Text

Emergency Plan Implementing Procedures for Cooper Nuclear Station
	ML031400336
Person / Time
Site:	Cooper
Issue date:	05/14/2003
From:	Hutton J; Nebraska Public Power District (NPPD)
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NLS2003056
	Download: ML031400336 (122)
	v • d • e

Nebraska Public Power District Always there when you need us NLS2003056 May 14, 2003 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

Emergency Plan Implementing Procedures Cooper Nuclear Station, NRC Docket 50-298, DPR-46 Pursuant to the requirements of 10 CFR 50, Appendix E, Section V, "Implementing Procedures,"

Nebraska Public Power District is transmitting the following Emergency Plan Implementing Procedures (EPIPs):

EPIP 5.7.1 Revision 30 "Emergency Classification" EPIP 5.7.17 Revision 31 "Dose Assessment" Should you have any questions concerning this matter, please contact me.

Sincerely, Hutton Plant Manager

/jr/nr Enclosures cc: Regional Administrator w/enclosure (2) NPG Distribution w/o enclosure USNRC - Region IV Senior Resident Inspector w/enclosure Records w/o enclosure USNRC COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 www.nppd.com

ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS Correspondence Number: NLS2003056 The following table identifies those actions committed to by Nebraska Public Power District (NPPD) in this document. Any other actions discussed in the submittal represent intended or planned actions by NPPD. They are described for information only and are not regulatory commitments. Please notify the NL&S Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.

COMMITTED DATE COMMITMENT OR OUTAGE NONE

,~~~~~~~~~~~II I PROCEDURE 0.42 l REVISION 12 l PAGE 14 OF 16

CNS OPRATIONMANUALUSE: REFERENCE EPIP A7 NA EFFECTIVE: 4/24/03 I E 5. APPROVAL: SORC/IQA EMERGENCY CLASSIFICATION DEPARTMENT: EP

1. PURPOSE ............................................................ 1

2. PRECAUTIONS AND LIMITATIONS ..................................... 1

3. REQUIREMENTS ......... . ........................................... 1

4. CLASSIFICATION AND DECLARATION ................................... 1

5. CLASSIFICATION GUIDANCE .......................................... 3

6. RECLASSIFICATION .................................................. 4 ATTACHMENT 1 EAL MATRIX ....................................... 5 ATTACHMENT 2 EMERGENCY ACTION LEVELS ....................... 7 ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS

. .................................................. 80 ATTACHMENT 4 EAL HARDCARDS .................................. 85 ATTACHMENT 5 INFORMATION SHEET ............................. 86

1. PURPOSE This procedure provides the formal set of threshold conditions necessary to classify an event at CNS into one of the four emergency classifications described in NUREG-0654 and the CNS Emergency Plan.

2. PRECAUTIONS AND LIMITATIONS

[ ] 2.1 The steps required by this procedure are in addition to the steps required to maintain or restore the station to a safe condition.

[ ] 2.2 If conflicts in personnel assignments or sequence of actions arise, first priority will be given to maintaining or restoring the station to a safe condition.

3. REQUIREMENTS

[ ] 3.1 An Emergency Operation Procedure has been initiated; or

[.1 3.2 An unusual occurrence has taken place at or near the site.

4. CLASSIFICATION AND DECLARATION

[ ] 4.1 After recognition of an off-normal event, Shift Supervisor shall:

[] 4.1.1 Compare the event to EALs in Attachments 1, 2, 3, and 4.

PROCEDURE 5.7.1 REVISION 30 PAGE 1 OF 87

[ ] 4.1.2 If more than one EAL of different classification levels is reached, i.e.,

an EAL for ALERT or an EAL for SITE AREA EMERGENCY, select EAL for most severe emergency classification.

[1 ] 4.1.3 If the event appears to meet an EAL, refer to Attachment 2 for further explanation and guidance.

[1 ] 4.1.4 If it is determined that an EAL is met:

[] 4.1.4.1 Assume Emergency Director responsibilities until relieved by another qualified Emergency Director.

[] 4.1.4.2 Declare the emergency.

[ ] 4.1.4.3 Record the emergency class, time of declaration, and EAL number in the Shift Supervisor's Log.

[ ] 4.1.4.4 Enter Procedure 5.7.2 and perform the actions directed.

[] 4.1.4.5 Continue to monitor and re-evaluate emergency classification per this procedure until the event is terminated-

[ ] 4.1.5 When relieved of Emergency Director duties by another qualified Emergency Director located in the EOF, the Shift Supervisor shall no longer be responsible for performance of actions specified in this procedure or Procedure 5.7.2.

[] 4.1.5.1 The Emergency Director may direct the Shift Supervisor to perform specific actions, such as activation of emergency alarm, which can only be performed from the Control Room.

[] 4.1.5.2 The Shift Supervisor shall bring to the attention of the Emergency Director, changing plant conditions which may affect the emergency classification.

PROCEDURE 5.7.1 I REVISION 30 l PAGE 2 OF 87

5. CLASSIFICATION GUIDANCE

[ ] 5.1 Four standardized emergency classes have been established; they are:

[] 5.1.1 NOTIFICATION OF UNUSUAL EVENT

[] 5.1.1.1 This classification is comprised of events in progress, or which have occurred, that indicate a potential degradation of the level of safety of the station. These types of events may progress to a more severe emergency classification if they are not mitigated.

No releases of radioactive material requiring off-site response or monitoring are expected unless further degradation of safety systems occurs.

[] 5.1.2 ALERT

[] 5.1.2.1 This classification is comprised of events in progress, or which have occurred, that involve an actual or potentially substantial degradation of the safety level of the station. At this classification level, minor releases of radioactivity may occur or may have occurred. Any releases expected to be limited to small fractions of EPA Protective Action Guideline exposure levels.

[] 5.1.3 SITE AREA EMERGENCY

[1 ] 5.1.3.1 This classification is comprised of events in progress, or which have occurred, which involve actual or potential major failure of plant functions needed for protection of the public. Releases are not expected to exceed EPA Protective Action Guidelines, except near the Site Boundary.

[] 5.1.4 GENERAL EMERGENCY

[] 5.1.4.1 This classification is comprised of events in progress, or which have occurred, that involve actual or imminent substantial core degradation or melting with a potential for the loss of primary containment integrity. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels off-site for more than the immediate site area.

[ ] 5.2 Possible events are divided into eight categories which are intended to bracket the Initiating Conditions listed in NUREG-0654, Revision 1, Appendix 1, as further defined and revised by Reference 3.3.6. The eight categories are:

[] 5.2.1 Radiological.

[] 5.2.2 Fission product barrier threat or loss.

I PROCEDURE 5.7.1 l REVISION 30 l PAGE 3 OF 87

[] 5.2.3 Operational.

[] 5.2.4 Power or alarms.

[ 5.2.5 Fire; flammable or toxic material.

[ ] 5.2.6 Security.

[ ] 5.2.7 Natural phenomenon.

[ 5.2.8 Other hazards.

5.3 Prompt recognition of the occurrence of one or more initiating events may prevent the situation from progressing to a classification of greater severity.

5.4 An emergency may warrant classification as a result of a combination of two or more events. Ensure each abnormal condition is evaluated against classification criteria.

5.5 The EAL Matrix (Attachments 1 and 4) is designed to assist in quickly locating the appropriate category of accident. The matrix is not to be used independently of the rest of the procedure when making classification decisions.

5.6 For classification purposes, grams, CCs, and milliliters are equivalent.

1 ,uCi/gm - 1 lCi/cc 1 tCi/ml

6. RECLASSIFICATION 6.1 An emergency may escalate to a higher classification if station conditions deteriorate or as a result of a combination of two or more events.

6.2 An emergency may be initially classified at one class and, upon further investigation or after corrective actions, may be reclassified or terminated.

6.3 If any GENERAL EMERGENCY has been declared, consultation with state authorities and the NRC should occur prior to reclassification or termination of the event.

6.4 Compare changing station conditions with the Emergency Action Levels in Attachment 2 and reclassify, as necessary.

I PROCEDURE 5.7.1 l REVISION 30 l PAGE 4 OF 87

ATTACHMENT 1 EAL MATRIX Emergency Class NOUE Alert 1.1.1 Uncontrolled. unmonitored radiological release of liquid outside the 1.2.1 Loss of control of radioacive materiat resulting Inarea radiation Protected Area. exceeding iOOX normal (or expected) levels within the Protected Area. Normal Isdetermined by trend recorder or other relevant Radiological 1.1.2 Off-Site Dose Assessment Manual (ODAM) lirits exceeded as Indicated by data.

a HIGH-HIGH alarm on a gaseous effluent radiological monitor which cannot be cleared within 30 minutes. 122 Gaseous effluent radiological monitors Indicate a release rate ten times the Off-Site Dose Assessment Manual (ODAM) limits.

without ndicaton of fuel cladding loss.

Fission 2.1.1 Steam Jet AirEjector radiation monitor reads > 1.5 E+3 mrem/hr or an 22.1 Loss of fuel cladding or Primary Coolant Boundary fission product Increase of 3.0 E+2 mrem/hr within a 30 minute period. barriers (refer to Attachment 3 for indicalion).

Product 2.1.2 Coolant sample activity exceeds 4 pCVgm DOSE EQUIVALENT 1-131.

Barrier 2.1.3 Any operaUonal RCS pressure boundary LEAKAGE; or unidentified Threat LEAKAGE exceeds S gpm: or total LEAKAGE exceeds 30 gpm averaged over a previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period; or unidentified LEAKAGE Increase of more or than 2 gpm within the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period InMODE 1.

Loss 3.1.1 Inability to meet the Action Statement associated with a Technical 3.2.1 Fuel handling accident on the refueling floor with release of Specification Limiting Conditon for Operation (LCO). radioactivity to secondary containment as indicated by HIGH alarm on refueling floor ARM #2. CAM. or Reactor Building venblaton monitor.

322 Evacuation of Control Room required or anticipaled with control of Operational shutdown systems estabtished from local stations.

32.3 Complete toss of capability to place or maintain the plant In MODE 4 or 5.

32.4 Failure of Reactor Protection System (RPS) to nitiate and complete a scram which brings the reactor suboritical.

4.1.1 Loss of ALL off-site power sources to vital busses F and G for 4.2.1 Loss of all AC power (on and off-site sources) to vital busses F Power ,15 minutes. and G during MODE 4 or 5.

or 4.1.2 Unplanned loss of most or a safety system annunciators. 4.2.2 Loss of all DC power sources resulting in loss of al ECCS capability for < 15 minutes.

Alarms 4.2.3 Unplanned loss of most or all safety system annunciators with a transient in progress.

Fire 5.1.1 Any fire within the Protected Area which takes longer than 10 minutes to extnguish.

52.1 A fire with a potential to cause degradation of a plant safety system required to be OPERABLE.

Flammable 5.12 Report or detection of toxic or flammable gases that could enter the 522 Report or detection of toxic or flammable gases within a Vital Area Toxic Protected Area Inamounts that will affect the health of plant personnel or can effect nornal operaton of the plan.

in concentrations that will be life threatening to plant personnel or wiltaffect the safe operation of the plant.

Security 6.1.1 Security threat. attempted entry, or attempted sabotage. 6.2.1 On-going security compromise.

7.1.1 Ground moton > 0.01g as Indicated by Control Room seismic monitoring 7.2.1 Ground motion > 0.1 g as ndicated by Control Room seismic panel. monitoring panel.

Natural 7.12 River level> 899'or 86r. 722 River level> 90Z or < 865.

Phenomenon 7.1.3 Tomado touching down within the Owner Controled Area. 72.3 Tomado touching down within the Protected Area.

7.1.4 Sustained wind speed > 74 mph. 72.4 Sustained wind speed > 95 mph.

8.1.1 Aircraft crash within the Protected Area. 82.1 Aircraft striking structures within the Protected Area.

8.12 Explosion within the Protected Area. 82.2 Missile mpacL from whatever source, within the Protected Area.

Other 8.1.3 Failure of a turbine rotatng component causing an automatic reactor scram 82.3 Known explosion damage to the facility affectng plant operation.

with release of radioacivity to the Turbine Bulding or which potentially Hazards affects safety systems. 8.2.4 Turbine failure causing casing penetration which creates serious radiological concems or damages plant safety systems.

