The Duane Arnold EAL scheme deviated from the NUMARC methodology by not grouping EALs under initiating conditions (ICs). The Duane Arnold EAL basis document did group the EALs under ICs; however, this arrangement was not majntained in the emergency implementing procedure used for classifying the emergency. The grouping of EALs under the ICs to which the EALs correspond allows the person classifying (and 'he people being notified of the classification) to more clearly understand the plant condition of concem.
IES Utilities Resoonse The NUMARC Initiating Conditions's have been grouped with their applicable DAEC EAL's.
1
EAL Recognition Category A Abnormal Rad Levels / Radiological Effluent Issue No. 2 NUMARC Initiating Condition (IC) AUl states:
AU)
Any unplanned Release ofGaseous or Liquia Radioactivity to the Environment that Erceeds Two Times the Radiological Technical Specificationsfor 60 Minutes or Longer.
NUMARC EALs associated with this IC include:
1.
A valid reading on one or more ofthefollowing monitors that exceeds the "value shown " (site specific monitors) indicates that the release may have exceeded the above criterion and indicates the need to assess the release with (site specipcprocedure):
(site-specipe list) 2.
Conprmed sample analysesfor gaseous or liquid releases indicates concentrations or release rates with a release duration of 60 minutes or longer in excess oftwo times (site-specipc technical specipcations).
A.
An EAL corresponding the NUMARC Example EAL 2 was not provided. No justification for this deviation was provided. (This comment also applies to the corresponding Duane Arnold Alert level EAL AAl.)
B.
In the Duane Arnold basis document for this EAL it is stated that:
The Low Level Radwaste Processing and Storage Facility (LLRPSF) is not considered as an accident release point since the radiation monitor automatically trips the building exhaust at the Technical Specipcation instantaneous release limit thus terminating the release..
The NUMARC basis states that this IC " represents an uncontrolled situation and hence, a potential degradation in the level of safety." In formulating theJAL's I
for this IC, it should not be presumed that safety systems will operate as designed.
f In fact it is the misoperation of this equipment which will cause the IC to be met.
Therefore, the Duane Arnold EAL scheme should include EAL's for the j
moriitored release paths. (This comment also applies to the corresponding Duane Arnold Alert level EAL AAl.)
2
1 IES Utilities Resoonse EAL's have been added to meet the NUMARC condition of concern. The DAEC EAL's read " Confirmed sample analyses for gaseous or liquid releases indicates concentrations in excess of 2 times ODAM limits for greater than 60 minutes" for AUl and " Confirmed sample analyses for gaseous or liquid releases indicates concentrations in excess of 200 times ODAM limits for greater than 15 minutes" for AAl.
In order to meet the NUMARC condition of concem for issue 2 B, EAL's have been added which read " Valid LLRSPF (Kaman) rad monitor reading above 9 E-4 pCi/cc for more than 60 minutes" for AUI and " Valid LLRSPF (Kaman) rad monitor reading above 9 E-2 Ci/cc for more than 15 minutes" for AAl.
)
Issue No. 3 NUMARC Initiating Condition (IC) AAl states.
AA1 Any unplanned Release ofGaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times the Radiological Technical Specificationsfor 15 Minutes or Longer.
A NUMARC EAL associated with this IC is:
3.
A valid reading on perimeter radiation monitoring system greater than 10.0 mR/hr sustainedfor 15 minutes or longer.
The equivalent Duane Arnold EAL is:
Validfleld survey reading outside the.s,te Icundary above 10 mR/hr.
(Dose assessment is NOTavailable.)
A.
The addition of the condition " dose assessment NOT available"is not appropriate because exceeding the survey results, in and ofitself, is indicative of a loss of control of radioactive material which meets the IC. (This comment also applies to Duane Arnold EAL's ASI and AGl.)
B.
The Duane Arnold EAL did not include the condition " sustained for 15 minutes or longer." Nojustification was provided for this deviation. (This same copment also applies to Duane Arnold EAL's ASI and AGl.)
IES Utilities Resoonse The phrase " dose assessment NOT available" has been removed and the phrase " sustained for 15 minutes or longer" has been added to the EAL's under IC's AAl, AS1, and AGl.
3
r-____
Issue No. 4 NUMARC IC AA2 states:
AA2 A fajor Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering ofIrradiatedFuel Outside the Reactor vi.
NUMARC EAL's associated with this IC include:
2.
Report ofvisual observation ofirradiatedfuel uncovered.
?
Water levelless than (site-specific)feetfor the Reactor Refueling Cavity that will result in Irradiated Fuel Uncovering.
The Duane Arnold EAL scheme includes the following EAL's:
1.
Uncontrolled loss ofreactor cavity orfuelpool water level resvits in a spentfuel assembly that is NOTfully covered by water.
OR 2.
Valid Fuel Pool water level indication (LI-3414) below 13 feet 9 inches A.
The Duane Arnold EAL #1 does not provide for the method of detection of the plant condition as is provided for in NUMARC EAL #2, i.e. " Report of visual observation ofirradiated fi:el uncovered." This concern may be the result of the Duane Arnold EAL scheme not including EALs under ICs.
B.
The Duane Arnold F AL scheme does not include an EAL corresponding the NUMARC EAL #3. Nojustification was provided for this deviation.
IES Utilities Resconse An EAL has been added as identified in the NUMARC document. The EAL reads
" Report of visual observation ofirradiated fuel uncovered." The phrase " Uncontrolled loss of reactor cavity..." has been removed.
An EAL that reads " Water level reading below 450", as indicated on LI4541 (floodup) for the Reactor Refueling Cavity that will result in Uncovering Irradiated Fuel."Jas been added.
Issue No. 5 NUMARC IC AA3 states:
Release ofRadioactive Afat: rial or Increases in Radiation Levels Within the Facility that Impedes Operation ofSystems Required to Afaintain Safe Operation or to Establish or Afaintain Cold Shutdown 4
NUMARC EAL's associated with this IC include:
1.
Valid (site-specific) radiation monitor readings GREA TER THAN 15 mR/hr in areas requiring continuous occupancy to maintain plant safetyfimetions.
2.
Valid (site-specific) radiation monitor readings GREA TER THAN
< site specific > values in areas requiring infrequent access to maintainplant safetyfunctions.
(Site-specific) list The corresponding Duane Arnold EAL is:
Dose ratesprevent occupancy or access to areas regi f.a to achieve or maintain safe shutdown.
A.
The Duane Arnold EAL scheme did not include EAL's corresponding to these NUMARC EALs for this IC. The condition provided in the Duane Arnold scheme is closely related to the NUMARC IC but does not contain site-specific thresholds for classifying the event.
IES Utilities Resnonse Two EAL's have been added. They are " Valid area radiation monitor (RE9162) reading greater than 15 mR/hr in the Control Room" and " Valid area radiation monitor (RE9168) 1 reading greater than 500 mR/hr at the Remote Shutdown Panel IC388."
Issue No. 6 i
1 AS1 Boundary Dose Resultingfrom a:t Actual or Imminent Release of Gaseous Radioactivity Exceeds ; JO mR Whole Body or 500 mR Child Thyroidfor the Actual or Projected Duration ofthe Release.
NUMARC Example EAL's associated with this IC include:
1.
A validreading on one or more ofthefollowing monitors that exceeds or is expected the value shown indicates that the release may have exceeded the above criterion and indicates the yed to assess the release with (site-specificprocedure):
4.
Field survey results indicate site boundary dose rates exceeding 100 mR/hr expected to continuefor more than one hour; or analyses offield survey samples indicate child thyroid dose commitment of500 mRfor one hour ofinhalation.
5
~
The Duane Arnold EAL corresponding to NUMARC EAL #4 is:
4.
Validfield survey reading outside the site boundary above 100 mR/hr.
A.
The Duane Arnold EAL scheme did not include the NUMARC condition for the child thyroid dose commitment.
B.
The Duane Arnold EAL scheme includes EAL's corresponding to NUMARC EAL #1. The Duane Arnold EAL basis document states that; 'In order to calculate suitable radiation monitor values as described :n the generic methodology, use of an assumed source term mixture, use of annual average meteorology, and rounding offis required." Insufficient detail was provided to determine whether the " assumed source term" met the guidance for the source term in the NUMARC basis for this EAL.
IES Utilities Resnonse Dose assessment using MIDAS is based on the EPA-400 methodology, e.g., use of Total Effective Dose Equivalent (TEDE) and Committed Dose Equivalent (CDE) Thyroid.
TEDE is somewhat different from whole body dose received from gaseous effluents, as determined by ODAM methodology, which forms the basis for the radiation monitor readings calculated in AUI. Whole body dose from gaseous effluents is in accordance with the generic methodology.
The gaseous effluent radiation monitors can only detect noble gases. The contribution of iodine's to TEDE and CDE Thyroid could therefore only be determined either by: (1) utilizing MIDAS, or (2) gaseous effluent sampling. DAEC EAL #4 is written in terms of TEDE and CDE Thyroid. Including the child thyroid dose commitment terminology is not applicable in this case.
The source terms used have been pre-loaded into MIDAS and are the default mixes associated with a loss of coolant accident (LOCA) and a control rod drop (CRD) as described in the UFSAR, incorporating EPA-400 guidelines. DAEC has recalculated and rewritten the EAL's associated with ASI and AGI to reflect the revised source term. The revised EAL's associated with ASI and AGI are " Valid field survey reading outside the site boundary >100 mR/hr or above 500 mR/hr CDE Thyroid." and " Valid field survey reading outside the site boundary >1,000 mR/hr or >5,000 mR/hr CDE Thyroid",
respectively.
l NUMARC Recognition Category F Fission Product Barrier Degradation Issue No. 7 The NUMARC EAL methodology includes a fission product barrier matrix for i
determining whether or not a barrier (fuel clad, reactor coolant system, or 6
i
containment) is lost or potentially lost and for classifying events based on the combination oflost or potentially lost barriers. The fission product barrier matrix provides multiple indications to operators to assess the status of each of the barriers.
Classification of an event is made by determining the combination of barriers which have either been lost or potentially lost. The NUMARC guidance specifies that the following combination of barriers is indicative of a Site Area Emergency.
Loss ofBOTH Fuel Cladand RCS OR Potential Loss ofBOTH Fuel Clad and RCS OR Potential Loss ofEITHER Fuel Clad OR RCS, and Loss ofANY AdditionalBarrier A.
The Duane Arnold EAL scheme also contains a fission product barrier matrix.
However, the Duane Arnold EAL scheme defines the combination of barriers which is indicative of a Site Area Emergency differently than the NUMARC guidance. The combination of barriers specified in the Duane Arnold EAL scheme for the Site Area Emergency is:
Loss or Potential Loss ofAny Two Barriers The Duane Arnold EAL basis document explains that using this combination of barriers makes the classification easier to understand and that no sequences are significantly affected by the simplified logic. Insufficient detail was provided m i
the Duane Arnold EAL basis document to verify that the Duane Arnold EAL l
scheme meets the intent of the NUMARC guidance. The comparison table provided did not identify which EALs were being compared and did notjustify the adequacy of those combinations which would result in a Site Area Emergency j
classification in the Duane Arnold EAL which would not have resulted in a Site Area Emergency classification in the NUMARC guidance.
IES Utilities Resoonse The potential loss of the primary containment based on radiation / core damage and RPV level indicators can only occur if there is a loss of both the fuel clad and RCS bayiers.
This is true because the primary containment barrier potential loss value is higher than the corresponding values for the same ir% tors that indicate loss of both the RCS and fuel clad barriers. For the primary cco.onnsnt atmosphere indicators, potential loss indicators of either torus press ire e u psig or elevated hydrogen levels can only result from significant core damage that would result from a loss of both the RCS and the fuel clad. Thus, for these conditions, existence of the thresholds for containment potential loss could only lead to a General Emergency declaration. That only leaves EC/OSS judgment indicators, Primary Containment Atmosphere loss indicators and the Leakage indicators to be considered.
7
.A
s By their very nature, the EC/OSS Judgment indicators arejudgment calls and use of the NUMARC generic logic or the DAEC logic would make no difference. That leaves the Leakage indicators and the remaining Primary Containment Atmosphere indicators.
Since the primary containment barrier indicators are all " loss" indicators, the existence of at least a " potential lo.ss" of either the fuel clad or the RCS barriers will always result in a Site Area Emergency whether or not the NUMARC logic or the logic used at the DAEC and other plants is applied.
Issue No. 8 The NUMARC EAL for the loss and potential loss of the fuel clad barrier based i
on reactor vessel water level indications are:
- Loss:
i Level LESS THAN (site-specific) value Potential Loss:
Level LESS THAN (site-specific) value 8
.m
~-
The corresponding Duane Arnold EAL's are:
Loss:
RPV Level below -30 inches and cannot be restored Potential Loss:
RP V Level below -15 inches and cannot be restored A.
The Duane Arnold EAL basis document did notjustify the addition of"cannot be restored" to these EALs. It is not clear why the loss or potential loss cannot be based on the level alone. The addition of the condition "cannot be restored" may cause confusion and/or delay classification. (this same comment also applies to the Loss of Reactor Coolant System Barrier EAL based upon reactor vessel level.)
IES Utilities Resoonse The phrase "cannot be restored" will be removed. This will be applied to a!! applicable level statements contained on the Fission Barrier Table.
Issue No. 9 The Duane Arnold EAL scheme includes the following EAL:
Core damage assessment determines at least 5% fuel clad damage The Duane Arnold EAL basis document states:
It is intended that determination ofbarrier loss be made whenever the indicator threshold (for the containment monitor) is reached until such time that core damage assessment is performed, at which time direct use ofcontainment rad monitor readings is no longer required A.
The Duane Arnold EAL scheme did not include a statement corresponding to the statement in the Duane Arnold basis document regarding the use of the containment rad monitor EAL. This may cause confusion when classifying an event.
9 A
IES Utilities Resoonse The Fuel Clad Barrier statements have been rewritten as follows:
[
Fuel damage assessment (PASAP 7.2) determines at least 5% fud clad damage OR Fuel damage is indicated by any of the following:
[
Valid drywell rad monitor reading above 7E+2 R/hr OR
[
Valid torus rad monitor reading above 3E+1 R/hr OR L
Coolant activity above 300uCi/gm DOSE EQUIVALENT I-131 Issue No.10 The NUMARC EAL for the potential loss of the reactor coolant system barrier i
based on RCS leak rate indications includes the following conditions:
Unisolable primary system leakage outside the drywell as indicated by area temp or area radalarm The corresponding Duane Arnold EAL is:
Unisolable primary system leakage outside the drywell as indicated by ARMS or in-plant radiological surveys A.
The Duane Arnold EAL basis document did notjustify not including the condition "as indicated by area temp"in the Duane Arnold EAL. (This comment also applies to the same Duane Arnold EAL listed under the Loss of Containment Barrier column of the fission product barrier table.)
IES Utilities Resoonse The phrase "as indicated by area temperatures" was added to all applicable locations on the Fission Barrier table.
Issue No.11 The NUMARC EAL for the potential loss of the RCS barrier based on drywell pressure indications is:
Pressure Greater than (site-specific) psig 10 M
i i
The NUMARC basis for this EAL states:
The (site-specific) drywellpressure is based on the drywell high pressure alarm setpoint and indicates a LOCA. A higher value may be used if supporting documentation is provided which indicates the chosen value is less than the pressure which would be reachedfor a 50 gpm Reactor Coolant System leak.
The corresponding Duane Arnold EAL is:
Drywell cooling operating AND drywellpressure above 2 psig 1
The Duane Arnold EAL basis document states:
There is no sigmficant deviationfrom the generic indicator. The (site-specific) valuefor this loss indicator corresponds to the drywell high pressure ECCS initiation signal setpoint of2.0 psig.
A.
The Duane Arnold EAL basis does not address why the Duane Arnold EAL uses the ECCS initiation drywell pressure setpoint instead of the alarm setpoint as specified in the NUMARC guidance.
B.
It is not clear whether drywell cooling operation may be automatically isolated when drywell pressure exceeds 2 psig and whether this may cause confusion when classifying the event.
IES Utilities Resnonse The DAEC design is that of a GE Mark I Containment. During reactor operation, drywell pressure is maintained between 0.5 and 1.0 psig. The high pressure alarm setpoint of 1.5 psig was not selected, as this is too close to the normal operating pressure band, and could be exceeded for reasons other than a RCS leak. Analysis at the DAEC shows that a 50 gpm RCS leak would result in a 2 to 3 psig pressure rise over a six minute time period.
Since a 2 psig rise would be above the ECCS initiation setpoint, it is more practical to use the ECCS initiation setpoint of 2 psig for this EAL. Drywell cooling does not automatically isolate at 2 psig in the drywell, at the DAEC. Therefore, no confusion would result when classifying the event.
Issue No.12 The NUMARC EAL's for the loss of the Containment barrier based on drywell pressure indications are:
Rapid unexplained decreasefollowing initial increase or Drywellpressure response not consistent with LOCA conditions A.
The Duane Arnold EAL scheme did not include these EAL's and the Duane Arnold EAL basis document did notjustify this deviation.
I1
1 IES Utilities Resconse l
The NUMARC EAL's for Primary Containment Barrier Loss for Drywell Pressure have been added as follows:
Rapid unexplained decrease following initial increase OR Drywell pressure response not consistent with LOCA conditions.
The basis information to support the above conditions has been inserted into the DAEC basis document to support these EAL's.
Issue No.13 The NUMARC EAL for the potential loss of the containment barrier based on reactor pressure vessel water level is:
Reactor vessel water level LESS THAN (site-specific) value and the maximum core uncovery time limit is in the UNSAFE region The corresponding Duane Arnold EAL is:
RPV Level below -40 inches AND no injections source available A.
The Duane Arnold EAL does not appear to meet the intent of the NUMARC EAL. Two concerns have been identified with the Duane Arnold EAL. One is that the term "not available' has not been defined and may cause confusion when classifying the event. The second concern is that even if the injection source is available, if the water level was to remain below 40 inches for a given amount of time, the barrier should be considered potentially lost. As stated in the NUMARC EAL basis: "if emergency operating procedures have been ineffective in restoring reactor vessel level within the maximum core uncovery time limit, there is not a success path.". Whether or not the procedures will be effective should be apparent within the time provided. The Emergency Director should make the declaration as soon as it is determined that the procedures have been, or will be, ineffective."
IES Utilities Resoonse The phrase "..AND no injection source available" has been removed, as this is just a restatement of given information in EOP's. Also the Maximum Core Uncovery Time Limit (MCUTL) is addressd in Q&A Fission Product Barriers - BWR #10. Which states that this is the improper use of this chart considering the input assumptions for the curve.
12
._____..-___.__..-._._._...m NUMARC Recognition Category H Hn7neds and Other Conditions Affecting Phnt Safety
~ Issue No.14
' NUMARC IC HUl includes the following EAL:
3.
Assessment by the control room that an event has occurred.
A.
The Duane Arnold EAL scheme did not include an EAL corresponding to this EAL and no justification was provided for this deviation.
IES Utilities Resnonse l
An EAL stating " Assessment by the control room that an event has occurred" along with supporting basis information has been included.
l Issue No.15 NUMARC IC HUI includes the following EAL:
4.
Vehicle crash into plant structures or systems withinprotected area boundary.
The NUMARC Basis for this EAL explains that:
Automobiles, trucks, andforklifts are also vehicles within the context of this EAL. The key is whether or not the vehicle canpotentially cause sigmficant damage to plant structures.
The corresponding Duane Arnold EAL is:
1
\\
7.
Vessel or vehicle collision with structures or equipment required for safe shutdown The Duane Arnold basis document states that:
DAEC EAL 7 addresses vessel (aircraft) or vehicle (truck or train) crashes with structures or equipment requiredfor safe shutdown i
A.
The Duane Arnold EAL did not define the term " structures or equipment required i
for safe shutdown." It is not clear that users of the EAL procedure will be able to ascepain what are the structures or equipment required for safe shutdown. (This comment applies to the other EALs under IC HU1.)
l B.
The Duane Arnold basis document deviates from the NUMARC guidance by specifically not including automobiles and forklifts as vehicles for this EAL.
13
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IES Utilities Resoonse The table located adjacent to HU2 titled " Systems & Equipment of Concern" is intended to illustrate the equipment or structures required for safe shutdown.
" Automobiles" and " forklifts" have been added to the basis document.
Issue No.16 NUMARC IC HUI includes the following EAL:
5.
Report byplantpersonnel ofan unanticipated explosion within protected area boundary resulting in visible damage to permanent structure or equipment.
The corresponding Duane Arnold EAL is:
3.
Vis damage ofstructures or equipment requiredfor safe s,
swn 5.
Explosion within plant protected area A.
Duane Amold EAL #3 is not specific as to the cause of the damage which would result in the Unusual Event classification. It is not clear whether " damage" to equipment from maintenance errors or operational errors would be classified under this EAL.
B.
Duane Arnold EAL #5 does not include the NUMARC condition of"resulting in visible damage.. " Nojustification was provided for this deviation.
IES Utilities Resoonse
]
EAL's #3 and #5 were combined into a single EAL consistent with the NUMARC EAL which states " Report of an unanticipated explosion within the protected area boundary resulting in visible damage to permanent structures or equipment".
Issue No.17 NUMARC IC HUl includes the following EAL:
6.
Report ofturbinefailure resulting in casing penetration or damage to turbine or generator seals The corresponding Duane Arnold EAL is:
6.
Turbinefailure causing observable casing damage 14
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A.
Contrary to the NUMARC EAL, the Duane Arnold EAL did not include the j
condition " damage to turbine or generator seals."
j IES Utilities Resoonse j
" Report of turbine failure resulting in casing penetration or damage to turbine or generator seals"has been added to the DAEC EAL's.
I; 8
(
Issue No.18 1
i l
NUMARC IC HU2 includes the following EAL:
i 1.
Fire in buildings or area contiguous to any ofthefollowing (site-specylc) areas not extinguished within 15 minutes ofcontrol room notylcation or veryication ofa comrolroom alarm:
1 (Site-specylc) list i
j The corresponding Duane Arnold EAL is:
I I
1.
Fire within safe shutdown area NOT extinguished within 15 minutes ojcontrolroom notylcation or verylcation ofcontrol
)
room alarm.
I i
j A.
The NUMARC EAL specifies " buildings or areas contiguous to...." The
)
corresponding Duane Arnold EAL limits the areas considered to only " safe shutdown area (s)." This same list of areas is included in the related Alert level EAL. The areas applicable under the Unusual Event EAL is broader than the -
areas applicable under the Alert level EAL.
IES Utilities Resoonse The revised DAEC EAL reads " Fire in buildings or areas contiguous to any of the follow' g areas not extinguished within 15 minutes of control room notification or m
verification of a control room alarm."
Reactor, Turbine, Control, Administrative / Security buildings I
Pump house Intake structure i
The basis information has been updated to identify the above facilities.
i 15
1 Issue No.19 NUMARC IC HU3 includes the following EAL:
2.
Report by local, county or State officialfor potential evacuation of site personnel based on offsite event The corresponding Duane Arnold EAL is:
2.
Notification ofnear site release that may require evacuation.
A.
The term "near site" is not defined in the Duane Arnold EAL. In addition, it is not clear that including this term is necessary for the Duane Arnold EAL to meet the intent of this NUMARC EAL.
IES Utilities Resoonse l
The DAEC EAL has been reworded as follows: "Regiort by local, county or State official for potential evacuation of site personnel based on an offsite event".
l hsue No. 20 l
l NUMARC IC HU4 includes the following EAL:
2.
Other security events as determinedfrom (site-specific) Safeguards Contingency Plan.
The corresponding Duane Amold EAL is:
1.
Suspected sabotage device discovered in plant switchyard.
The Duane Amold EAL basis document states:
Other (site-specific) security events ofconcern at DAEC include discovery ofa suspected sabotage device in the plant switchyard, which is located outside the protected area.
A.
It is not clear from the information provided whether Duane Arnold EAL #2 includes all the applicable security events in the Duane Arnold Safeguards Contingency Plan.
IES Utilities Resconse The information provided is complete and consistent with the DAEC Security Contingency Plan, Revision 33. Due to the nature of the safeguards information contained in that plan it is exempt from public disclosure pursuant to 10CFR73.21.
16
- -_ - ~.-. -.- -. -
Issue No. 21 4
NUMARC IC HAl includes the following EAL:
i 3.
Report ofany visible ' structural damage on any ofthefollowing plant i
structures:
Reactor Building
[
Intake Building Ultimate Heat Sink h
Refueling Water Storage Tank Diesel Generator Building
{
Turbine Building
\\
Condensate Storage Tank ControlRooms Other (Site-Specific) Structures a
i A.
The Duane Arnold EAL scheme did not include an EAL corresponding to this
. EAL and did notjustify this deviation.
j-1 i.
1 IES Utilities Response j
~
The DAEC EAL corresponding to NUMARC IC HAl EAL #3 is; " Report to control j
room of damage affecting safe shutdown areas". The DAEC basis document addresses j
this in Section H par 9. The DAEC basis document also identifies all locations l
applicable to this EAL m the table labeled " Safe Shutdown Areas".
I i
Issue No. 22 j
i NUMARC IC HAl includes the following EAL:
)
-l 3.
Vehicle crash affectingplant vital areas
}
The corresponding Duane Arnold EAL is:
6.
Vessel or vehicle collision affecting ability to achieve or maintain i
The Duane Arnold EAL deviates from the NUMARC EAL by includingihe condition that the collision affects the ability to achieve or maintain safe j
shutdown. Nojustification was provided for this deviation. It may be difficult to make a definitive determination whether the vehicle collision did affect the ability i
to achieve or maintain safe shutdown. It is not appropriate to delay classification in order to make this determination. (This comment also applies to Duane Arnold
,l -
EAL HA1, #5) 4 17
~,.
.-_._,..m
IES Utilities Responsg HAl is intended to be a " confirmed" collision affecting a plant safe shutdown area (vital area). The EAL has been rewritten as follows: " Vehicle crash affecting plant vital areas".
Issue No. 23 NUMARC IC HA4 includes the following EAL:
2.
Other security events as determinedfrom (site-specipc) Safeguards Contingency Plan.
A.
The Duane Arnold EAL scheme did not include an EAL corresponding to this EAL. The Duane Arnold EAL basis document states that: " Based on information provided by DAEC Security, generic EAL 2 is unnecessary at DAEC." It is not clear what, if any, security events were considered in making this determination.
This comment also applies to the corresponding Site Area Emergency IC HSI.
For HS1 it appears that a sabotage device discovered in the plant vital are as should be included as an EAL.
IES Utilities Resoonse
" Sabotage device discovered in the plant protected area" as an EAL under HA4, has been added. The EAL " Sabotage device discovered in the plant vital area" has also been added under H5.'.
Issue No. 24 NUMARC IC HA5 includes the following EAL:
1.
Entry into (site-specipc) procedurefor control room evacuation.
The corresponding Duane Arnold EAL is:
Control Room evacuation procedures have been initiated Contrary to the NUMARC guidance, the specific Duane Arnold proceduIe for A.
control room evacuation is not identified in the EAL.
IES Utilities Resconse A reference to Abnormal Operating Procedure 915," Shutdown Outside Control Room",
was incorporated into this EAL.
18
Issue No. 25 NUMARC IC HS2 includes the following EAL:
1.
Thefollowing conditions exist:
a.
Control room evacuation has been initiated AND 1
b.
Control ofthe plant cannot be establishedper (site-specific) procedure within (site-specific) minutes.
The corresponding Duane Arnold EAL is:
1 Control room has been evacuated AND control ofplantfrom Remote j
Shutdown Panel 1C388 NOT established within 20 minutes.
\\
The basis for the NUMARC EAL states:
(Site-specific) timefor transfer based on analysis or assessments as to how quickly control must be reestablished without core uncovering and/or core damage. This time should not exceed 15 minutes. (emphasis added) j The Duane Arnold basis document states:
Operator control within 20 minutes wou d not impact the integrity ofthe l
fuel clad, the reactorpressure vessel, and the primary containment.
A.
The Duane Amold EAL basis did notjustify why the time limit to classify this event should exceed 15 minutes. For instance, the Duane Arnold basis did not describe why more than 15 minutes is needed for determining whether control is established at the remote shutdown panel IES Utilities Resnonse i
Support has been added to the DAEC basis document explaining why the basis for 20 minutes vs.15 minutes, as recommended by the generic document. DAEC physically has satellite panels associated with the remote shutdown panel at various locations within the plant. General Electric Report MDE-44-0386," Safe Shutdown Appendix R Ana,1ysis for i
DAEC", March 83, identifies that for the DAEC, it is not possible to reposition all the required switches at all the remote locations in less than 15 minutes Therefor: DAEC has allowed 20 minutes to line up the remote shutdown panel.
19 i
Issue No. 26 NUMARC IC HG1 contains the following EAL's:
1.
Loss ofphysical control ofthe control room due to security event.
OR 2.
Loss ofphysical control ofthe remote shutdown capability due to security event.
The corresponding Duane Arnold EAL's are:
1.
Loss ofphysical control ofthe control room OR 2.
Loss ofphysical control ofremote shutdown capability A.
Contrary to the ' b 34 ARC guidance, the Duane Arnold EAL's do not include the condition "due to security event." Therefore an event where the control room must be evacuated for reasons other than due to a security event may erroneously be classified under this EAL.
IES Utilities Resnonse Due to the layout of the DAEC EAL table, the words,"...due to security event" is unnecessary because the associated EAL's are identified within the " Security" event type.
This is consistant with current plant practice and consistent with the NUREG - 0654 based EAL tables, currently in use at the DAEC.
NUMARC Recognition Category S System Malfunction hsue No. 27 NUMARC IC EAL SUI contains the following EAL's:
1.
Thefollowing conditions exist:
Loss ofpower to (site-specylc) transformersfor greater a.
than 15 minutes AND b.
At lease (site-specific) emergency generators are supplying power to emergency busses.
20
The corresponding Duane Arnold EAL is:
Loss ofofsite power lasting more than 15 minutes A.
Contrary to the NUMARC guidance, the Duane Arnold EAL does not identify site-specific transformers, loss of power to which constitutes " loss of all offsite power".