8.1.4 Other conditions existing which Inthe Judgement of the Emergency Director warrant declaraton of an Usual Event. 82.5 Other conditions existing which in the Judgement of the Emergency Director warrant declaration of an Aiert.

PROCEDURE 5.7.1 REVISION 30 PAGE 5 OF 87

ATTACHMENT 1 EAL MATRIX Site Area Emergency General Emergency 1.3.1 Radiological gaseous effluent releases resulting In Total Effeclive Dose Equivalent (TEDE) 1.4.1 Radiological gaseous eMuent releases resulting In Total Effective Dose projection at or beyond the Site Boundary of > 0.1 REM. Equivalent (TEDE) dose at or beyond the Site Boundary of 1 REM.

1.32 Radiological gaseous effluent releases resulting In Comrrilled Dose Equivalent (CDE) (thyroid) 1.A2 Radiological gaseous effluent releases resulting in Comnitted Dose projection at or beyond the Site Boundary of > 0.5 REM. Equivalent (CDE) (thyroid) dose at or beyond the Site Boundary of 5 REM.

2.3.1 Degraded core with a possible loss of coolable geometry as Indicated by: 2.4.1 Loss of any TWO of THREE fission product barriers AND the potential exists for loss of the THIRD. The fission product barriers are defined as A.1 Greater than or equal to 20% gap activity as determined by Cheristry. follows (refer to Attachment 3 for Indication):

OR A.2 Prinary Containment radiaton monitors read > 1.0 E+4 REMhr. A. Fuel Cladding.

AND B. Primary Coolant Boundary.

B.1 High core plate Dp for the corresponding core flow. C. Primary Containment.

OR B.2 Inability to Insert In-core detectors.

2.3.2 Known loss of coolant accident greater than makeup capadty.

2.3.3 Loss of any TWO fission product barriers. The fission product barriers are defined as follows (refer to Altachment 3 for indication):

A. Fuel Cadding.

B. Primary Coolant Boundary.

C. Primary Containment.

3.3.1 Major damage to irradiated fuel or fuel pool water level below the top of the spent fuel. 3.4.1 Failure of the Reactor Protection System (RPS) or altemate rod Insertion or SLC to bring the reactor subcritical which could result in a core 3.32 Evacuafion of the Control Room accompanied by the Inability to locally control shutdown meltdown vwithsubsequent containment failure fikely.

systems within 15 minutes.

3.4.2 Other plant conditions exisL from whatever source, which make a 3.3.3 Complete loss of all available means to place or maintain the plant In MODE 3. release of large amounts of radioactivity in a short time possible (e.g..

any core melt situation).

3.3.4 Failure of the Reactor Protection System (RPS), including Altemate Rod nserion (ARI), to bring the reactor subcritical.

4.3.1 Loss of all AC power (on and orffsite sources) for more than 15 minutes with the Reactor In 4.4.1 Total loss of all AC power (on and off-site sources) with the Inability to MODE 1. 2 or 3. keep the core covered.

4.3.2 Loss of an DC power sources required for ECCS operation for more than 15 rrnutes.

4.3.3 Inability to monitor a significant transient In progress.

5.3.1 Fire compromising the funcions of safety systems. 5.4.1 Any major intemal or extemal fire substanially beyond the design basis which could cause massive common damage to plant systems.

6.3.1 Imminent loss of physical control of the staUon. 6.4.1 Loss of physical control of the staton.

7.3.1 Ground motion > 0.1 g as Indicated on the Control Room seisrric moniltoring panel AND reports 7.4.1 Any major natural phenomenon substantially beyond the design basis of major plant damage. which could cause massive common damage to plant syslems.

7.32 Sustained wind speed > 100 mph.

73.3 Flood which renders mulUple ECCS systems Inoperable when they are required to be OPERABLE.

7.3.4 Low river level which results In compiete toss of the Service Water System.

8.3.1 Aircraft crash affecUng vital areas with the plant In MODE 1. 2, or 3. 8.4.1 Other conditons existing which In the judgement of the Emergency Director warrant declaraton of a General Emergency (.e_ any core melt 8.32 Missile or explosion damage to safe shutdown equipmentwith the plant In MODE 1, 2, or 3. situaUon).

8.3.3 Other conditions existing which In thejudgement of the Emergency Directorwarrant dectaration of a Site Area Emergency.

PROCEDURE 5.7.1 REVISION 30 PAGE 6 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 1.1.1 NOUE TEXT Uncontrolled, unmonitored radiological release of liquid outside the Protected Area.

APPLICABILITY ALL EXAMPLE Unisolable leak from a condensate storage tank into the discharge canal.

MEMO The actual dose is generally not the primary concern; it is the degradation in plant control implied by the fact that the release was not isolated. To be conservative, it is to be assumed that any radiologically contaminated liquid released off-site in an uncontrolled, unmonitored fashion has the potential to exceed RETS limits. Therefore, any uncontrolled, unmonitored release of radioactive liquid outside the Protected Area will meet this EAL.

REFERENCES NUREG-0654: N.02 I PROCEDURE 5.7.1 l REVISION 30 l PAGE 7 OF 87 l

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 1.1.2 NOUE TEXT Off-Site Dose Assessment Manual (ODAM) limits exceeded as indicated by a HIGH-HIGH alarm on a gaseous effluent radiological monitor which cannot be cleared within 30 minutes.

APPLICABILITY ALL EXAMPLE Turbine Building KAMAN alarms. "TG BLDG VENT HIGH-HIGH RAD" annunciator is received. Release is verified, but cannot be stopped.

MEMO The HIGH-HIGH alarm in the text of this EAL refers to the normal range KAMAN.

Each gaseous effluent stream has two alarm setpoints. Under normal circumstances, the high alarm will come in first allowing operator action to stop or reduce the release.

The HIGH-HIGH alarm is set at (or near) the RETS release rate limit. Because the RETS limit (being based on a yearly continuous dose projection) is extremely conservative, the 30 minute delay in verifying the alarm and attempting to clear it is justified.

Reduce power or isolate systems as appropriate. If alarm is valid, and release cannot be reduced to below RETS release rate limits or terminated in 30 minutes, declare.

REFERENCES NUREG-0654: N.02 I PROCEDURE 5.7.1 l REVISION 30 l PAGE 8 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 1.2.1 ALERT TEXT Loss of control of radioactive material resulting in area radiation exceeding 1000X normal (or expected) levels within the Protected Area. Normal is determined by trend recorder or other relevant data.

APPLICABILITY ALL EXAMPLE Radiography source becomes uncoupled and lost. RP survey indicates direct radiation has increased by > 1000 times.

MEMO By themselves, indications of increased levels of radiation only meet the NOUE class description; however, when combined with "loss of control" a higher classification is warranted. Non-essential personnel should be assembled off-site. Additional manpower or other resources will likely be needed. The ALERT classification is appropriate.

The operative phrase in this EAL is "loss of control". Combined with this is the phrase "or expected levels". For most plant evolutions increases of radiation can be estimated, most within a factor of 1000. If, in the judgement of those concerned, control has been lost, AND radiation levels increase beyond 1000X normal or expected levels, declare.

REFERENCES NUREG-0654: A.06 NUREG-0654: A.12 I PROCEDURE 5.7.1 l REVISION 30 - PAGE 9 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 1.2.2 ALERT TEXT Gaseous effluent radiological monitors indicate a release rate ten times the Off-Site Dose Assessment Manual (ODAM) limits without indication of fuel cladding loss.

APPLICABILITY ALL EXAMPLE Operating at 100% power AOG is lost. ERP KAMAN reading goes to 1.13 E+7 iLCi/sec.

MEMO This ERP KAMAN reading will exceed ten times the ODAM instantaneous limit. Rely on the PMIS "ten times ODAM Limit Exceeded" flag.

If there are any indications that the fuel cladding is not intact (fuel has been uncovered, SJAE monitors > 1.5 E+4 mrem/hr, PASS sample, Primary Containment radiation monitors > 2.5 E+3 REM/hr, or other) the iodine component will result in a higher dose and may also warrant a higher classification.

NOTE - Radiation release resulting in an ALERT is an EOP entry condition.

REFERENCES NUREG-0654: A.15 PROCEDURE 5.7.1 REVISION 30 l PAGE 10 OF 87 l

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 1.3.1 SITE AREA EMERGENCY TEXT Radiological gaseous effluent releases resulting in Total Effective Dose Equivalent (TEDE) projection at or beyond the Site Boundary of > 0.1 rem.

APPLICABILITY ALL EXAMPLE ARW KAMAN reads 5 E+6 pCiIsec. With default wind speed (8 mph) and stability class (D), Standby Gas Treatment is not in the release path, the core is not degraded, secondary containment is bypassed, and the reactor not shutdown, an integrated dose for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months at 1 mile of > 0.1 REM TEDE is projected.

MEMO If a release greater than license limits is under way, or suspected, and anv dose assessment model or methodology indicates a Site Boundary integrated TEDE dose of

> 0.1 rem, classify and follow applicable procedures. This is the conservative response.

Conservative is defined as that action which yields the greatest possible protection of the public from radiological consequences.

This EAL is related to integrated dose; therefore, the estimated length of release is critical to obtain an accurate integrated dose projection. As conditions change, dose projections should be re-calculated.

REFERENCES NUREG-0654: S.13 I PROCEDURE 5.7.1 l REVISION 30 l PAGE 11 OF87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 1.3.2 SITE AREA EMERGENCY TEXT Radiological gaseous effluent releases resulting in Committed Dose Equivalent (CDE)

(thyroid) projection at or beyond the Site Boundary of > 0.5 REM.

APPLICABILITY ALL EXAMPLE ERP KAMAN reads 2 E+6 tiCi/sec. The core has been uncovered (dose assessment question on core degraded = YES). SBGT is not in the path. The reactor has been shutdown for 30 minutes and secondary containment has been bypassed. With default wind speed (13 mph) and stability class (D), a CDE dose > 0.5 rem over 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is projected.

MEMO If a release greater than license limits is under way, or suspected, and anv dose assessment model or methodology indicates a Site Boundary integrated CDE dose of

> 0.5 rem, classify and follow applicable procedures. This is the conservative response.

Conservative is defined as that action which yields the greatest possible protection of the public from radiological consequences.

REFERENCES NUREG-0654: S.13 PROCEDURE 5.7.1 l REVISION 30 l PAGE 12 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 1.4.1 GENERAL EMERGENCY TEXT Radiological gaseous effluent releases resulting in Total Effective Dose Equivalent (TEDE) dose at or beyond the Site Boundary of 1 REM.

APPLICABILITY ALL EXAMPLE Turbine Building KAMAN reads 2 E+8 ILCi/sec. With default wind speed (8 mph) and stability class (D), Standby Gas Treatment is not in the release path, the core is not degraded, secondary containment is not bypassed, the release is expected to last 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and the reactor not shutdown a TEDE dose > 1 REM is projected at or beyond 1 mile.

MEMO If a release greater than license limits is under way, or suspected, and anv dose assessment model or methodology indicates a Site Boundary TEDE dose of 1 rem or greater, classify and follow applicable procedures. This is the conservative response.

Conservative is defined as that action which yields the greatest possible protection of the public from radiological consequences.

NUREG-0654 requires that a GENERAL EMERGENCY be declared when EPA Protective Action Guidelines are projected to be exceeded off-site.

REFERENCES NUREG-0654: G.01 PROCEDURE 5.7.1 l REVISION 30 l PAGE 13 OF 87 l

ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 1.4.2 GENERAL EMERGENCY TEXT Radiological gaseous effluent releases resulting in Committed Effective Dose (CDE)

(thyroid) dose at or beyond the Site Boundary of 5 REM.

APPLICABILITY ALL EXAMPLE Turbine KAMAN reads 2.6 E+6 pCi/sec. The core has been uncovered (dose assessment question on core degraded = YES). With wind default wind speed (8 mph) and stability class (D), Standby Gas Treatment is not in the release path, secondary containment is bypassed, the reactor is not shutdown, and the release is expected to last 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , a CDE dose at or beyond 1 mile is projected to be > 5 REM.

MEMO If a release greater than license limits is under way, or suspected, and anv dose assessment model or methodology indicates a Site Boundary CDE dose rate of 5 rem/hr or greater, classify and follow applicable procedures. This is the conservative response. Conservative is defined as that action which yields the greatest possible protection of the public from radiological consequences.

NUREG-0654 requires that a GENERAL EMERGENCY be declared when EPA Protective Action Guidelines are projected to be exceeded off-site.

REFERENCES NUREG-0654: G.01 I PROCEDURE 5.7.1 l REVISION 30 l PAGE 14 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 2.1.1 NOUE TEXT Steam Jet Air Ejector radiation monitor reads > 1.5 E+3 mrem/hr or an increase of 3.0 E+2 mrem/hr within a 30 minute period.

APPLICABILITY ALL EXAMPLE RM-150A reads > 1.5 E+3 mrem/hr.

MEMO These numbers correspond to some fuel damage. They do not reflect a LOSS of the fuel cladding.

REFERENCES NUREG-0654: N.03A I PROCEDURE 5.7.1 l REVISION 30 l PAGE 15 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 2.1.2 NOUE TEXT Coolant sample activity exceeds 4.0 liCi/gm DOSE EQUIVALENT I-131.

APPLICABILITY ALL EXAMPLE Rx coolant sample results indicate 5.0 [iCi/gm DOSE EQUIVALENT I-131.

MEMO 0.2 piCi/gm DOSE EQUIVALENT I-131 is the Tech Spec limit. The limit may be increased up to 4.0 [tCi/gm DOSE EQUIVALENT I-131 or less for a maximum of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to allow a reasonable time for temporary coolant activity increases (iodine spikes or crud bursts) to be cleaned up with the normal processing systems. If at any time the DOSE EQUIVALENT I-131 > 4.0 iLCi/gm DOSE EQUIVALENT I-131, it must be determined at least once every four (4) hours and all the main steam lines must be isolated with 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . See LCO 3.4.6 for details.

REFERENCES NUREG-0654: N.03B Tech Spec 3.4.6 NOTE - For purposes of reactor coolant samples:

1 iiCi/ml - lCi/cc - lCi/mg dose equivalent I-131 I PROCEDURE 5.7.1 l REVISION 30 l PAGE 16 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 2.1.3 NOUE TEXT Operational RCS pressure boundary LEAKAGE; or unidentified LEAKAGE exceeds 5 gpm; or total LEAKAGE exceeds 30 gpm averaged over a previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period; or unidentified LEAKAGE increase of more than 2 gpm within the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period in MODE 1.

APPLICABILITY MODE 1, 2, or 3.

EXAMPLE Sump integrators indicate leakage from the primary coolant boundary of 7 gpm unidentified.

MEMO This leak rate constitutes entry into a LCO; however, this case will not wait for inability to meet associated action statement(s); therefore, declare a NOUE upon confirmation of the leak rate.

REFERENCES NUREG-0654: N.05 I PROCEDURE 5.7.1 l REVISION 30 l PAGE 17 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 2.2.1 ALERT TEXT Loss of Fuel Cladding or Primary Coolant Boundary fission product barriers (refer to Attachment 3 for indication).

APPLICABILITY Per Technical Specifications EXAMPLE Reactor Recirculation pump seizure leading to fuel cladding failure.

PASS sample results show > 300 txCi/gm DOSE EQUIVALENT I-131.

-OR Loss of Coolant Accident.

MEMO Refer to Attachment 3 for indications of lost fission product barriers to ensure that only one barrier is lost. Loss of two barriers is a SITE AREA EMERGENCY (EAL: 2.3.3), loss of two barriers with the potential loss of the third is a GENERAL EMERGENCY (EAL: 2.4.1).

REFERENCES NUREG-0654: A.01 NUREG-0654: A.04 NUREG-0654: A.05 NUREG-0654: A.09 NUREG-0654: N.06 I PROCEDURE 5.7.1 l REVISION 30 l PAGE 18 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 2.3.1 SITE AREA EMERGENCY TEXT Degraded core with a possible loss of coolable geometry as indicated by:

A.1 20% gap activity as determined by Chemistry.

OR A.2 Primary Containment radiation monitors read > 1.0 E+4 REM/hr.

AND B.1 High core plate Dp for the corresponding core flow (see EAL: 2.3.1A).

OR B.2 Inability to insert in-core detectors.

APPLICABILITY ALL EXAMPLE Drywell radiation monitors read 2 E+4 REM/hr following a transient. Traversing In-Core Probes cannot be inserted by any machine into the reference channel.

MEMO The term "degraded core" in the EAL text refers to a significantly degraded core (e.g.,

20% clad failure).

Could lead to further core degradation due to overheating.

Reference Dp vs. Core Flow Chart, Figure 1, on EAL 2.3.1A (next page).

REFERENCES NUREG-0654: S.02 I PROCEDURE 5.7.1 l REVISION 30 l PAGE 19 OF87

( ( (

11S 0

Core Plate dP vs Core Flow (for determination of degraded core) 26 O

24 C 22 20

-2j 18 0 16 CL 14 IL a

.9 12 L1 0

8 8

6 z 4 co O) 0 2 0

30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 110 5-7-1A.SCAN Core Flow (% of 73.5 m#/hr) 0

-D It 00

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 2.3.2 SITE AREA EMERGENCY TEXT Known loss of coolant accident greater than makeup capacity.

APPLICABILITY ALL EXAMPLE LOCA greater than RCIC capacity with HPCI inop and inability to depressurize.

MEMO This EAL is a combination of loss of one fission product barrier (RPV) and other major failures. It therefore meets the class description for SITE AREA EMERGENCY of NUREG-0654.

Follow Emergency Operating Procedures (EOPs). If all means to maintain level in the reactor fail, declare.

REFERENCES NUREG-0654: S.01 PROCEDURE 5.7.1 l REVISION 30 l PAGE 21 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 2.3.3 SITE AREA EMERGENCY TEXT Loss of any TWO fission product barriers. The fission product barriers are defined as follows:

A. Fuel Cladding.

B. Primary Coolant Boundary.

C. Primary Containment.

APPLICABILITY Per Technical Specifications.

EXAMPLE Steam line break outside primary containment without isolation from the Control Room.

OR 100 gpm leak into Primary Containment following fuel failure (> 300 ,uCi/gm DOSE EQUIVALENT I-131).

OR Primary Containment isolation failures allowing a direct flow path to the environment such as failures of both MSIVs to close with open valves downstream to the turbine or to the condenser.

MEMO TWO, and only two, fission product barriers must meet the criteria for being considered lost. If there is only one barrier lost, see EAL: 2.2.1. If there is the potential for loss of the third barrier a GENERAL EMERGENCY shall be declared on EAL: 2.4.1.

See Attachment 3 for indications of loss or potential loss of fission product barriers.

REFERENCES NUREG-0654: S.04 PROCEDURE 5.7.1 l REVISION 30 l PAGE 22 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 2.4.1 GENERAL EMERGENCY TEXT Loss of any TWO of THREE fission product barriers AND the potential exists for the loss of the THIRD. The fission product barriers are defined as follows:

A. Fuel Cladding.

B. Primary Coolant Boundary.

C. Primary Containment.

APPLICABILITY Per Technical Specifications.

EXAMPLE LOCA with core damage and drywell pressure is nearing design pressure, OR two MSIVs on the same steam line cannot be isolated from the Control Room and chemistry data trends indicate fuel cladding is deteriorating.

MEMO See Attachment 3 for indications of loss or potential loss of fission product barriers.

REFERENCES NUREG-0654: G.02 NUREG-0654: G.06 PROCEDURE 5.7.1 REVISION 30 PAGE 23 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 3.1.1 NOUE TEXT Inability to meet the action statement associated with a Technical Specification Limiting Condition for Operation (LCO).

APPLICABILITY Per Technical Specifications.

EXAMPLE Following discovery that one of the 125 volt batteries is inoperable, the battery was not restored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , nor was MODE 3 achieved within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

MEMO Declaration of NOUE is warranted by failure to meet the action statement of a Limiting Condition for Operation (LCO). This constitutes a condition outside that analyzed by Technical Specifications. The NOUE may not be terminated until the action statement has been met. This varies; reference the Tech Specs.

REFERENCES NUREG-0654: N.08 NUREG-0654: N.09 NUREG-0654: N.15 PROCEDURE 5.7.1 1 REVISION 30 l PAGE 24 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 3.2.1 ALERT TEXT Fuel handling accident on the refueling floor with release of radioactivity to secondary containment as indicated by HIGH alarm on refueling floor ARM #2, CAM, or Reactor Building ventilation monitor.

APPLICABILITY ALL EXAMPLE Dropped fuel bundle, bubbles appear near the impact zone, ARM #2 alarms.

MEMO For major damage, see EAL: 3.3.1.

REFERENCES NUREG-0654: A.12 I PROCEDURE 5.7.1 l REVISION 30 l PAGE 25 OF 87 l

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 3.2.2 ALERT TEXT Evacuation of Control Room required or anticipated with control of shutdown systems established from local stations.

APPLICABILITY ALL EXAMPLE Electrical fire in the Control Room causes evacuation. ASD accomplished.

MEMO Do not delay alternate shutdown. Declare ALERT and note time. Make required notifications as soon as possible. If control of shutdown systems cannot be accomplished within 15 minutes, EAL: 3.3.2 applies.

This EAL does not say that all actions associated with ASD shall be completed in order to avoid the higher EAL pertaining to Control Room evacuation (EAL: 3.3.2). If the reactor successfully scrams, level and pressure are being controlled, and no impediments to the associated ASD activities are being encountered, this emergency classification is appropriate. If impediments are being encountered in completing critical ASD functions and more than 15 minutes expire, EAL: 3.3.2 is met.

REFERENCES NUREG-0654: A.20 I PROCEDURE 5.7.1 l REVISION 30 l PAGE 26 OF 87 l

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 3.2.3 ALERT TEXT Complete loss of all capability to place or maintain the plant in MODE 4 or MODE 5.

APPLICABILITY Irradiated fuel in the vessel.

EXAMPLE Loss of both LPCI Subsystems following a scram from startup.

MEMO Loss of MODE 4 capability while at power would be adequately covered by Tech Specs, but does not warrant an ALERT.

Follow appropriate procedures. Attempt alternate means of cooling if required. If all means to place or maintain the reactor < 212°F fail, declare. Monitor plant for indications of other EAL thresholds.

REFERENCES NUREG-0654: A.10 I PROCEDURE 5.7.1 l REVISION 30 l PAGE 27 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 3.2.4 ALERT TEXT Failure of Reactor Protection System (RPS) to initiate and complete a scram which brings the reactor subcritical.

APPLICABILITY Reactor critical.

EXAMPLE RPS initiated scram with half the control rods not full in (hydraulic lock caused by an undrained scram discharge volume). Continued power generation.

MEMO A failure of RPS in this EAL is a failure of either the automatic trip systems or the manual scram pushbuttons to initiate and complete a scram which brings the reactor subcritical. If ARI also fails, see EAL 3.3.4. Subcritical is defined as all but one rod full-in, all rods inserted to or beyond Position 02, OR a qualified Reactor Engineer has determined reactor will remain subcritical under all conditions without boron injection.

REFERENCES NUREG-0654: A.11 PROCEDURE 5.7.1 l REVISION 30 l PAGE 28 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 3.3.1 SITE AREA EMERGENCY TEXT Major damage to irradiated fuel or fuel pool water level below the top of the spent fuel.

APPLICABILITY ALL EXAMPLE Shipping cask head dropped on spent fuel. Several fuel bundles prepared for shipment (de-channeled) are crushed.

MEMO Major fuel damage is defined as "affecting more than ten irradiated fuel bundles". It is anticipated that no fuel handling accident associated with normal fuel handling could cause this EAL to be met. Only large objects (such as fuel shipping casks) dropped on fuel, or uncovery of the fuel could meet this EAL.