IES Utilities
" Loss of Offsite Power" is consistent with the DAEC Operation's Department terminology for the conditions of SUI. The NUMARC conditions as outlined in the NUMARC example EAL's would be less clear to the operators at the DAEC.
1 4
Issue No. 29 NUMARC IC SU3 includes the following EAL:
1.
Thefollowing conditions exist:
a.
Loss ofmost or all (site-specific) annunciators associated with safety systemsfor greater than 15 minutes.
AND b.
Compensatory non-alarming indications are available AND c.
In the opinion ofthe Shift Supervisor, the loss ofthe annunciators or indicators requires increased surveillance to safely operate the unit (s)
AND d
Annunciator or indicator loss does not resultfrom planned action.
The corresponding Duane Arnold EAL is:
Unplanned Loss ofannunciators on panels 1C03,1C04, and 1C05 lasting more than 15 minutes AND compensatory non-alarming indications are available.
A.
The Duane Arnold EAL is inconsistent with the NUMARC guidance in that it specifies loss of all annunciators. The Duane Arnold EAL basis document states that the annunciators share a common power supply and therefore it is not necessary to include the condition of"most annunicators." It is not clear that there is no event which could result in a loss of most annunciators and no reason was given for why a loss of most annunciators would not :neet the intent of the NUMARC guidance.
21
IES Utilities Resconse DAEC rewrote the EAL as follows:
Unplanned loss of most annunciators on Panels IC03, IC04, and IC05 lasting more than 15 minutes.
AND Compensatory non-alarming indications are available.
Issue No. 29 NUMARC IC SU4 includes the following EAL's:
1.
(Site-specific) radiation monitor readings indicatingfuel clad degradation greater than technical specification limits.
2.
(Site-specific) coolant sample activity value indicatingfuel clad degradation greater than technical specification limits.
The corresponding Duane Amold EAL's are:
1.
Valid Pretreat RM-4104 rad monitor reading above 4E+3 mR/hr 2.
Coolant activity above 1.2 pCi/ml DOSE EQUIVALENTI-131 A.
The Duane Amold basis document describes the technical specification basis used for developing these EAL's. It is not clear why technical specification 3.6.b.2 was used for the basis for Duane Amold EAL #1 whereas technical specification 3.6.b.1 was used for the basis for Duane Amold EAL #2.
IES Utilities Resoonse The DAEC basis document has veen revised to reflect that both DAEC EAL's are based on Technical Specification 3.6.B.I.a.
22 i
l Issue No. 30 NUMARC IC SU5 contains the following EAL:
1.
Thefollowing conditions exist:
a Unidentified orpressure boundary leakage greater than 10 epm OR
}
b.
Identyled leakage greater than 25 gpm i
l The corresponding Duane Arnold EAL is:
Unidentyledleakage above 10 GPAf OR Total RCSleakage above 35 GPAf OR Alain steam line break as determinedfrom annunciators orplant personnelreport A.
The Duane Arnold EAL is not consistent with the NUMARC guidance in that it does not specify a value for pressure boundary leakage. No justification was provided for this deviation.
B.
The Duane Arnold EAL is not consistent with the NUMARC guidance in that it
[
specifies 35 gpm for the total RCS leakage instead of 25 gpm as is specified in the NUMARC guidance. The NUMARC guidance states that this IC is included as an Unusual Event because it may be a precursor of more serious conditions. The Duane Arnold basis document does not address why a 25 gpm total RCS is not indicative of a potential degradation of the level of safety at Duane Arnold and therefore is not an Unusual Event.
IES Utilities Resnonse This EAL has been revised to be consistent with the NUMARC guidance. The EAL now reads:
Unidentified or pressure boundary leakage greater than 10 gpm.
OR Identified leakage greater than 25 gpm 23
1 Issue No. 31 NUMARC IC SU6 contains the following EAL's:
1.
Loss ofall (site-specific list) onsite communication capability affecting the ability to perform routine operations.
2.
Loss ofall (site-specific) offsite communications capability The corresponding Duane Amold EAL's are:
1.
Loss ofALL onsite electronic communication niethods 2.
Loss ofALL electronic communication methods with government agencies A.
Contrary to the NUMARC guidance, a site-specific list of communication capabilities was not included in these EAL's. The concern with this deviation is that the user of the classification procedure may not be readily able to ascertain whether the EAL's are met or not because of the lack of site-specific information.
d IES Utilities Resconse Consistent with the NUMARC basis information, the EAL's have been revised to state:
1)
Loss of all onsite telephone and radio communication methods (PABX, direct-ring, UHF, and radiological survey radio systems.)
2)
Loss of ALL electronic communication methods with government agencies (PABX, direct-ring, ENS, microwave and police radio),
i 1
Issue No. 32 NUMARC IC SU7 contains the following EAL:
1 1.
Either ofthefollowing conditions exist:
Unplanned Loss of Vital DCpower to required D'C busses a.
based on (site-specific) bus voltage indications.
AND
\\
b.
Failure to restore power to at least one required DC bus within 15 minutesfrom the time ofloss.
l The corresponding Duane Arnold EAL is:
Complete Loss of125 VDC lasting more than 15 minutes 24
A.
Contrary to the NUMARC guidance, the Duane Arnold EAL did not specify the a7plicable vital buses in this EAL and did not specify what voltage level
.:onstitutes a loss of DC. The concern with this deviation is that classification may be delayed or an event improperly classified due to the lack of specific information. (This comment also applies to IC SS3) i IES Utilities Resoonse l
The DAEC EAL for SU7 has been revised to read "The following conditions exist:
Unplanned Loss of Div 1 and Div 2125 VDC busses based on bus voltage less j
than 105 VDC indicated.
AND Failure to restore power to at least one required 125 VDC bus within 15 minutes from time ofloss.
The basis has been revised as follows: There is no significant deviation from the generic EAL, Unplannedloss of Div 1 and Div 2125 VDC busses excludes scheduled maintenance and testing activities. Under the conditions of concern, AOP 302.1," Loss of 125 VDC Power", would be entered. The DAEC.EAL's address the loss of both divisions of the 125 VDC systems consistent with AOP 302.1.
The 125 VDC system is divided into two intpendent divisions - Division I (1DI) and Division II (ID2)- each supplied by separate AC and DC (battery) power supplies. Loss of both 125 VDC Divisions could compromise the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations. These EAL's are intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss. If this loss results in i
the inability to maintain cold shutdown, the escalation to an Alert will be per SA3 "RCS temperature rise that is not allowed by procedures that will result in RCS temperature above 212 F".
Bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipm:nt and may be indicated by the illumination of annunciators "125 VDC System 1 Trouble" on 1C08A A-9 and/or "125 VDC System 2 Trouble" on IC08B A-4.
The EAL for SS3 has been rewritten as follows " Unplanned Loss of Div I and Div II 125 VDC Busses Lasting More Than 15 Minutes".
25
Issue No. 33 NUMARC IC SA3 contains the following EAL:
1.
Thefollowing conditions exist:
a.
Loss of(site-specipc) technical specipcation required functions to maintain coldshutdown.
AND b.
Temperature increase that either:
Exceeds technical specipcation coldshutdown temperature limit OR Results in uncontrolled temperature rise approaching cold shutdown technical specipcation limit.
The corresponding Duane Amold EAL is:
RCS temperature rise that is not allowed byprocedures or Tech Specs that will result in RCS temperature above 212 F.
)
A.
The Duane Arnold EAL deviates from the NUMARC EAL by including the condition that the temperature rise is "not allowed by procedures or Tech Specs" i
rather than "the loss of tech spec functions." The concern is that the conditions specified in the Duane Arnold EAL will make classifying events more difficult and that some events classified under the NUMARC EAL scheme may not be classified under the Duane Amold EAL scheme, i
B.
The Duane Amold EAL deviates from the NUMARC guidance by not including the condition of" uncontrolled temperature rise approaching cold shutdown technical specification limit." This may result in delaying classifications. This deviation was not justified in the Duane Amold EAL basis document.
IES Utilities Resnonse The revised wording of this EAL is:
Loss of decay heat removal systems required to maintain cold shutdown AND Temperature rise that exceeds 212'F OR Uncontrolled temperature rise approaching 212'F 26
The loss of monitoring and removal of decay heat during shutdown conditions is currently governed by DAEC's procedure AOP 149," Loss of Decay Heat Removal."
The DAEC EAL is written to imply a RCS temperature rise above 212 F that is not allowed under plant procedures. This corresponds to the inabiliiy to maintain required temperature conditions for Cold Shutdown. " Uncontrolled" means that system temperature increase is not the result of planned actions by the plant staff. Minor cooling interruptions occurring at the transition between Hot Shutdown and Cold Shutdown or temperature changes that are permitted to occur during the establishment of alternate core cooling would not require an unnecessary declaration of an Alert.
Issue No. 34 i
NUMARC IC SS2 includes the following EAL:
i I
(Site-specific) indications exist that automatic and manual scram were not successfid The corresponding Duane Arnold EAL is:
All control rods NOTinserted to at leastposition 02 AND boron injection with SBLCis required.
A.
The Duane Arnold EAL deviates by including the condition that " boron injection with SBLC is required." This condition may result in delaying classification. If the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed then conditions exist that lead to imminent loss or potential loss of both fuel clad and the RCS and therefore a Site Area Emergency classification is warranted. It is not appropriate to wait until boron injection is procedurally mandated to classify the event.
IES Utilities Resoonse This EAL has been reworded as follows:
Failure of automatic and manual scram AND Power remains above 5%
OR Boron injection is required.
This change addresses the issue where an automatic and manual scram are not considered successful if actions away from the reactor control console are required to scram the reactor. Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed. These conditions could lead to imminent loss or potential loss of both fuel clad and RCS.
27
t Issue No. 35 NUMARC IC SS4 states:
Complete loss ofFunction Needed to Achieve or Maintain Hot Shutdown NUMARC EAL's associated with this IC include:
l 1
1.
Complete loss ofany (site-specific)Jimetion requiredfor hot shutdown
)
The corresponding Duane Arnold EAL is:
i Adequate core cooling conditions CANNOT be achieved or maintained OR Reactor CANNOT be brought subcritical A.
The Duane Arnold EAL does not include plant specific indication for determining whether adequate core cooling conditions exist. This could make this EAL difficult to use.
IES Utilities Response This EAL has been reworded as follows:
EOP Graph 4, Heat Capacity Limit is exceeded j
OR Reactor CANNOT be brought suberitical This EAL addresses complete loss of functions, including ultimate heat sink and reactivity control, required for hot shutdown with the reactor at pressure and temperature.
Under these conditions, there is an actual major failure of a system intended for the protection of the public. If the main condenser is unavailable and the Torus is threatened, this would be a pbnt condition that would correspond to an actual major failure of a system intended for the protection of the public. This condition would also impact
" adequate core cooling" conditions. The reactivity condition criteria is addressed by maintenance of required shutdown margin.
)
28 1
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Issue No. 36 NUMARC IC SGI contains the following EAL:
1, Prolongedloss ofall offsite and onsite ACpower as indicated by:
Loss ofpower to (site-specipc) transformers.
a.
AND b.
Failure of(site-specipc) emergency diesel generators to supplypower to emergency busses.
i AND c.
At least one ofthefollowing conditions exists:
Restoration ofat least one emergency bus within (site-specipc) hours is NOTlikely OR (Site-specipc) indication ofcontinuing degradation h
ofcore cooling based on Fission Product Barrier l
monitoring.
The corresponding Duane Arnold EAL is:
Loss of Voltage on Buses 1A3 and 1A4 and ANYofthefollowing Restoration ofpower to either Bus 1A3 or IA4 is NOT likely within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> RPVlevelremains indeterminate RPYlevelremains below -30 inches 1
A.
The terms " remains indeterminate" and " remains below -30 inches" are not defined in the Duane Arnold EAL. Using undefined terms such as these may result in confusion when classifying an event. In addition,if a station blackout condition occurred and water level reached the top of active fuel, plant cof tions di warrant classifying the event as a General Emergency without waiting to determine if the level is going to " remain" less than top of active fuel.
i IES Utilities Resnonse i
j
" Loss of Voltage on Buses I A3 and 1 A4" addresses the NUMARC EAL statements of
" Loss of Power to (site-specific) transformers AND failure of(site-specific) emergency i
diesel generators to supply power to emergency busses." " Restoration of power to either Bus l A3 or 1 A4 is NOT likely within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />" is synonymous with " Restoration of at 29 I
r F
least one emergency bus within (site-specific) hours is NOT likely". The DAEC removed the word " remains" from the EAL table under SGl. The EAL wording has bee'n revised i
to read, "RPV level indeterminate, RPV level below 15 inches. This wording is synonymous with "(site-specific)" indication of continuing degradation of core cooling l
based on Fission Product Barrier monitoring. DAEC has stated the EAL's in terms that are familiar to our operators by extracting those words directly from our EOP's.
The EOP basis document for DAEC identifies that adequate core cooling is assured for-the DAEC at a level of 15 inches.
Issue No. 37 i
NUMARC IC SG2 contains the following EAL's:
1.
(Site-specific) indications exist that automatic and manual scram were not successful AND 2.
Either ofthefollowing:
I (Site-specific) indications exist that the core cooling is a.
extremely challenged
?
OR b.
(site-specific) indications exist that heat removal is extremely challenged.
The corresponding Duane Arnold EAL is:
1 Entry into A TWS EOP-RP Y Control is required and BOTH ofthe following:
Reactorpower is expected to remain above 5% or CANNOT be I
determined l
AND Main condenser is NOTavailable A.
It is not clear that the condition of the main condenser not being available,is a sufficient indication of an extreme challenge to heat removal. The NUMARC EAL guidance state that "For BWRs (site-specific) considerations include inability to remove heat via the main condenser, or via the suppression pool or torus (e.g. due to high pool water temperature). The Duane Arnold EAL did not include indications regarding heat removal via the suppression pool.
30
i 4
- No condition equivalent to the NUMARC condition "(Site-specific) indications B.
o-exist that the core cooling is extremely challenged" was provided in the Duane Arnold EAL. The NUMARC guidance states,"For BWRs, the extreme challenge of the ability of cool the core is intended to mean that the reactor vessel water level is below 2/3 coverage of active fuel." The Duane Arnold EAL did not include a comparable EAL for this condition."
C.
Contrary to the NUMARC guidance the Duane Arnold EAL includes the
~
condition " Reactor power is expected to remain above 5% or CANNOT be determined." Furtherjustification is needed to determine whether tne addition of this condition meets the intent of the NUMARC EAL. In addition, the term "is l
expected to remain above 5%"is not defined in the Duane Arnold EAL.
i l
IES Utilities Resggmig l
This EAL has been reworded as follows:
i i
Entry into ATWS EOP-RPV Control is required AND RPV level cannot be maintained above -30 inches OR l
EOP Graph 4 Heat Capacity Limit is exceeded.
The NUMARC condition, (Site-specific) indications exist that automatic and manual j
scram were not successful is addressed by " Entry into ATWS EOP-RPV Control is required."
4 j
The NUMARC condition, (Site-specific) indications exist that the core cooling is j
extremely challenged is addressed by "RPV level cannot be maintained above -30 inches."
The NUMARC condition, (Site-specific) indications exist that heat removal is extremely challenged is addressed by "EOP Graph 4 Heat Capacity Limit is exceeded" I
i i
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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC revieu)
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PAGE i of iv TABLE OF CONTENTS 4
EFFECTIVE DATE: TBD i
TABLE OF CONTENTS INTRODUCTION................
.................1-1 DEFINITIONS.................
.... D-1 ORGANIZATION OF BASIS INFORMATION..........
.......................O-1 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT CATEGORY l
,A, AUI Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment That Exceeds Two Times the
\\.j Radiological Technical Specifications Foc 60 Minutes or Longer..
.. A-1 AU2 Unexpected increase in Plant Radiation..
..A-5 AAl Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times Radiological Technical Specifications for 15 Minutes or Longer..
..A-8 AA2 Major Damage to Irradiated Fuel or Loss of Water Level that lias or Will Result in the Uncovering ofIrradiated Fuel Outside the Reactor Vessel.
.. A-12 AA3 Release of Radioactive Material or increases in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or to Maintain Cold Shutdown.
.. A-15 ASI Site Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mrem TEDE or 500 mrem CDE Thyroid for the Actual or Projected Duration of the Release..
.. A-17 AGI Site Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity that Exceeds 1,000 mrem TEDE or 5,000 mrem CDE Thyroid for the Actual or Projected Duration of the Release..
.. A-21 FISSION PRODUCT BARRIER DEGRADATION CATEGORY FUI Any Loss or Any Potential Loss of Primary Containment Barrier..
. F-1 FA1 Any Loss or Any Potential Loss of Either Fuel Clad Or RCS Barrier..
.F-2 O
FSI Loss Or Potential Loss of Any Two Barriers..
.F-3 FGI Loss of Any Two Barriers AND Potential Loss of the Third Barrier.
.F-5
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
(
)
x.s PAGE ii of iv TABLE OF CONTENTS EFFECTIVE DATE: TBD FISSION PRODUCT BAIUUER DEGRADATION CATEGORY (continued)
FUEL CLAD BARRIER INDICATORS..
.F-6 RCS BARRIER INDICATORS.,
.F-12 PRIMARY CONTAINMENT BARRIER INDICATORS.
. F-21 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY CATEGORY 11U1 Natural and Destructive Phenomena Affecting the Protected Area.
..H-1 11U2 Fire Within Safe Shutdown Areas Not Extinguished Within 15 Minutes of Detection.
.11-4 11U3 Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant.,
.11-5 t
i V
ilU4 Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant..
. 11-6 11U5 Other Conditions Existing Which in the Judgment of the EC/OSS Warrant Declaration of an Unusual Event...11-7 ilAI Natural and Nstructive Phenomena AfTecting the Plant Vital Area.
. 11-8 IIA 2 Fire Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown.
. 11-1 2 11A3 Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown..
. 11-1 5 IIA 4 Security Event in a Plant Protected Area..
. 11-1 7 IIA 5 Control Room Evacuation lias Been Initiated.
. 11-1 8 IIA 6 Other Conditions Existing Which in the Judgment of the EC/OSS Warrant Declaration of an Alert..
.1-1-19 IIS1 Security Event in a Plant Vital Area.
. 11-2 0 IIS2 Control Room Evacuation lias Been Initiated and Plant Control Cannot Be Established..
. 11-2 1 l
1153 Other Conditions Existing Which in the Judgment of the EC/OSS Warrant Declaration of a Site Area Emergency 11-2 3 11G1 Security Event Resulting in Loss Of Ability to Reach and Maintain Cold Shutdown.
. 11-2 4 A
IIG2 Other Conditions Existing Which in the Judgment of the EC/OSS Warrant Declaration of a General Emergency 11-2 5
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Duare Arnold Energy Center EMURGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview) i
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TABLE OF CONTENTS i
EFFECTIVE DATE:TBD
(
SYSTEM MALFUNCTION CATEGORY
[
SUI Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes...
..S1 i
{
SU2 Inability to Reach Required Shutdown Within Technical Specification Limits.......
..S-2
).
SU3 Unplanned Loss of All Safety System Annunciation or Indication in the Control Room for Greater Than 15 Minutes
{
... S-3 l
i SU4 Fuel Clad Degradatior
.... S-5 i
i SUS RCS Leakage..
... S-8 SU6 Unplanned Loss of All Onsite or Offsite Communications Capabilities.
.S.10 i
SU7 Unplanned Loss of Required DC Power During Cold Shutdown or Refuel Mode For Greater Than 15 Minutes... S-12 l
SA1 Loss of All OiTsite Power and Loss of All Onsite AC Power to Essential Busses During Cold Conditions....
.S-14 SA2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint lias Been Exceeded and Manual Scram Was Successful...
.S-15 SA3 Inability to Maintain Plant in Cold Shutdown...-...
S-17 t
i SA4 Unplanned Loss of Most or All Safety System Annunciation or indicction in Control Room With Either (1) a.
Significant Transient in Progress, or (2) Compensatory Non-Alarming Indicators are Unavailable...
...S-19 SAS AC Power Capability to Essential Busses Reduced to a Single Power Source for Greater Than 15 Minutes Such That Any Additional Single Failure Would Result in Station Blackout..
.5-21 SSI Loss of All Offsite Power and Loss of All Gas!te AC Power to Essential Dusses..
.S-22 SS2 Failure of Reactor Protection System Instrumentation to Complete or Initiate on Automatic Reactor Scram Once a Reactor Protection System Setpoint lias Been Exceeded and Manual Scram Was NOT Successful..
..S-23 SS3 Loss of All Vital DC Power..
.. S-24 SS4 Complete Loss of Function Needed to Achieve or Mair.tain 110t Shutdown.
.S-25 i
SS$ Loss of Water Level in the Reactor Vessel That lias or Will Uncover Fuel in the Reactor Vessel.
.S-27 SS6 Inability to Monitor a Significant Transient in Progress..
.S-28
.. S-30 SG1 Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power...
Duane Arnold Energy Center l
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f SYSTEM MALFUNCTION CATEGORY (continued) i 4
i SG2 Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core =
.S-32
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i Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC revieu) qV PAGE I-l of 2 INTRODUCTION EFFECTIVE DATE: TBD l
l INTRODUCTION IES Utilities has revised the Duane Arnold Energy Center (DAEC) Emergency Plan to incorporate guidance from NUMARC/NESP-007, Revision 2 (January 1992), Methodology for Development of Emergency <fcrion Levels. The NUMARC (now Nuclear Energy Institute or NEI) guidance was developed to replace Emergency Action Levels (EAL) guidance contained in NUREG-0654/ FEMA-REP-1 (Revision i
l), Criteriafor Preparation and Evaluation ofRadiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, November 1980. The NEI-sponsored methodology was used to develop a set of generic EAL guidelines, together with the basis, so that they could be used and adapted by each utility in a consistent manner. The NRC has endorsed use of the NEI generic guidance as an acceptable alternative method to NUREG-0654 for developing plant-specific EALs in Regulatory Guide 1.101, " Emergency Planning and Preparedness for Nuclear Power Reactors," Revision 3, August 1992.
i O This Regulatory Guide further states that: " Licensees may use either NUREG-0654/ FEMA-REP-1 or V NUMARC/NESP-007 in developing their EAL scheme but may not use portions of both methodologies."
This EAL basis document was developed to: (1) provide clear documentation of how NEI generic guidance was applied in the development of DAEC upgraded EALs, (2) provide justification of any exceptions or additions to NEI generic guidance as it is applied to DAEC, and (3) facilitate the regulatory approval of the upgraded EALs that is required under 10 CFR 50 Appendix E.
Although there are many similarities, there are some basic differences from the previous EALs based on NUIEG-0654 guidance. These include:
- 1. Events that are explicitly covered under 10 CFR 50.72 such as one-hour or four-hour reports are no longer classified under the Unusual Event emergency classification. Items such as contaminated injured person transported off-site, partial communications losses, meteorological measurement losses, shutdown within the equirements of technical specifications, and inadvertent actuation of ECCS are no longer treated as emergencies because they are explicitly defined in 10 CFR 50.72 as "non-emergency" conditions to report.
- 2. Precursor conditions are explicitly included in the Unusual Event emergency classification. This includes EALs addressing RCS leakage and loss of off-site power.
O v
-... - ~ - - -. -.. - -
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- 3. Conditions such as fire, explosion, gas releases, flooding, low river water level, tornado, or earthquake i
i can be directly escalated only up to the Alert emergency classification. Escalation to Site Area
{
_ Emergency or General Emergency is based on degraded system response as would be determined by fission product barrier, loss of AC power, or projected efiluent release EALs.
j
- 4. Core damage sequences are addressed by determining their level of challenge to each of the three l
primary tission product barriers - fuel clad, reactor coolant system, and the primary containment. The i
level of challenge is determined in accordance with the Emergency Operating Procedures (EOPs),
[
Integrated Plant Operating Instructions (IPOls), Abnormal Operating Procedures (AOPs) and core 1
damage assessment methodology.
This allows the operations crew to readily recognize the j
corresponding emergency classification and allows for ready escalation to Site Area Emergency or General Emergency as conditions may worsen.
i
- 5. Offsite radiological releases that can be expected to exceed Environmental Protection Agency (EPA)
,'s Protective Action Guide (PAG) levels for inhalation doses - 1,000 mrem TEDE or 5,000 mrem CDE.
~
Thyroid - will result in declaration of a General Emergency.
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Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevicu)
PAGE D-1 of 10 E DAH: ED DEFINITIONS DEFINITIONS AC-Altemating Current AffL>cting (in regard to events such as fire, flood, or missiles) - Causing degraded equipment performance as determined by physical observation or by indications in the Control Room or at local control stations.
Alert - Events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guide (PAG) exposure levels.
All - Initiating Condition applies to all Technical Specification operating modes as well as defueled operation.
I b AOP - Abnormal Operating Procedure APRAI-Average Power Range Monitor ARAI-Area Radiation Monitor ATWS-Anticipated Transient Without Scram i
Barrier - Same as " Fission Product Barrier", below.
Barrier Aftmitoring Ability - This is a judgment factor in determining whether a fission product barrier is lost or potentially lost. Decreased ability to monitor a barrier results from a loss of/ lack of reliable indicators, including instrumentation operability concems, readings from portable instrumentation, and consideration for offsite monitoring results.
Becquerel-A measurement of radioactive decay rate equal to one disintegration per second.
BOP - Balance of Plant BWR - Boiling Water Reactor O()
CAAI-Continuous Air Monitor CDE-Committed Dose Equivalent es defined in 10 CFR 20.1003
.. - - - ~..
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Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
O PAGE D-2 of 10 EFFECTIVE DATE: TBD i
DEFINITIONS i
i i
i CEDE - Committed Effective Dose Equivalent as defined in 10 CFR 20.1003 4
CFM-Cubic Feet per Minute 1
2 CFS-Cubic Feet per Second i
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i Cold condition - As defined in Technical Specification 1.0, this refers to the condition where the reactor coolant temperature is less than or equal to 212 F.
j Coldshutdown - As defined in Technical Specification 1.0, the reactor is in the shutdown mode, the reactor coolant temperature is less than or equal to 212 F, and the reactor is vented to atmosphere.
I Compensatory non-alarming indications - Information displayed in the main control room including analog and digital parameter displays, trend recorders, the Safety Parameter Display System (SPDS), and the plant process computer.
Control-As applied to remote shutdown capability, this is the ability to manipulate plant parameters without reliance on control room devices or instrumentation using components and methods specified by Abnormal Operating Procedure 915, Shutdown Outside Control Room.
j i
CPS-Counts Per Second CRD - Control Rod Drive CSCS-Core Standby Cooling System CST-Condensate Storage System
)
Curie (Ci) - A measurement of radioactive decay rate equal to 3.70E+10 disintegration's per second (becquerels).
1
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CW-Circulating Water DAEC-Duane Amold Energy Center DC-Direct Cu rent P
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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevien)
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PAGE D-3 of 10 EFFECTIVE DATE: TBD DEFINITIONS DEG - Dose Equivalent Dominant accident sequences - These will lead to degradation of all fission product barriers. Dominant accident sequences leading to core damage at DAEC include complete loss of 125 VDC, loss of decay heat removal, ATWS with failure of Standby 1.iquid Control, prolonged station blackout, and loss of offsite power with early HPCI/RCIC failure.
DW-Drywell EC-Emergency Coordinator ECCS-Emergency Core Cooling System
(')'N EDE - Effective Dose Equivalent as defined in 10 CFR 20.1003 L
Emergency Action Level (EAL) - A pre-determined, site-specific, observable threshold for a plant Initiating Condition that places the plant in a given Emergency Class. An EAL can be: an instrument reading, an equipment status indicator, a measurable parameter (on-site or offsite), a discrete observable event, results of analyses, entry into specific emergency operating procedures, or another phenomenon which, if it occurs, indicates entry into a particular Emergency Class.
Emergency Class - Same as " Emergency Classification Level" below.
Emergency Classification Level-These are taken from 10 CFR 50, Appendix E. They are, in escalating order: (Notification of) Unusual Event (UE), Alert, Site Area Emergency (SAE), and General Emergency j
(GE).
j EOP - Emergency Operating Procedure EPA - Environmental Protection Agency EPIP - Emergency Plan Implementing Procedure ESF-Engineered Safety Features
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() ESS-Engineered Safety Systems Establish - Make anangemcnts for a stated condition, e.g., establish communications with control room.
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
,\\Ni PAGE D-4 of 10 A E: TBD DEFINITIONS ESil'- Emergency Service Water Fission Product Barrier - One of the three principal barriers to uncontrolled release of radionuclides: Fuel Clad, Reactor Coolant System (RCS), and the Primary Containment.
FP - Fuel Pool Fuel Clad (Barrier) - The zirconium alloy tubes that contain the fuel pellets.
General Emergency (GE) - Events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Releases can reasonably be expected to exceed EPA Protective Action Guide (PAG) exposure levels offsite for more O than the immediate site area.
L/
GPM-Gallons Per Minute GSil'- General Service Water Hot shutdown - As defined in Technical Specification 1.0, the reactor is in shutdown mode and the reactor coolant temperature is greater than 212 F.
Hot standby condition - As defined in Technical Specification 1.0, this refers to operation with the reactor coolant temperature greater than 212 F, reactor pressure vessel less than 1055 psig, and the mode switch in Startup position.
HPCI-High Pressure Coolant Injection (system).
Identified Leakage - Identified Leakage shall be:
- a. Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or
- b. Leakage into the containment atmosphere from sources that are both specifically
/7 located and known not to interfere with the operation of the leakage detection U
systems.
IDLH-Immediately Dangerous to Life and Health
Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
PAGE D-5 of 10 EFFECTIVE DATE: TBD DEFINITIONS Inadvertent - Accidental or unintentional, e.g., the event occurred because procedures were not strictly adhered to.
Imminent - No tumaround in safety system performance is expected and escalation to a higher emergency classification level is expected to occur within two hours.
Implement - Commence a required program or series of procedures.
In service - A component or system in the appropriate configuration for nomial operation and is considered operable as defined in the Technical Specifications.
hulicator - The name for the row on the fission barrier table that is used for convenient grouping of p%_/ similar symptoms.