REFERENCES NUREG-0654: S.10 PROCEDURE 5.7.1 l REVISION 30 l PAGE 29 OF 87 l

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 3.3.2 SITE AREA EMERGENCY TEXT Evacuation of the Control Room accompanied by the inability to locally control shutdown systems within 15 minutes.

APPLICABILITY ALL EXAMPLE Electrical fire in the control room causes evacuation. Shutdown systems are not responding properly from the ASD panel.

MEMO An ALERT should have been declared on EAL: 3.2.1 upon evacuation of the Control Room. When local control cannot be achieved in 15 minutes, a SITE AREA EMERGENCY shall be declared.

REFERENCES NUREG-0654: S.18 PROCEDURE 5.7.1 REVISION 30 PAGE 30 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 3.3.3 SITE AREA EMERGENCY TEXT Complete loss of all available means to place or maintain the plant in MODE 3.

APPLICABILITY MODE 1, 2, or 3.

EXAMPLE Shutdown margin cannot be maintained.

MEMO Could lead to fuel cladding failure.

Carefully monitor plant parameters for indications of fission product barrier loss.

Attempt alternate means of heat removal. If all means of heat removal fail, declare.

Escalation of this EAL to a General Emergency is based on actual or imminent substantial core degradation or melting with potential for loss of primary containment.

REFERENCES NUREG-0654: S.08 PROCEDURE 5.7.1 l REVISION 30 l PAGE 31 OF87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 3.3.4 SITE AREA EMERGENCY TEXT Failure of the Reactor Protection System (RPS), including Alternate Rod Insertion (ARI), to bring the reactor subcritical.

APPLICABILITY Reactor critical.

EXAMPLE Low reactor water level scram with hydraulic lock on all the north HCUs. Half the rods remain un-inserted. Continued power generation.

MEMO If any scram signal and initiation of ARI fails to bring the reactor subcritical, a SITE AREA EMERGENCY based on this EAL exists.

Subcritical is defined as all but one rod full-in, all rods inserted to or beyond Position 02, OR a qualified Reactor Engineer has determined reactor will remain subcritical under all conditions without boron injection.

Escalation of this EAL to a GENERAL EMERGENCY is based on actual or imminent substantial core damage or melting with potential for loss of primary containment.

REFERENCES NUREG-0654: Appendix 1, SITE AREA EMERGENCY, Step 9.

PROCEDURE 5.7.1 l REVISION 30 l PAGE 32 OF 87 l

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 3.4.1 GENERAL EMERGENCY TEXT Failure of the Reactor Protection System (RPS) or alternate rod insertion or SLC to bring the reactor subcritical which could result in a core meltdown with subsequent containment failure likely.

APPLICABILITY Reactor critical.

EXAMPLE All methods to shut down the reactor fail.

MEMO Subcritical is defined as all but one rod full-in, all rods inserted to or beyond Position 02, OR a qualified Reactor Engineer has determined reactor will remain subcritical under all conditions without boron injection or cold shutdown boron per EOPs cannot be injected. All methods to shut down the reactor have failed. If heat sink is lost fuel will eventually be degraded or melt. Loss of heat sink will also degrade the Primary Containment integrity.

REFERENCES NUREG-0654: G.06A PROCEDURE 5.7.1 l REVISION 30 _ PAGE 33 OF 87l

ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 3.4.2 GENERAL EMERGENCY TEXT Other plant conditions exist, from whatever source, which make a release of large amounts of radioactivity in a short time period possible (e.g., any core melt situation).

APPLICABILITY ALL EXAMPLE Event in progress or which has occurred, that involves actual or imminent substantial core degradation or melting with the potential for the loss of Primary Containment integrity.

MEMO Attempt to classify under more specific EALs. If none apply and the potential for large releases or core melt exists, declare.

REFERENCES NUREG-0654: G.04 NUREG-0654: G.06 PROCEDURE 5.7.1 l REVISION 30 l PAGE 34 OF 87 l

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 4.1.1 NOUE TEXT Loss of ALL off-site power sources to vital busses "F" and "G" for > 15 minutes.

APPLICABILITY ALL EXAMPLE Tornado drops all lines feeding the plant. Diesel generators start and load properly.

Lightning strike results in loss of SSST with degraded voltage on the ESST (1FS/1GS autoclosure not permitted) for > 15 minutes.

MEMO The NSST should not be considered a source of off-site power.

The SSST must be supplied by T2 to be considered a source of off-site power.

REFERENCES NUREG-0654: N.07 PROCEDURE 5.7.1 l REVISION 30 l PAGE 35 OF 87 l

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 4.1.2 NOUE TEXT Unplanned loss of most or all safety system annunciators.

APPLICABILITY Reactor critical.

EXAMPLE Complete failure of all annunciators while at power.

MEMO If a transient is also in progress, see EAL: 4.2.3.

REFERENCES NUREG-0654: A.14 I PROCEDURE 5.7.1 l REVISION 30 _ PAGE 36 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 4.2.1 ALERT TEXT Loss of all AC power (on and off-site sources) to vital Busses "F" and "G" during MODE 4 or 5.

APPLICABILITY MODE 4 or 5.

EXAMPLE Loss of all off-site AC power while in MODE 4 or 5. DGs fail to start.

MEMO Being in MODE 4 or 5, reduces the risk for core damage or other fission product barrier challenge caused by the loss of power.

See EAL: 4.3.1 for loss of power when the reactor is hot.

REFERENCES NUREG-0654: A.07 I PROCEDURE 5.7.1 l REVISION 30 l PAGE 37 OF87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 4.2.2 ALERT TEXT Loss of all DC power sources resulting in loss of all ECCS capability for < 15 minutes.

APPLICABILITY ALL EXAMPLE Any loss of DC power that results in a complete loss of ECCS capability for

< 15 minutes.

MEMO If the loss of ALL ECCS capability is the result of a loss of DC power (either 125 VDC or 250 VDC; or a combination of the two) the EAL is met.

If the complete loss of ECCS capability as a result of the loss of DC power lasts

Ž15 minutes, refer to EAL 4.3.2 (SAE).

REFERENCES NUREG-0654: A.08 PROCEDURE 5.7.1 l REVISION 30 _ PAGE 38 OF 87 l

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 4.2.3 ALERT TEXT Unplanned loss of most or all safety system annunciators with a transient in progress.

APPLICABILITY Reactor critical.

EXAMPLE Complete failure of all safety system annunciators while at power and a transient is in progress.

MEMO Similar to EAL: 4.1.2 except this EAL includes a transient in progress.

The USAR definition of "transient" is an abnormal operational transient includes the events following a single equipment malfunction or a single operator error that is reasonable expected during the course of planned operations. Power failures, pump trips, and rod withdrawal errors are typical of the single malfunctions or errors initiating the events in this category.

Loss of all annunciators in the Control Room would also likely be classifiable under an EAL for loss of DC.

REFERENCES NUMARC/NESP-007: SA4 I PROCEDURE 5.7.1 l REVISION 30 l PAGE 39 OF 87 l

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 4.3.1 SITE AREA EMERGENCY TEXT Loss of all AC power (on and off-site sources) for more than 15 minutes with the Reactor in MODE 1, 2, or 3.

APPLICABILITY MODE 1, 2, or 3.

EXAMPLE Tornado drops all lines feeding the plant while at power. Both diesel generators fail to start and cannot be started within 15 minutes (i.e., Station Blackout > 15 minutes).

MEMO Either RCIC or HPCI, are capable of injecting water to the vessel independent of AC power. Loss of all other means to inject water to the vessel for an extended period of time meets the class description for SITE AREA EMERGENCY listed in NUREG-0654.

REFERENCES NUREG-0654: S.06 PROCEDURE 5.7.1 REVISION 30 l PAGE 40 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 4.3.2 SITE AREA EMERGENCY TEXT Loss of all DC power sources required for ECCS operation for more than 15 minutes.

APPLICABILITY ALL EXAMPLE Any loss of DC power that results in a complete loss of ECCS capability for

Ž15 minutes.

MEMO If the loss of ALL ECCS capability is the result of a loss of DC power (either 125 VDC or 250 VDC; or a combination of the two) for Ž15 minutes, the EAL is met.

REFERENCES NUREG-0654: S.07 PROCEDURE 5.7.1 l REVISION 30 l PAGE 41 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 4.3.3 SITE AREA EMERGENCY TEXT Inability to monitor a significant transient in progress.

APPLICABILITY ALL EXAMPLE Complete failure of all annunciators while at power, a significant transient in progress, and inability to monitor key parameters via other instrumentation.

MEMO Similar to EAL: 4.2.3 except this EAL includes the inability to monitor the transient using redundant instrumentation.

A significant transient includes responses to automatic or manually initiated functions, such as; scrams, runbacks involving > 25% thermal power changes, ECCS injections, or thermal power oscillations of 10% or greater.

REFERENCES NUMARC/NESP-007: SS6 I PROCEDURE 5.7.1 l REVISION 30 l PAGE 42 OF 87 l

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 4.4.1 GENERAL EMERGENCY TEXT Total loss of all AC power (on and off-site sources) with the inability to keep the core covered.

APPLICABILITY ALL EXAMPLE HPCI and RCIC fail during a station blackout. Level drops below 0" (FZ).

MEMO Failure to keep the core covered combined with a loss of all AC indicates failure of steam driven pumps. Without cooling the core will degrade, Primary Containment could heat up and potentially fail.

I REFERENCES NUREG-0654: G.06A I PROCEDURE 5.7.1 l REVISION 30 l PAGE 43 OF 87 l

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 5.1.1 NOUE TEXT Any fire within the Protected Area which takes longer than 10 minutes to extinguish.

APPLICABILITY ALL EXAMPLE Fire brigade is unable to extinguish a fire in the turbine lube oil reservoir room within 10 minutes from receipt of report or alarm in the Control Room.

MEMO Time is measured from the time the report or alarm of a fire is received in the Control Room.

REFERENCES NUREG-0654: N.10 Meacham to ERO, "Clarification of Certain Emergency Action Levels (EALs)",

CNSS900421 August 7, 1990.

Telecon Krumland/Hayden to Spitzberg (NRC IV), "EAL Interim Guidance - Memo",

August 22, 1990.

I PROCEDURE 5.7.1 l REVISION 30 l PAGE 44 OF 87 l

ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 5.1.2 NOUE TEXT Report or detection of toxic or flammable gases that could enter the Protected Area in amounts that will affect the health of plant personnel or can effect normal operation of the plant.

APPLICABILITY ALL EXAMPLE Bulk hydrogen delivery truck regulator fitting is broken during unloading and cannot be isolated.

MEMO Certain spills or releases may require notification of EPA or other agencies.

REFERENCES NUREG-0654: N.14D PROCEDURE 5.7.1 l REVISION 30 l PAGE 45 OF 87 l

ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 5.2.1 ALERT TEXT A fire with a potential to cause degradation of a plant safety system required to be OPERABLE.

APPLICABILITY ALL EXAMPLE A fire in NE Reactor Building 903' during Power operations with the potential to damage cables.

MEMO This EAL is intended to apply to a fire which could directly affect any (one or more) plant safety system(s). Implicit in this interpretation is that plant conditions are such that the potentially affected safety system should be OPERABLE. For example, during MODE 4 or 5, HPCI is not required to be OPERABLE. Therefore, a fire in the HPCI Room would not necessarily threaten a required safety system. A large fire in the same area, however, that constituted a threat to the "B" and "D" RHR Pumps would meet the threshold for this EAL.

The threshold of the EAL would also be met if, while at power, a fire occurred in the HPCI Room which threatened the OPERABILITY of the system. This is true even if HPCI was inoperable at the time (under the required Technical Specification LCO),

since HPCI should be OPERABLE while at power.

On the other hand, a small fire (e.g., a smoldering rag or burning piece of paper),

which does not constitute a threat to a safety system, does not meet the intent of this EAL.

REFERENCES NUREG-0654: A.13 Meacham to ERO, "Clarification of Certain Emergency Action Levels (EALs)",

CNSS900421, August 7, 1990.

I PROCEDURE 5.7.1 - REVISION 30 l PAGE 46 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 5.2.2 ALERT TEXT Report or detection of toxic or flammable gases within a Vital Area in concentrations that will be life threatening to plant personnel or will affect the safe operation of the plant.

APPLICABILITY ALL EXAMPLE CO 2 pre-discharge alarm on DG Room #1 received. Personnel evacuate room out different doors. Upon exit, all personnel cannot be accounted for.

MEMO To meet the class description for an ALERT, the condition must indicate an actual or potential substantial degradation of the level of safety of the plant (NUREG-0654, Appendix 1) or be life threatening to personnel.

If personnel are not in the affected area nor required to enter, or must remain in the affected area but have adequate protection (to safely operate or shutdown the plant),

this EAL is not met.

REFERENCES NUREG-0654: A.18D PROCEDURE 5.7.1 REVISION 30 PAGE 47 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 5.3.1 SITE AREA EMERGENCY TEXT Fire compromising the functions of safety systems.

APPLICABILITY ALL EXAMPLE A fire in the Cable Spreading Room affecting the function of HPCI while required to be OPERABLE.

MEMO This EAL applies to a fire which compromises the active function (e.g., low pressure injection or automatic depressurization) of a safety system or multiple safety systems.

In reviewing EAL: 5.2.1 and 5.3.1, it is important to note that EAL: 5.2.1 covers the potential for degradation of nuclear safety, while EAL: 5.3.1 is recognition that an actual degradation has occurred. Additionally, the statements made regarding system OPERABILITY for EAL: 5.2.1 also apply to EAL: 5.3.1.

This EAL is intended to apply to a fire which could directly affect any (one or more) plant safety system(s). Implicit in this interpretation is that plant conditions are such that the potentially affected safety system should be OPERABLE. For example, during MODE 4 or 5, HPCI is not required to be OPERABLE. Therefore, a fire in the HPCI Room would not necessarily threaten a required safety system. A large fire in the same area, however, that constituted a threat to the "B" and "D" RHR pumps would meet the threshold for this EAL.

REFERENCES NUREG-0654: S.11 Meacham to ERO, "Clarification of Certain Emergency Action Levels (EALs)",

CNSS900421, August 7, 1990.

I PROCEDURE 5.7.1 l REVISION 30 l PAGE 48 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 5.4.1 GENERAL EMERGENCY TEXT Any major internal or external fire substantially beyond the design basis which could cause massive common damage to plant systems.

APPLICABILITY ALL EXAMPLE A fire in Critical Switchgear Rooms, where both rooms are involved, result in loss of CS, RHR, SW, etc.

MEMO None.

REFERENCES NUREG-0654: G.07 PROCEDURE 5.7.1 REVISION 30 l PAGE 49 OF 87 l

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 6.1.1 NOUE TEXT Security threat, attempted entry, or attempted sabotage.

APPLICABILITY ALL EXAMPLE A credible bomb threat.

MEMO I A confirmed "Red, Site Specific and credible" threat warning from the NRC should be considered a Security threat.

As determined by the Security Contingency Plan or procedures.

REFERENCES NUREG-0654: N.12 I PROCEDURE 5.7.1 l REVISION 30 l PAGE 50 OF 87 l

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 6.2.1 ALERT TEXT On-going security compromise.

APPLICABILITY ALL EXAMPLE Armed intruders within the Protected Area.

MEMO As determined by the Security Contingency Plan or procedures.

REFERENCES NUREG-0654: A.16 I PROCEDURE 5.7.1 l REVISION 30 l PAGE 51 OF 87l

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 6.3.1 SITE AREA EMERGENCY TEXT Imminent loss of physical control of the station.

APPLICABILITY ALL EXAMPLE Large number of armed intruders in the station.

MEMO None.

REFERENCES NUREG-0654: S.14 I PROCEDURE 5.7.1 l REVISION 30 l PAGE 52 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 6.4.1 GENERAL EMERGENCY TEXT Loss of physical control of the station.

APPLICABILITY ALL EXAMPLE I Armed intruder(s) in the Control Room or Alternate Shutdown Panel.

MEMO I Loss of either the Control Room or Alternate Shutdown Panel would be considered a I loss of physical control of the station.

REFERENCES NUREG-0654: G.03 PROCEDURE 5.7.1 REVISION 30 l PAGE 53 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 7.1.1 NOUE TEXT Ground motion > 0.01g as indicated by Control Room seismic monitoring panel.

APPLICABILITY ALL EXAMPLE I Noticeable seismic shock felt in Control Room.

MEMO I If the Seismic Monitoring Panel (SMA-3) is not available and a noticeable seismic I shock is felt, a seismic event > 0.1g is assumed to have occurred. Refer to EAL 7.2.1 or 1 7.3.1 if major plant damage has occurred.

I Personnel on upper floors of buildings noticeably feel a 0.01g earthquake and see I suspended objects swing. At 0.1g, some heavy furniture may move and personnel on I upper floors may have difficulty standing.

Attempt to rule out "false" causes for alarm (i.e., heavy equipment operation).

REFERENCES NUREG-0654: N.13A PROCEDURE 5.7.1 REVISION 30 PAGE 54 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 7.1.2 NOUE TEXT River level > 899' or < 867'.

APPLICABILITY ALL EXAMPLE Flood, river level 900' MSL.

MEMO Flood of record per USAR is 900.8'.

REFERENCES NUREG-0654: N.13B PROCEDURE 5.7.1 T REVISION 30 l PAGE 55 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 7.1.3 NOUE TEXT Tornado touching down within the Owner Controlled Area.

APPLICABILITY ALL EXAMPLE Tornado striking north Training Building.

MEMO Consider performing assembly and accountability after danger has passed. If tornado touches down within the Protected Area, see EAL: 7.2.3.

REFERENCES NUREG-0654: N.13C PROCEDURE 5.7.1 l REVISION 30 l PAGE 56 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 7.1.4 NOUE TEXT Sustained wind speed > 74 mph.

APPLICABILITY ALL EXAMPLE Severe sustained winds from a thunderstorm. MET indicates sustained winds of 80 mph.

MEMO CNS' version of "hurricane" listed in NUREG-0654 initiating condition.

These are sustained winds, not gusts.

REFERENCES NUREG-0654: N.13D PROCEDURE 5.7.1 l REVISION 30 l PAGE 57 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 7.2.1 ALERT TEXT Ground motion > 0. ig as indicated by Control Room seismic monitoring panel.

APPLICABILITY ALL EXAMPLE Earthquake.

MEMO If the Seismic Monitoring Panel (SMA-3) is not available and a noticeable seismic I shock is felt, a seismic event > 0.1g is assumed to have occurred. Refer to EAL 7.3.1 if major plant damage has occurred.

I Personnel on upper floors of buildings noticeable feel a 0.01g earthquake and see I suspended objects swing. At 0.1g, some heavy furniture may move and personnel on I upper floors may have difficulty standing.

This EAL is the Operating Basis Earthquake (OBE) for CNS per the USAR.

I REFERENCES NUREG-0654: A.17A I PROCEDURE 5.7.1 l REVISION 30 l PAGE 58 OF 87 l

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 7.2.2 ALERT TEXT River level > 902' or < 865'.

APPLICABILITY ALL EXAMPLE Ice jam upstream causes river level to drop below 865'.

MEMO These levels equate to "near design levels" specified in NUREG-0654 initiating condition. This could result in "potential substantial degradation" to safety systems as found in the ALERT class description of NUREG-0654.

REFERENCES NUREG-0654: A.17B I PROCEDURE 5.7.1 r REVISION 30 l PAGE 59 OF87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 7.2.3 ALERT TEXT Tornado touching down within the Protected Area.

APPLICABILITY ALL EXAMPLE Tornado striking Security, Craft Change, and the NRC/Ambulance Buildings.

MEMO Ensure tornado has passed before conducting assembly and accountability.

REFERENCES NUREG-0654: A.17C I PROCEDURE 5.7.1 l REVISION 30 l PAGE 60 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 7.2.4 ALERT TEXT Sustained wind speed > 95 mph.

APPLICABILITY ALL EXAMPLE MET indicates sustained winds of 96 mph.

MEMO Equates to "hurricane winds beyond design basis level" specified in NUREG-0654 initiating condition.

These are sustained winds, not gusts.

REFERENCES NUREG-0654: A.17D I PROCEDURE 5.7.1 l REVISION 30 l PAGE 61 OF87

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 7.3.1 SITE AREA EMERGENCY TEXT Ground motion > 0.1g as indicated on the Control Room seismic monitoring panel AND reports of major plant damage.

APPLICABILITY MODE 1, 2, or 3.

EXAMPLE Visible crack on Drywell following an earthquake.

MEMO I If the Seismic Monitoring Panel (SMA-3) is not available and a noticeable seismic I shock is felt, a seismic event > 0.1g is assumed to have occurred.

Personnel on upper floors of building noticeable feel a 0.01g earthquake and see suspended objects swing. At 0.1g, some heavy furniture may move and personnel on upper floors may have difficulty standing.

This EAL represents the Safe Shutdown Earthquake (SSE) from the USAR. The SSE for CNS is 0.2g. CNS has no active instrumentation beyond 0.1g. Whether equipment damage is considered "major plant damage" is based on the judgement of SS/ED.

Equipment damage that places the plant in condition not addressed by Technical Specifications (e.g., T.S. LCO 3.0.3) should be considered major plant damage.

REFERENCES NUREG-0654: S.15A PROCEDURE 5.7.1 l REVISION 30 l PAGE 62 OF 87 l

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 7.3.2 SITE AREA EMERGENCY TEXT Sustained wind speed > 100 mph.

APPLICABILITY MODE 1, 2, or 3.

EXAMPLE Sustained MET indicates wind speed of 100 mph.

MEMO This is a sustained wind speed, not gusts.

CNS instrumentation only goes to 100 mph, not beyond.

REFERENCES NUREG-0654: S.15C PROCEDURE 5.7.1 l REVISION 30 l PAGE 63 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 7.3.3 SITE AREA EMERGENCY TEXT Flood which renders multiple ECCS Systems inoperable when they are required to be OPERABLE.

APPLICABILITY MODE 1, 2, or 3.

EXAMPLE HPCI quad flooded (affecting HPCI and RHR function).

MEMO The SITE AREA EMERGENCY class description refers to plant functions needed to protect the public. If systems were impacted, but not needed, CNS would maintain the ALERT.

REFERENCES NUREG-0654: S.15B PROCEDURE 5.7.1 l REVISION 30 l PAGE 64 OF 87 l

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 7.3.4 SITE AREA EMERGENCY TEXT Low river level which results in complete loss of the Service Water System.

APPLICABILITY All EXAMPLE SWPs cavitate due to low river level.

MEMO Service water is always needed as the ultimate heat sink for the plant. Its loss meets the class description for SITE AREA EMERGENCY found in NUREG-0654.

Follow the procedures for maximizing water level in E Bay. This EAL is complete loss.

Service Water operation which does not meet Tech Specs, but provides some cooling should be classified as an ALERT on EAL: 7.2.2.

REFERENCES NUREG-0654: S.15B I PROCEDURE 5.7.1 REVISION 30 l PAGE 65 OF 87 l

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 7.4.1 GENERAL EMERGENCY TEXT Any major natural phenomenon substantially beyond the design basis which could cause massive common damage to plant systems.

APPLICABILITY ALL EXAMPLE Earthquake which causes immediate, massive, and obvious damage to many plant systems.

MEMO None.

REFERENCES NUREG-0654: G.07 PROCEDURE 5.7.1 l REVISION 30 PAGE 66 OF 87 l

ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 8.1.1 NOUE TEXT Aircraft crash within the Protected Area.

APPLICABILITY ALL EXAMPLE Small aircraft crashes within the Protected Area, but does not strike any structures.

MEMO An airplane crash must be within the Protected Area to meet the NOUE classification description of NUREG-0654.