/nitiate - Take action to begin a process Initiating Condition (IC) - One of a predetermined subset of nuclear power plant conditions where either the potential exists for a radiological emergency or such an emergency has occurred.
/PE-Individual Plant Examination IPOI-Integrated Plant Operating Instruction IRM-Intermediate Range Monitor Isolate - Remove from service by closing ofTthe flow path kV-Kilovolt (s)
LCO - Limiting Condition for Operation LLRPSF-Low Level Radwaste Processing and Storage Facility
,q LOCA - Loss of Coolant Accident V
Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC revieu)
U PAGE D-6 of 10 EFFECTIVE DATE: TBD DEFINITIONS Loss (of a fission product barrier) - A severe challenge to a fission product barrier exists such that the barrier is considered incapable of performing its safety function.
LPCI-Low Pressure Coolant Injection AICC-Motor Control Center i
AICUTL - Maximum Core Uncovery Time Limit J
Aficrocurie ( Ci) - One millionth of a curie, i.e.,3.7E+4 disintegration's per second (becquerels).
AllDAS - Meteorological Information and Dose Assessment System, primary method for detecting and quantifying gaseous releases at the DAEC.
(*%
U Afillicurie (mci) - One thousandth of a curie, i.e.,3.7E+7 disintegration's per second (becquerels).
Afillirem (mrem) - One thousandth af a tem AIPfl-Miles Per Hour mR - milliroentgen, i.e., one thousandth of a roentgen (R)
AfSIV-Main Steam Isolation Valve A/SL - Main Steam Line NEI-Nuclear Energy Institute (formerly NUMARC)
Notification of Unusual Event (NOUE) - Same as " Unusual Event", below.
NPSII-Net Positive Suction llead NUAIARC - Nuclear Utility Management and Resources Council (now NEI)
OBE-Operating Basis Earthquake
%j ODAAf-Offsite Dose Assessment Manual l
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC review) gU PAGE D-7 of 10 EFFECTIVE DATE: TBD DEFINITIONS Operating Mode - As defined by Technical Specification Table 1.0-1, Operating Mode describes the operating status of the unit. Mode designations (and the associated Reactor Mode Switch Positions) used at DAEC are: RUN/ POWER OPERATION (Run), STARTUP (Startup or Refuel), HOT SIIUTDOWN (Shutdown), COLD SIlUTDOWN (Shutdown), and REFUELING (Shutdown or Refuel).
Operable - A system is considered capable of performing its function in accordance with the applicable Technical Specification requirements. Implicit in this definition is the assumption that all auxiliary equipment required for the system is also operable.
OSS-Operations Shift Supervisor PAG - Protective Action Guide
(~} Planned - Loss of a component or system due to expected events such as scheduled maintenance and O'
testing activities.
Potential Loss (of a fission product barrier) - A challenge to a fission product barrier exists such that the barrier is considered degraded in its ability to perform its safety function.
Primary Containment (Barrier) - The drywell, the torus, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves.
PSIG - Pounds per Square Inch Gauge RB - Reactor Building RBCCW-Reactor Building Closed Cooling Water (system)
RCIC - Reactor Core Isolation Cooling (system)
RCS-Reactor Coolant System RCS Barrier - The reactor coolant system pressure boundary including the reactor pressure vessel and all reactor coolant system piping up to and including the outermost isolation valves.
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Recognition Category - A logical grouping ofInitiating Conditions, e.g., System Malfunctions.
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Rem - Unit of radiation dose as defined in 10 CFR 20.1004
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC review) n PAGE D-8 of 10 EFFECTIVE DATE: TBD DEFINITIONS Required - Action taken (such as entry into emergency operating procedure) is neither optional nor merely suggested; rather, it is imperative based on existing conditions.
RHR - Residual licat Removal (system)
RHRSW-Residual Heat Removal Service Water (system)
Roentgen (R) - Unit ofionizing radiation energy absorbed in a cubic centimeter of air RPV-Reactor Pressure Vessel RWCU-Reactor Water Clean-Up (system)
, p V SBDG - Standby Diesel Generator SBGT-Standby Gas Treatment (system)
SBLC-Standby Liquid Control (system)
SBO - Station Blackout S/D - Shutdown SDC - Shutdown Cooling SDV-Scram Discharge Volume Shutdown - As defined in Technical Specification 1.0, the reactor is in a shutdown condition when the reactor mode switch is in the Shutdown position.
Sigmficant transient - (See also, "Trensient", below.) Includes response to automatic or manually initiated j
functions such as scrams, runbacks involving greater than 25% thermal power change, ECCS injections, or thermal power oscillations of 10% or greater.
A V Site Area Emergency (SAE) - Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public. Any releases are not expected to result in
Duane Arnold Energy Center
,o EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC revicw)
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V PAGE D-9 of 10 EFFECTIVE DATE: TBD DEFINITIONS exposure levels which exceed EPA Protective Action Guide (PAG) exposure levels except near the site boundary.
SPDS-Safety Parameter Display System SRM-Startup Range Monitor SRO - Senior Reactor Operator SRV-Safety-Relief Valve Sustained wind speed-Baseline wind speed measured by meteorological tower that does not include gusts (G) TAF-Top of Active Fuel (344.5 inches above bottom of RPV)
TEDE - Total Efrective Dose Equivalent as defined in 10 CFR 20.1003 Total Leakage - Total leakage is the sum ofIdentified Leakage and Unidentified Leakage.
Transient - A condition that: (1) is beyond the expected steady-state fluctuations in temperature, pressure, power level, or water level, and (2) is beyond the normal manipulations of the Control Room operating crew, and (3) is expected to require actuation of fast-acting automatic control or protection systems to bring the reactor to a new safe, steady-state condition.
TSC-Technical Support Center Uncontrolled - Condition is not the result of planned actions by the plant staff in accordance with procedures.
Unisolable - Actions taken from the Main Control Board or locally are not successful in eliminating the leakage path.
Unidenti/ led Leakage - Unidentified Leakage shall be all leakage which is not identified leakage.
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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu) 7 i
<V PAGE D-10 of 10 EFFECTIVE DATE: TBD DEFINITIONS Unplanned - Used to preclude the declaration of an emergency where a component or system has been removed intentionally from service (e.g., for maintenance and/or testing activities). As used in the context of radioactive releases, " unplanned" includes any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit.
Unusual Event (UE) - Events are in process or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
VAC-Volt (s) Alternating Current VI)C-Volt (s) Direct Current r%U Valid - Indication is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.
WEC-Water Effluent Concentration t"T N)
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
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PAGE O-1 of 3 ORGANIZATION OF BASIS INFCRMATION EFFECTIVE DATE: TBD ORGANIZATION OF BASIS INFORMATION The format of the EAL Basis information was developed to address training needs, to facilitate NRC approval, and to facilitate future revisions and 10 CFR 50.54(q) evaluations. Each EAL Basis is organized in the following manner:
Initiating Condition Identifier For consistency, DAEC has chosen to make its Initiating Condition (IC) identifiers identical to those used in NEI document NUMARC/NESP-007. The EAL Technical Basis information is organized by generic IC
/~'s b identifier number and name.
NUMARC/NESP-007 organized the generic information into four Recognition Categories. These are:
A - Abnormal Rad Levels / Radiological Effluent F - Fission Product Barrier Degradation H - Hazards and Other Conditions Affecting Plant Safety S - System Malfunctions For the A, H, and S recognition categories, all EAL basis information is organized by IC identifier in escalating emergency class order from Unusual Event through General Emergency. For the F recognition category, the initiating conditions are the combinations of fission product barrier losses and potential losses that correspond to each emergency classification level. The individual indicators used on the fission barrier table are separately discussed below. The generic IC identifiers use two letters followed by one number.
The first letter corresponds to the event category as shown above. The second letter corresponds to the emergency classification level for the IC:
U -(Notification of) Unusual Event A - Alert S - Site Area Emergency G - General Emergency The number designates whether the IC is the first, second, third, etc., IC for that recognition category under p)
(,
that emergency classification. 1or example, SU2 is the designator for the second System Malfunction recognition category IC in the Unusual Event classification, etc. Generic information is quoted directly from NUMARC/NESP-007 Revision 2, dated January,1992. Changes from the NUMARC/NESP-007 l
Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu) nV)
PAGE O-2 of 3 ORGANIZATION OF BASIS thf0RMATION EFFECTIVE DATE: TBD text are denoted by caret marks (< >). Such changes are based on correction of typographical errors such as those mentioned in the NUMARC Questions and Answers dated June 1993, reflect changes made in 10 CFR Part 20, or to put the information in proper context.
Event Tyne This is the label of the applicable row for the EAL Table shown in EPIP-1.1, Determination ofEmergency Action Levels. The event type lists the general area of concem and includes Offsite Rad Conditions, Onsite Rad Conditions, Natural Disasters, Fire, Other llazards and Failures, Security, Control Room Evacuation, EC/OSS Judgment, Loss of Power, RPS Failure, Inability to Maintain Shutdown Conditions, Instrumentation / Communication, Coolant Activity, and Coolant Leak. This structure was chosen to be consistent with the previous EAL presentation which is already familiar to the Emergency Coordinators and Operations Shift Supervisors. It is also permissible to organize the generic information in this manner (l based on the response to Question 5 contained in the NUMARC Methodologyfor Development of G'
Emergency Action Levels NUMARC/NESP-007 Revision 2 Questions and Answers June 1993.
Annlicable Onerating Modes The applicable operating modes for each Initiating Condition / Emergency Action Level is then listed based on NUMARC/NESP-007 mode descriptions. The DAEC EALs use the operating modes defined in Technical Specifications Table 1.0-1. These are:
1 - Run/ Power Operation 4 - Cold Shutdown 2 - Startup 5 - Refueling 3 - llot Shutdown To conserve space, the EAL displays use "Run" to mean "Run/ Power Operation" and "S/D" as an abbreviation for "Shutdo vn."
Operating mode applicability ofEALs is based on the operating mode that the plant was in immediately before the event sequence leading to entry into the emergency classification.
For example, events / conditions addressed by EAl.s applicable to Run mode are expected to lead to reactor trip which should bring the plant to llot Shutdown. Ilowever, the appropriate emergency classification would still be based on the applicable EALs for Run/ Power Operation for these events / conditions. If"ALL" operating O) modes are specified f>r the EAL, then the EAL applies to all modes identified above plus defueled b
conditions.
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I Duane Amold Energy Center EMERGENCY ACTIO'N LEVEL BASES DOCUMENT Rev. 2 (for NRC revieu) ob PAGE O-3 of 3 ORGANIZATION OF BASIS INFORMATION EFFECTIVE DATE: TBD Generic Example EAL(s)
The generic example EALs are then listed. When more than one is provided, logic phrasing is used to describe whether all EALs are suggested or whether at least one EAL should be chosen.
DAEC EAL Information This contains the plant-specific information used to implement the generic EALs. This section will also include the basis, as appropriate, for deviation from generic EALs. For example, DAEC does not have a Independent Spent Fuel Storage Installation for on-site dry storage of spent fuel. Thus, DAEC does not have EALs corresponding to the generic guidance for this item. As appropriate, description of any supporting calculations, their underlying bases and assumptions, and their results are included in this section.
pU'i References The references used to develop the DAEC EAL Information are lis:ed here, as appropriate.
- 2. Fission Product Barrier Table Indicators The basis information for the fission barrier table indicators is organized similarly to the other basis information described above. For each barrier - fuel clad, RCS, and primary containment - basis information is organized by " Indicator." The indicator is the name for the row on the fission barrier table and is used for convenient grouping of similar symptoms, similar to the " Event Type" used for the A, H, and S EALs described above. Indicators include Radiation / Core Damage, RPV L.evel, Leakage, Primary Containment Atmosphere, and EC/OSS Judgment.
After the DAEC Indicator, the applicable generic BWR fission product barrier indicators are then displayed, showing both the generic loss and potential loss conditions, as applicable. Next displayed is the appropriate DAEC infbrmation and references. These are displayed in the same manner as the A, II, and S recognition category basis information described above.
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EMERGENCY PLAN IMPLEMENTING PROCEr ~ 'E No. EPIP - 1.1
- v. 2 PAGE 1 of 1 (For NRC A..ww)
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I ABNORMAL RAD LEVEISRADIOACTIVE EITLUENT EFFECTIVE DATE TBD g
EVENT TYPE Uf# USUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY AUf AAf ASf AGf Any Unplanned Release of Gaseous or Any Unplar.ned Release of Gaseous or Site Boundary Dose Resutung from an Site Boundary Dose Resulting from an Liquid Radioactivity to the Environment Liquid Radioactivity to the Environment Actual or Imminent Release of Gaseous Actual or Imminent Release of Gaseous That Exceeds Two Times the That Exceeds 200 Times the Radioactivity Exceeds 100 mrem TEDE Radioactivity Exceeds 1000 mrem TEDE Radiological Technical Specifications Radiological Technical Specifications or 500 mrem CDE Thyroid for the Actuait of 5000 mrem CDE Thyroid for the For 60 Minutes or Longer For 15 Minutes or Longer or Projected Duration of the Release l Actual or Projected Duration of the Release Vahd Reactor Building or Turbine Building Vahd Reactor Building or Turtnne Building Vahd Reactor Building or Turtine Building Vatid Reactor Bui; ding or Turbine Building ventilation (Kaman) rad monitor readmg ventitation (Kaman) rad monitor reading ventitation (Kaman) rad monitor reading ventilation (Kaman) rad monitor reading above 1 E-3 pCVcc for more than 60 above 3 E-2 pCVcc for more than 15 above 6 E-2 pCVcc for more than 15 above 6 E-1 pCVcc for more than 15 minutes.
minutes.
minutes (Dose assessment not available) minutes. (Dose assessment not available)
-R OR OR Vahd Offgas Stack (Kaman) rad monitor Valid Offgas Sta > u. aman) rad monitor Valid Offgas Stack (Kaman) rad monitor Valid Offgas Stack (Kaman) rad monitor reading above 6 E-1 pCVcc for more than reading above 2 E pCVcc for more than reading above 4 E+1 pCVcc for rwre than reading above 4 E+2 pCVcc for more than 60 minutes.
15 minutes.
15 maiutes. (Dose assessment not 15 minutes. (Dose assessment not OR OR available) available)
Valid LLRPSF (Kaman) rad monitor reading Valid LLRPSF (Kaman) rad monitor reading above 9 E-4 pCVcc for more than above 9 E-2 pCycc for more than 60 msnutes.
15 minutes.
OR OR OFFSITE RAD Vald GSW rad monitor reading above 3E+3 Vald GSW.ad monitor reading above 3E+5 CONDITIONS CPS for more than 60 minutes.
CPS for more than 15 msnutes.
OR OR Valid RHRSW & ESW rad monitor reading Valid RHRSW & ESW rad monitor reading above 8E+2 CPS for more than 60 minutes.
above 8E+4 CPS for more than 15 msnutes.
OR OR Vald RHRSW & ESW Discharge Canal rad Vald HHRSW & ESW Discharge Canal rad monitor reading above 1E+3 CPS for more monitor reading above 1E+5 CPS for more than 60 minutes.
than 15 minutes.
OR OR Confirmed sample anatyses for gaseous or Confirmed sample analyses for gaseous or {
houid releases indicates concentrations in hqud releases indicates concentrations in excess of 2 times ODAM hmits for greater excess of 200 times ODAM limits for greater than 60 menutes.
than 15 minutes.
OR OR OR Valid field survey reading outside the site Valid field survey reading outside the site Vahd field survey reading outside the site boundary >10 mR,hr or >50 mFLhr CDE boundary >100 mRhr or >500 mR5r CDE boundary >1.000 mR/hr or >5.000mRhr Thyroid.
Thyroid.
CDE Thyroid.
OR OR OR OR Dose assessment determines hourty dose Dose assessment determines hourty dose Dose assessment determines integrated Dose assessment determines integrated outside the site boundary above 0.1 mrem outside the site boundary above 10 mrem accident dose projection outside the site accident dose projection outside the site TEDE.
TEDE.
boundary above 100 mrem TEDE or above boundary above 1,000 mrom TEDE or 500 mrem CDE Thyroid.
above 5.000 mrem CDE Thyrod.
Op. Modes: ALL Op. Modes: ALL Op. Modes: ALL Op. Modes: ALL AU2 AA2 Unexpected increase in Plant Radiation Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel Uncontrolled loss of reactor cavity or fuel Report of ANY of the following:
pool water level with all spent fuel e
Valid ARM HI RAD alarm for the assembhes remaining water covered as Refuehng Floor North End, Refueling
_indi_catad_by CW? of tha fo&wnT h x PA4TM U L @ n.h
Report to control room Area, or Spent Fuei Storage Area w
Valid Refueling Floor North End, Valid fuel poollevelindication (U-e e
ONSITE RAD 3413) below 36 feet ar.d lowenng Refuehng Flocr Scuth End, or New Fuel CONDITIONS e
VLlid WR GEMAC Flcedup indicatbn Storage Area ARM Readng atxwe 10 (U-4541) coming on sc.te.
mRhr Valid Spent Fusi Storage Area ARM e
OR Readog above 100 mRMr Unexpected ARM reading offscale high or OR above 1000 times normal readog.
Report of visual observation of Irradiated Fuel uncovered Op. Modem-ALL OR Water level readog below 450" as indcated on U4541 (floodup) for the Reactor Hefueling Cavity that will result in irradiated Fuel uncovering OR Valid Fuel Pool water level indication (U-3413) below 16 feet.
Op. Modes: ALL AA3 Release of Radioective Material or increases in Redution Levels Within the Facility Ttut Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or to Maintain Cold Shutdown Vaild area radiaton monitor (RE9162) readng grester than 15 mRhr in the Control Room.
OR Vaild area radiation monitor (RE9168) readng greater than 500 mRMr at the Remote Shutdown Panel,1C388.
Op. Modes: ALL
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O ABNORMAL RAD LEVELS /RADIOLr telLL EFFLUENT CATEGORY O
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
PAGE A-1 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY AUI Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment That Exceeds Two Times the Radiological Technical Specifications For 60 Minutes or Longer i
EVENT TYPE: Offsite Rad Conditions OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2 or 3 or 4)
- 1. A valid reading on < site-specific > monitors that < > indicatu that the release may have exceeded <2 x site-specific technical specifications for 60 minutes or longer.>
d
- 2. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates with a release duration of 60 minutes or longer in excess of two times (site-specific technical specifications).
- 3. Valid reading on perimeter radiation monitoring system greater than 0.10 mR/hr above normal background for 60 minutes [for sites having telemetered perimeter monitors].
- 4. Valid indication on automatic real-time dose assessment capability greater than (site-specific vdue) for 60 minutes or longer [for sites having such capability].
Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.
The primary methodfor declaration is by means of Jose assessment using the MIDAS computer model.
This is listed as DAEC EAL.l. However, if the monitor readings are sustairedfor longer than 60 minutes and the required dose assessments cannot be completed within this period, then the declaration must be made based on the validreading.
The approach taken for calculation of gaseous radioactive ellluent EAL setpoints includes use of the ODAM Table 3-2 source term computed by BWR-GALE for the DAEC Base Case. The release is p
assumed to be from a single release point. Multiple release points would be difficult to present as explicit V
EAL threshold values and in any case, are addressed by off-site dose assessment by MIDAS, which is the preferred method for detennining this condition. The calculation methods for setpoint determination are from ODAM Section 3.4 and are based on Regulatory Guide 1.109 methodology. The table below lists the
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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
PAGE A-2 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY i
results of the gaseous eflluent EAL calculations. The Kaman extended range capability is used because the General Electric Offgas Stack monitor has a limited range.
Gaseous Effluent EALs Offgas Stack Kaman 9/10 Turbine Bldg (Kaman 1/2) and Reactor Bldg (Kaman 3/4,5/6,7/8) i l
Maximum flow (CFM) 10,000 72,000 Release Limits Concentration Release Rate Concentration Release Rate (pCi/cc)
(pCi/sec)
( Ci/cc)
(pCi/sec)
Tech Spec 3.2E-1 1.5E+6 6.2E-4 2.1 E+4 1
Unusual Event (2 x TS) 6.4E-1 3.0E+6 1.2E-3 4.2E+4 iO Alert (60 x TS) 1.9E+1 8.9E+7 3.7E-2 1.3 E+6 v
LLRPSF Kaman 12 Maximum flow (CFM) 99,000 Release Limits Concentration Release Rate (pCi/cc)
(pCi/sec) 4 Tech Spec 4.5E-4 2.1 E+5 Unusual Event (2 x TS) 9.0E-4 4.2E+5 l
Alert (200 x TS) 9.0E-2 4.2E+7 The oft-gas stack is treated as an elevated release and the turbine building and reactor building vents are treated as mixed-mode releases. The ground level setpoints are taken from the default setpoint calculations from the quarterly surveillance tests performed by DAEC Chemistry technicians. Reactor Building, Turbine Building, LLIU'SF (Low Level Radwaste Processing and Storage Facility) and Offgas Stack
~
Noble Gas Monitor alarm setpoints are calculated based on achieving the Tech Spec instantaneous release limit, assuming annual average meteorology as defined in the ODAM. The Tech Spec Limit currently j
corresponds to a reactor building or turbine building ventilation alarm setpoint of 6.2 E-04 pCi/ce. The monitor alarm setpoint can be periodically adjusted but typically does not vary by much. The DAEC EAL therefore addresses valid radiation levels exceeding 2 times the alarm setpoint for greater than 60 minutes.
Rounded off, this corresponds to 1 E-3 Ci/cc. The corresponding ofTgas stack monitor value is 0.64 1
pCi/ce, rounded off to 6 E-1 pCi/cc. The Tech Spec Limit currently for the LLRPSF building ventilation alarm setpoint is 4.5 E-04 pCi/cc. The DAEC EAL therefore addresses valid radiation levels exceeding 2 times the alarm setpoint for greater than 60 minutes. This corresponds to 9 E-4 Ci/cc.
Technical specification setpoints for radioactive liquid radiation monitors are 10 times the 10 CFR 20 Appendix B, Table 2 Water Ellluent Concentration (WEC) limits. It is the policy of DAEC to process all
Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC revicu) v PAGE A-3 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY liquid radwaste so that no release of radioactive liquid to the environment is allowed. The radwaste effluent line which could be used as a batch release mechanism has a trip function that prevents exceeding the DAEC release limit, however, an EAL has been provided. The other pathways to the environment (RHRSW - to cooling tower, RIIRSW - to discharge canal) have radiation monitors with readouts going to the Control Room. These systems could become contaminated if heat exchanger leaks develop; however, historically this has not occurred in the service water systems at DAEC. These monitors are displayed on panels IC02 and IC10.
Reactor water is the likely source of contamination through the service water systems as opposed to floor drain, detergent drain, and chemical waste discharge. The floor drain and detergent drains go to Radwaste Processing and would be batch released to the Radwaste effluent discharge line (if such a release were to q occur). The chemical discharge sump is normally a radioactivity clean system and is tested by Chemistry Q to ensure no contamination prior to discharging to the canal.
The setpoints for the three service water radiation effluent monitors vary because of differences in detector efficiencies and background. Setpoints based on the same reactor water sample are listed below to show the differences. The rounded off readings will be used for the EALs for ease of reading the monitor scales.
Monitor TS Limit Reading UE Level Alert Level GSW 1,555 CPS 1.5E+3 CPS 3E+3 CPS 3E+5 CPS RHRSW & ESW to cooling 413 CPS 4E+2 CPS 8E+2 CPS 8E+4 CPS tower RHRSW & ESW to 507 CPS 5E+2 CPS lE+3 CPS lE+5 CPS Discharge Canal There are no significant deviations from the generic EALs. However, DAEC does not have a telemetered radiation monitoring system. As an alternative, use of field instruments was considered. It is not practical to establish an EAL based on field survey readings of 0.1 mR/hr for greater than 60 minutes because field instruments in use for emergency response do not have a threshold of detection to meet such criteria. Thus, DAEC does not have an EAL corresponding to generic EAL 3.
O
Duane Arnold Energy Center c
EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
PAGE A-4 of 24 f
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY l
llourly Whole Body Dose Corresponding to 2 x ODAM Limit for Gaseous Release ODAM limit = 500 mrem / year Whole Body Dose 2 x ODAM limit = [2 x 500 mrem / year]/8760 hours / year = 0.114 mrem Whole Body in one hour Rounded off to 0.1 mrem Dose assessment using MIDAS is based on the EPA-400 methodology, e.g., use of Total Effective Dose Equivalent (TEDE). This is somewhat different from whole body dose from gaseous effluents determined p
by ODAM methodology which forms the basis for the radiation monitor readings calculated in accordance
(
with the generic methodology. The gaseous efiluent radiation monitors can only detect noble gases. The contribution ofiodine's to TEDE could therefore only be determined either by: (1) utilizing MIDAS, or (2) gaseous effluent sampling. DAEC EAL 4 is written in terms of TEDE and the gaseous efiluent radiation monitor readings are determined based on ODAM.
4
REFERENCES:
- l. OfTsite Dose Assessment Manual Section 6.1.2 and 7.1.2 Bases
- 2. Emergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and Protective Action
- 3. Radiation Protection Calculation No. 95-001-C, Emergency Actions Levels Based on Efiluent Radiation Monitors, January 24,1995
- 4. UFSAR Section 11.5, Process and Efiluent Radiation Monitoring and Sampling Systems
- 5. EPA 400-R-92-001, Ahmual ofProtective Action Guides and Protective Actionsfor Nuclear Incidents i
- 6. NUA{ ARC Afethodologyfor Development ofEmergency Action Levels NUAfARC/NESP-007 Revision 2 Questions andAnswers, June 1993
(.
Duane Amold Energy Center c
EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
PAGE A-5 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY AU2 Unexpected Increase in Plant Radiation < >
EVENT TYPE: Onsite Rad Conditions OPERATING MODE APPLICABILITY. All EXAMPLE EMERGENCY ACTION LEVELS:
(1 or 2 or 3 or 4)
- 1. (Site-specific) indication of uncontrolled water level decrease in the reactor refueling cavity with all irradiated fuel assemblies remaining covered by water.
- 2. Uncontrolled water level decrease in the spent fuel pool < > with all irradiated fuel assemblies CN remaining covered by water.
V
- 3. (Site-specific) radiation reading for irradiated spent fuel in dry storage.
j
- 4. Valid Direct Area Radiation Monitor readings increases by a factor of 1000 over normal
- levels.
Nonnal levels can be considered as the highest reading in the past twenty-four hours excluding the current peak value.
There are no significant deviations from the generic EALs. DAEC does not have a spent fuel transfer canal or on-site dry storage of spent fuel.
Uncontrolled means that the condition is not the result of planned actions by the plant staffin accordance with procedures. Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.
There are three methods to determine water level decreases of concem. The first method is by report to the control room. The other methods include use of the Floodup level indicator and the spent fuel pool level indicator. These are further described below.
During preparation for reactor cavity flood up prior to entry into refuel mode, reactor vessel level Q
instrument LI-4541 (WR GEMAC, FLOODUP) on control room panel 1C04 is placed in service by I&C personnel connecting a compensating air signal after the reference leg is disconnected from the reactor head. Normal refuel water level is above the top of the span of this flood up level indicator. A valid indication (e.g., not due to loss of compensating air signal or other instrument channel failure) of reactor
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview) n i
(G PAGE A-6 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY cavity level coming on span for this instrument is used at DAEC as an indicator of uncontrolled reactor cavity level decrease.
DAEC Technical Specifications require a minimum of 36 feet of water in the spent fuel pool. During refueling, the gates between the reactor cavity and the refueling cavity are removed and the spent fuel pool level indicator LI 3413 is used to monitor refueling water level. Procedures require that a normal refueling water level be maintained at 37 feet 5 inches. A low level alarm actuates when spent fuel pool level drops below 37 feet 1 inch. Symptoms ofinventory loss at DAEC include visual observation of decreasing water levels in reactor cavity or spent fuel storage pool, Reactor Building (RB) fuel storage pool radiation monitor or refueling area radiation monitor alarms, observation of a decreasing trend on the spent fuel pool water level recorder, and actuation of the spent fuel pool low water level alarm. To eliminate minor level perturbations from concem, DAEC uses LI 3413 indicated water level below 36 feet and lowering.
b i
Increased radiation levels can be detected by the local refueling floor area radiation monitors, the refueling j
floor Continuous Air Monitor (CAM) alarm, refueling areas radiation monitors, fuel pool ventilation exhaust monitors, and by Standby Gas Treatment (SGBT) System automatic start. Applicable area radiation monitors include those that are displayed on Panel IC02 and alarmed on Panel IC04B. The DAEC EAL has also been written to reflect the case where an ARM may go offscale high prior to reaching i
1,000 times the normal reading.
NOTE: On Annunciator Panel IC04B, the indicators listed below are expected alarms during pre-planned transfers of highly radioactive material through the affected area. If an IIP Technician is present, sending an Operator is not required. Radiation levels other than those expected should be promptly investigated.
The indicators are high radiation alarms from the Hot Laboratory or Administrative Building, the new fuel storage area, and the radwaste building.
4
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Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
PAGE A-7 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY i
REFERENCES:
- 1. Alarm Response Procedure (ARP) IC048, Reactor Water Cleanup and Isolation
- 2. Tecimical Specification 3.9C, Spent Fuel Pool Water Level
- 3. Emergency Plan Implementing Procedure (EPIP) 3.1, Inplant Radiological Monitoring, Attachment 1, ARM Locations
- 4. Emergency Operating Procedures (EOP) Basis Document, Breakpoints for RC/L & L
- 6. Integrated Plant Operating Instruction (IPOI) 8, Outage and Refueling Operations
- 7. Fuel & Reactor Component Handling Procedure (F&RCIIP) 5, Procedure for Moving Core Components Between Reactor Core and Spent Fuel Pool, Within the Reactor Core, or Within the Spent Fuel Pool
- 8. NUAMRC Metnodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions and Answers, June 1993 O
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
PAGE A-8 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY AAl Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times Radiological Technical Specifications for 15 Minutes or Longer EVENT TYPE: Offsite Rad Conditions OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS:
(1 or 2 or 3 or 4)
- 1. A valid reading on < site-specific > monitors that < > indicates that the release may have exceeded <200 D
x site-specific technical specifications for 15 minutes or longer.>
- 2. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates in excess of(200 x site-specific technical specifications) for 15 minutes or longer.
- 3. A valid reading on perimeter radiation monitoring system greater than 10.0 mR/hr sustained for 15 minutes or longer. [for sites having telemetered perimeter monitors]
- 4. Valid indication on automatic real-time dose assessment capability greater than (200 x site-specific Technical Specifications value) for 15 minutes or longer. (for sites having such capability]
Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.
The primary methodfor declaration is by means ofdose assessment using the MIDAS computer model.
This is listed as DAEC EAL 4. However, ifthe monitor readings are sustainedfor longer than 15 minutes and the required dose assessments cannot be completed within this period, then the declaration must be made based on the valid reading.
O
Duane Amold Energy Center n
EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview) b PAGE A-9 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY l
Gaseous Effluent EALs OITgas Stack Kaman 9/10 Turbine Bldg (Kaman 1/2) and Reactor Bldg (Kaman 3/4,5/6,7/8)
Maximum flow (CFM) 10,000 72,000 Release Limits Concentration Release Rate Concentration Release Rate (pCi/cc)
(pCi/sec)
( Ci/cc)
(pCi/sec)
Tech Spec 3.2E-1 1.5E+6 6.2E-4 2.1 E+4 Unusual Event (2 x TS) 6.4E-1 3.0E+6 1.2E-3 4.2E+4 Alert (60 x TS) 1.9E+1 8.9E+7 3.7E-2 1.3 E+6 LLRPSF Kaman 12 Maximum flow (CFM) 99,000 Release Limits Concentration Release Rate
( Ci/cc)
( Ci/sec)
Tech Spec 4.5E-4 2.1 E+5 Unusual Event (2 x TS) 9.0E-4 4.2E+5 l
Alert (200 x TS) 9.0E-2 4.2E+7 The off-gas stack is treated as an elevated release and the turbine building and reactor building vents are treated as mixed-mode releases. The ground level setpoints are taken from the default setpoint calculations i
from the quarterly surveillance tests performed by DAEC Chemistry technicians. Reactor Building, Turbine Building, LLRPSF (Low Level Radwaste Processing and Storage Facility) and Offgas Stack t
Noble Gas Monitor alarm setpoints are calculated based on achieving the Tech Spec instantaneous release limit assuming annual average meteorology as defined in the ODAM. The Tech Spec Limit currently corresponds to a reactor building or turbine building ventilation alarm setpoint of 6.2 E-4 pCi/ce. The monitor alarm setpoint can be periodically adjusted but typically does not vary by much. For the Offgas j
Stack, Reactor Building and Turbine building KAMAN monitor readings, DAEC chose to multiply the i
technical specification concentration by a factor of 60 (instead of 200) in order to allow for a logical step I
progression in monitor setpoints from the AUl through AAl to ASl. The DAEC EAL therefore addresses valid radiation levels exceeding 60 times the alarm setpoint for greater than 15 minutes. Rounded off, this corresponds to 3 E-2 pCi/cc. The corresponding ofTgas stack monitor value is 19.2 pCi/ce, rounded off to 2 E+1 Ci/cc. The Tech Spec Limit currently for the LLRPSF building ventilation alarm setpoint is 4.5 E-04 pCi/cc. The DAEC EAL therefore addresses valid radiation levels exceeding 200 times the alarm setpoint for greater than 15 minutes. This corresponds to 9 E-2 pCi/cc.
Technical specification setpoints for radioactive liquid radiation monitors are 10 times the 10 CFR 20 Appendix B, Table 2, Water Efiluent Concentration (WEC) limits. It is the policy of DAEC to process all
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
J PAGE A-10of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY liquid radwaste so that no release of radioactive liquid to the environment is allowed. The radwaste effluent line which could be used as a batch release mechanism has a trip function that prevents exceeding the DAEC release limit, and therefore no EAL limits are provided. The other pathways to the environment (RiiRSW - to cooling tower, RHRSW - to discharge canal) have radiation monitors with readouts going to the Control Room. These systems could become contaminated if heat exchanger leaks develop; however, historically this has not occurred in the service water systems at DAEC. These monitors are displayed on panels IC02 and IC10.
Reactor water is the likely source of contamination through the service water systems as opposed to floor drain, detergent drain, and chemical waste discharge. The floor drain and detergent drains go to Radwaste Processing and would be batch released to the Radwaste effluent discharge line (if such a release were to occur). The chemical discharge sump is normally a radioactivity clean system and is tested by Chemistry w
to ensure no contamination prior to discharging to the canal.
The setpoints for the three service water radiation effluent monitors vary because of differences in detector efficiencies and background. Setpoints based on the same reactor water sample are listed below to show the differences. The rounded off readings will be used for the EALs for ease of reading the monitor scales.
Monitor TS Limit Reading UE Level Alert Level GSW 1,555 CPS 1.5E+3 CPS 3E+3 CPS 3E+5 CPS RiiRSW & ESW to coohng 413 CPS 4E+2 CPS 8E+2 CPS 8E+4 CPS tower RiiRSW & ESW to 507 CPS 5E+2 CPS 1E+3 CPS 1E+5 CPS Discharge Canal DAEC does not have a telemetered radiation monitoring system. As an alternative, DAEC uses valid field survey readings outside the site boundary greater than 10 mR/hr or greater than 50 mR/hr CDE Thyroid.
Hourly Whole Body Dose Corresponding to 200 x ODAM Limit for Gaseous Release ODAM limit = 500 mrem / year Whole Body 200 x ODAM limit = [200 x 500 mrem / year]/8760 hours / year = 11.4 mrem Whole Body in one hour Rounded off to 10 mrem
1 Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
/,s V
PAGE A-11 of 24 ABNORMAL RAI) LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY Dose assessment using MIDAS is based on the EPA-400 methodology, e.g., use of Total Effective Dose Equivalent (TEDE). This is somewhat different from whole body dose from gaseous emuents determined j
by ODAM methodology which forms the basis for the radiation monitor readings calculated in AUl in accordance with the generic methodology. The gaseous emuent radiation monitors can only detect noble gases. The contribution ofiodine's to TEDE could therefore only be determined either by: (1) utilizing MIDAS, or (2) gaseous emuent sampling. DAEC EAL 4 is written in terms of TEDE and the gaseous j
emuent radiation monitor readings are determined based on ODAM.
REFERENCES:
- 1. Offsite Dose Assessment Manual Section 6.1.2 and 7.1.2 Bases
- 2. Emergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and Protective Action s
- 3. Radiation Protection Calculation No. 95-001-C, Emergency Actions Levels Based on Emuent Radiation Monitors, January 24,1995
- 4. UFSAR Section 11.5, Process and Emuent Radiation Monitoring and Sampling Systems
- 5. EPA 400-R-92-001, Ahmual ofProtective Action Guides andProtective Actionsfor Nuclear Incidents
- 6. NUAfARC Afethodologyfor Development ofEmergency Action Levels NUAfARC/NESP-007 Revision 2 Questions and Answers, June 1993 im
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
V PAGE A-12 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY AA2 Major Damage to Irradiated Fuel or Loss of Water Level that lias or Will Result in the Uncovering ofIrradiated Fuel Outside the Reactor Vessel EVENT TYPE: Onsite Rad Conditions OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS:
(1 or 2 or 3 or 4)
- 1. < Valid < Site-specific > radiation monitor readings for the refuel floor area, fuel handling area, and the fuel bridge area.>
q
- 2. Report of Visual observation ofirradiated fuel uncovered.
(_)
- 3. Water Level less than (site-specific) feet for the Reactor Refueling Cavity that will result in Irradiated Fuel Ursovering.
i
- 4. Wate. Level less than (site-specific) feet for the Spent Fuel Pool < > that will result in Irradiated Fuel uncovering.
Valid means that the reading is from instrumentation determined to be operable in accordace with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results. Valid alarms are solely due to damage to irradiated fuel or loss of water level that has or will result in the uncovering ofirradiated fuel.
There are no significant deviations from the gencric EALs. Increased radiation levels can be detected by the local radiation monitors, in-p! ant radiological surveys, new fuel and spent fuel storage area radiation monitor alarms displayed on panel IC04B, fuel pool ventilation exhaust monitors, and by Standby Gas i
Treatment (SBGT) System automatic start. Appli,:able area radiation monitors include RT 9163, RT 9164, RT 9153, and RT 9178. These monitors are located in the north end of the refuel floor, the south end of the refuel floor, the new fuel vault area, and near the spent fuel pool, respectively.
Per ARP IC04B, the applicable area radiation monitor alanns actuate when radiation levels increase above Q
100 mR/hr in the spent fuel pool area or above 10 mR/hr in the other three areas of concem. If a valid actuation of these alarms were to occur, the refueling floor would be immediately evacuated. Thus, a report of a fuel handling accident with either valid actuation of the fuel area alarms on panel IC04B or with
~
l Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
O PAGE A-13 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY measured rediation levels in the spent fuel pool or north fuel area are used to address the generic concern consistent with DAEC design and procedures.
During preparation for reactor cavity flood up prior to entry into refuel mode, reactor vessel level instrument LI-4541 (WR GEMAC, FLOODUP) on control room panei IC04 is placed in service by I&C personnel connecting a compensating air signal after the reference leg is disconnected from the reactor head. Normal refuel water level is above the top of the span of this flood up level indicator. A valid on-scale indication (e.g., not due to loss of compensating air signal or other instrument channel failure) from this instrument can be used to determine uncontrolled loss of water level in the reactor cavity.
During refueling, the gates between the reactor cavity and the refueling cavity are removed and the spent p
fuel pool level indicator LI 3413 is used to monitor refueling water level. This measures the common
(
water level in the reactor cavity and the fuel pool. The bottom of the fuel transfer slot between the spent fuel pool and the reactor cavity is 16 feet above the bottom of the spent fuel pool. The top of the active fuel in the spent fuel storage racks is slightly less than 13 feet 9 inches above the bottom of the spent fuel pool.
Therefore, postulated failures which drain the reactor cavity through the reactor vessel cannot uncover fuel in the spent fuel storage racks. However, valid indication of spent fuel pool level less than 16 feet would indicate that spent fuel in the storage racks may potentially become uncovered.
F&RCHP 5 requires that upon a loss of water level situation, that the refueling crew on the refueling floor shall discharge any fuel assembly on the fuel grapple as follows:
if a fuel assembly is currently being withdrawn from a slot in the core or spent fuel pool, immediately e
reinsert it into that slot.
if a fuel assembly is being transferred and is still over or near the core, insert it into the closest available e
slot in the core.
If a fuel assembly is being transferred and is over or near the spent fuel pool, insert it into the closest available slot in the spent fuel racks.
Following these actions, the refueling floor is to be evacuated of all personnel. The DAEC EAL is written to address the generic concem that a spent fuel assembly was not fully covered by water. This can either be by visual observation of an uncovered spent fuel assembly or by trending fuel pool level in the control room if a spent fuel assembly could not be placed in a safe storage location specified by F&RCHP 5 as described above.
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC review) n\\
iG PAGE A-14 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY
REFERENCES:
- 1. Alarm Response Procedure (ARP) IC04B, Reactor Water Cleanup and Isolation
- 2. Technical Specification 3.9C, Spent Fuel Pool Water Level
- 3. Emergency Operating Procedures (EOP) Basis Document, Breakpoints for RC/L & L
- 4. Emergency Plan Implementing Procedure (EPIP) 3.1, Inplant Radiological Monitoring, Attachment 1, ARM Locations
- 6. Integrated Plant Operating Instruction (IPOl) 8, Outage and Refueling Operations
- 7. Fuel & Reactor Component IIandling Procedure (F&RCHP) 5, Procedure for Moving Core Components Between Reactor Core and Spent Fuel Pool, Within the Reactor Core, or Within the Spent
,m Fuel Pool
()
- 8. Bechtel Drawing C-492, Reactor Building - Reactor Well, Spent Fuel & Dryer-Separator Pool General Arrangement, Rev. 6
- 9. Bechtel Drawing C-493, Reactor Building - Spent Fuel Liner Plan Elevations and Details, Sheet 1, Rev.6
- 10. Holtec International Drawing No.1045, Rack Construction - Spent Fuel Storage Racks, Rev. 3 l 1.NUAblRC Afethodologyfor Development ofEmergency stcrion Levels NUAb1RC/NESP-007 Revision 2 Questions andifnswers, June 1993 OV
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC revieu) g(
PAGE A-15 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY i
AA3 Release of Radioactive Material or Increases in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or to Maintain Cold Shutdown EVENT TYPE: Onsite Rad Conditions OPERATING MODE APPLICABILITY: All
)
EXAMPLE EMERGENCY ACTION LEVELS:
(1 or 2)
- 1. Valid (site-specific) radiation monitor readings GREATER THAN (site-specific) values in areas f
requiring continuous occupancy to maintain plant safety functions < >
- 2. Valid (site-specific) radiation monitor readings GREATER THAN (site-specific) values in areas requiring infrequent access to maintain plant safety functions < >
Valid means that the reading is from instrmnentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.
l There are no significant deviations from the generic EALs. Per the UFSAR, the control room is the only j
area that is required to be continuously occupied to achieve and maintain safe shutdown following design basis accidents. DAEC EAL 1 is directly applicable to NUMARC EAL 1. However, the capability exists for plant shutdown from outside the main control room in the event that the control room becomes uninhabitable using remote shutdown panel IC388. DAEC EAL 2 is directly applicable to NUMARC EAL2.
The EC/OSS shoidd determine the cause of the increase in radiation levels and review other EifLs for upplicability. Expected increases in monitor readings due to controlled evolutions (such as lifting the steam dryer during refueling) do not result in emergency declaration. Nor should momentary increases due to events such as resin transfers or controlled movement of radioactive sources result in emergency d declaration. In-plant radiation level increases that would result in emergency declaration, are also unplanned, e.g., outside the limits established by an existing radioactive discharge permit.
l Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
PAGE A-16 of 24 ABNORMA L RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY
REFERENCES:
- 1. Alarm Response Procedure (ARP) IC04B, Reactor Water Cleanup and Isolation
- 2. Abnormal Operating Procedure (AOP) 913, Fire
- 3. Abnormal Operating Procedure (AOP) 914, Security
- 4. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room
- 6. Integrated Plant Operating Instruction (IPOI) 8, Outage and Refueling Operations
- 7. Emergency Plan Implementing Procedure (EPIP) 3.1, Inplant Radiological Monitoring
- 8. UFSAR Section 6.4, Habitability Systems
- 9. Bechtel Calculation DA-4, Project Number 265-002, Control Room Habitability,9/3/80 10.NUAfARC Afethodologyfor Development ofEmergency Action Levels NUAIARC/NESP-007 Revision 2 (j
Questions and Answers, June 1993 i
O m
Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu) gb PAGE A-17 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD l
CATEGORY l
1 ASI Site Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 < mrem TEDE> or 500 < mrem CDE>
Thyroid for the Actual or Projected Duration of the Release EVENT TYPE: Offsite Rad Conditions OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2 or 3 or 4)
- 1. A valid reading on < site-specific > monitors <for greater than 15 minutes which corresponds to an A
ofTsite dose of 100 mrem or 500 mrem Thyroid in an hour >.
.b
- 2. A valid reading sustained for 15 minutes or longer on perimeter radiation monitoring system greater than 100 mR/hr. [for sites having telemetered perimeter monitors].
- 3. Valid dose assessment capability indicates dose consequences greater than 100 < mrem TEDE> or 500
< mrem CDE> thyroid.
- 4. Field survey results indicate site boundary dose rates exceeding 100 < mrem >/hr expected to continue for more than one hour; or analyses of field survey samples indicate <CDE> thyroid of 500 < mrem > for one hour ofinhalation.
Valid means that the reading is from instrumentation detemlined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.
There are no significant deviations from the generic EALs.
DAEC's Meteorological Information and Dose Assessment System (MIDAS) was utilized to determine the KAMAN monitor limits. Eight separate combinations of release point, source term, meteorological j
conditions and equipment status were analyzed. Pathways considered were the offgas stack, the turbine building exhaust vent and a single reactor building exhaust vent. Multiple release points were not p
considered. In this same vein, it was assumed that only one of the three reactor building vents is on during d
the release.
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview) f-b PAGE A-18 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY i
The source temis used have been pre-loaded into MIDAS and are the default mixes associated with a loss of coolant accident (LOCA) and a control rod drop (CRD). The LOCA mix was used in conjunction with a release via the ofTgas stack while the CRD mix was used for releases via the turbine or reactor building vents. The source term for a release via the offgas stack is further impacted by the status of the standby gas treatment system. The status of that system was also taken into consideration.
Based of 1995 data (NG-96-0987), the atmospheric stability was classified as Pascal E 33% of the time.
Consequently, both classifications were evaluated. Based on the same report, the most common wind speeds were:
Pascal Class Altitude Speed (mnh)
D 156' 8 - 12 nQ D
33' 8 - 12 E
156' 8 -12 E
33' 4-7 Though the temperature setting has no impact on the MIDAS calculations, a value must be entered in order for the program to run. Consequently, the temperature was arbitrarily set at 50 F.
The rain estimate was set at zero, to eliminate any on site washout of radioactive material.
For the first MIDAS runs a ICi/cc concentration was assumed. The results of these runs were then normalized to the limits, thus generating a theoretical KAMAN limit. Additional MIDAS runs were made with these theoretical limite as input to verify the normalization process.
In addition to the total integrated dose, MIDAS calculates a peak whole body DDE rate resulting from the plume and a peak thyro'd CDE rate resulting from inhalation. Because the ASI and AGl KAMAN limits are to be based on a one hour exposure, establishing concentration limits so these peak values match the NUMARC limits is acceptable.
pO
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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview) i PAGE A-19 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY 4
l i
Site Area Emergency General Emergency initiating Condition ASI AGl Valid Turbine or Reactor Building ventilation rad monitor (KAMAN) reading for more than 15 0.06 Ci/cc 0.6 Ci/cc minutes above:
Valid OfTgas Stack ventilation rad monitor (KAMAN) reading for more than 15 minutes 40 pCi/cc 400 Ci/cc above:
The primary methodfor declaration is by means ofJose assessment using the MIDAS computer model.
Ilowever, if the monitor readings are sustained for longer than 15 minutes and the required dose Q assessments cannot be completed within this period, then the declaration must be made based on the valid 1 0 reading.
DAEC does not have a telemetered radiation monitoring system. As an attemative, DAEC uses valid field q
survey readings outside the site boundary greater than 100 mR/hr or greater than 500 mR/hr CDE Thyroid.
Dose assessment using MIDAS is based on the EPA-400 methodology, e.g., use of Total Effective Dose Equivalent (TEDE) and Committed Dose Equivalent (CDE) Thyroid. TEDE is somewhat different from whole body dose from gaseous effluents detennined by ODAM methodology which forms the basis for the radiation monitor readings calculated in AUI. These factors can introduce differences that are at least as large as those introduced by using TEDE versus whole body dose. The gaseous effluent radiation monitors can only detect noble gases. The contribution ofiodine's to TEDE and CDE Thyroid could therefore only be determined either by: (1) util zing the source term mixture in MIDAS, or (2) gaseous effluent sampling.
Therefore, DAEC EAL 4 is written in terms of TEDE and CDE Thyroid.
4 i
REFERENCES:
- 1. Offsite Dose Assessment Manual, Section 6.1.2 and 7.1.2, Bases
- 2. Emergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and Protective Action
- 3. Radiation Protection Calculation No. 95-001-C, Emergency Actions Levels Based on Eftluent Radiation Monitors, January 24,1995
- 4. Radiation Engineering Calculation No. 96-007-A, Determination of DAEC Radioactive Release V
Initiating Conditions for ASI & AGI Emergency Classifications, July 3,1996
- 5. UFSAR Section 11.5, Process and Effluent Radiation Monitoring and Sampling Systems
- 6. EPA 400-R-92-001, Manual ofProtective Action Guides and Protective Actionsfor Nuclear Incidents
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Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
O-PAGE A-20 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL, EFFLUENT EFFECTIVE DATE: TBD CATEGORY
- 7. NUMARC Methodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions andAnnvers, June 1993 I
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J Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview) b PAGE A-21 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY AGI Site Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity that Exceeds <1,000 mrem TEDE> or <5,000 mrem CDE> Thyroid for the Actual or Projected Duration of the Release < >
EVENT TYPE: Offsite Rad Conditions OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2 or 3 or 4)
- 1. A valid reading on < site-specific > monitors <for greater than 15 minutes which corresponds to en p
offsite dose of 1,000 mrem or 5,000 mrem Thyroid in an hour >.
V
- 2. A valid reading sustained for 15 minutes or longer on perimeter radiation monitoring system greater than 1,000 mR/hr. (for sites having telemetered perimeter monitors].
- 3. Valid dose assessment capability indicates dose consequences greater than 1,000 < mrem TEDE> or 5,000 < mrem CDE> thyroid.
- 4. Field survey results indicate site boundary dose rates exceeding 1,000 < mrem >/hr expected to continue for more than one hour; or analyses of field survey samples indicate <CDE thyroid > of 5,000 < mrem >
for one hour ofinhalation.
Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.
There are no significant deviations from the generic EAI.s.
DAEC's Meteorological Information and Dose Assessment System (MIDAS) was utilized to determine the KAMAN monitor limits. Eight separate combinations of release point, source term, meteorological conditions and equipment status were analyzed. Pathways considered were the ofTgas stack, the turbine building exhaust vent and a single reactor building exhaust vent. Multiple release points were not considered. In this same vein, it was assumed that only one of the three reactor building vents is on during
(
the release.
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
P AGE A-22 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY The source terms used have been pre-loaded into MIDAS and are the default mixes associated with a loss of coolant accident (LOCA) and a control rod drop (CRD). The LOCA mix was used in conjunction with a release via the ofTgas stack while the CRD mix was used for releases via the turbine or reactor building vents. The source term for a release via the offgas stack is further impacted by the status of the standby gas treatment system. The status of that system was also taken into considwtion.
Based of 1995 data (NG-96-0987), the atmospheric stability was classi;ied as Pascal E 33% of the time.
Consequently, both classifications were evaluated. Based on the same report, the most common wind speeds were:
Pascal Class Altitude Sneed (mnh)
D 156' 8-12 D
33' 8 - 12 E
156' 8 - 12 E
33' 4-7 Though the temperature setting has no impact on the MIDAS calculations, a value must be entered in order for the program to run. Consequently, the temperature was arbitrarily set at 50 F.
The rain estimate was set at zero, to eliminate any on site washout of radioactive material.
For the first MIDAS runs a ICi/cc concentration was assumed. The results of these runs were then normalized to the limits, thus generating a theoretical KAMAN limit. Additional MIDAS runs were made with these theoretical limits as input to verify the normalization process.
In addition to the total integrated dose, MIDAS calculates a peak whole body DDE rate resulting from the plume and a peak thyroid CDE rate resulting from inhalation. Because the ASI and AGl KAMAN limits are to be based on a one hour exposure, establishing concentratioa limits so these peak values match the NUMARC limits is acceptable.
O
Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevicw)
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V PAGE A-23 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY Site Area Emergency General Emergency Initiating Condition ASI AGl Valid Turbine or Reactor Building ventilation rad monitor (KAMAN) reading for more than 15 0.06 Ci/cc 0.6 pCi/cc j
minutes above:
Valid Offgas Stack ventilation rad monitor (KAMAN) reading for more than 15 minutes 40 Ci/cc 400 Ci/cc abwe:
The preferred methodfor declaration is by means ofdose assessment using the MIDAS comjmter model and is therefore is listed as DAEC EAL 4. However, ifthe monitor readings are sustainedfor longer than 15 minutes and the required Jose assessments cannot be completed within this period, then the declaration must be made based on the valid reading.
DAEC does not have a telemetered radiation monitoring system. As an attemative, DAEC uses valid field survey readings outside the site boundary greater than 1,000 mR/hr or greater than 5,000 mR/hr CDE Thyroid.
Dose assessment using MIDAS is based on the EPA-400 methodology, e.g., use of Total Effective Dose Equivalent (TEDE) and Committed Dose Equivalent (CDE) Thyroid. TEDE is somewhat different from whole body dose from gaseous effluents determined by ODAM methodology which forms the basis for the radiation monitor readings calculated in AUl. These factors can introduce differences that are at least as large as those introduced by using TEDE versus whole body dose. The gaseous ellluent radiation monitors can only detect noble gases. The contribution ofiodine's to TEDE and CDE Thyroid could therefore only be determined either by: (1) utlizing the source term mixture in MIDAS, or (2) gaseous effluent sampling.
Therefore, DAEC EAL 4 is written in terms of TEDE and CDE Thyroid.
REFERENCES:
- 1. OfTsite Dose Assessment Manual, Section 6.1.2 and 7.1.2, Bases
- 2. Emergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and Protective Action
- 3. Radiation Protection Calculation No. 95-001-C, Emergency Actions Levels Based on Effluent Radiation Monitors, Jamlary 24,1995 n
- 4. Radiation Engineering Calculation No. 96-007-A, Determination of DAEC Radioactive Release Initiating Conditions for ASI & AGl Emergency Classifications, July 3,1996
- 5. UFSAR Section 11.5, Process and Effluent Radiation Monitoring and Sampling Systems
- 6. EPA 400-R-92-001, Ahmual ofProtective Action Guides and Protective Actionsfor h'aclear incidents
.1 i
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Duane Arnold Energy Center.
i EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (forNRCreview) i i
PAGE A-24 of 24 5
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT l
EFFECTIVE DATE: TBD CATEGORY J
i l
l
- 7. NUAMRC MethodologyJbr Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions andAnswers, June 1993 l
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FISSION HARRIER TABLE IITI TIVEDATE:TBD INDICATORS FUEL CLAD BARRIER RCS BARRIER i
Loss Loss L Fuel damage assessment (PASAP 72) determines at L Valid drywell rad monitor reading above 5 R/hr after 1 least 5% fuel clad damage reactor shutdown
{
OR Fuel damage is indicated by any of the following:
Potentialloss-None RADIATION /
L Valid drywell rad monitor reading above 7E+2 R/hr OR CORE DAMAGE L Valid torus rad monitor reading above 3E+1 R/hr OR L Coolant activity above 300pCVgm DOSE EQUIVALENT l-131 PotentialLoss -None Loss Loss L RPV Level below A0 inches L RPV Level below 15 inches RPV LEVEL PotentialLoss PotentialLoss-None P RPV Level below 15 inches Loss - None PotentialLoss P RCS Leakage is above 50 GPM i
OR P Unisolable primary system leakage outside the drywa i
LEAKAGE None as indicated by area temps or ARMS Loss L Drywell pressure above 2 psig and not caused by a j loss of DW Cooling PRIMARY CONTAINMENT None ATMOSPHERE Any condition which in the EC/OSS's judgment indicatec Any condition which in thc EC/OSS's judgment indicates EC/OSS loss or potentialloss of the fuel clad barrier due to:
loss or potentialloss of the RCS barrier due to:
- Imminent barrier degradation Imminent barrier degradation
- Degraded fission barrier monitoring capability Degraded fission barrier monitonng capability IMMINENT No turnaround in safety system performance is expected and escalation to General Emergency conditions is expected within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> NOTE: Step f; for allindacators, move from left to right across table, marking all applicable "L's* and *P's' for each barrier, based on plantindications. TheX "L's* and "P's" marked on Barrier Table to the Logic Diagram (at right). "L's* and *P's'should be marked for each affected barrier (working top to bottom) 0@
Step 3, an *L* or ~P" marked for each associated barrier will constitute a Logic Iinput. When coincidence is met, then the EAL can be declared.
(/
L = Loss (of a fission product barrier)- A severe challenge to a fission product barrier exists such tnat the burier is considered incapable of performing its i P = Potentialloss (of a fission product barrier)- A challenge to a fission product barrier exists such that the barrier is considered degraded in its ability to B s.--
PRIMARY CONTAINMENT B ARRIER ONE BARRIER AFFECTED Loss - None L
P L
P L
P PotentialLoss f
P Valid drywell rad monitor reading above 3E+3 R/hr CLAD RCS CNTMT P Valid torus rad monitor te ding above 1E+2 R/hr E
l FUI OR UNUSUAL P Coro damage assessment determines at least 20%
is t EVENT o o fuel clad damage EMEMI I
T 1/2 fat ALERT Loss - None TWO BARRIERS AFFECTED r-QQff[f&$
PotentialLoss L
P L
P L
P P RPV Level below -40 inches hhh CLAD RCS C,NTMT I
I
~
L Frilure of both isolation valves and a downstream AlsO Avallable ort E
EIR Apertura Card p:thway to the envirornnent exists OR 2/3 L Unisolab:e primary system leakage outside the drywell FS1 cs indicated by area temps or ARMS 1
EMERGENCY SITE AREA OR L Primary containment venting performed per EOPs THREE BARRIERS AFFECTED o
PotentialLoss-None L
P L
P L
P Loss - None CLAD RCS CNTMT L R:pid unexplained decrease following initial increase l
l OR I
Drywell pressure response not consistent with LOCA
- o o o conditions EEEEll PotentialLoss P Torus pressure reaches 53 psig OR LOSSOF AT No LE AST 2 P Drywell or torus Ha CANNOT be determined to be BARRIERS?
below 6% AND Drywell or torus 0 CANNOT be 2
determined to be below 5%
YES FG1 GENERAL Any condition which in the EC/OSS's judgment indicates EMERGENCY loss or potential loss of the primary containment barrier due to:
o Imminent barrier degradation Op. Modes: Run, Startup, Hot S/D o
Degraded fission barrier monitoring capability step 2, tr n:cribe all 10 flowchirt.
if ty fur'.ction.
riorm its saf;ty function.
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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
'~)
PAGE F-1 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY FUI Any Loss or Any Potential Loss of < Primary > Containment < Barrier >
t EVENT TYPE: See Fission Barrier Table 4
OPERATING MODE APPLICABILITY: Run, Startup,llot Shutdown EXAMPLE EMERGENCY ACTION LEVELS:
i See the naion Barrier Table indicators discussed later in this section.