REFERENCES NUREG-0654: N.14A PROCEDURE 5.7.1 l REVISION 30 l PAGE 67 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 8.1.2 NOUE TEXT Explosion within the Protected Area.

APPLICABILITY ALL EXAMPLE Gasoline storage tank explodes.

MEMO An explosion includes all sudden, violent, and rapid releases of energy. "Detonation" and "Deflagration" are releases of chemical energy which qualify as "Explosions". Also included is the rapid release of mechanical energy, i.e., pressure.

The source or location of the explosion must be within the Protected Area to meet the NOUE class description of NUREG-0654. An explosion on the Owner Controlled Area (OCA) does not meet the NOUE class description of NUREG-0654.

The rapid release of mechanical energy may result in the generation of a missile (see EAL: 8.2.2).

REFERENCES NUREG-0654: N.14C PROCEDURE 5.7.1 l REVISION 30 l PAGE 68 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 8.1.3 NOUE TEXT Failure of a turbine rotating component causing an automatic reactor scram with release of radioactivity to the Turbine Building or which potentially affects safety systems.

APPLICABILITY ALL EXAMPLE Low pressure rotor fails. Radioactivity is released to the Turbine Building prior to MSIV closure.

MEMO A reactor scram (from whatever cause) does not meet the NOUE class description unless there is an associated release of radioactivity or safety systems are potentially affected.

If the radiological release is considered to be serious or safety systems are actually degraded, see EAL: 8.2.4.

REFERENCES NUREG-0654: N.14E I PROCEDURE 5.7.1 l REVISION 30 l PAGE 69 OF 87

ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 8.1.4 NOUE TEXT Other conditions existing which in the judgement of the Emergency Director warrant declaration of an Unusual Event.

APPLICABILITY ALL EXAMPLE Event in progress or which has occurred, that indicate a potential degradation of the level of safety of the station. The event may progress to a more severe emergency classification if it is not mitigated.

MEMO For events of minor safety significance, but which warrant notification of authorities.

Attempt to classify under more specific EALs. If none apply, declare under this one.

REFERENCES NUREG-0654: N.15 PROCEDURE 5.7.1 REVISION 30 1 PAGE 70 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 8.2.1 ALERT TEXT Aircraft striking structures within the Protected Area.

APPLICABILITY ALL EXAMPLE Aircraft striking the Elevated Release Point (ERP).

MEMO None.

REFERENCES NUREG-0654: A.18A I PROCEDURE 5.7.1 l REVISION 30 l PAGE 71 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS I CLASSIFICATION EAL: 8.2.2 ALERT TEXT Missile impact, from whatever source, within the Protected Area.

APPLICABILITY ALL EXAMPLE Helicopter drops unknown objects onto the Turbine Building roof.

MEMO "Missile" is not defined by NUREG-0654. It is assumed that any large projectile is a missile.

REFERENCES NUREG-0654: A.18B I PROCEDURE 5.7.1 l REVISION 30 l PAGE 72 OF 87 l

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 8.2.3 ALERT TEXT Known explosion damage to the facility affecting plant operation.

APPLICABILITY ALL EXAMPLE Hydrogen explosion in hydrogen seal oil pump (Iron Horse) room causing turbine trip.

MEMO An explosion includes all sudden, violent, and rapid releases of energy. "Detonation" and "Deflagration" are releases of chemical energy which qualify as "Explosions". Also included is the rapid release of mechanical energy, i.e., pressure.

The rapid release of mechanical energy may result in the generation of a missile (see EAL: 8.2.2).

An explosion affecting operation could also have caused damage not yet discovered which could be of safety significance.

REFERENCES NUREG-0654: A.18C PROCEDURE 5.7.1 1 REVISION 30 PAGE 73 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 8.2.4 ALERT TEXT Turbine failure causing casing penetration which creates serious radiological concerns or damages plant safety systems.

APPLICABILITY ALL EXAMPLE Portion of the turbine rotor penetrates casing. Other failures result in serious radiological concerns.

MEMO Extension of EAL: 8.1.4. Turbine casing penetration alone does not meet the ALERT class description of NUREG-0654.

Serious radiological concerns would also likely be classifiable under other EALs.

REFERENCES NUREG-0654: A.18E PROCEDURE 5.7.1 l REVISION 30 l PAGE 74 OF 87 l

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 8.2.5 ALERT TEXT Other conditions existing which in the judgement of the Emergency Director warrant declaration of an ALERT.

APPLICABILITY ALL EXAMPLE An event in progress, or which has occurred, that involved an actual or potentially substantial degradation of the safety level of the station. Minor releases of radioactivity may occur or may have occurred.

MEMO Attempt to classify under other more specific EALs. If none apply, declare on this one.

REFERENCES NUREG-0654: A.19 PROCEDURE 5.7.1 REVISION 30 PAGE 75 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 8.3.1 SITE AREA EMERGENCY TEXT Aircraft crash affecting vital areas with the plant in MODE 1, 2, or 3.

APPLICABILITY MODE 1, 2, or 3.

EXAMPLE Airplane crash into 1001' (Reactor Building 5th floor) while at power.

MEMO None.

REFERENCES NUREG-0654: S.16A PROCEDURE 5.7.1 l REVISION 30 l PAGE 76 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 8.3.2 SITE AREA EMERGENCY TEXT Missile or explosion damage to safe shutdown equipment with the plant in MODE 1, 2, or 3.

APPLICABILITY MODE 1, 2, or 3.

EXAMPLE A high pressure nitrogen cylinder is dropped and its valve assembly is sheared off, it becomes a "missile" damaging several HCUs.

MEMO An explosion includes all sudden, violent, and rapid releases of energy. "Detonation" and "Deflagration" are releases of chemical energy which qualify as "Explosions". Also included is the rapid release of mechanical energy, i.e., pressure.

The rapid release of mechanical energy may result in the generation of a missile (see EAL: 8.2.2).

REFERENCES NUREG-0654: S.16B I PROCEDURE 5.7.1 l REVISION 30 l PAGE 77 OF 87l

ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 8.3.3 SITE AREA EMERGENCY TEXT Other conditions existing which in the judgement of the Emergency Director warrant declaration of a SITE AREA EMERGENCY.

APPLICABILITY ALL EXAMPLE Events in progress or have occurred, which involve actual or potential major failure of plant functions needed for the protection of the public.

MEMO Attempt to classify under other more specific EALs. If none apply and there is actual or likely major failures of plant equipment needed for the protection of the public, declare on this one.

REFERENCES NUREG-0654: S.17 PROCEDURE 5.7.1 l REVISION 30 PAGE 78 OF 87

I ATTACHMENT 2 EMERGENCY ACTION LEVELS CLASSIFICATION EAL: 8.4.1 GENERAL EMERGENCY TEXT Other conditions existing which in the judgement of the Emergency Director warrant declaration of a General Emergency (i.e., any core melt situation).

APPLICABILITY ALL EXAMPLE Event in progress or which has occurred, that involves actual or imminent substantial core degradation or melting with a potential for the loss of Primary Containment integrity.

MEMO Attempt to classify on other more specific EALs. If none apply and there is the possibility of release of large quantities of radioactive material in a short period of time, declare under this one.

REFERENCES NUREG-0654: G.07 PROCEDURE 5.7.1 l REVISION 30 l PAGE 79 OF 87l

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS BARRIER POTENTIAL LOSS (1) LOSS (2)

1. 1500 mrem/hr on SJAE monitor 1. 15,000 mrem/hr on SJAE monitor.

[EAL: 2.1.1].

2. Reactor Coolant sample

2. Coolant sample activity > 300 [tCi/gm DOSE

> 4.0 Ci/gm DOSE EQUIVALENT EQUIVALENT I-131.

1-131 [EAL: 2.1.2].

3. LOCA with DW radiation monitor

3. LOCA with DW radiation monitor reading > 2500 REMlhr.

Fuel Cladding reading > 250 REM/hr.

4 Non-LOCA with DW radiation monitor reading > 115 REM/hr.

5. Reactor water level below 0" (FZ) or cannot be determined.

6. Main steam line radiation monitor 2 Hi-Hi alarm setpoint.

1. Operational RCS pressure boundary 1. Reactor water cannot be restored LEAKAGE; or unidentified and maintained above 0" (FZ) or LEAKAGE exceeds 5 gpm; or total cannot be determined.

LEAKAGE exceeds 30 gpm averaged over a previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2. Drywell pressure > 2 psig due to Primary Coolant period; or unidentified LEAKAGE RCS leakage.

Boundary increase of more than 2 gpm within the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period in 3. Primary coolant leak > 50 gpm.

MODE 1.

4. Safety or Relief valve stuck open AND Suppression Pool Temperature 110°F.

1. Primary Containment pressure > 25 1. Inability to isolate primary psig and increasing. containment.

Primary 2. Loss of all cooling capabilities. 2. Loss of Primary Containment Containment structural integrity.

OPERABILITY 3. Hydrogen concentration > 4%.

4. Unexplained drop inDrywell3. Drywell pressure 2 56 psig.

4. Unexplaied dropin Drywell pressure or rise in nitrogen makeup. 4. Hydrogen concentration > 15%.

(1) Applies to classification only when combined with two actual losses, or if a separate EAL is indicated by a bracketed [ ] EAL #.

(2) Single fission product barrier loss (Fuel Cladding or Primary Coolant Boundary) is an ALERT, loss of two barriers (any two) is a SITE AREA EMERGENCY, loss of two barriers with potential for loss of the third barrier is a GENERAL EMERGENCY.

PROCEDURE 5.7.1 l REVISION 30 l PAGE 80 OF 8

I ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS NOTE 1 - An emergency class may be declared on a potential loss or on an actual loss, but equating multiple potential losses to an actual loss is not acceptable. That is, two potential losses do not equal one actual loss. Only when a potential loss is combined with the actual loss of two barriers does the potential loss of the barrier change an emergency classification (i.e., from a SITE AREA EMERGENCY to a GENERAL EMERGENCY).

NOTE 2 - Paragraph numbers below correspond to those in the table on the previous page.

FUEL CLADDING - POTENTIAL LOSS

1. Based on 0.1% cladding failure (NEDC 02-004).

2. Based on Technical Specification 3.4.6. See Technical Specification bases.

3. Derived from Attachment 7 of Procedure 5.7.17. This attachment in turn comes from NEDO 22215. This value (250 REM/hr) approximates 0.1% fuel cladding failure with a LOCA environment in the DW.

FUEL CLADDING - LOSS

1. Based on 1% cladding failure (NEDC 02-004).

2. From NUREG-0654, Initiating Condition Appendix 1, ALERT, Step l.b, requires reactor water coolant analysis.

3. Derived from Attachment 7 of Procedure 5.7.17. This attachment in turn comes from NEDO-22215 and is valid for LOCA conditions. This number (2500 rem/hr) approximates 1% fuel cladding failure.

4. Based on 1% clad failure during Non-LOCA conditions in the DW. Refer to NEDC 02-009.

5. Cladding integrity cannot be guaranteed if fuel is not covered with water. Note this EAL says below 0" (FZ). If level is intentionally lowered to 0" (FZ) (but not below) per EOPs, this EAL does not apply. If level falls below 0" (FZ) accidentally, even for a short time, this EAL does apply and the barrier shall be declared lost. If RPV level cannot be determined (unknown), the barrier shall be considered lost.

6 Based on analysis for Design Bases Control Rod Drop Accident (DBCRDA). Fuel cladding failure resulting from DBCRDA will result in MSL Radiation Monitor Hi-Hi alarm setpoint being reached. Refer to Tech Spec Bases 3.3.6.1/2.d for MSL Radiation Monitor Hi-Hi alarm setpoint bases.

I PROCEDURE 5.7.1 1 REVISION 30 l PAGE 81 OF 87

I ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS l PRIMARY COOLANT BOUNDARY - POTENTIAL LOSS

1. Technical Specification leak rate limit. Refer to Technical Specification 3.4.4.

PRIMARY COOLANT BOUNDARY - LOSS

1. If RPV water level cannot be restored and maintained above 0" (FZ), then the primary coolant boundary shall be assumed to be lost. If RPV water level cannot be determined (unknown), then the barrier shall be considered lost.

2. Drywell pressure > 2 psig with corollary indications (DW temperature, humidity) should be considered a loss of Primary Coolant Boundary. Loss of Drywell cooling that results in > 2 psig should not be a loss of Primary Coolant Boundary (NOTE: Using ideal gas law, DW temperature would be 202° at 2 psig and 2420 at 3 psig if due to loss of cooling).

3. From NUREG-0654, Initiating Condition, Appendix 1, ALERT, 5.

4. Technical Specifications require a SCRAM when Suppression Pool Average Temperature reaches 110°F. Below this point, the reactor is considered in a safe condition even with relief valves stuck open.

PRIMARY CONTAINMENT - POTENTIAL LOSS

1. Represents a degrading trend representative of loss of control of some parameter affecting containment pressure. At this value (approximately half that of the loss value) the potential exists for loss.

2. Primary containment's design temperature is 281°F. Loss of all cooling capabilities may result in approaching this design limit.

3. Derived from NUREG/BR-0150, RTM-93 Table on page B-19. This is the beginning of the flammability region for a dry atmosphere.

4. Indicates a possible leak from primary containment.

PRIMARY CONTAINMENT - LOSS

1. From NUREG-0654, Initiating Condition Appendix 1, ALERT, 4.

I PROCEDURE 5.7.1 l REVISION 30 l PAGE 82 OF 87 l

I ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS

2. Number 1 Loss indicator, above, refers to Primary Containment Isolation System (i.e.,

valves and associated logic). This indicator is intended to expand upon PCIS to include any indication that the containment's integrity is not intact. Also, valves other than PCIS may be used to isolate containment and restore the barrier.

3. 56 psig is the design pressure for containment. At or above this pressure, the containment is to be considered lost.

4. Derived from NUREG/BR-0150, RTM-93 Table on page B-19. This is the beginning of the detonation region for a dry atmosphere.

ISOLATION VALVE FAILURES The following apply to determining fission product barrier loss in response to Primary I Containment Isolation Valve (PCIV) failures:

I NOTE - Both valves in a line must fail to be considered a loss of the barrier(s).

I 1. The barrier(s) should be considered lost if ANY of the following exist:

I

Attempted manual isolation from the Control Room failed.

Line remains un-isolated following a Group Isolation AND subsequent attempt to I isolate from the Control Room is unsuccessful.

I 2. Valves other than PCIS may be used to isolate containment and restore the barrier.

3. If an Operator must leave the Control Room to close a valve, the barrier(s) shall be considered lost until a valve can be closed manually.

4. If the line penetrates PC and also communicates with the RPV, then two barriers are to be considered lost (EAL: 2.3.3 - SITE AREA EMERGENCY).

5. If either of the valves in a line are subsequently closed manually, then the barrier is to be considered restored and the emergency may be reclassified, as appropriate.

A special case exists concerning SDV vent and drain valves when a scram occurs. When a scram occurs, these valves are supposed to close. While the scram inlet and outlet valves remain open (before the scram is reset) the water/steam isolated by these valves communicates directly to the reactor. The design fission product barriers (RPV and PC) have effectively "moved" from the scram valves to the vent and drain valves. If these valves fail, they therefore meet the criteria for loss of two of three fission product barriers (EAL: 2.3.3 - SITE AREA EMERGENCY).

I PROCEDURE 5.7.1 l REVISION 30 l PAGE 83 OF 87

I ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS l A special case also exists concerning operation of HPCI and RCIC to support Emergency Operating Procedures (5.8 series). If HPCI or RCIC were to isolate on high temperature during operation to support the EOPs, the EOPs allow you to install jumpers to bypass the isolation and restart the system. This is allowed even if a leak from the steam supply is causing the high temperature condition. If a leak does in fact exist and the isolation valves are opened, this would constitute a loss of two fission product barriers (EAL: 2.3.3 -

SITE AREA EMERGENCY). These barriers would be Reactor Coolant System and Primary Containment. The justification for the loss of the barriers is that you are releasing steam from the Reactor Coolant System to the atmosphere of the secondary containment. If the valves were reclosed, the fission product barriers would once again be considered intact.

Another issue was raised concerning the loss of a barrier due to local leak rate testing results. Local leak rate test results are not applicable to these EALs and valve position (i.e., can the valve be closed) will be the sole basis for declaring a barrier lost.

I PROCEDURE 5.7.1 l REVISION 30 1 PAGE 84 OF 87l

I ATTACHMENT 4 EAL HARDCARDS I Information contained in Attachment 1, EAL Matrix, and Attachment 3, Fission Product Barriers-Indication of Loss Table, may be reformatted and placed on HARDCARDS similar to EOP Flowcharts. These EAL HARDCARDS will be controlled per this attachment. This information will be word for word but may be formatted differently using different font sizes or color backgrounds to assist the visual presentation.

Each EAL HARDCARD will be labeled with a EAL HARDCARD Revision data box that will list the latest revision and the date of the revision of the HARDCARD. This data will match the information below:

EAL HARDCARD Revision Data Procedure HARDCARD Revision # Date of Last HARDCARD Revision i EPIP 5.7.1, Revision 3 4/14/2003 Attachment 4 It is not necessary that the HARDCARD revision number be revised with each revision of this procedure. However, if the HARDCARD is revised, or, if Attachment 1 or 3 are revised, then Attachment 4 must be revised to reflect the new EAL HARDCARD Revision Data with the new information.

EAL HARDCARD distribution will be made to following locations:

EAL HARDCARD Locations:

1. Control Room

2. Simulator

3. Emergency Operations Facility

4. Technical Support Center

5. Alternate Emergency Operations Facility

6. Emergency Preparedness Office I PROCEDURE 5.7.1 l REVISION 30 l PAGE 85 OF 87 l

I ATTACHMENT 5 INFORMATION SHEET REFERENCES 1.1 TECHNICAL SPECIFICATION 1.1.1 Bases 3.3.6.1/2.d, Main Steam Line Radiation - High.

1.1.2 Section 3.6, Containment Systems.

1.2 CODES AND STANDARDS 1.2.1 10CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors.

1.2.2 NPPD Emergency Plan For CNS.

1.2.3 NUREG-0654, Revision 1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants.

1.2.4 NUREG/BR-0150, Volume 1, Revision 3, November 1993, Response Technical Manual.

1.2.5 Environmental Protection Agency EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, May 1992.

1.3 PROCEDURES 1.3.1 Instrumentation Operating Procedure 4.12, Seismic Instrumentation.

1.3.2 Emergency Plan Implementing Procedure 5.7.2, Shift Supervisor EPIP.

1.3.3 Emergency Plan Implementing Procedure 5.7.16, Release Rate Determination.

1.3.4 Emergency Plan Implementing Procedure 5.7.17, Dose Assessment.

1.4 MISCELLANEOUS 1.4.1 NRC Inspection Reports: 87-25, 88-29, 91-27, 92-14, and 93-24.

1.4.2 Letter CNSS900421 from Meacham to ERO, dated August 7, 1990, Clarification of Certain Emergency Action Levels (EALs).

PROCEDURE 5.7.1 l REVISION 30 l PAGE 86 OF 87 l

ATTACHMENT 5 INFORMATION SHEET 1.4.3 Telecon Krumland/Hayden to Spitzberg (NRC IV), dated August 22, 1990, EAL Interim Guidance.

1.4.4 Telecon Hayden/Dean to Terc (NRC IV), dated April 22, 1992, Spent Fuel EAL 3.3.1.

1.4.5 Letter NSD940202 from G. R. Smith to G. R. Horn, Commitments from 1/31/94 Enforcement Conference.

1.4.6 Memorandum from Richard L. Emch, Jr., Acting Chief of Emergency Preparedness Branch, Division of Radiation Safety and Safeguards, Office of Nuclear Reactor Regulation, to James H. Joyner (Region 1),

William E. Cline (Region 2), John A. Grobe (Region 3), and Blaine Murray (Region 4), dated July 11, 1994.

Subject:

Branch Position on Acceptable Deviations to Appendix 1 to NUREG-0654/FEMA-REP-1.

1.4.7 NEDC 00-099, Core dp vs. Flow Curve for Determination of Degraded Core.

1.4.8 RCR 2001-0871, Action #2. Revised memo field of EAL 4.1.1 to discuss need to classify if power to both S/U and EMER XFMR is lost

> 15 minutes.

1.4.9 DD 10154409, Clarify term "degraded core" for EAL 2.3.1.

1.4.10 RCR 2002-0660, Action #1, Clarify that the loss of DC power that results in a loss of ECCS injection capability is the intent of EALs 4.2.2 and 4.3.2.

1.4.11 RCR 2001-1272, Action #1.

1.4.12 RCR 2002-0448, Action #1, Clarify definition of "major equipment damage" in EAL 7.3.1.

1.4.13 RCR 2003-0051, Action #2, Clarify Expected Response to Loss of RPV Level Trend.

PROCEDURE 5.7.1 l REVISION 30 l PAGE 87 OF 87 l

.,I USE: REFERENCE CNS OPERATIONS MANUAL EFFECTIVE: 4/24/03 EPIP PROCEDURE 5.7.17 APPROTVL: S4/QA DOSE ASSESSMENT OWNER: R. J. FISCHER DEPARTMENT: EP

1. PURPOSE ............................................................ 1

2. PRECAUTIONS AND LIMITATIONS ..................................... 2

3. REQUIREMENTS ....... .............................................. 2

4. COMPUTER DOSE PROJECTION (CNS-DOSE) ............................ 3

5. HAND-CALCULATED DOSE PROJECTION (CENTERLINE) ................. 5

6. HAND-CALCULATED DOSE PROJECTION (NON-CENTERLINE) ............ 8

7. CORRELATING OFF-SITE SAMPLE RESULTS WITH DOSE PROJECTIONS©

.................................................................... .12

8. CORE DAMAGE ESTIMATE USING IN-CONTAINMENT HI-RANGE RADIATION MONITORS .......................................................... 14 ATTACHMENT 1 HAND-CALCULATED DOSE PROJECTION (NON-CENTERLINE) .15 ATTACHMENT 2 TRANSIT TIMES AND EFFECTIVE AGES OF NOBLE II GASES AT RECEPTOR SITES .17 ATTACHMENT 3 HAND-CALCULATED DOSE PROJECTION (CENTERLINE)

.................................................. 18 ATTACHMENT 4 CORRELATING OFF-SITE SAMPLE RESULTS WITH DOSE PROJECTIONS .......... ................. 20 ATTACHMENT 5 METEOROLOGICAL AND RADIOLOGICAL DATA SOURCES FOR CNS-DOSE ........... ................ 21 ATTACHMENT 6 PMIS SYSTEM ACCESS AND USE .................... 22 ATTACHMENT 7 CORE DAMAGE ESTIMATION ....................... 29 ATTACHMENT 8 INFORMATION SHEET ........................... 30 04 1. PURPOSE

[ ] 1.1 HD This procedure provides instructions for performing a dose projection using the CNS-DOSE Computer Program.

[ ] 1.2 This procedure provides a manual backup method for performing dose 0__ [ ] 1.3 assessment.

This procedure provides instructions for making a rapid gross estimation of core damage based on in-containment high range radiation monitor readings for primary containment LOCA events.

[ ] 1.4 This procedure provides instructions for obtaining meteorological data from alternate sources if the primary sources are not available. The general order of preference will be PMIS, National Weather Service, and then the use of historically determined default values.

PROCEDURE 5.7.17 REVISION 31 PAGE 1 OF 33

2. PRECAUTIONS AND LIMITATIONS

[ ] 2.1 Actual dose rates will vary as a function of:

[] 2.1.1 The total curies released.

[1 ] 2.1.2 Release rate.

[] 2.1.3 The duration of the release.

[] 2.1.4 The isotopic mixture of the release.

[] 2.1.5 Meteorological conditions.

[ ] 2.2 Update and refine dose calculations upon significant changes in one or more of the above parameters.