DAEC INFORMATION:
i3 V See the Fission Barrier Table indicators discussed later in this section. The entry conditions for this i
2 Initiating Condition are shown by the logic chart located to the right of the Fission Barrier Table.
REFERENCES:
See the Fission Barrier Table indicators discussed later in this section.
l
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I Duane Arnold Energy Center l
EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
PAGE F-2 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY FAI Any Loss or Any Potential Loss of Either Fuel Clad Or RCS < Barrier >
EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EXAMPLE EMERGENCY ACTION LEVELS:
See the Fission Barrier Table indicators discussed later in this section.
DAEC INFORMATION:
See the Fission Barrier Table indicators discussed later in this section. The entry conditions for this initiating Condition are shown by the logic chart located to the right of the Fission Barrier Table.
REFERENCES:
1 See the Fission Barrier Table indicators discussed later in this section.
O
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC review) 4 PAGE F-3 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY FS1 < Loss Or Potential Loss of Any Two Barriers >
EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EXAMPLE EMERGENCY ACTION LEVELS:
See the Fission Barrier Table indicators discussed later in this section.
i I
DAEC INFORMATION:
The entry conditions for this initiating Condition are shown by the logic chart located to the right of the Fission Barrier Table. DAEC uses " Loss Or Potential Loss of Any Two Barriers." This logic is simplified from the generic logic based on the following considerations:
- 1. Human Factors - It is easier to understand and to remember the escalation from Alert to Site Area
~
Emergency to General Emergency using the simpler logic.
- 2. Comnrehensiveness - A comparison was made of the combinations of barrier losses and potential j
losses corresponding to Site Area Emergency between the DAEC logic and the NUMARC/NESP-007 logic. All six generic barrier loss / potential loss combinations are addressed in the DAEC logic that j
l addresses 12 combinations of barrier loss / potential loss.
No sequences addressed by the NUMARC/NESP-007 logic are significantly affected by the simplified logic when applied to a BWR.
See the table below.
4
REFERENCES:
See the Fission Barrier Table indicators discussed later in this section.
l
Duane Arnold Energy Center i
EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevien) l l
PAGE F-4 of 27 FISSION PRODUCT IMRRIER DEGRADATION EFFECTIVE DATE: TBD l
CATEGORY l
COMPARISON OF DAEC FS1 BARRIER COMBINATIONS WITil NUMARC/NESP-007 FS1 BARRIER COMBINATIONS FUEL CLAD llARRIER RCS BARRIER PRIMARY CONTAINMENT BARRIER LOSS POTENTIAL LOSS POTENTIAL LOSS POTENTIAL LOSS LOSS LOSS 1.
D, N D, N 2.
D, N D, N 3-D D
4-D D
5.
D, N D, N 6-D, N D, N 7-D, N D, N 8.
D D
9-D D
10 D
D 11 D, N D, N i2 9
9 l
D - Barrier status addressed by DAEC simplified logic (Loss Or Potential Loss of Any Two Barriers)
,o N -Barrier status addressed by NUMARC/NESP-007 generic logic (Loss of BOTil Fuci Clad AND RCS
! (v)
OR Potential Loss of BOTil Fuel Clad AND RCS OR Potential Loss of EITIIER Fuel Clad OR RCS AND Loss of ANY Additional Barrier)
Duane Arnold Energy Center
,f 3 EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (/brNRCreview)
O PAGE F-5 of 27 FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY 1
FG1 Loss of Any Two Barriers AND Potential Loss of <the> Third Barrier l
EVENT TYPE: See Fission Banier Table OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EXAMPLE EMERGENCY ACTION LEVELS:
See the Fission Barrier Table indicators discussed later in this section.
DAEC INFORMATION:
p(
See the Fission Barrier Table indicators discussed later in this section. The entry conditions for this Initiating Condition are shown by the logic chart located to the right of the Fission Barrier Table.
REFERENCES:
See the Fission Barrier Table indicators discussed later in this section.
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4 I
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC review)
O)
PAGE F-6 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY
)
FISSION BARRIER: Fuel Clad 1
DAEC INDICATOR: Radiation / Core Damage GENERIC INDICATOR:
Drywell Radiation Monitoring i
LOSS - < Valid > Drywell Rad Monitor Reading GREATER THAN (site-specific) R/hr POTENTIAL LOSS - Not Applicable i
DAEC INFORMATION:
Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed j
i on the control panels, reports from plant personnel, coolant sampling or radiological survey results.
There is no significant deviation from the generic " loss" indicator. Per NUMARC/NESP-007, the (site-specific) reading is a value which indicates release into the drywell of reactor coolant with elevated activity corresponding to about 2% to 5% fuel clad damage. This activity level is well above that expected from iodine spiking. It is intended that determination ofbarrier loss be made whenever the indicator threshold is reached umtil such time that core damage assessment is performed, at which time direct use of containment rad monitor readings is no longer required.
As documented by NG-88-0966, General Electric performed a study to predict dose rate readings from fuel damage calculations for emergency planning. The calculations were performed to obtain ganuna ray dose rates at the locations of the containment atmospheric monitoring system radiation detectors in the drywell and torus locations for assumed releases of gap activity form the core. These calculations were based on
" nominal" estimates of fuel rod gap fission product inventory fractions, which are considered to be more appropriate for determining a minimum threshold reading than inventory assumptions found in the NRC Regulatory Guides. The Regulatory Guide inventory assumptions applicable to dose assessments are larger and therefore non-conservative for determination of this EAL threshold. Two separate cases were evaluated. In the first case, the released activity was assumed to be contained in the drywell atmosphere.
This case is considered representative of conditions following a line break in which activity is released directly into the drywell. In the second case, the released activity was assumed to be contained in the torus.
This could be applied for an event which results in vessel isolation and blowdown to the suppression chamber. The results for each case were provided for each case in the form of gamma ray dose rate versus time profiles for assumed releases of 100% and 20% of the gap activity from the core. The dose rate calculations were carried out independent of any specific information on details of construction or response
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevicu) v 1
PAGE F-7 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY l
3 characteristics of the detector systems. The figures show a drywell reading of about 2.9 x 10 R/hr or a 2
ton.s reading of about 1.1 x 10 R/hr associated with 20% gap release at two hours after shutdown. Scaling this down to 5% gap release:
Calculation of Drywell and Torus Monitor Readings Assuming 5% Gap Release 3
NG-88-0966 value 20% Gap Release at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for drywell = 2.9 x 10 R/hr 3
2 Drywell reading = 2.9 x 10 R/hr x [5 % / 20 %] = 7.25 x 10 R/hr, round off as 7 E+2 R/hr 2
NG 88-0966 value 20% Gap Release at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for torus = 1.1 x 10 R/hr O
2
()
Torus reading = 1.1 x 10 R/hr x [5 % / 20 %) = 2.75 x 10' R/hr, round off as 3 E+1 R/hr The results are rounded off for ease of reading the respective radiation monitors' scales. The two hour point was picked because it allows ample time for the Technical Support Center to be operational and core damage assessment to begin. These indicators correspond to about 2.5% gap release if they occur immediately after shutdown. Thus, the indicators address the 2%-5% fuel clad damage range of concem described by the generic guidance.
REFERENCES:
- 1. Office Memo NG-88-0966, G.E. Fuel Damage Documentation / Dose Rate Calculations,03/18/88 O
j
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevicu) pb PAGE F-8 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY l
FISSION BARRIER: Fuel Clad DAEC INDICATOR: Radiation / Core Damage GENERIC INDICATOR:
Primary Coolant Activity Level LOSS - Coolant activity GREATER THAN (site-specific) value POTENTIAL LOSS - Not Applicable j
l DAEC INFORMATION:
There is no significant deviation from the generic indicator. Consistent with the generic methodology, s
's DAEC uses a coolant activity value of 300 pCi/gm li3i equivalent. This value is well above that expected for iodine spikes and would indicate fuel clad damage has occurred.
REFERENCES:
i
- 1. Post Accident Sampling and Analysis Procedure (PASAP) 7.2, Fuel Damage Assessment i
yV
i Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (forNRCrevieu) qO PAGE F-9 of 27 FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: TBD a
CATEGORY FISSION HARRIER: Fuel Clad DAEC INDICATOR: Radiation / Core Damage GENERIC INDICATOR:
Other (Site-Specific) Indications LOSS - (Site-specific) as applicable POTENTIAL LOSS -(Site-specific) as applicable l
DAEC INFORMATION-i As a site-specific loss indicator, DAEC uses determination of at least 5% fuel clad damage, which is p) consistent with the containment rad monitor reading indicators described previously. This can
(,
determined from the appropriate fuel damage assessment procedures.
No other reliable indications of Fuel Clad " loss" or " potential loss" could be determined.
REFERENCES:
- 1. Post Accident Sampling and Analysis Procedure (PASAP) 7.2, Fuel Damage Assessment I
l
Duane Amold Energy Center jeg EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC revieu)
V PAGE F-10 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY FISSION BARRIER: Fuel Clad DAEC INDICATOR: RPV Level GENERIC INDICATOR:
Reactor Vessel Water Level LOSS - Level LESS THAN (site-specific) value POTENTIAL LOSS - Level LESS TIIAN (site-specific) value DAEC INFORMATION:
p There are no significant deviations from the generic indicators. The generic loss indicator is based on a
(/ (site-specific) value that corresponds to the minimum value to assure core cooling without further degradation of the fuel clad. DAEC uses the Minimum Steam Cooling RPV Water Level of-30 inches.
This is defined to be the lowest RPV water level at which the covered portion of the reactor core will j
generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1500 F. Consistent with the EOPs, an indicated RPV level below -30 inches that cannot be restored is used.
The potential loss indicator corresponds to the (site-specific) water level at the top of the active fuel (TAF).
Consistent with the EOPs, an indicated RPV level below 15 inches that cannot be restored is used.
REFERENCES:
- 3. Emergency Operating Procedure (EOP) Basis, Curves and Limits, C5, Minimum Steam Cooling RPV Water Level O
Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
J PAGE F-11 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DA,FE: TBD CATEGORY FISSION BARRIER: Fuel Clad DAEC INDICATOR: EC/OSS Judgment GENERIC INDICATOR-Emergency Director Judgment Any condition which in the judgment of the Emergency Director that indicates LOSS or POTENTIAL LOSS of the FUEL CLAD barrier DAEC INFORMATION:
n There is no significant deviation from the generic indicator. Per EPIP 7.1, Emergency Coordinator Duties,
(
) the Emergency Coordinator / Operations Shift Supervisor (EC/OSS) performs the emergency director function at DAEC. EC/OSS considerations for determining whether any barrier " Loss" or " Potential Loss" include imminent barrier degradation, degraded barrier monitoring capability, and consideration of dominant accident sequences.
\\
Imminent means that no turnaround in safety system performance is expected and that General Emergency conditions can be expected to occur within two hours. Imminent fission barrier degradation must be considered by the EC/OSS to assure timely declaration of a General Emergency and to better assure that offsite protect!ve scJons can be effectively accomplished. Degraded barrier monitoring capability from loss of/ lack of reliable indicators must also be considered by the EC/OSS when determining if a fission barrier loss or potential loss has occurred. This assessment should also include consideration for instrumentation operability, portable instrumentation readings, and offsite monitoring results. Dominant accident sequences can lead to loss of all Fission Barriers. Based on the IPE, the dominant accident sequences leading to core damage at DAEC include complete loss of 125 VDC, loss of decay heat removal, ATWS with failure of Standby Liquid Control, prolonged station blackout, and loss of offsite power with early HPCI/RCIC failure. The EC/OSS should also consult System Malfunction EALs, as appropriate, to assure timely emergency classification declaration.
REFERENCES:
- 1. Emergency Plan Implementing Procedure (EPIP) 7.1, Emergency Coordinator Duties D
- 2. Duane Arnold Energy Center Individual Plant Examination (IPE) November 1992 J
Duane Arnold Energy Center EMERGENCY ACTION L.EVEL BASES DOCUMENT Rev. 2 (for NRC revieu) n tO'
, PAGE F-12 of 27 FISSION PRODUCT HARRIER DEGRADATION EF."ECTIVE DATE: TBD CATEGORY FISSION HARRIER: RCS DAEC INDICATOR: Radiation / Core Damage GENERIC INDICATOR:
Drywell Radiation Monitoring LOSS - < Valid > Drywell Rad Monitor Reading GREATER THAN (site-specific) R/hr POTENTIAL LOSS - Not applicable DAEC INFORMATION:
Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications er has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, coolant sampling, or radiological survey results.
There is no significant deviation from the generic indicator. This loss indicator is based on conditions afler reactor shutdown to assure that it is not misapplied, i.e., to exclude readings due to N-16 effects which are typically 5 to 8 R/hr at full power conditions.
The (site-specific) value for this loss indicator corresponds to instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within Technical Specifications) into the drywell atmosphere. The reading will be less than that specified for the loss indicator for Radiation / Core Damage that applies to the Fuel Clad barrier. Thus, this indicator would be indicative of a RCS leak only. If the radiation monitor reading increased to that value specified by the Radiation / Core indicator applying to the Fuel Clad barrier, fuel damage would also be indicated.
As documented by NG-88-0966, General Electric performed a study to predict dose rate readings from fuel damage calculations for emergency planning. The calculations were performed to obtain gamma ray dose rates at the locations of the containment atmosphere monitoring system radiation detectors in the drywell and torus locations for assumed releases of gap activity form the core. These calculations were based on
" nominal" estimates of fuel rod gap fission product inventory fractions, which are considered to be more appropriate for determining a minimum threshold reading than inventory assumptions found in the NRC Regulatory Guides. The Regulatory Guide inventory assumptions applicable to dose assessments are O'v larger and therefore non-conservative for determination of this EAL threshold. Two separate cases were evaluated. In the first case, the released activity was assumed to be contained in the drywell atmosphere.
This case is considered representative of conditions following a line break in which activity is released directly into the drywell. In the second case, the released activity was assumed to be contained in the torus.
m
,_ _ _ _ m.
m
_m.
m Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (forNRCreview)
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PAGE F-13 of 27 FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY 1
l i
This could be applied for an event which results in vessel isolation and blowdown to the suppression chamber. The results for each case were provided for each case in the form of gamma ray dose rate versus time profiles for assumed releases of 100% and 20% of the gap activity from the core. The dose rate 4
calculations were carried out independent of any specific information on details of construction or response characteristics of the detector systems. The figures show a drywell reading of about 2.1 x 10 R/hr i
associated with a 100% gap release immediately after shutdown. Assuming 99.99% fuel clad integrity l
(0.01% gap release) and uniform dispersal of radionuclides into the drywell immediately after shutdown, a drywell monitor reading is calculated:
1 Calculation of Drywell Monitor Reading Assuming 0.01% Gap Release 4
NG-88-0966 value for 100% Gap Release at 0.01 minutes = 2.1 x 10 R/hr (2.1 x 10 ) R/hr x [(1 x 10-2 ) percent /100 percent] = (2.1) x 10" R/hr = 2.1 x 10 R/hr = 2 R/hr 4
3 To assure an indicator that is readily discernible on the drywell radiation monitor scale, DAEC uses a valid reading above 5 R/hr after reactor shutdown.
REFERENCES:
- 1. Office Memo NG-88-0966, G.E. Fuel Damage L)ocumentation/ Dose Rate Calculations,03/18/88
- 2. Technical Specification 3.2E, Drywell Leak Detection Instrumentation 1
i
- U
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview) q V
PAGE F-14 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY FISSION BARRIER: RCS DAEC INDICATOR: RPV Level GENERIC INDICATOR:
Reactor Vessel Water Level LOSS - Level LESS THAN (site-specific) value POTENTIAL LOSS - Not applicable DAEC INFORMATION:
There is no significant deviation from the generic indicator. This (site-specific) loss indicator corresponds to the water level at the top of the active fuel (TAF). Consistent with the EOPs, an indicated RPV level below 15 inches that cannot be restored is used.
REFERENCES:
- 1. Emergency Operating Procedures (EOP) Basis, Breakpoints N})
i
Duane Arnold Energy Center EMERLENCY. ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC revieu) q Y-.)
PAGE F-15 of 27 FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY l
i FISSION HARRIER: RCS DAEC INDICATOR: Leakage GENERIC INDICATOR:
RCS Leak Rate LOSS - < Valid > (site-specific) indication of Main Steamline Break POTENTIAL LOSS - RCS leakage GREATER THAN 50 GPM inside the drywell OR unisolable primary system leakage outside drywell as indicated by < valid > area temp or area rad monitor alarm DAEC INFORMATION:
0
(,/
Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.
There are no significant deviations from the generic potential loss indicators applying to RCS leakage and indications of unisolable primary system leakage. Please note that RCS leakage inside the drywell excludes Safity-Relief Valve (SRV) discharge through the SRV discharge piping into the torus below the water line. SRV lealage is addressed by SU5, RCS Leakage.
Unisolable primary system leakage outside the drywell includes leakage through portions of the main steam lines, portions of the Reactor Water Cleanup System (RWCU), and through the Scram Discharge Volumes (SDVs) detected per EOP 3. It is possible to have relatively small amounts ofleakage result in radiation monitor alarms, therefore it is treated as a potential loss of the RCS barrier and loss of the Primary Containment barrier (see the discussion under Primary Containment Leakage indicator).
DAEC does not use the generic " loss" indicator for main steam line break. NUMARC Methodologyfbr Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions and Answers, June 1993, discloses that the main steam line break with isolation does not have to be included as a fission barrier table indicator. This event can be appropriately classified in the System Malfunction Recognition Category. This event was classified as a RCS barrier loss indicator in the generic guidance because this event typically results in a puff release with dose consequences greater than 10 millirem whole body, i.e.,
g offsite dose consequences consistent with declaration of an Alert in accordance with AAl, Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times Radiological Technical Specifications for 15 Minutes or Longer. However, UFSAR Section 15.6.6, Table 15.6-1, Steam-Line Break - Radiological Effects for Puff Release at 47 Meters, Total Dose, shova a maximum
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC review)
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R PAGE F-16 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY l
dose of 0.58 mrem (5.8E-04 rem) passing cloud whole body dose using conservative assumptions.
Therefore, because this event at DAEC has dose consequences similar to those of AUl, Any Unplanned Release of Gaseous or Liquid Radioactivity to the Enviromnent that Exceeds 2 Times Radiological Technical Specifications for 60 Minutes or Longer, it has been added as an Unusual Event EAL in SUS, RCS Leakage.
REFERENCES:
- 1. Alarm Response Procedure (ARP) IC04B, Reactor Water Cleanup and Recirculation
- 2. Alarm Response Procedure (ARP) IC04C, Reactor Water Cleanup and Recirculation
- 3. Emergency Operating Procedure (EOP) 3, Secondary Containment Control
- 4. UFSAR Section 15.6.6, Loss-of-Coolant-Accident g~)
- 5. NUAMRC Methodologyfor Development ofEmergency Action Levels NUAMRC/NESP-007 Revision 2
(
Questions and Answers, June 1993 i
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l Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
PAGE F-17 of 27 I
FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY l
FISSION HARRIER: RCS DAEC INDICATOR: Primary Containment Atmosphere GENERIC INDICATOR:
Drywell Pressure LOSS - < Valid > Pressure <Readim> GREATER THAN (site-specific) psig POTENTIAL LOSS - Not applicable DAEC INFORMATION:
Valid means that the reading is from instrumentation determined to be opemble in accordance with the (g Technical Specifications or has been verified by other independent methods such as indications displayed i
on the control panels, reports from plant personnel, or radiological survey results.
There is no significant deviation from the generic indicator. The (site-specific) value for this loss indicator corresponds to the drywell high pressure ECCS initiation signal setpoint of 2.0 psig. DAEC also specifies that drywell cooling is operating to assure that the indicator is not misapplied to conditions that do not indicate RCS leakage into the drywell, i.e., the drywell pressure increase is not due to loss of drywell cooling.
DAEC uses a GE Mark I Containment. During reactor operation, with drywell cooling in operation and the
]
drywell inerted, the normal operating pressure in the drywell is between 0.5 and 1.0 psig. Analysis at the DAEC shows that a 50 gpm RCS leak would result in a 2 to 3 psig pressure rise over a six minute time period. Since a 2 psig rise would place DAEC above the ECCS initiation setpoint,( 2 psig) it is necessary to select the DAEC ECCS initiation setpoint of 2 psig to indicate an actual loss of the RCS. Drywell cooling is not isolated at the 2 psig ECCS initiation setpoint, therefor further pressure rise would be l
indicative of a RCS leak.
l l
REFERENCES:
l
- 1. Emergency Operating Procedures (EOP) Bases, Breakpoints
- 3. Emergency Operating Procedures (EOP)-2, Primary Containment Control G
1
i Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
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PAGE F-18 of 27 FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY J
FISSION HARRIER: RCS DAEC INDICATOR: Primary Containment Atmosphere GENERIC INDICATOR:
Emergency Director Judgment Any condition which in the judgment of the Emergency Director that indicates LOSS or POTENTIAL 4
LOSS of the RCS barrier DAEC INFORMATION:
There is no significant deviation from the generic EAL. Per EPIP 7.1, Emergency Coordinator Duties, the Emergency Coordinator / Operations Shift Supervisor (EC/OSS) performs the emergency director ftmetion at DAEC. EC/OSS considerations for determining whether any barrier " Loss" or " Potential Loss" include imminent barrier degradation, degraded barrier monitoring capability, and consideration of dominant accident sequences.
Imminent means that no turnaround in safety system performance is expected and that General Emergency conditions will occur within two hours. Imminent fissio.i barrier degradation must be considered by the j
EC/OSS to assure timely declaration of a General Emergency and to better assure that ofTsite protective actions can be effectively accomplished. Degraded barrier monitoring capability from loss of/ lack of reliable indicators must also be considered by the EC/OSS when determining if a fission barrier loss or potential loss has occurred. This assessment should also include consideration for instrumentation operability, portable instrumentation readings, and o1Tsite monitoring results.
Dominant accident sequences can lead to loss of all Fission Barriers. Based on the IPE, the dominant accident sequences leading to core damage at DAEC include complete loss of 125 VDC, loss of decay heat removal, ATWS with failure of Standby Liquid Control, prolonged station blackout, and loss of offsite power with early llPCI/RCIC failure. The EC/OSS should also consult System Malfunction EALs, as appropriate, to assure timely emergency classification declaration.
For the RCS barrier, the EC/OSS shmdd also consider safety-relief udres (SRVs) open or cycling. If an SRV is stuck open or is cycling and no other emergency conditions exist, an emergency declaration may not be appropriate. IIowever, ifthefuelis damaged and the SR Vis allowingfission products to escape into primary containment. a loss ofRCS should be determined as having occurred. The EC/OSS should also consult SUS, RCS Leakage, to determine if RCS leakrige oceeds the threshold required for declaradon of an Unusual Event.
f Duane Arnold Energy Center l
EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview) 1 PAGE F-19 of 27 FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: TBD l
CATEGORY e
REFERENCES:
j.
- 1. Emergency Plan Implementing Procedure (EPIP) 7.1, Emergency Coordinator Duties
- 2. Duane Amold Energy Cen+er Individual Plant Examination (IPE) November 1992
- 3. NUMARC Afethodologyfor Development ofEmergency Action Levels NUhfARC/NESP-007 Revision 2 l
Questions andAnswers, June 1993
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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
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PAGE F-20 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD i
CATEGORY FISSION BARRIER: RCS DAEC INDICATOR: None GENERIC INDICATOR:
Other (Site-Specific) Indications LOSS - (Site-specific) as applicable POTENTIAL LOSS -(Site-specific) as applicable DAEC INFORMATION:
3 Other indicators were also considered. No other reliable indicators of RCS barrier " loss" or " potential loss" gj could be determined.
REFERENCES:
None O
Cuane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
O, PAGE F-21 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY FISSION HARRIER: Primary Containment DAEC INDICATOR: Radiation / Core Damage GENERIC INDICATOR:
Significant Radioactive Inventory in Containment LOSS - Not applicable POTENTIAL LOSS - Containment Rad Monitor reading GREATER TilAN (site-specific) R/hr DAEC INFORMATION:
There is no significant deviation from the generic indicators. The " potential loss" (site-specific) indicator value corresponds to at least 20% fuel clad damage with release into the primary containment. This indicator corresponds to loss of both the Fuel Clad and RCS barriers with Potential Loss of the Primary Containment barrier, and would result in declaration of a General Emergency. The basis for the 20% fuel clad damage threshold is described under the 20% core damage assessment indicator. It is intended that determination ofbarrierpotentialloss he made whenever the indicator thresholdis reached until such time that core damage assessment is performed, at which time direct use ofcontainment rad monitor readings is no longer required.
l As documented by NG-88-0966, General Electric performed a study to predict dose rate readings from fuel damage calculations for emergency planning. The calculations were performed to obtain gamma ray dose rates at the locations of the containment atmospheric monitoring system radiation detectors in the drywell and torus locations for assumed releases of gap activity form the core. These calculations were based on
" nominal" estimates of fuel rod gap fission product inventory fractions, which are considered to be more appropriate for determining a minimum threshold reading than inventory assumpt'ons found in the NRC Regulatory Guides. The Regulatory Guide inventory assumptions applicable to dose assessments are larger and therefore non-conservative for determination of this EAL threshold. Two separate cases were evaluated. In the first case, the released activity was assumed to be contained in the drywell atmosphere.
This case is considered representative of conditions following a line break in which activity is released directly into the drywell. In the second case, the released activity was assumed to be contained in the torus.
This could be applied for an event which results in vessel isolation and blowdown to the suppression chamber. The results for each case were provided for each case in the form of gamma ray dose rate versus i
time profiles for assumed releases of 100% and 20% of the gap activity from the core. The dose rate calculations were carried out independent of any specific information on details of construction or response 3
characteristics of the detector systems. The figures show a drywell reading of about 2.9 x 10 R/hr and a torus reading of about 1.1 x 10 R/hr associated with 20% gap release at two hours af ter shutdown. These
l Duane Arnold Energy Center -
EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
PAGE F-22 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD l
CATEGORY 1
values are rounded to 3 E+3 R/hr and 1 E+2 R/hr, respectively. The two hour point was picked because it l
allows ample time for the Technical Support Center to be operational and core damage assessment to begin.
i
REFERENCES:
I I
- 1. Office Memo NG-88-0966, G.E. Fuel Damage Documentation / Dose Rate Calculations,03/18/88 l
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Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevien)
O PAGE F-23 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY FISSION BARRIER: Primary Containment DAEC INDICATOR: Radiation / Core Damage GENERIC INDICATOR:
Other (Site-Specific) Indications LOSS - (Site-specific) as applicable POTENTIAL LOSS - (Site-specific) as applicable DAEC INFORMATION:
As a site-specific " potential loss" indicator, DAEC uses determination of at least 20% fuel clad damage,
() which is consistent with the level of fuel damage indicated by the drywell and torus radiation monitor readings above. This can be determined using appropriate fuel damage assessment procedures. Regardless ofwhether primary containment integrity is challenged, it is possiblefor sigmficant radioactivity within the primary containment to result in EPA PAG plume exposure levels being exceeded even assuming that the primary containmen; is within technical specification allowable leakage rates. With or without primary containment challenge, however, a major release of radioactivity requiring off-site protective actions from core damage is not possible unless a major failure of the fuel clad barrier allows radioactive material to be released from core into the reactor coolant. NUREG-1228 indicates that such conditions do not exist when the amount of fuel clad damage is less than 20%.
Other indicators were also considered. No other reliable indicators for Primary Containment " loss" or
" potential loss" could be determined.
REFERENCES:
- 1. Post Accident Sampling and Analysis Procedure (PASAP) 7.2, Fuel Damage Assessment
- 2. NUREG-1228, Source Term Estimations During incident Response to Severe Nuclear Power Plant Accidents, October 1988 4
V
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC review) c.
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PAGE F-24 of 27 FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY FISSION BARRIER: Primary Containment DAEC INDICATOR: RPV Level GENERIC INDICATOR:
Reactor Vessel Water Level LOSS -Not applicable POTENTIAL LOSS
<RPV> level less than (site-specific) value and <no injection source is available>
DAEC INFORMATION:
O) The underlying concem for this indicator is a threshold that represents significant uncoverin
(
and imminent core damage. Imminent means that no tumaround in safety system performance would be expected and that General Emergency conditions would be expected within two hours.
Consistent with the underlying concem, the DAEC indicator addresses conditions where the water level is below the Minimum Zero-Injection RPV Water Level of-40 inches with no injection source available.
The Minimum Zero-Injection RPV Water Level is defined to be the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any fuel clad temperature in the uncovered portion of the core from exceeding 1800 F. The Minimum Zero-Injection RPV Water Level is utilized to preclude significant fuel clad damage and hydrogen generation for as long as possible when no sources of RPV makeup water are available.
Thus, for RPV water level below -40 inches, if no source of injection water was available, water levels would continue to decrease and the fuel clad temperature would be expected to continue to rise. Due to large uncertainties in severe accident progression, it should be assumed that severe core melt is imminent if this condition were to occur. It would not be acceptable to delay the declaration of the General Emergency and issuance of Protective Action Recommendations beyond this point.
REFERENCES:
- 1. Emergency Operating Procedure (EOP) Bases Document, Curves and Limits
- 3. NUMARC Methodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions and Answers, June 1993
l Duane Arnold Energy Center p
EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
U PAGE F-25 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY FISSION BARRIER: Primary Containment DAEC INDICATOR: Leakage GENERIC INDICATOR:
Containment Isolation Valve Status After Containment Isolation Signal LOSS - Failure of both valves in any one line to close AND downstream pathway to the environment exists OR Intentional venting per EOPs DIl unisolable primary system leakage outside drywell as indicated by < valid > area temp or area rad alarm POTENTIAL LOSS - Not applicable DAEC INFORMATION:
Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.
The " loss" indicators used at DAEC directly correspond to the generic indicators. Venting of the primary containment can be performed in accordance with EOP 2 irrespective of the offsite radioactivity release rate that will occur and by defeating isolation interlocks as necessary. The consequences of not doing so may be the loss of primary containment integrity, core damage, and an uncontrolled radioactive releace much greater than might otherwise occur. Primary containment venting is performed only as necessary to reduce and t':en maintain torus pressure below the Primary Containment Pressure Limit (PCPL) of 53 psig.
Unisolable primary system leakage outside the drywell includes leakage through portions of the main steam lines, portions of the Reactor Water Cleanup System (RWCU), and through the Scram Discharge Volumes (SDV's) detected per EOP 3. It is possible to have relatively small amounts ofleakage result in radiation monitor alarms, therefore it is treated as a " potential loss" of the RCS (see the discussion under RCS Barrier Leakage indicator) and " loss" of the Primary Containment.
REFERENCES:
- 1. Emergency Operating Procedure (EOP) 2, Primary Containment Control
- 2. Emergency Operating Procedure (EOP) 3, Secondary Containment Control
- 3. Emergency Operating Procedures (r )P) Bases, Breakpoints
l l
i Duane Arnold Energy Center
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EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
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%d PAGE F-26 of 27 l
FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE:,fBD CATEGORY FISSION BARRIER: Primary Containment DAEC INDICATOR: Primary Containment Atmosphere GENERIC INDICATOR:
Drywell Pressure LOSS - Rapid unexplained decrease following initial increase OR Drywell pressure response not consistent with LOCA conditions POTENTIAL LOSS - (site-specific) PSIG OR explosive mixture exists DAEC INFORMATION:
There are no significant deviations from the generic indicators. The " loss" indicators used at DAEC directly correspond to the generic indicators.
The first " potential loss" (site-specific) indicator is torus pressure of 53 psig, which is the Primary Containment Pressure Limit (PCPL) used in the EOPs. The second " potential loss" indicator is based on determination of explosive mixture in accordance with the EOPs. DAEC EOPs require control of drywell and torus atmosphere gas concentrations to less than 6% H and less than 5% O to assure that an explosive 2
2 mixture does not exist. This " potential loss" indicator is written to be consistent with the EOPs.
REFERENCES:
i
- 1. Emergency Operating Procedure (EOP) 2, Primary Containment Control j
- 2. Emergency Operating Procedure (EOP) PCH - Primary Containment Hydrogen i
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Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
O PAGE F-27 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY FISSION BARRIER: Primary Containment DAEC INDICATOR: EC/OSS Judgment GENERIC INDICATOR:
Emergency Director Judgment Any condition which in the judgment of the Emergency Director that indicates LOSS or POTENTIAL LOSS of the RCS barrier DAEC INFORMATION:
There is no significant deviation from 6e generic indicator. Per EPIP 7.1, Emergency Coordinator Duties, 7
the Emergency Coordinator /Operationn Shift Supervisor (EC/OSS) performs the emergency director function at DAEC. EC/OSS consideratians for determining whether any barrier " Loss" or " Potential Loss" include imminent barrier degradation, degraded barrier monitoring capability, and consideration of dominant accident sequences.
hnminent means that no turnaround in safety system performance is expected and General Emergency conditions will occur within two hours. Imminent fission barrier degradation must be considered by the EC/OSS to assure timely declaration of a General Emergency and to better assure that offsite protective actions can be effectively accomplished. Degraded barrier monitoring capability from loss of/ lack of reliable indicators must also be considered by the EC/OSS when detennining if a fission barrier loss or potential loss has occurred. This assessment should also include consideration for instrumentation operability, portable instrumentation readings, and offsite monitoring results.
Dominant accident sequences can lead to loss of all Fission Barriers. Based on the IPE, the dominant accident sequences leading to core damage at DAEC include complete loss of 125 VDC, loss of decay heat removal, ATWS with failure of Standby Liquid Control, prolonged station blackout, and loss of offsite power with early IIPCI/RCIC failure. The EC/OSS should also consult System Malfunction EALs, as appropriate, to assure timely emergency classification declaration.
REFERENCES:
- 1. Emergency Plan Implementing Procedure (EPIP) 7.1, Emergency Coordinator Duties O
- 2. Duane Arnold Energy Center Individual Plant Examination (IPE) November 1992 Y
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EMERGENCY PLAN IMPLEMENTING PROCEDURE No. EPIP - 1.1 Rev.2 PAGE 1 of I (ForNRC Review)
IIAZARDS AND OTIIER CONDITIONS AFFECTING PLANT SAITTY HTTICTIVE DATE 'mD EVENT TYPE UNUSUAL EVENT ALERT SITE AREA EMERGENCY l
GENERAL EMERGENCY
{
HU1 HA1 Natural and Destructive Phenomena Natural and Destructive Phenomena Affecting the Protected Area Affecting the Plant Vital Area Earthquake detected per AOP 901 Earthquake peak horizontal acceleration Safe Shutdown Areas Earthquake.
above t 0.06 Gravity.
OR OR Categofy Area Report of tomado toucNng down within plant Report of tomado striking plant vital area.
Electrical Switchyard,1G31 DG and Day Tank Rooms, o '*"'"Y*#
P' '**'# ***
on Power 1G21 DG and Day Tank Rooms, Assessment by the mntrol room that an event Report to contml room of damage affecting Battery Rooms, Essential Switchgear Rooms, has occurd safe shutdown areas.
Cable Spreading Room OR OR Vehicle crash into plant structures or systems Vehicle crash affecting plant vital areas.
Heat Sink /
Torus Room, inteke Structure, Pumphouse within protected area boundary.
Coo [ ant Supply OR OR Containment Drywell, Torus NATURAL Report of ar, unanttipated exploson within the Sustained wind speed above 95 MPH.
DISASTERS protected area boundary resutting in visible Emergency NE, NW, SE Comer Rooms, HPCI Room, d mage to stmetwes or equipn ent Systems RCIC Room, RHR Valve Room, North CRD OR OR Turbine failure resulting in casing penetration Missiles affecting safe shutdown areas.
Area, South CRD Area or damage to turtine or generator seats.
Other Control Building, Remote Shutdown Panel ea, Panet M Area, EM Room River level above 757 River level above 767 t.
OR OR Any area water level above Max Normal Water level above Max Safe Operating Limit Operating Limit.
in 2 or more areas AND Reactor shutdown is required.
Water Level Operating Limits OR OR Max Normal Max Safe Rrver level below 725 feet 6 inches.
River lever below 724 feet 6 inches.
Room Area Indicator Operating Operating Op. Modes: ALL Op. Modes: ALL Limit (inches)
Limit (inches)
HU2 I
HA2 HPCI Room Area LI3768 6
24 Fire Within Safe Shutdown Areas Not Fire Affecting the Operability of Plant Safety RCIC Room Area LI3769 G
18 Extinguished Within 15 Minutes of Systems Required to Establish or Maintain Detection Safe Shutdown A RHR Comer LI3770 6
23 Fire in buildings or areas contiguous to any of Fire or explosion in any of the following areas:
Reactor, turtine, control, admWsecunty B RHR Comer LI3771 6
23 the fonowing areas not extinguished within 15 Intake structure Room NW Area minutes of control room notrreaton or FIRE venfication of a control room alarm:
Pump house Torus Area LI3772 12 24 Reactor, turtine, control, admin'secunty AND Intake structure Affected system parameter indications show Pump house degraded performance or plant personnel report visible damage to pernanent structures or equipment within the specified area.
Systems & Equipment of Concem 0" "$*' #*
U" "[ **
=
an n D ell / Torus)
RHR/ Core Spray /SRV's Release of Toxic or Flammable Gases Release of Toxic or Flammable Gases Deemed Detrimentat to Safe Operation of Within a Facility Structure Which HPCl/RCIC the Plant Jeopardizes Operation of Systems RHRSW/ River Water /ESW
=
Required to Maintain Safe ogwations or to
M Offsite AC Power e
OTHER Toxic or flammable gas release affecting Toxic cr flammable gas making safe shtf.down e Instrument AC HAZARDS AND normal operatiort areas untnhatAta orinaccessible.
FAILURES OR e DC Power Report by local, county cr State official for Remote Shutdown Capability e
potential evacuation of site personnel based on offsite event.
Op Modes: ALL Op. Modes: ALL HU4 HA4 HSf HGf Confirmed Security Event Which Indicates Security Event in a Plant Protected Area Security Event in a Plant Vital Area Security Event Resulting in Loss of a Potential Degradation in the Level of Safety of the Plant Ablity to Reach and Maintain Cold Shutdown Suspected sabotage device descovered within Intrusion into plant protected area by a hostile intrusion into plant vital area by a hostile Loss of physical control of the Control SECURITY plant protected area and outside plant vital force.
force.
Room.
area.
OR OR OR OR Suspected sabotage device discovered in plant Sabotage device discovered in the plant Sabotage device discovered in the plant vital loss of physical control of remote shutdown switchyard.
protected area.
area.
capability.
Op. Modes: ALL Op. Modes: ALL Op. Modes: ALL Op. Modes: ALL HA5 HS2 ContrS Room Evacuation Has Been Control Room Evacuation Has Been Initiated initiated and Plant Control Cannot Be None Established None CONTROL ROOM Control room evacuation initiated per Control room has been evacuated AND EVACUATION AOP 915. Shutdown Outside Control Room.
control of plant from Remote Shutdown Panet 1C388 NOT estabhshed within 20 minutes.
Op. Maria =: ALL Op. Mc-h_ ALL HU5 HA6 HS3 HG2 Other Conditions Existing Which in the Other Conditions Existing Which in the Other Conditions Existing Which in the Other Conditions Existing Which in the Judgment of the EC/OSS Warrant Judgment of the EC/OSS Warrant Judgment of the EC/OSS Warrant Judgment of the EC/OSS Warrant
~
Declarationof anUnusualEvent Declaration of an Alert Declaration of a Site Area Ew=Wy Declaration of a General EW=vcucy Other conditons exist which in the judgment of Other conditions exist which in the judgment of Other conditions extst which in the judgment Other conditions exist which in the judgment O
the EC/OSS indicate potential degradation of the EC/OSS indicate that plant systems may of the EC/OSS Indicate actual or hkety major of tne ECf0SS indicate EITHER:
thelevelof safetyof the plant.
be degraded and that increased monitoring of failures of plant functions needed for Actual or imminent sutstantialcore EC/OSS plant functions is warranted.
protection of the public.
degradation with potential for loss of JUDGMENT cantainment.
Potential for uncontrolled radionuchde releases which can reasonably be expected to exceed EPA PAG plume exposure levels outside the site boundary.
Op. Modes: ALL Op. Modes: ALL Op. MA. ALL Op. Modes: ALL O
M T>
C$t.
E>
0m2 n
en b li tt)
U B if THH RE OCm R$
TO
=
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l
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IIAZARDS AND OTIIER CONDITIONS AFFECTING PLANT SAFETY CATEGORY I
O
l Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
Y PAGE H-1 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY HUI Natural and Destructive Phenomena Affecting the Protected Area
)
EVENT TYPE: Natural Disasters, Other Hazards and Failures OPERATING MODE APPLICABILITY: All j
EXAMPLE EMERGENCY ACTION LEVELS:
(1 or 2 or 3 or 4 or 5 or 6 or 7)
- 1. (Site-Specific) method indicates felt earthquake.
- 2. Report by plant personnel of tomado striking within protected area boundary.
- 3. Assessment by the control room that an event has occurred.
o
- 4. Vehicle crash into plant structures or systems within protected area boundary.
V
- 5. Report by plant personnel of an unanticipated explosion within the protected area boundary resulting in visible damage to permanent structures or equipment.
- 6. Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.
- 7. (Site-Specific) occurrences.
There are no significant deviations from the generic EALs. EAL 1 addresses earthquakes that are detected in accordance with AOP 901. For DAEC, a minimum detectable earthquake that is indicated on panel IC35 is an acceleration greater than 0.01 Gravity. DAEC EAL 2 addresses report of a tomado striking within the protected crea or within the plant switchyard. DAEC EAL 3 allows for the control room to determine that and event has occurred and take appropriate action based on personal assessment as opposed to verification. No attempt is made to assess the actual magnitude of the damage. Such damage can be due to collision, tomadoes, missiles, or any other cause. Damage can be indicated by report to the control room, physical observation, or by Control Room / local control station instrumentation. Such items as scorching, cracks, dents, or discoloration of equipment or structures required for safe shutdown are addressed by this EAL. DAEC EAL 4 addresses a vehicle (automobile, aircraft, forklift, truck or train) crash that may potentially damage plant structures containing functions and systems required for safe shutdown of the plant. This does not include vehicle crashes with each other or damage to office or warehouse structures. Escalation to Alert under HAl would occur if damage was suflicient to affect the c
ability to achieve or maintain safe shutdown, e.g., damage made required equipment inoperable or structural damage was observed such as bent supports or pressure boundary leakage.
DAEC EAL 5 addresses explosions within the protected area. As used here, an explosion is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts
Duane Amold Energy Center p
EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevien)
PAGE H-2 of 25 IIAZARDS AND OTilER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY 1
4 significant energy to near-by structures or equipment. Damage can be indicated by report to the control I
room, physical observation, or by Control Room / local control station instrumentation. Such items as 1
scorching, cracks, dents, or discoloration of equipment or structures required for safe shutdown are i
addressed by this EAL. The EC/OSS needs to consider the security aspects of the explosion, if applicable.
DAEC EAL 6 addresses turbine failure causing observable damage to the turbine casing or to the seals of the generator.
EALs 7 through 9 address site-specific occurrences of concem. These concerns include external flood I
water levels, internal flooding, and low river water level affecting the ultimate heat sink. DAEC EAL 7 addresses the observed effects of flooding in accordance with AOP 902. Plant site finished grade is at elevation 757.0 ft Personnel doors and railroad and truck openings at or near grade would require
- q protection in the event of a flood above elevation 757.0 ft Therefore, EAL 7 uses a threshold of flood
- V water levels above 757.0 ft DAEC EAL 8 addresses internal flooding can be due to system malfunctions, component failures, or repair activity mishaps (such as failed freeze seal) that can threaten safe operation of the plant. Therefore, this EAL is based on a valid indication that the water level is higher than the maximum normal operating limits.
l The Maximum Normal Operating Limits are defined as the highest values of the identified parameter l
expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly. Exceeding these limits is an entry condition into EOP 3, Secondary Containment Control and may be an indication that water from a pdmary system is discharging into secondary containment. Exceeding the maximum nonnal operating limit is interpreted as a potential degradation in the level of the safety of the plant and is appropriately treated as an Unusual Event emergency classification. The rnaximum normal operating water level limits are taken from AOP 902 and EOP 3 and are shown in the table below:
Maximum Operating Limits - Water Levels Affected Location Indicator Maximum Normal OL Maximum Safe OL HPCI Room Area LI3768 6 inches 24 inches 3
RCIC Room Area LI3769 6 inches 18 inches A RHR Corner Room SE Area L13770 6 inches 23 inches B RHR Corner Room NW Area LI 3771 6 inches 23 inches Torus Area L13772 12 inches 24 inches EAL 9 addresses the effects of low river water level. The intake structure for the safety-related water supply systems (river water, RHR service water, and emergency service water) is located on the west bank
Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevien) 1
'd PAGE H-3 of 25 l
IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY 1
of the Cedar River. An overflow-type barrier across the river was designed and constructed in accordance with Seismic Category I criteria to intercept the streambed flow and divert it to the intake structure. This makes the entire flow of the river available to the safety-related water supply systems. A minimum flow of 13 cubic feet per second (cfs) from a minimum 1000-year river flow of 60 cfs must be diverted. The top of the barrier wall is at elevation 725 ft 6 in. River water level below this level represents a potential degradation in the level of safety of the plant and is addressed by EAL 9.
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 901, Earthquake
- 2. Abnormal Operating Procedure (AOP) 902, Flood
- 3. Abnormal Operating Procedure (AOP) 903, Tomado 7
(
- 4. Emergency Operating Procedure (EOP)-3, Secondary Containment Control
- 5. EOP Basis Document, EOP-3, Secondary Containment Control
- 6. UFSAR Chapter 3, Design of Structures, Components, Equipment, and Systems
- 7. Bechtel Drawing BECH-M017, Equipment Location - Intake Structure Plans at Elevations, Rev. 6 ym 4
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
PAGE H-4 of 25 IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY HU2 Fire Within Protected Area BoundaryNot Extinguished Within 15 Minutes of Detection EVENTTYPE: Fire OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVEL:
- 1. Fire in buildings or areas contiguous to any of the following (site-specific areas) areas not extinguished 1
within 15 minutes of control room notification or verification of a control room alarm:
(Site-specific) list i
There is no significant deviation from the generic EAL. The purpose of this EAL is to address the magnitude and extent of fires that may be potentially significant precursors to damage to safety systems.
This includes such items as fires within the administration building, and security building (buildings contiguous to the reactor building, turbine building and control building), yet, excludes fires in the warehouse or construction support center, waste-basket fires, and other small fires of no safety consequence.
Per AOP 913, the location of a fire can be determined by observing XL3 alarm messages, Zone Indicating Unit [ZiU) alarms, or fire annunciators on panels IC40 and IC40A. The location of a fire can also be j
determined by verbal report of the person discovering the fire. Verification of the alarm in this context means those actions taken to determine that the control room alarm is not spurious.
REFERENCES:
~
- 1. Abnormal Operating Procedure (AOP) 913, Fire
- 2. Abnormal Operating Procedure (AOP) 914, Security (q)
i a
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EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevicu)
PAGE II-S of 25 IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY HU3 Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant i
EVENT TYPE: Other llazards and Failures OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS:
(1 or 2) i
- 1. Report or detection of toxic or flammable gases that could enter within the site area boundary in amounts that can affect normal operation of the plant.
/
- 2. Report by Local, County or State Officials for potential evacuation of site personnel based on offsite v
event.
There is no significant deviation from the generic EALs. This IC is based on releases in concentrations within the site boundary that will affect the health of plant personnel or affecting the safe operation of the plant with the plant being within the evacuation area of an offsite event (i.e., tanker truck accident releasing toxic gases, etc.) The evacuation area is as determined from the DOT Evacuation Tables for Selected llazardous Materials, in the DOT Emergency Response Guide for liazardous Materials.
For the purposes of this IC, CO (such as is discharged by the fire suppression system) is not toxic. CO 2
2 can be lethal if it reduces oxygen to low concentrations that are immediately dangerous to life and health (IDLll). CO discharge into an area is not basisfor emergency classification under this IC unless: (1) 2 ifccess to the afected area is required, and (2) CO concentration results in conditions that make the area 2
uninhabitable or inaccessible (i.e., IDLH).
REFERENCES:
- 1. UFSAR. Section 2.2, Nearby Industrial, Transportation, and Military Facilities
- 2. UFSAR Section 6.4,liabitability Systems
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu) 7 PAGE H-6 of 25 I
IIAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY i
HU4 Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant EVENTTYPE: Security OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS:
(1 or 2)
- 1. Bomb device discovered within plant Protected Area and outside the plant Vital Area.
- 2. Other security events as determined from (site-specific) Safeguards Contingency Plan.
There is no significant deviation from the generic EALs. Security events which do not represent at least a potential degradation in the level of safety of the plant are reported under 10 CFR 73.71 or in some cases under 10 CFR 50.72. The term " suspected sabotage device" is used in place of " bomb device" for consistency with the DAEC Safeguards Contingency Plan.
Other (site-specific) security events of concem at DAEC include discovery of a suspected sabotage device in the plant switchyard, which is located outside the protected area.
Suspected sabotage devices discovered within the plant Vital Area would result in escalation via other Security Event ICs.
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 914, Security Events C)
V'
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC revien)
PAGE 11-7 of 25 IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY HU5 Other Conditions Existing Which in the Judgment of the <EC/OSS> Warrant Declaration of an Unusual Event EVENT TYPE: EC/OSS Judgment OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVEL:
- 1. Other conditions exist which in the judgment of the En'ergency Director indicate a potential degradation of the level of safety of the plant.
[d DAEC EALINFORMATION:
There is no significant deviation from the generic EAL.
Per EPIP 7.1, the Emergency Coordinator / Operations Shin Supervisor (EC/OSS) is the title for the emergency director function at DAEC. The EAL addresses conditions that fall under the Notification of Unusual Event emergency classification description contained in NUREG-0654, Appendix 1 that is retained under the generic methodology.
REFERENCES:
- 1. Emergency Plan Implementing Procedure (EPIP) 7.1, Emergency Coordinator Duties
- 2. NUREG-0654/ FEMA-REP-l, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Revision 1, 0ctober 1980, Appendix 1 OV
1 Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
PAGE H-8 of 25 j
llAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY HA1 N.itural and Destructive Phenomena Affecting the Plant Vital Area EVENT TYPE: Natural Disasters, Other Hazards and Failures OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS:
(1 or 2 or 3 or 4 or 5 or 6 or 7)
L (Site-Specific) method indicates Seismic Event greater than Operating Basis Earthquake (OBE).
- 2. Tornado or high winds striking plant vital areas: Tornado or high winds greater than (site-specific) mph
]
strike within protected area boundary.
i
- 3. Report of any visible structural damage on < site-specific structures >
3
- 4. (Site-Specific) indications in the control room.
- 5. Vehicle crash affecting plant vital areas.
- 6. Turbine failure generated missiles result in any visible structural damage to or penetration of any of the following < site-specific areas >
- 7. (Site-Specific) occurrences.
There are no significant deviations from the generic EALs. For the events ofconcern here, the key issue is not the wind speed, earthquake intensity, etc., but whether there is resultant damage to equipment or structures required to achieve or maintain safe shutdown, regardless of the cause. Determination of damage affecting the ability to achieve or maintain safe shutdown can be indicated by reports to the control room, physical observation or by Control Room / local control station instrumentation.
EAL 1 addresses OBE events that are detected in accordance with AOP 901. For DAEC, the OBE is associated with a peak horizontal acceleration of
- 0.06 Gravity. DAEC EAL 2 addresses report of a tomado striking a plant vital area. DAEC EAL 3 addresses a report to the control room of damage affecting safe shutdown areas. The reported damage can be from tomadoes, high winds, flooding, missiles, collisions, or any other cause.
DAEC EAL 4 addresses vehicle (automobile, aircraft, forklift, truck or train) confirmed crashes affecting plant vital areas. This does not include vehicle crashes with each other or damage to office or warehouse structures. DAEC EAL 5 addresses sustained high wind speeds as measured by the 33-Foot or 156-Foot elevations on the Meteorological Tower. Sustained wind speed means the baseline wind speed measured by meteorological tower that does not include gusts. The design basis wind speed is 105 miles per hour.
l Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
PAGE H-9 of 25 IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY i
j Ilowever, the meteorological instrumentation is only capable of measuring wind speeds up to 100 miles per hour. nus the alert level for sustained high wind speed,95 miles per hour, is selected to be on-scale for j
the meteorological instrumentation and to conservatively account for potential measurement errors. DAEC EAL 6 addresses missiles affecting safe shutdown areas. Such missiles can be from any cause, e.g.,
tornado-generated; turbine, pump or other rotating machinery catastrophic failure; or generated from an exp'osion.
Per AOPs 913 and 914, the following areas are identified as safe shutdown areas and are shown on the EAL tables. This table is dispkged as an aid to the Emergency Coordinator in determining appropriate 1
areas ofconcern.
i p
Safe Shutdown Areas Q
Category Area Electrical Switchyard,1G31 DG and Day Tank Rooms,1G21 DG and Day Tank Rooms, Power Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room i
Heat Sink /
Torus Room, Intake Structure, Pumphouse Coolant Supply Containment Drywell, Torus Emergency NE, NW, SE Corner Rooms, HPCI Room, RCIC Room, RHR Valve Room, North Systems CRD Area, South CRD Area l
Other Control Building, Remote Shutdown Panel 1C388 Area, Panel 1C56 Area, SBGT Room 1
DAEC EALs 7,8, and 9 address site-specific occurrences of concern. These concerns include external i
flood water levels, internal flooding, and low river water level affecting the ultimate heat sink. DAEC EAL 7 addresses river water levels exceeding design flood water levels. All Seismic Category I structures and non-seismic structures housing Seismic Category I equipment are designed to withstand the hydraulic head resulting from the " maximum probable flood" to which the site could be subjected. The design flood water is at elevation 767.0 ft. Major equipment penetrations in the exterior walls are located above elevation 767.0 ft. Openings below the flood level are either watertight or are provided with means to control the inflow of water in order to ensure that a safe shutdown can be achieved and maintained.
- p Consideration has also been given to providing temporary protection for openings in the exterior walls up j
j V to flood levels of 769.0 ft All buildings were also checked for uplift (buoyancy) for a flood level at elevation 767.0 ft, and the minimum factor of safety used was 1.2. Therefore, DAEC EAL 7 uses as its threshold flood water levels above 767 feet.DAEC EAL 8 addresses intemal flooding consistent with the
Duane Amold Energy Center i
EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
PAGE H-10 of 25 IIAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY
't 3
requirements of EOP 3, Secondary Containment Control. If RPV pressure reduction will have no efTect on leakage into secondary containment, then EOP 3 requires that reactor shutdown be performed in accordance with Integrated Plant Operating Instruction (IPol) 3,4, or 5 as necessary if the water level exceeds its maximum safe operating limits in two or more areas. IfIU)V pressure reduction will decrease leakage into secondary containment then this is due to leakage from the primary system, which is addressed by the Fission Barrier Table indicators and System Malfunction EALs, and is not addressed here.
Therefore, EAL 8 addresses conditions in which water level in two or more areas is above Maximum Safe i
Operating Limits and reactor shutdown is required.
Required means that the reactor shutdown was procedurally mandated by EOP 3 and is not merely performed as a precaution or inadvertently. Maximum Safe Operating Limits are defined as the highest parameter value at which neither (1) equipment necessary for safe shutdown of the plant will fail nor (2)
] hp personnel access necessary for the safe shutdown of the plant will be preclude j
be due to system malfunctions, component failures, or repair activity mishaps (such as failed freeze seal) i i
that can threaten safe operation of the plant. This includes water intrusion on equipment that is not designed to be submerged (e.g., motor control centers).
i The maximum safe operating water level limits are taken from EOP 3 and are shown on the table below:
i Maximum Operating Limits - Water Levels Affected Location Indicator Maximum Normal OL Maximum Safe OL HPCI Room Area LI3768 6 inches 24 inches RCIC Room Area LI 3769 6 inches 18 inches A RHR Comer Room SE Area LI3770 6 inches 23 inches B RHR Comer Room NW Area L13771 6 inches 23 inches Torus Area L13772 12 inches 24 inches DAEC EAL 9 addresses the effects of low river water level. The intake structure for the safety-related water supply systems (river water, RHR service water, and emergency service water) is located on the west bank of the Cedar River. The overflow weir is at elevation 724 feet 6 inches. River level at or below this elevation will result in all river flow being diverted to the safety related water supply systems. The top of the intake structure around the pump wells is at elevation 724 feet. If the river water level dropped to this level, the pump suction would have no continuous supply. Therefore, this EAL uses a threshold of water level below 724 feet 6 inches as a potential substantial degradation of the ultimate heat sink capability.
. _ - ~..._
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview) lO PAGE H-11 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 901. Earthquake
- 2. Abnormal Operating Procedure (AOP) 902, Flood
- 3. Abnormal Operating Procedure (AOP) 903, Tomado
- 4. Abnormal Operating Procedure (AOP) 913, Fire
- 5. Abnormal Operating Procedure (AOP) 914, Security Events
- 6. UFSAR Chapter 3, Design of Structures, Components, Equipment, and Systems
- 7. Bechtel Drawing BECH-M017, Equipment Location - Intake Structure Plans at Elevations, Rev. 6
- 8. EOP Basis Document, EOP 3 - Secondary Containment Control
- 9. Emergency Operating Procedure (EOP) 3, Secondary Containment Control j
O O
I Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC review)
PAGE 11-12 of 25 IIAZARDS AND OTilER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY IIA 2 Fire Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown EVENT TYPE: Fire OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVEL:
- 1. The following conditions exist:
- a. Fire or explosion in < site-specific areas >
r-AND
' b. Affected system parameter indications show degraded performance or plant personnel report visible damage to permanent structures or equipment within the specified area.
There is no s gnificant deviation from the generic EAL. Of particular concem for this EAL are fires that i
may be deter.ed in the reactor building, control building, turbine building, pumphouse, and intake structure as shown in Tabs 1 and 3 of AOP 913. Damage from fire or explosion can be indicated by physical observation, or by Control Room / local control station instrumentation.. No attempt is made in this EAL to assess the actual magnitude ofthe damage.
Per AOP 913, the location of a fire can be determined by observing XL3 alarm messages, Zone Indicating Unit (ZlU) alarms, or fire annunciators on panels IC40 and IC40A.
O
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevien)
A
- PAGE11-13 of 25 IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY 4
This table is displayed as an aid to the Emergency Coordinator in determining appropriate areas of concern.
Systems & Equipment of Concern Reactivity Control e
Containment (Drywell/ Torus) e RIiR/ Core Spray /SRV's e
HPCI/RCIC e
e RIIRSW/ River Water /ESW Onsite AC Power /EDG's e
Offsite AC Power e
Remote Shutdown Capability e
NOTE:
Scope of Systems and Equipment of concern established by review of Appendix R Safe Shutdown credited systems. Only those systems directly affecting safe shutdown or heat removal are hsted for consideration, due to fire damage. Support Systems and equipment such as liVAC and specific instrumentation, while included in Appendix R analysis is not considered an immediate threat to the ability to shutdown the plant and remove decay heat.
With regard to explosions, only those explosions ofsufficient force to damage permanent structures or identified equipment requiredfor safe operation, should be considered. As used here, an explosion is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts significant energy to near-by structures and materials. The occurrence of the explosion with reports of evidence of damage (e.g., deformation, scorching) is sufficient for the declaration. The EC/OSS also needs to consider any security aspects ofthe explosions, ifapplicable.
1 Per the UFSAR, the control room is the only area that is required to be continuously occupied to achieve and maintain safe shutdown following design basis accidents. liowever, the capability exists for plant shutdown from outside the main control room in the event that the control room becomes uninhabitable. If j
g the control room becomes uninhabitable, remote shutdown panel IC388 is utilized in accordance with AOP 915.
I l
P Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
PAGE 11-14 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 913, Fire
- 2. Abnormal Operating Procedure (AOP) 914, Security Events
- 3. Abnonnal Operating Procedure (AOP) 915, Shutdown Outside Control Room
- 4. UFSAR Section 6.4, liabitability Systems i
i O
O
Duane Arnold Energy Center i
EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
PAGE H-15 of 25 IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY IIA 3 Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown 2
EVENT TYPE: Other Hazards and Failures OPERATING MODE APPLICAHILITY: All EXAMPLE EMERGENCY ACTION LEVELS:
(1 or 2)
- 1. Report or detection of toxic gases within a Facility Structure in concentrations that will be life
/G threatening to plant personnel.
b
- 2. Report or detection of flammable gases within a Facility Structure in concentrations that will affect the safe operation of the plant.
There is no significant deviation from the generic EALs. This IC, in addition to IC HA5 below, also addresses entry of toxic gases that may result in control room evacuation in accordance with AOP 915.
For the purposes of this 'C, CO (such as is discharged by the fire suppression system) is not toxic. CO 2
2 can be lethal ifit reduces oxygen to low concentrations that are immediately dangerous to life and health (IDLH). CO discharge into an area is not basisfor emergency classification under this IC unless: (1) 2 Access to the affected area is required, and (2) CO concentration residts in conditions that make the area 2
uninhabitable or inaccessible (i.e., IDLH).
Per the UFSAR, the control room is the only area that is required to be continuously occupied to achieve and maintain safe shutdown following design basis accidents. However, the capability exists for plant shutdown from outside the main control room in the event that the control room becomes uninhabitable. If the control room becomes uninhabitable, remote shutdown panel IC388 is utilized to achieve and maintain cold shutdown.
O t
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
!v PAGE H-16 of 25 IIAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY f
1 4
Per AOPs 913 and 914, the following areas are identified as safe shutdown areas. This table is displayed as an aid to the Emergency Coordinator in determining appropriate areas ojconcern.
Safe Shutdown Areas a
Category Area j
Electrical Power Switchyard, IG31 DG and Day Tank Rooms, IG21 DG and Day Tank Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room Heat Sink / Coolant Supply Torus Room, Intake Structure, Pumphouse Containment Drywell, Torus i
Emergency Systems NE, NW, SE Comer Rooms, HPCI Room, RCIC Room, RHR Valve Room, North CRD Area, South CRD Area J
Other Control Building, Remote Shutdown Panel IC388 Area, Panel IC56 Area,
.)
SBGT Room
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 913, Fire
- 2. Abnormal Operating Procedure (AOP) 914, Security Events
- 3. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room
- 4. UFSAP, Section 6.4, Habitability Systems J
- O
1 Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview) b PAGE 11-17 of 25 IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY IIA 4 Security Event in a Plant Protected Area EVENTTYPE: Security 1
OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS:
(1 or 2)
- 1. Intrusion into plant protected area by a hostile force.
i
)
- 2. Other security events as determined from (site-specific) Safeguards Contingency Plan.
There is no significant deviation from generic EALs.
This class of security events represents an escalated threat to plant safety above that contained in the Unusual Event. For the purposes of this EAL a civil disturbance which penetrates that protected area boundary can be considered a hostileforce. Under this EAL, adversaries within the protected area are not yet affecting nuclear safety systems, engineered safety features, or reactor shutdown capability that are located within the vital area. Intrusion into a vital area by a hostile force will escalate the event to a Site Area Emergency.
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 914, Security Events
/GU
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
'es
(>I PAGE H-18 of 25 IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLA.NT SAFETY CATEGORY IIA 5 Control Room Evacuation Has Been Initiated EVENT TYPE: Control Room Evacuation OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVEL:
- 1. Entry into (site-specific) procedure for control room evacuation.
/~N i )
There is no significant deviation from the generic EAL. The applicable procedure for control room i
evacuation at DAEC is AOP 915.
REFERENCES:
i
- 1. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room
- 2. UFSAR Section 6.4,IIabitability Systems i
.V
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
PAGE H-19 of 25 IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY IIA 6 Other Conditions Existing Which in the Judgment of the <EC/OSS> Warrant Declaration of an Alert EVENT TYPE: EC/OSS Judgment OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVEL:
- 1. Other conditions exist which in the Judgment of the Emergency Director indicate that plant safety systems may be degraded and that increased monitoring of plant functions is warranted.
p.,
There is no significant deviation from the generic EAL.
Per EPIP 7.1, the Emergency Coordinator / Operations Shift Supervisor (EC/OSS) is the title for the emergency director function at DAEC. The EAL addresses conditions that fall under the Alert emergency classification description contained in NUREG-0654, Appendix 1.
REFERENCES:
- 1. Emergency Plan Implementing Procedure (EPIP) 7.1, Emergency Coordinator Duties
- 2. NUREG-0654/ FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support ofNuclear Power Plants, Revision 1, 0ctober 1980, Appendix 1
i Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu) s PAGE H-20 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY HS1 Security Event in a Plant Vital Area EVENT TYPE: Security OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS:
(1 or 2) 1
- 1. Intrusion into plant vital area by a hostile force.
- 2. Other security events as determined from (site-specific) Safeguards Contingency Plan.
f~'..
V DAEC EALINFORMATION:
There is no significant deviation from generic EAL 1.
This class of security events represents an escalated threat to plant safety above that contained in HA4, Security Event in a Plant Protected Area, in that a hostile force has progressed from the Protected Area to the Vital Area. Under the condition ofconcern here, the adversaries are considered to be in a position to directly and negatively affect nuclear safety systems, engineered safety features, or reactor shutdoun capability.
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 914, Security Events f~h O
l Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevien)
N.]
PAGE H-21 of 25 IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY M
HS2 Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established EVENT TYPE: Control Room Evacuation OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVEL:
- 1. The following conditions exist:
- a. Control room evacuation has been initiated.
O AND l ()
- b. Control of the plant cannot be established per (site-specific) procedure within (site-specific) minutes.
There is no significant deviation from the generic EAL. The applicable procedure for control room I
evacuation at DAEC is AOP 915. Based on the results of the analysis described below, DAEC uses 20 minutes as the site-specific time limit for establishing control of the plant. DAEC has satellite panels associated with the remote shutdown panel at various locations through out the plant. It physically takes an operator longer than 15 minutes to lineup all the controls at the various panels. Control of the plant from outside the control room is assumed when the controls are transferred to remote shutdown panel IC388 in accordance with AOP 915.
The EC/OSS is expected to make a reasonable, informedjudgment within the 20 minute time limit that control of the plantfrom the remote shutdown panel has been established. The intent of the EAL is that control ofimportant plant equipment and knowledge ofimportant plant parameters has been achieved in a timely manner. Primary emphasis should be placed on those components and instruments that provide protection of and information about safety functions. At a minimum, consistent with the Appendix R safe j
shutdown analysis described above, these safety functions include reactivity control, maintaining reactor water level, and decay heat removal.
l General Electric performed analyses to demonstrate compliance with the requirements of 10 CFR 50 Appendix R for DAEC. The evaluation of Reactor Coolant Inventory was performed using the GE evaluation model (SAFE). The SAFE code determines if the rea: tor coolant inventory is above the TAF duiing the safe shutdown operation. If core uncovery occurs, the fuel clad integrity evaluation is performed l
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC revieu) g PAGE H-22 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY i-4 4
by determining the duration of the core uncovery and the resulting peak cladding temperature (PCT). The PCT calculations were performed by incorporating the SAFE output into the Core Heatup Analysis code (CHASTE). The details of these calculations are provided in Section 4 of the final report for DAEC Appendix R analyses (" Safe Shutdown Appendix R Analyses for Duane Arnold Energy Center", MDE 036).
1 The required analyses include evaluation of the safe shutdown capability of the remote shutdown system for various control room fire events assuming: (1) no spurious operation of equipment, (2) spurious operation of a safety-relief valve (SRV) for 20 minutes, (3) spurious operation of a SRV for 10 minutes, and (4) spurious leakage from a one-inch line. The analyses show that the worst case spurious operation of SRV or isolation valves on a one-inch liquid line (high-low pressure interface) will not afTect the safe i
shutdown ability of the remote shutdown system t -r DAEC in case of a fire requiring control roorn l
evacuation before the identified time limit for the necessary operator actions at the auxiliary shutdown panels. For the limiting cases of worst case spurious leakage from a one-inch line and spurious operation of a SRV, operator control within 20 minutes would not impact the integrity of the fuel clad, the reactor pressure vessel, and the primary containment.
4
REFERENCES:
l
- 1. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room i
- 2. General Electric Report MDE-44-0386, Safe Shutdown Appendix R Analysisfor DAEC, March 1986
^
- 3. UFSAR Section 6.4, Habitability Systems
- 4. NUMARC Methodologfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions and Answers, June 1993 i
i i
O
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
O PAGE H-23 of 25 IIAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY HS3 Other Conditions Existing Which in the Judgment of the <EC/OSS> Warrant Declaration of Site Area Emergency EVENT TYPE: EC/OSS Judgment OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVEL:
- 1. Other conditions exist which in the Judgment of the Emergency Director indicate actual or likely major failures of plant functions needed for protection of the public.
(3
's_,/
There is no significant deviation from the generic EAL, Per EPIP 7.1, the Emergency Coordinator / Operations Shill Supervisor (EC/OSS) is the title for the i
emergency director function at DAEC. The EAL addresses conditions that fall under the Site Area Emergency classification description contained in NUREG-0654, Appendix 1.
REFERENCES:
- 1. Emergency Plan Implementing Procedure (EPIP) 7.1, Emergency Coordinator Duties
- 2. NUREG-0654/ FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Revision 1, October 1980, Appendix 1 O
. ~.
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (/br NRC revien) 6 PAGE11-24 of 25
^
IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY 1
HG1 Security Event Resulting in Loss Of Ability to Reach and Maintain Cold Shutdown i
EVENTTYPE: Security OPERATING MODE APPLICABILITY: All i
EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2)
- 1. Loss of physical control of the control room due to security event.
l
- 2. Loss of physical control of the remote shutdown capability due to security event.
i l
There are no significant deviations from the generic EALs. The EALs encompass conditions under which l
a hostile force has taken physical control of vital area required to reach and maintain safe shutdown. This also includes areas where any switches that transfer control of safe shutdown equipment to outside the control room are located.
l
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 914, Security Events
- 2. UFSAR Section 6.4, Habitability Systems l
l l
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu) i)
l G PAGE11-25 of 25 h
IIAZARDS AND OTilER CONDITIONS AFFECTING EFFECTIVE DATE: TBD l
PLANT SAFETY CATEGORY l
HG2 Other Conditions Existing Which in the Judgment of the <EC/OSS> Warrant Declaration of General Emergency EVENT TYPE: EC/OSS Judgment OPERATING MODE APPLICAHILITY: All EXAMPLE EMERGENCY ACTION LEVEL:
- 1. Other conditions exist which in the Judgment of the Emergency Director indicate: (1) actual or imminent substantial core degradation with potential for loss of containment, or (2) potential for l A uncontrolled radionuclide releases. These releases can reasonably be expected to exceed EPA PAG U
plume exposure levels outside the site boundary.
l There is no significant deviation from the generic EAL.
Per EPIP 7.1, the Emergency Coordinator / Operations Shift Supervisor (EC/OSS) is the title for the t
emergency director function at DAEC. The EAL addresses conditions that fall under the General Emergency classification description contained in NUREG-0654, Appendix 1 and is consistent with FG1, l
Loss of Any Two Barriers AND Potential Loss of Third Barrier, and AG1, Site Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity that Exceeds 1000 mrem TEDE or 5000 l
mrem CDE Thyroid for the Actual or Projected Duration of the Release.
REFERENCES:
- 1. Emergency Plan Implementing Procedure (EPIP) 7.1, Emergency Coordinator Duties l
- 2. NUREG-0654fFEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency l
Response Plans and Preparedness in Support of Nuclear Pmrer Plants, Revision 1, October 1980, Appendix 1 O
f p
a s
1 I
EMERGENCY PLAN IMPLEMENTING PROCEDURE No. EPIP - 1.1 Rev.2 PAGE 1 of (ForNRCReview}
SYSTEM MALFUNCTION EHTfTIVE DATE-TBD EVENT TYPE UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENER AL EMERGENCY SUf SAf SSf SGf Loss of All Offsite Power to Essential Loos of M Offsite Power erwt Loss of M Loss of M Offsite Power and Loss of M Prolonged Loss of M Offsite Power and Busses for Greater Then 15 Minutes Onsite AC Power to Essentail Busses Onsite AC Power to Essential Busses Prolonged Loss of M Onsite AC Power During Cold Conditions Loss of Oftsite Power Lasting More Than 15 Loss of Voltage on Buses 1 A3 and 1 A4 Loss d Voltage on Buses 1 A3 and 1 A4 lasting Loss of Voltage on Buses 1 A3 and 1 A4 and ANY d Minutes.
lasting more than 15 rnmutes.
more than 15 rrunutes.
the followng.
Restoration d power to either Bus 1 A3 or 1 A4 is Op. Modes: ALL
_Op. Modes: Cold S/D, Refuel. Defueled Op. Modes: Run, Startup, Hot S/D NOT likely withm 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
RPV level indetemuraste SU7 SAS SSJ Unplanned Loss of Required DC Power AC Power Capability to Essential Busses Loss of All Vital DC Power RPV Level below +15 inches.
During Cold Shutdown or Refuel Mode Reduced to a Single Power Source for For Greater Then 15 Minutes Greater Then 15 Minutes Such That Any LOSS OF POWER Additional Single Failure Would Result in Station Blackout Unplanned Loss of Div 1 and Dnr 2 Only one AC power source remams Unptanned Loss d Dev 1 and Div 2 125 VDC busses based on bus voltage less available to supply Bus 1 A3 ce Bus 1 A4 AND 125 VDC busses Lasting More Than 15 than 105 VDC mdcated if it is lost. a Station Blackout will occur.
Mmutes.
AND Failure to restore power to at least one required 125 VOC bus withm 15 mnutes from time of loss.
Op. Modes: Cold SiD, Refuel Op. Modes: Run, Startup, Hot S/D Op. Modes: Run, Startup, Hot S/D Op. Modes: Run, Startup, Het S/D SA2 SS2 SG2 Failure of Reactor Protection System Failure of Reactor Protection System Failure of the Reactor Protection System to Instrumentation to Complete or Initiate Instrumentation to Complete or Initiate on Complete en Anutomatic Scram and Manual en Automatic Reactor Scram Once a Automatic Reactor Scram Once a Reactor Scram was NOT successful and There is Reactor Protection System Setpoint Has Protection System Setpoint Has Been Indication of an Extreme Challenge to the Ability Been Exceeded and Manual Scram Was Exceeded and Manual Scram Was NOT to Cool the Core Successful Successful None RPS FArLURE Failure of automatic scram.
Failure of automatic and manual scram Entry into ATWS EOP-RPV Control is required AND AND Power remams above 5%
RPV level cannot be mantamed above -30 inches.
OR OR Baron intactionTequired.
EOP Graph 4 Heat Capacity Limit is exceeded Op. Modes: Run, Startup Op. Modes: Run, Startup Op. Modes: Run, Startup SU2 SA3 554 Inability to Reach Required Shutdown inability to Maintain Plant in Cold Complete Loss of Function Needed to W~ thin Technical Specification Limits Shutdown Achieve or Maintain Hot Shutdown r
Plant NOT brought to regtsred made unthm loss of decay heat removal systeras EOP Graph 4 Heat Capacity Limit is appicable LCO Action Statement Trne Limits.
required to maintain cold shutdown exceeded AND OR Temperature nse that exceeds 212'F.
Reactor CANNOT be brought subcntical See Fission Bamer Table OR INABILITY TO MAINTAIN Uncontrolled tempe sture nse approachmg Op. Modes: Run, Startup, Hot S/D SHUTDOWN CONDITIONS 212*F.
ss5 Loss of Water Level in the Reactor Vessel That Has or Will Uncover Fuel in the Reactor Vessel NO coolmg method kned up or available AND RPV Level below 15 inches.
Op. Modes: Run, Startup, Hot SD Op. Modes: Cold SD, Refuel Op. Modes: Cold SD, Refuel SU3 SA4 SS6 Unplanned Loss of A!! Safety System Unplanned Loss of Most or All Safety inability to Monitor c Significant Transient Annunication or Indication in the Control System Annunication or Indication in in Progress Room for Groeter Then 15 Minutes Control Room With Either (1) o signifiesnt Transient in Progress, or (2)
Compensatory Non-Alarming Indicators are Unavailable See Fisson Bamer Table Unplanned loss of most annunciators on Unplanned loss of rnost arw unciators on Significant transient in progress and BOTH of panels 1CO3 IC04 and 1C05 lasting rnore panels 1C03.1C04 and 1C05 lasting more the folkwng.
than 15 rnenutes AND compensatory non-than 15 minutes and EITHER:
Loss of annunciators on panels 1C03, alarming indicatons are available.
s,gneficant transient in progress.
1C04 and 1C05 AND Loss of compematory non-alamwng Loss of compensatory non-alarming INSTRUMENTATION I Op. MW: Run, M, M SD indcatons.
Wtons.
COMMUNICATION SU8 Unplanned Loss of All Onsite or Offsite Communications Capabilities Loss of ALL onsite telephone and rado comrnurucation methods (PABX. direct-nng.
UHF, and radological survey rado systems).
OFI Loss of ALL electrorwc commurucation methods wrth govemment agencies (PABX, direct-nng. ENS, trucrowave and pohce rado)
Op. Modes: ALL Op. Modes: Run, Startup, Hot SD Op. Modes: Run, Startup, Hot SD SU4 Fuel Cted Degradation Vahd Pretreat RM-4104 rad monitor reading above 4E4 mRhr COOLANT ACTIVTTY OR See Fasson Bamer Table See Fisson Bamer Table See Fission Bamer Table Coolant activity above 1.2 pCi/mi DOSE EOUtVALENT l-131 Op. Modes: Run, Startup, Hot SO SUS RCS Leakage Urudentrfied or pressure boundary leakage O
COOLANT LEAKAGE greater than 10 GPM.
OR See Fisson Bamer Table See Fission Bamer Table See Frsson Bamer Table r
identified leakage greater than 25 GPM.
Main steam line brea s determined from I
annunciators or plant personnel report.
l Op. Modes: Run, Startup, Hot S/D C>
>F T D-vo Om2 e
Eye
> "U CD g
ag mHH ost CCm P
Ra 00 C. o s
m l
1 l
,__,_._._m._
..,...s...
,.__au
__u.m4.
- - -m#s..mam e.---em-A.
O i
1 1
O SYSTEM MALFUNCTION CATEGORY i
l l
lO
Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
V PAGE S-1 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SUI Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: All i
EXAMPLE EMERGENCY ACTION LEVEL:
L The following conditions exist:
- a. L,oss of power to (site-specific) transformers for greater than 15 minutes.
AND
- b. At least (site-specific) emergency generators are supplying power to emergency busses.
V There is no significant deviation from the generic EAL. This event is aprecursor ofa more serious Station Blackout condition andis thus considered as a potential degradation ofthe level ofsafety ofthe plant. It is possible to be operating within Technical Specification LCO Action Statement time limits and make a j
declaration ofan Unusual Event in accordance with this EAL.
Under the conditions of concern, entry into AOP 301, Loss of Essential Electrical Power, would be made under Tab 3, Loss of Offsite Power. Indications / alarms related to loss of offsite AC power are displayed on control room panel IC08 and are listed in the procedure under " Probable Indications." Under these conditions, Essential 4160V Buses l A3 and 1 A4 would indicate zero volts until A diesel generator 1G-31 4kV breaker l A311 and B diesel generator 1G-214kV breaker l A411, respectively, close for each bus.
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 301, Loss of Essential Electrical Power
- 2. UFSAR Section 8.2, Offsite Power System
- 3. NUAfARC Alethodologyfor Development ofEmergency Action Levels NUAfARC/NESP-007 Revision 2 Questions andAnswers, June 1993 O
I v
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
V PAGE S-2 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SU2 Inability to Reach Required Shutdown Within Technical Specification Limits EVENT TYPE: Inability to Maintain Shutdown Conditions OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EXAMPLE EMERGENCY ACTION LEVEL:
- 1. Plant is not brought to required operating mode within (site-specific) Technical Specifications LCO Action Statement Time.
O There is no significant devbtion from the generic EAL. LCO Action Statement time limits for placing the O
unit in the required OPCON are provided in the DAEC Tecimical Specifications.
An immediate Notification of an Unusual Event is required when the plant is not brought to the required i
OPCON within the Technical Specifications LCO Action Statement time limits. Declaration of an Umtsual Event is based on the time at which the LCO-specified action statement time period elapses under the site Technical Specifications and is not related to how long a condition may have existed.
REFERENCES:
- 1. DAEC Technical Specifications O
N.Y
Duane Arnold Energy Center f
EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview) gV PAGE S-3 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SU3 Unplanned Loss of All Safety System Annunciation or Indication in the Control Room for Greater Than 15 Minutes i
EVENT TYPE: Instrumentation / Communication OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EXAMPLE EMERGENCY ACTION LEVEL:
- 1. The following conditions exist:
- a. Loss of most annunciators < > associated with safety systems for greater than 15 minutes.
l AND l
- b. Compensatory non-alarming indications are available.
g AND
- c. In the opinion of the < Operations > Shift Supervisor, the loss of aimunciators or indicators requires increased surveillance to safely operate the unit < >.
AND 1
- d. Annunciator or indicator loss does not result from planned action.
Control room panels IC03, IC04, and IC05 contain the annunciators associated with safety systems at DAEC. Therefore, the DAEC EAL addresses unplanned loss of most annunciators on these panels.
Compensatory non-alarming indications includes the plant process computer, SPDS, plant recorders, or l
plant instrument displays in the control room. Unplanned loss of annunciators or indicators excludes i
scheduled maintenance and testing activities, dnder the conditions of concern, entry into AOP 302.2, Loss of Alarm Panel Power, would be made. The procedure requires alerting operators on shift to the nature of the lost annunciation. It further requires that operators be attendant and responsive to abnormal indications that relate to those systems and components that have lost annunciation. Therefore, the generic criterion related to specific opinion of the Operations Shift Supenisor that additional operating personnel will be required to safely operate the unit is not included in the DAEC EAL because the concern is addressed by the AOP.
- O Annunciators on IC03, IC04, and IC05 share a common power supply from 125 VDC Division I that is d
fed through circuit breaker 1D13.
l l
1 Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview) i PAGE S-4 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD l
i Indications ofloss of annunciators associated with safety systems include:
i 125 VDC charger, battery, or system annunciators on control room panel IC08 e
Loss of" sealed in" annunciators at affected panels e
Failure of affected annunciator panels shiftily testing by plant operators j
Expected alarms are not received e
4 e
Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to panels 1C03, I
1C04, and 1C05)
REFERENCES:
i 3
- 1. Operating Instruction (01) No. 317.2 Annunciator System
- 2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power
! (Q3
- 3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power 4
i I
i j
i t
T 4
)
2 i
1 4
lO 4
.o l
1 1
I Duane Arnold Energy Center i
EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
PAGE S-5 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SU4 Fuel Clad Degradation EVENT TYPE: Coolant Activity OPERATING MODE APPLICABILITY: Run, Startup, Hot S/D EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2)
- 1. (Site-Specific) < valid > radiation monitor readings indicating fuel clad degradation greater than Technical Specification allowable limits.
- 2. (Site-Specific) coolant sample activity value indicating fuel clad degradation greater than Technical Specification allowable limits.
O DAEC EALINFORMATION:
U There are no significant deviations from the generic EALs. These EALs are precursors ofmore serious fuel clad degradation and are thus considered as indicating a potential degradation ofthe level ofsafety of the plant. Thus, it is possible to be operating within Technical Specification LCO Action Statement time limitsfor iodine spikes and make a declaration ofan Unusual Event. DAEC mode applicability for these EALs are consistent with the Tech Specs.
EAL 1 addresses valid pretreat rad monitor exceeding (RM-4104) above 4E+3 mR/hr. The calculation supporting this value is described below. Valid means that the pretreat rad monitor reading is detennined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or coolant sampling results. This reading would be displayed on Control Room panels IC-02 and IC-10 on pretreat rad recorder RR-4104.
As specified in the generic methodology, DAEC EAL 2 addresses coolant samples exceeding technical specification 3.6.B.I.a, coolant activity less than or equal to 1.2 Ci/ml dose equivalent I-131.
Radiological Engineering Calculation 94-014A and UFSAR Table 15.4-1 were reviewed to determine a suitable EAL threshold for the pretreat rad monitor reading corresponding to the Tech Spec 3.6.B.I.a coolant activity limit of 1.2 Ci/ml of dose equivalent I-131. Using the condenser noble gas source term for the control rod drop accident of 2.38 E +06 Curies shown on UFSAR Table 15.4-1 and the condenser p(/
free volume of 55,000 cubic feet, an initial noble gas concentration in the condenser offgas line is determined. Because the offgas flow rate is very small (about 50 standard cubic feet per minute) compared to the total condenser free volume, dilution of the condenser noble gas concentration due to offgas flow is
Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview) g iV; 1
PAGE S-6 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD not considered in the calculation shown below. Decrease in the noble gas source term due to decay of short-lived noble gas radioisotopes and offgas flow dilution effects are addressed by rounding down the value calculated as shown below.
Calculation 94-014A used an exposure rate method based on using a source term consisting of a defined mixture of noble gases and iodine from the control rod drop accident as described in the DAEC UFSAR, Section 15.4. The calculation assumed that the activity is released instantly and immediately reached in equilibrium with the reactor coolant inventory l Using this calculation, using dose correction factors (DCFs) for child thyroid dose from Reg. Guide 1.109, and adjusting for the specific gravity (0.736) of saturated water at 1050 psia (fluid conditions assumed in the calculation) to adjust for standard conditions, the I-131 dose equivalent (in units of Ci/ml assuming I cc equals 1 ml) is determined for this event. This result is then linearly scaled for rad monitor readings corresponding to the Tech Spec 3.6.B.I.a allowable primary coolant activity of 1.2 Ci/ml I-131 dose equivalent, i.e., the relative mixture of noble gases and iodine is assumed to remain constant. I-129 is ignored because it has no effect on the calculation result.
O Isotope DCF (mrem /pci)
Concentration ( Ci/cc)
Correction Factor [DCF
/
l-131 DEQ (pCi/cc)
DCFoul / 0.736 l-131 4.39 E-03 1.6 E+01 1.4 E+00 2.2 E+01 1-132 5.23 E-05 2.2 E+01 1.6 E-02 3.6 E-01 1-133 1.04 E-03 3.1 E+01 3.2 E-01 1.0 E+01 1-134 1.37 E-05 3.4 E+01 4.2 E-03 1.4 E-01 1-135 2.14 E-04 2.9 E+01 6.6 E-02 1.9 E+00 l
3.4 E+01 TOTAL Therefore, for this event, a coolant activity of 34 pCi/cc I-131 dose equivalent is calculated. Scaling the results for 1.2 pCi/cc I-131 dose equivalent, a suitable condenser source term and corresponding initial concentration in the offgas flow is then determined. This is then converted to a pretreat rad monitor reading by use of the monitor efficiency factor:
Pretreat Rad Monitor (RM-4104) Reading NG concentrationow% = NG concentration X[1.2 Ci/cc /34 pCi/cc]
= [2.38 E +6 Ci x 1 E+6 pCi /Ci]/ [5.5 E+4 ft' x 2.83 E+4 cc/ft'] X [1.2 Ci/cc/34 pCi/cc]
= 1529 Cl x 0.0353 = 54.0 pCi/cc Pcetreat rad monitor reading = NG concentration X Rad monitor efficiency j
,O%
l
()
Rad monitor efficiency = 89.2 mR/hr / pCi/cc, therefore:
Pretreat rad monitor reading = 89.2 X 54.0 = 4800 mR/hr To account for isotopic decay and dilution effects of offgas flow, round down to 4E+03 mR/hr.
Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
(3 PAGE S-7 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD The calculation results were also reviewed to determine if suitable values for the main steam line (MSL) radiation monitors could be developed. As shown above, the rod drop accident corresponds to coolant activity of 34 pCi/cc I-131 dose equivalent. As determined by the reference calculation, this corresponds to a MSL radiation monitor reading of about 5.7 R/hr. Scaling the results for 1.2 pCi/ml I-131 dose equivalent:
MSL Reading Corresponding to 1.2 pCl/ml 1-131 dose equivalent
((1.2 pCi/cc] / [34 Ci/cc)) X 5.7 R/hr = 0.2 R/hr = 200 rnR/hr 200 mR/hr is at the lower end of the normal MSL monitor readings during full power Because this value is not distinguishable and hydrogen water chemistry system malfunctions that result in increased production of N-16 can also result in increased main steam line radiation levels, it is not appropriate at m
(
) DAEC to use the main steam line monitor readings.
REFEIENCES:
- 1. Abnormal Operating Procedure (AOP) 672.2, Offgas Radiation / Reactor Coolant High Activity
- 2. Technical Specification 3/4.6.B. Coolant Chemistry
- 3. Radiological Engineering Calculation No. 94-014A, Main Steam Line Radiation Monitor Setpoint Calculation, August 29,1994 j
- 4. Surveillance Test Procedure (STP) No. 46B001, Reactor Coolant Gamma and Iodine Activity S. Annunciator Response Procedure (ARP) IC03A, Reactor and Containment Cooling and Isolation
- 6. Annunciator Response Procedure (ARP) IC05B, Reactor Control
- 7. NUMARC Methodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions andAnswers, June 1993 pv
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview) g k
PAGE S-8 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SU5 RCS Leakage EVENT TYPE: Coolant Leak OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown, Cold Shutdown EXAMPLE EMERGENCY ACTION LEVELS < >: < (1 or 2 or 3) >
< l.>
Unidentified or pressure boundary leakage greater than 10 gpm.
<2.>
Identified leakage greater than 25 gpm.
<OR>
<3.>
< Valid (site-specific) indication of Main Steamline Break >
EALs 1 and 2 are precursors of more serious RCS harrier challenges and are thus considered as a potential degradation of the level of safety of the plant. Thus, it is possible to be operating within TechnicalSpecification LCO Action Statement time limits and make a declaration ofan Unusual Event in accordance with these EALs. Creditfor the action statement time limit should only be given when leakage exceeds technical specification limits but has not yet ereeded the Unusual Event EAL thresholds described above. In addition, indication of main steam line break ;has been added here as previously discussed in the basis for Fission Barrier Table RCS Barrier EAL 1, RCS Leak Rate, and is further discussed below. This is in accordance with NUMARC Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 2 Questions and Answers, June 1993, Fission Product Barrier-BWR section, response to question 4 which states that the main steam line break with isolation can be classified under System Malfunctions.
Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.
The UAEC Tech Spec Section 3.6.C.1 coolant system leakage LCO limits are: (1) 5 gpm unidentified leakage, (2) 2 gpm increase in unidentified leakage within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, and (3) 25 gpm total leakage.
Total leakage is defined as the sum ofidentified and unidentified leakage.
DAEC EAL 1 uses the generic value of 10 GPM for unidentified leakage or pressure boundary leakage.
The 10 gpm value for the unidentified or pressure boundary leakage was selected as it is observable with
1 Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
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PAGE S-9 ef 33 SYSTEM MALFUNCTION CATEGORY EFFEJTIVE DATE: TBD 1
i normal control room indications. DAEC EAL 2 uses identified leakage set at a higher value due to the
)
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' lesser significance ofidentified leakage in comparison to unidentified or pressure boundary leakage.
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REFERENCES:
- 1. Technical Specification 3.6.C, Coolant Leakage
- 2. Surveillance Test Procedure No. (STP) 42A001, Reactor Coolant System Leak Rate Calculation
- 4. Alarm Response Procedure (ARP) IC04B, Reactor Water Cleanup and Recirculation
- 5. Alarm Response Procedure (ARP) IC04C,. Reactor Water Cleanup and Recirculation
- 6. UFSAR Section 5.2.5, Detection of Leakage through Reactor Coolant Pressure Boundary
- 7. UFSAR Section 15.6.6, Loss-of-Coolant-Accident
- 8. NUAMRC Methodologyfor Development ofEmergency Action Levels NUAMRC/NESP-007 Revision 2 Questions andAnswers, June 1993 i
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Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
PAGE S-10 of 33 SYSTEM MALFUNCTION CATEGGin' EFFECTIVE DATE: TBD SU6 Unplanned Loss of All Onsite or Offsite Communications Capabilities EVENT TYPE: Instrumentation / Communication OPERATING MODE APPL.ICABILITY: All EXAMPLE EMERGENCY ACTION LEVEL:
- 1. Either of the following conditions exist-
- a. Loss of all (site-specific list) onsite communications capability affecting the ability to perform routine operations.
- b. Loss of all (site-specific list) offsite communications capability.
/
There is no significant deviation from the generic EAL. The communications methods used at DAEC are described in the Emergency Plan. In-plant and extemal agency telephone communication methods include PABX lines, direct-ring lines, and NRC telephones which are extensions for the Emergency Notification System. There is also a microwave system to provide backup emergency telephone communications.
The availability of one method of ordinary offsite communication is sufficient to inform state and local authorities of plant problems. This EAL is intended to be used only when extraordinary means (relaying of informationfrom radio transmissions, individuals being sent to ofsite locations, etc.) are being utili:ed to make communications possible.
The DAEC plant operations radio system is a UlIF system with consoles located in the Control Room, Technical Support Center, Operational Support Center, and the Central Alarm Station. Hand-held transceivers are used in this system to provide simplex communications within the plant and onsite. The DAEC Radiological Survey Radio System is an 800 MHz trunked / conventional repeater system that provides base-to-portable communications throughout the DAEC EPZ. A secondary high-band system provides back-up capability for the 800 MHz radio. Consoles are located in the Technical Support Ccnter and the Emergency Operations Facility at the IES Tower. The DAEC Security (backup radiological survey) Radio System provides base-to-portable security communication within the plant and with the Linn f}
County Sheri1Ts Office using a mobile relay (repeater) type base station and two VHF frequencies. Control U
consoles are located in the Secondary Alarm Station, Central Alarm Station, Security Control Point, Technical Support Center, and Emergency Operations Facility. The DAEC also has a base station licensed for operation in the Police Radio Service on the law enforcement state-wide, point-to-point VHF
c Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
PAGE S-11 of 33 4
SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD frequency. The transmitter and one control console are located at the Secondary Alann Station and in the Central Alarm Station. This station is for communications with lowa Department of Public Safety radio station, Linn County Sheriffs oflice, and the Benton County Sheriffs office. This point-to-point channel is i
also used by the Linn County Emergency Management and other public-safety organizations throughout the state oflowe
REFERENCES:
- 1. Emergency Plan, Section F, Emergency Communications
- 2. NUAfARC Afethodologfor Development ofEmergency Action Levels NUAIARC/NESP-007 Revision 2 Questions andAnswers, June 1993 O
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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieW U
1 PAGE S-12 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD i
SU7 Unplanned Loss of Required DC Power During Cold Shutdown or Refuel < >
Mode For Greater Than 15 Minutes EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Cold Shutdown, Refuel EXAMPLE EMERGENCY ACTION LEVEL:
l
- 1. <T>he following conditions exist:
- a. Unplanned Loss of Vital DC power to required DC busses based on (site-specific) bus voltage indications.
AND
- b. Failure to restore power to at least one required DC bus within 15 minutes from time ofloss.
There is no significant deviation from the generic EAL. Unplanned loss of Div. I and Div. 2.125 VDC busses excludes scheduled maintenance and testing activities. Under the conditions of concern, AOP 302.1, Loss of 125 VDC Power, would be entered. The DAEC EAL's address the loss of both divisions of l
the 125 VDC systems consistent with AOP 302.1.
The 125 VDC system is divided into two independent divisions - Division I (1D1) and Division II (ID2) -
each with separate AC and DC (battery) power supplies. Loss of both 125 VDC Divisions could compromise the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations. These EAL's are intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss. If this loss results in the inability to maintain cold shutdown, the escalation to an Alert will be per SA3 "RCS temperature rise that
)
is not allowed by procedures or Technical Specifications that will result in RCS temperature above 212 F" Bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment and may be indicated by the illumination of annunciators "125 VDC System 1 Trouble" on IC08A A-9 and/or"125 VDC System 2 Trouble" on IC08B A-4.
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i Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu) 4 PAGE S-13 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power
- 2. Abnormal Operating Procedure (AOP) 388, Loss of 250 VDC Power
- 3. Technical Specification 3.8.B, DC Power Systems
- 4. UFSAR Section 8.3, Onsite Power Systems
- 5. UFSAR Table 8.3-6, Plant Battery System - DC Power, Instrumentation, and Control, Principle DC Loads (125V)
- 6. ARP IC08A A-9
- 7. ARP IC08B A-4 1
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Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (TorNRCreview) i lL; PAGE S-14 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD i
SAI Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Besses During Cold < Conditions >
EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Cold Shutdown, Refuel, Defueled EXAMPLE EMERGENCY ACTION LEVEL:
- 1. The following conditions exist:
- a. Loss of power to (site-specific) transformers.
AND
- b. Failure of(site-specific) emergency generators to supply power to emergency busses.
I.
AND
- c. Failure to restore power to at least one emergency bus within 15 minutes from the time ofloss of I
both offsite and onsite AC power.
There is no significant deviation from the generic EAL. Under the conditions of concern, entry into AOP 301.1, Station Blackout, would be made under Tab 1. Indications / alarms related to station blackout are displayed on control room panel 1C08 and are listed in the procedure under " Probable Indications."
At DAEC, the Essential Buses of concern are the 4160V Buses l A3 and 1 A4. Each of these buses feed their associated 480V and 120V AC buses through step down transformers.
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 301.1, Station Blackout
- 2. Abnormal Operating Procedure (AOP) 301, Loss of Essential Electrical Power
- 3. Technical Specifications Section 3.8, Auxiliary Electrical Systems OU
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
V PAGE S-15 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SA2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful EVENT TYPE: RPS Failure OPERATING MODE APPLICABILITY: Run,Startup EXAMPLE EMERGENCY ACTION LEVEL:
- 1. (Site-specific) indication (s) exist that indicate that reactor protection system setpoint was exceeded and automatic scram did not occur, and a successful manual scram occurred.
The DAEC EAL is written in terms of failure of automatic scram. IPOI 5 specifies manual scram insertion immediately following any automatic scram signal, and therefore separately specifying successful manual scram is not required. The reactor is considered successfully shutdown if either: (1) all control rods are inserted to least position 02, or (2) it has been determined that the reactor will remain shutdown under ALL conditions without boron. If these conditions are not achieved, entry into the ATWS - RPV Control EOP will be made where additional manual actions to be performed at panel IC05 are specified to quickly shutdown the reactor. These actions include reducing recirculation pumps to minimum speed and inserting Alternate Rod Insertion (ARI).
If the mode switch is in Startup and the rods arefidly inserted (i.e., the reactor is shutdown) prior to the automatic signalfailure, then declaration ofan Alert would not be required in this case, the event would be reported under 10 CFR 50. 72 (b) (2) (I) as afour hour report.
The condition of concern is failure of the automatic protection system to 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. In the generic EAL, reactor protection system setpoint being exceeded (rather than limiting safety system setpoint being exceeded) is specified to emphasize that failure of the automatic protection system to complete the scram following generation of a scram signal is the issue of concern.
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
PAGE S-16 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD
REFERENCES:
1, Integrated Plant Operating Instruction (IPOI) No. 5, Reactor Scram
- 4. NUAfARC Afethodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions and Answers, June 1993 O
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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevicu) g i
PAGE S-17 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SA3 Inability to Maintain Plant in Cold Shutdown EVENT TYPE: Inability to Maintain Shutdown Conditions OPERATING MODE APPLICABILITY: Cold Shutdown, Refuel
~
EXAMPLE EMERGENCY ACTION LEVEL:
- 1. Loss of < decay heat removal systems required > to maintain cold shutdown AND Temperature increase that either:
Exceeds Technical Specification cold shutdown temperature limit
(
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Results' in uncontrolled temperature rise approaching cold shutdown technical specification limit.
Under the conditions of concern for EAL 1, AOP 149, Loss of Decay Heat Removal, would be entered under Tab 1, Loss of Shutdown Cooling. Indications / alarms related to loss of shutdown cooling are displayed on control room panels IC03 and IC05 and are listed in the procedure under " Probable Indications." The procedure requires that shutdown cooling be re-established. If this cannot be done, then the following actions are to be performed:
4 Re-establish primary and/or secondary containment.
Increase reactor water level to between 240 inches and 250 inches to improve natural circulation.
Start one reactor recirculation pump, if available.
Monitor reactor temperatures per STP 46A003 noting that some points will lag behind the bulk coolant temperature.
Notify llealth Physics to begin increased monitoring of the Reactor Building.
Evaluate plant conditions to determine need to achieve or maintain the plant in cold shutdown.
Initiate alternate means of shutdown cooling which includes feed and bleed to radwaste or condenser, e
feed and bleed to the torus through the SRVs, using the reactor water cleanup heat exchanger, or reactor cavity flood up and use of fuel pool cooling.
b The procedure provides curves of maximum water heat up rates which provide an upper bound of the heatup until an estimated time to boil calculation can be completed by Engmeenng.
Duane Arnold Energy Center l
EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
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PAGE S-18 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD The DAEC EAL is written to imply a RCS temperature rise above 212 F that is not allowed by plant procedures. This corresponds to the inability to maintain required temperature conditions for Cold Shutdown. " Uncontrolled" means that system temperature increase is not the result of planned actions by the plant staff. The wording is also intended to eliminate minor cooling interruptions occurring at the transition between Hot Shutdown and Cold Shutdown or temperature changes that are permitted to occur during establishment of alternate core cooling so that an unnecessary declaration of an Alert does not occur. The uncontrolled temperature rise is necessa y to preserve the anticipatory philosophy of NUREG-0654 for events starting from temperatures much lowoc than the cold shutdown temperature limit.
REFERENCES:
- 1. Abnormal Operating Procedure ( AOP) 149, Loss of Decay Heat Removal
(]
- 2. DAEC Technical Specifications
(.)
- 3. Surveillance Test Procedure (STP) 46A003, Heatup and Cooldown Rate Log
- 4. NUREG 1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the UnitedStates, September 1993
- 5. NUMARC Methodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions and Answers, June 1993 fN_)
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
J PAGE S-19 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SA4 Unplanned Loss of Most or All Safety System Annunciation or Indication in Control Room With Either (1) a Significant Transient in Progress, or (2)
Compensatory Non-Alarming Indicators are Unavailable EVENT TYPE: Instrumentation / Communication OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EXAMPLE EMERGENCY ACTION LEVEL:
- 1. The following conditions exist:
- a. Loss of most< > annunciators associated with safety systems for greater than 15 minutes.
AND gtg
- b. In the opinion of the < Operations > Shift Supervisor, the loss of all annunciators or indicators requires increased surveillance to safely operate the unit < >.
AND
- c. Annunciator or Indicator loss does not result from planned action.
AND
- d. Either of the following:
A significant plant transient in progress.
OR Compensatory non-alarming indications are unavailable.
=
Control room panels IC03, IC04, and IC05 contain the annunciators associated with safety systems at DAEC. Therefore, the DAEC EAL addresses unplanned loss of annunciators on these panels.
Compensatory non-alarming indications includes the plant process computer, SPDS, plant recorders, or plant instrument displays in the control room. Unplanned loss of annunciators or indicators excludes scheduled maintenance and testing activities. Sigmficant transient includes response to automatic or manually initiated ftmetions such as scrams, runbacks involving greater than 25% thermal power change, ECCS injections, or thermal power oscillations of 10% or greater.
Under the conditions of concern, entry into AOP 302.2, Loss of Alami Panel Power, would be made. The j
J procedure requires aleiting operators on shift to the nature of the lost annunciation. It further requires that j
operators be attendant and responsive to abnormal indications that relate to those systems and components j
that have lost annunciation. Therefore, the generic criterion related to specific opinion of the Operations J
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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
PAGE S-20 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD i
Shift Supervisor that additional operating personnel will be required to safely operate the unit is not j
included in the DAEC EAL because the concern is addressed by the AOP.
j l
Annunciators on IC03, IC04, and IC05 share a common power supply from 125 VDC Division I that is l
fed through circuit breaker ID13. Therefore, DAEC does not specify a loss of"most" annunciators as specified in the generic methodology.
Indications ofloss of annunciators associateo a h safety systems include:
125 VDC charger, battery, or system annunciators on control room panel 1C08 l
e Loss of" sealed in" annunciators at afTected panels e
Failure of alTected annunciator panels shiftily testing by plant operators e
l l_
Expected alarms are not received e
l (]
e Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to panels 1C03, J (>
1C04, and IC05)
REFERENCES:
l
- 1. Operating Instruction (OI) No. 317.2 Annunciator System
- 2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power l
- 3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power i
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Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCrevieu)
PAGE S-21 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD 1
l SA5 AC Power Capability to Essential Busses Reduced to a Single Power Source for Greater Than 15 Minutes Such That Any Additional Single Failure Would Result in Station Blackout EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Run, Startup,110t Shutdown EXAMPLE EMERGENCY ACTION LEVEL:
- 1. The following conditions exist:
- a. Loss of power to (site-specific) transformers for greater than 15 minutes.
AND Q
- b. Onsite power capability has been degraded to one (train of) emergency bus (ses) powered from a single onsite power source due to los s of; (Site-specific list)
The DAEC EAL is written to address the underlying concern,i.e., only one AC power source remains and ifit is lost, a Station Blackout will occur. Under the conditions of concem, entry into AOP 301, Loss of Essential Electrical Power, would be made under Tab 1, Loss of One Essential 4160V Bus, and/or under Tab 3, Loss of Offsite Power. Indications / alarms related to degraded AC power are displayed on control room panel 1C08 and are listed in the procedure under " Probable Indications."
At DAEC, the Essential Buses of concem are 4160V Buses l A3 and l A4. Each of these buses feed their associated 480V and 120V AC busses through step down transformers. Onsite power sources at DAEC include the A and B Diesel Generators, IG-31 and IG-21, respectively.
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 301, Loss of Essential Electrical Power
- 2. UFSAR Chapter 8 Electrical Power
- 3. Technical Specifications Section 3.8. Auxiliary Electrical Systems
_m Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview) k PAGE S-22 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SSI Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EXAMPLE EMERGENCY ACTION LEVEL:
- 1. Loss of all offsite and onsite AC power as indicated by:
Loss of power to (site-specific) transformers.
a.
AND
- b. Failure of(site-specific) emergency generators to supply power to emergency busses.
AND I (~]
- c. Failure to restore power to at least one emergency bus within <l5 minutes > minutes from the time C'
ofloss of both offsite and onsite AC power.
There is no significant deviation from the generic EAL. In accordance with the generic guidance, DAEC is using a threshold of 15 minutes for Station Blackout to exclude transient or momentary power losses.
Under the conditions of concern, entry into AOP 301.1, Station Blackout, would be made under Tab 1.
Indications / alarms related to station blackout are displayed on control room panel IC08 and are listed in 4
the procedure under " Probable Indications."
At DAEC, the Essential Buses of concem are the 4160V Buses l A3 and 1A4. Each of these buses feed their associated 480V and 120V AC buses through step down transformers.
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 301.1, Station Blackout
- 2. Technical Specifications Section 3.8, Auxiliary Electrical Systems
- 3. UFSAR Chapter 8, Electric Power
Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
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U PAGE S-23 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SS2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful EVENTTYPE: RPS Failure j
OPERATING MODE APPLICABILITY: Run, Startup EXAMPLE EMERGENCY ACTION LEVEL:
- 1. (Site-specific) indications exist that automatic and manual scram were not successful.
,_.s
/
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The DAEC EAL addresses conditions where failure of an automatic scram has occurred and manual actions performed at panel IC05 to quickly shutdown the reactor do not meet the success criteria ofIPOl 5 and the ATWS - RPV Control EOP, where power remains above 5% or boron injection is required.
Under the conditions of concem for this EAL, the reactor may be producing more heat than the maximum decay heat load for which the safety systems are designed. A Site Area Emergency is indicated because conditions exist that lead to imminent loss or potential loss of the primary containment and the fuel clad.
In addition, if the SRV's are open, the RCS is no longer capable of retaining fission products and therefore is not acting as a fission product barrier. Although this EAL may be viewed as redundant to the Fission Barrier Table, its inclusion is necessary to better assure timely recognition and emergency response.
REFERENCES:
4
- 1. Integrated Plant Operating Instruction (IPOI) No. 5, Reactor Scram
- 3. NUMARC Methodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions and Answers,3une 1993 O
Duane. Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 2 (for NRC revicu)
OV PAGE S-24 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SS3 Loss of All Vital DC Power EVENTTYPE: LossofPower OPERATING MODE APPLICAHILITY: Run, Startup, Hot Shutdown EXAMPLE EMERGENCY ACTION LEVEL:
- 1. Loss of All Vital DC Power based on (site-specific) bus voltage indications for greater than 15 minutes.
There is no significant deviation from the generic EAL. Under the conditions of concem, AOP 302.1, Loss p
of 125 VDC Power, would be entered under Tab 3, Complete Loss of 125 VDC. Consequently, the DAEC EAL addresses loss of both divisions of the 125V DC system consistent with AOP.
At DAEC, the 250V/125V DC Systems ensure power is available for the reactor to be shutdown safely and maintained in a safe condition. The 125V System is divided into two independent divisions - Division I and Division 11 - with separate DC power supplies. These power supplies consist of two separate 125V batteries and chargers serving systems such as RCIC, RIIR, EDGs, and HPCI.
Complete loss of both 125V DC Divisions could compromise the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations.
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power
- 2. Abnormal Operating Procedure (AOP) 388, Loss of 250 VDC Power l
- 3. Technical Specification 3.8.B DC Power Systems L
- 4. UFSAR Section 8.3, Onsite Power Systems
- 5. UFSAR Table 8.3-6, Plant Battery System - DC Power, Instrumentation, and Control, Principle DC Loads (125V)
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PAGE S-25 of 33 i
SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD t
SS4 Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown EVENT TYPE: Inability to Maintain Shutdown Conditions OPERATING MODE APPLICABILITY: Run, Startup,llot Shutdown.
EXAMPLE EMERGENCY ACTION LEVEL:
- 1. <EOP Graph 4 Ileat Capacity Limit is exceeded >
<O R>
<2. Reactor CANNOT be brought subcritical.>
This EAL addresses complete loss of functions, including ultimate heat sink and reactivity control, required for hot shutdown with the reactor at pressure and temperature. Under these conditions, there is an actual major failure of a system intended for protection of the public. The reactivity condition criteria is addressed by maintenance of required shutdown margin. Ifinadvertent criticality could not be eliminated q'
by perfbrming the actions of AOP 255.1, AOP 255.2, or the ATWS EOP, it corresponds to a failure of a system intended for the protection of the public and thus classification as a Site Area Emergency is warranted.
This EAL represents an escalation from the conditions of concem in SA3, Inability to Maintain Cold Shutdown, because the reactor is at operating pressure and temperature and decay heat levels are higher.
Per DAEC Technical Specifications, the following systems are necessary to achieve or maintain Ilot Shutdown conditions:
Reactor Protection ystem Instrumentation (T.S. 3.1) s Obre and Containment Cooling Systems Instrumentation (T.S. 3.2)
Reactivity Control (T.S. 3.3) e Standby Liquid Control System (T.S. 3.4) e Core and Containment Cooling Systems (T.S. 3.5) e
/^
Primary System Boundary (T.S. 3.6)
\\
Auxiliary Electrical Systems (T.S. 3.8)
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PAGE S-26 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD Loss of instrumentation is addressed by SS6, Inability to Monitor a Significant Transient in Progress, below. The Auxiliary Electrical System is addressed by SS1, Station Blackout, and SS3, Loss of 125V DC, above and are therefore not covered here. Failure of the primary system boundary is covered by the Fission Barrier Table and SUS, RCS Leakage, above.
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 149, Loss of Decay Heat Removal
- 2. Abnormal Operating Procedure (AOP) 255.1, Control Rod Movement / Indication Abnormal
- 3. Abnormal Operating Procedure (AOP) 255.2, Power / Reactivity Abnormal Change
- 6. Emergency Operating Procedure ALC - Alternate Level Control
(]
- 8. NUAfARC Afethodologyfor Development ofEmergency Action Levels NUAfARC/NESP-007 Revision 2 Questions and Answers, June 1993 O
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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
PAGE S-27 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SS5 Loss of Water Level in the Reactor Vessel That flas or Will Uncover Fuel in the Reactor Vessel EVENT TYPE: Inability to Maintain Shutdown Conditions OPERATING MODE APPLICABILITY: Cold Shutdown, Refuel EXAMPLE EMERGENCY ACTION LEVEL:
- 1. Loss of Reactor Vessel Water Level as indicated by:
- a. Loss of all decay heat removal cooling as determined by (site-specific) procedure.
AND
- b. (Site-specific) indicators that the core is or will be uncovered.
(3 O
There is no significant deviation from the generic EAL. The DAEC EAL is written in terms of the general concern that no cooling water source is lined up or available for injection into the RPV and water level is decreasing below the top of the active fuel (TAF). Under the conditions of concern for EAL 1, AOP 149, Loss of Decay IIcat Removal, would be entered under Tab 1, Loss of Shutdown Cooling.
Indications / alarms related to loss of shutdown cooling are displayed on control room panels IC03 and IC05 and are listed in the procedure. Consistent with the value used in the EOPs, the EAL uses an indicated RPV level of 15 inches for the water level corresponding to TAF.
REFERENCES:
- 1. Abnormal Operating Procedure (AOP) 149, Loss of Decay lieat Removal
Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT O
Rev.2 (forNRCreview) v PAGE S-28 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SS6 Inability to Monitor a Significant Transient in Progress EVENT TYPE: Instrumentation / Communication OPERATING MODE APPLICAHILITY: Run, Startup,Ilot Shutdown EXAMPLE EMERGENCY ACTION LEVEL:
- 1. The following conditions exist:
- a. Loss of < > annunciators asscciated with safety systems.
AND
- b. Compensatory non-alamiing indications are unavailable.
AND
- c. Indications needed to monitor (site-specific) safety functions are unavailable.
AND
- d. <Significant> transient in progress.
The DAEC EAL is written in terms of a sigmficant transient in progress with loss of both safety system annunciators and loss of compensatory non-alarming instrumentation. The DAEC EAL structure, which addresses all the key points in the generic EAL, better assures that the condition of concem for this EAL i
will be readily recognized.
)
Sigmficant transient includes response to automatic or manually initiated functions such as scrams, f
runbacks involving greater than 25% thermal power change, ECCS injections, or thermal power j
oscillations of 10% or greater.
1 Control room panels IC03, IC04, and IC05 contain the annunciators associated with safety systems at DAEC. Annunciators on IC03,1C04, and 1C05 share a common power supply from 125 VDC Division 1 that is fed through circuit breaker 1D13. Therefore, DAEC does not specify a loss of"most" annunciators i
as specified in the generic methodology.
Compensatory non-a/ arming indications include the plant process computer, SPDS, plant recorders, or plant instrument displays in the control room. These indications are needed to monitor (site-specific) v/
safety functions that are of concem in the generic EAL.
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
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i PAGE S-29 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD Indications ofloss of annunciators associated with safety systems include:
1 125 VDC charger, battery, or system annunciators on control room panel IC08 e
Loss of" sealed in" annunciators at affected panels e
Failure of affected annunciator panels shiflily testing by plant operators e'
Expected alarms are not received Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to panels 1C03, e
1C04, and 1C05)
REFERENCES:
- 1. Operating Instruction (OI) No. 317.2, Annunciator System
- 2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power
- 3. Abnonnal Operating Procedure (AOP) 302.2, Loss ofAlarm Panel Power l
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Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview) g PAGE S-30 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD i
SGI Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power i
EVENT TYPE: Loss of Power OPERATING MODE APPLICAHILITY: Run, Startup,llot Shutdown EXAMPLE EMERGENCY ACTION LEVEL:
- 1. Prolonged loss of all offsite and onsite AC power as indicated by:
- a. Loss of power to (site-specific) transformers.
AND
- b. Failure of(site-specific) emergency diesel generators to supply power to emergency busses.
AND j
- c. At least one of the following conditions exist:
Restoration of at least one emergency bus within (site-specific) hours is NOTlikely OR (Site-Specific) Indication of continuing degradation of core cooling based on Fission Product Barrier monitoring.
I There is no significant deviation from the generic EAL. Under prolonged Station Blackout (SBO) conditions, fission product barrier monitoring capability may be degraded. Although it may be difficult to predict when power can be restored, it is necessary to give the EC/OSS a reasonable idea of how quickly a General Emergency should be declared based on the following considerations:
Are there any present indications that core cooling is already degraded to the point where a General Emergency is IMMINENT (i.e., loss of two barriers and a potential loss of the third barrier)?
If there are presently no indications of degraded core cooling, how likely is it that power can be restored prior to occurrence of a General Emergency?
The first part of this EAL corresponds to the threshold conditions for Initiating Condition SS1, Station Blackout - namely, entry into AOP 301.1, Station Blackout. The second part of the EAL addresses the conditions that will escalate the SBO to General Emergency. Occurrence of any of the following is sufficient for escalation: (1) SBO coping capability exceeded, or (2) loss of drywell cooling that continues
Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview) 7O l
PAGE S-31 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD to make RPV water level measurements unreliable, or (3) indications ofinadequate core cooling. Each of these conditions is discussed below:
1.
SBO Coning Canability Exceeded i
DAEC has a SBO coping duration of four hours. 7he likelihood ofrestoring at least one emergency bus should be based on a realistic appraisal ofthe situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing publicprotective actions.
2.
RPV Water I evel Measurements Remaining Unreliable Flashing of the reference leg water will result in erroneously high RPV water level readings giving a false CN indication of actual water inventory and potentially indicating adequate core cooling when it may not exist.
U EOP Graph 1, RPV Saturation Temperature, defmes the conditions under which RPV level instrument leg boiling may occur.
3.
Indications ofInadeauate Core Cooling DAEC uses the RPV level that is used for the Fuel Clad EAL 2 " potential loss" condition. This is RPV level below +15 inches.
REFERENCES:
1.
Abnormal Operating Procedure (AOP) 301.1, Station Blackout 2.
Letter NG-92-0283, John F. Franz, Jr. to Dr. Thomas E. Murley, Response to Safety Evaluation by NRC-NRR " Station Blackout Evaluation Iowa Electric Light and Power Company Duane Arnold Energy Center," February 10,1992 3.
Emergency Operating Procedure (EOP)1 - RPV Control 4.
Emergency Operating Procedure (EOP) ALC - Altemate Level Control t
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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.2 (forNRCreview)
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O PAGE S-32 of 33 i
SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SG2 Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core EVENT TYPE: RPS Failure i
OPERATING MOL)E APPLICAHILITY: Run,Startup i
EXAMPLE EMERGENCY ACTION LEVEL:
- 1. The following conditions exist:
i
- a. (Site-specific) indications exist that automatic and manual scram were NOT successful.
AND O
- b. Either of the following:
V (Site-specific) indication exists that the core cooling is extremely challenged.
OR (Site-specific) indication exists that heat removal is extremely challenged.
Automatic and manual scram are not considered successful if action away from the reactor control console l
is required to scram the reactor. Consistent with the EOPs, the ATWS conditions of concern in this EAL are reactor power that is expected to remain above 5% or that is indeterminate.
Escalation to the General Emergency classification requires extreme challenge to core or containment cooling, i.e., imminent barrier loss. If the Main Condenser is available for steam release from the reactor, sufficient heat removal capability exists, llowever, without the main condenser being available as a heat sink, heat removal capability under these conditions is insufficient and a threat to the Fuel Clad barrier exists. In addition, the SRV's will lifl and thus the RCS barrier will not retain fission products. Eventually, the torus water will be heated to the point whem the containment function will become ineffective. Thus, the resultant combination of barrier conditior.s warrants a declaration of a General Emergency if ATWS reactor power-level control methods are inefYective in reducing reactor power level.
REFERENCES:
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