[ ] 2.3 Should a release occur which necessitates rapid decision making concerning the recommendation of protective actions, the guidance contained in Procedure 5.7.20 should be followed.

[ ] 2.4 Attachment 7 should be used to estimate core damage only in cases where the high range in-containment radiation monitors are exposed to coolant or steam (i.e., only for primary containment LOCA situations). For other accident sequences, a Reactor Coolant System (RCS) sample and Core Damage Assessment Program (CORDAM) must be used. The Post-Accident Sampling System (PASS) may be used, as required, to obtain the RCS sample.

[ ] 2.5 If the needed KAMAN monitor(s) is (are) inoperable, Release Rate Determinations shall be performed using Procedure 5.7.16.

3. REQUIREMENTS

[ ] 3.1 Ensure following equipment and materials are available, as needed:

[] 3.1.1 COMPUTERIZED DOSE PROJECTION (CNS-DOSE)

[1 ] 3.1.1.1 Computer terminals.

[] 3.1.1.2 Computer printers.

[ 3.1.2 MANUALLY CALCULATED DOSE PROJECTION

[] 3.1.2.1 Environs map.

[1 3.1.2.2 X/Q isopleths (off-centerline only).

PROCEDURE 5.7.17 REVISION 31 PAGE 2 OF 33

[] 3.1.2.3 Scientific calculator.

[ ] 3.2 A release of airborne radioactive material has or may occur.

[ ] NOTE 1 - When Meteorological or Radiological data needed to perform dose assessment is unavailable or "unhealthy", refer to Attachment 5 for alternate sources of data. Health "quality codes" are defined in Attachment 6.

[ NOTE 2 - If the user is not familiar with the use of PMIS, Attachment 6 provides an overview and instructions on access and selected use of PMIS.

4. COMPUTER DOSE PROJECTION (CNS-DOSE)

[ ] 4.1 To start the dose projection program on a PMIS terminal, enter the turn-on code "DOSE" on a terminal logged into either the Primary or Backup System.

[ ] 4.2 The dose projection program can also be run on a non-PMIS terminal.

However, this is reserved for personnel having access to an account on the computer and familiar with its use. To start the dose projection program on a non-PMIS terminal, on either PMIS computer, login to an account that has privileges to run PMIS software and run program

[NPPD.EXECUTE]NPDOSEZ.

[ ] 4.3 Each time the program is started or the "New Sample" option is selected, new data will be loaded into the program. Verify that Field 1 correctly indicates the origin of the release and the data displayed is "healthy" and correct.

Health "quality codes" are defined in Attachment 6. Alternate sources of meteorological and radiological data needed to run CNS-DOSE or perform a hand-calculation are found in Attachment 5.

[] 4.4 Determine if SGT is in the effluent stream and if it is functional. Consult with Radiological, Operations, and Engineering personnel for this determination, if available.

[ ] 4.5 Estimate the duration of release (consult with Operations and/or Engineering for this time estimate) in hours. If the estimated duration of release cannot be determined, use the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months default value.

PROCEDURE 5.7.17 l REVISION 31 l PAGE 3 OF 33

NOTE - The Iodine to Noble Gas ratio is very dependent on the answer to the core degraded question and has a significant impact on the resultant dose projection calculations. The core is considered to be degraded if any of the following listed conditions are met OR if they were met and have subsequently dropped below the condition threshold. The answer to the core degraded question is coordinated between Radiological Protection, Chemistry, Operations, and Engineering, if available.

4.6 Determine if the core is degraded (fuel cladding loss) as indicated by any of the following conditions:

[] 4.6.1 15,000 mrem/hr on SJAE monitor.

[1 ] 4.6.2 Reactor Coolant Sample > 300 Ci/gm Dose Equivalent I-131.

[] 4.6.3 LOCA with DW Rad Monitor reading > 2500 REM/hr.

[] 4.6.4 Non-LOCA with DW Rad Monitor reading > 115 REM/hr.

[] 4.6.5 Main Steam Line Radiation Monitor Readings 2 Hi-Hi Alarm Setpoint.

[] 4.6.6 Reactor water level below 0" FZ (Fuel Zone) or cannot be determined.

4.7 DETERMINE IF RELEASE PATHWAY IS THROUGH REACTOR BUILDING

[] 4.7.1 If release bypasses Reactor Building (i.e., direct venting of drywell or a release from the Turbine Building), then enter N.

[1] 4.7.2 If release is through Reactor Building, then enter Y.

] 4.8 Make corrections or changes, as necessary.

[ ] 4.9 Use the ENTER key to accept data and move to the next field.

[ ] 4.10 Press the RESULTS option to display the dose projections.

[ 4.11 Select either the PRINT or HARD COPY option to make a hard copy of the results.

] 4.12 Select the "New Sample" or "Edit" option to return to the previous display and obtain new data or make additional changes.

[ ] 4.13 Exit the program by entering "Q" or pressing the "CANC" key on PMIS terminals.

[ ] 4.14 Select the "Help" option for additional program operational information.

I PROCEDURE 5.7.17 l REVISION 31 l PAGE 4 OF 33

5. HAND-CALCULATED DOSE PROJECTION (CENTERLINE)

NOTE - This method reflects the methodology used in the CNS-DOSE Program. It gives only downwind dose values for plume centerline at distances of 1, 2, 5, and 10 miles from the site. For calculating doses at specific receptor locations, the method in Section 6 is used.

5.1 Obtain release rate from effluent KAMAN monitor digital readout in ,Ci/sec and record value in Block 1 on Attachment 3. If KAMAN is inoperable, complete the appropriate attachment of Procedure 5.7.16 and record the noble gas release rate value (Ci/sec) in Block 1 on Attachment 3.

NOTE - The answer to the question concerning the status of the Standby Gas Treatment System has a significant impact on the resultant dose projection calculation. The answer to this question is coordinated with Radiological, Operations, and Engineering personnel, if available.0 5.2 Determine if SGT is in the effluent stream.

[] 5.2.1 If SGT is in the effluent stream, enter 0.01 in Block 2 of Attachment 3.

[ ] 5.2.2 If SGT is not in the effluent stream, enter 1 in Block 2 of Attachment 3.

NOTE - The Iodine to Noble Gas ratio is very dependent on the answer to the core degraded question and has a significant impact on the resultant dose projection calculations. The core is considered to be degraded if any of the following listed conditions are met OR if they were met and have subsequently dropped below the condition threshold. The answer to the core degraded question is coordinated between Radiological Protection, Chemistry, Operations, and Engineering, if available.

5.3 Determine if the core is degraded (fuel cladding loss) as indicated by any of the following conditions:

[ ] 5.3.1 15,000 mrem/hr on SJAE monitor.

[] 5.3.2 Reactor Coolant Sample > 300 ,uCi/gm Dose Equivalent I-131.

[ 5.3.3 LOCA with DW Rad Monitor reading > 2500 REM/hr.

[] 5.3.4 Non-LOCA with DW Rad Monitor reading > 115 REM/hr.

[] 5.3.5 Main Steam Line Radiation Monitor Readings 2 Hi-Hi Alarm Setpoint.

PROCEDURE 5.7.17 REVISION 31 PAGE 5 OF 33

5.3.6 Reactor water level below 0" FZ (Fuel Zone) or cannot be determined.

5.3.7 If core is degraded, obtain the Iodine to Noble Gas ratio from Table 1 of Attachment 3 and enter that value in Block 3 of Attachment 3.

5.3.8 If core is not degraded, enter 1.86E-7 in Block 3 of Attachment 3.

[5.4 Obtain the Noble Gas energy factor (MeV/dis) based on time since reactor shutdown in hours from Table 2 on Attachment 3 and enter this value in Block 4 on Attachment 3.

[5.5 Obtain the wind speed in miles per hour (mph) from PMIS and record the value in Block 5 of Attachment 3. If wind speed is not available from PMIS, call the National Weather Service (NWS) in Valley, NE and request an estimate of wind speed at CNS for the appropriate elevation. The telephone number for the NWS may be found in the Emergency Telephone Directory -

Federal TAB.

5.5.1 If the release is from the ERP, use wind speed at the 100 meter level.

If 100 meter data is unavailable, use the 60 meter data. If wind speed is unavailable from PMIS, and the NWS cannot be contacted, then use the historical default wind speed value of 13 mph.

5.5.2 If the release is from any other source, use the wind speed at the 10 meter level. Either MET tower 10 meter level is acceptable. If 10 meter data is unavailable, use the 60 meter data. If wind speed is unavailable from PMIS, and the NWS cannot be contacted, then use the historical default wind speed value of 8 mph.

[5.6 Determine the atmospheric stability class ("A" through "G") from PMIS and record in Block 6 on Attachment 3. If the stability class cannot be obtained from PMIS and the National Weather Service cannot be contacted, use "D" as the default stability class.

5.6.1 If using temperatures from the NWS to develop delta-T-based stability class, request the temperatures (10 meter (M) and 100 M) in degrees Centigrade. Determine degrees Centigrade (C) delta-T and the appropriate stability class using the following formula and table:

100 M C - 10 M C = delta-T C PROCEDURE 5.7.17 1 REVISION 31 l PAGE 6 OF 33

[] 5.7 DETERMINE IF RELEASE PATHWAY IS THROUGH REACTOR BUILDING

[] 5.7.1 If release bypasses Reactor Building (for example, direct venting of drywell or a release from the Turbine Building), then enter 1 in Block 7 on Attachment 3.

[] 5.7.2 If release is through the Reactor Building, then enter 0.5 in Block 7 on Attachment 3.

[ ] 5.8 Obtain TEDE Noble Gas Dose Conversion Factor from Table 3 of Attachment 3 and record in Block 8 on Attachment 3.

[ ] 5.9 Obtain TEDE Iodine Dose Conversion Factor from Table 3 of Attachment 3 and record in Block 9 on Attachment 3.

[] 5.10 Obtain CDE Iodine Dose Conversion Factor from Table 3 of Attachment 3 and record in Block 10 on Attachment 3.

[ ] 5.11 Compute TEDE "sub-calculation" value and record in Block 11 of Attachment 3.

[(Block l)(Block 4)(Block 8 + [(Block 1)(Block 2)(Block 3)(Block 7)(Block 9)1 (Block 5)

[ ] 5.12 Using the appropriate release point (ERP or other) and stability class (Block 6), obtain the mixing factors (/Qs) for distances 1, 2, 5, and 10 miles from Table 4 on Attachment 3 and record in Block 12 of Attachment 3.

[] 5.13 Compute the TEDE dose rate for each distance and record values in Block 13 on Attachment 3.

(Block 11) x (Block 12)

[] 5.14 Estimate the duration of the release (consult with Operations and/or Engineering for this time estimate) in hours and record value in Block 14 on Attachment 3. If the estimated duration of release cannot be determined, use 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months as a default value.

[ ] 5.15 Compute integrated TEDE doses for each distance and record values in Blocks 15 on Attachment 3.

(Block 13) x (Block 14)

[] 5.16 Compute CDE "sub-calculation" value and record in Block 16 of Attachment 3.

(Block 1)(Block 2)(Block 3)(Block 7)(Block 10)

(Block 5)

PROCEDURE 5.7.17 _ REVISION 31 l PAGE 7 OF 33

[] 5.17 Compute the CDE dose rate for each distance and record values in Block 17 on Attachment 3.

(Block 16) x (Block 12)

[ ] 5.18 Compute the CDE dose for each distance and record values in Block 18 on Attachment 3.

(Block 17) x (Block 14)

[ ] 5.19 Refer to Procedure 5.7.1 to determine if an emergency should be declared due to radiological effluent (dose rate or integrated dose to a member of the public) calculated at or beyond 1 mile.

[ ] 5.20 Refer to Procedure 5.7.20 to determine if any protective action recommendations should be made to off-site authorities.

[ ] 5.21 Recalculate dose projections whenever conditions change significantly.

[] 5.22 Record name, time, and date at the bottom of Attachment 3.

6. HAND-CALCULATED DOSE PROJECTION (NON-CENTERLINE)

[ ] 6.1 Obtain release rate from effluent KAMAN monitor digital readout in IiCi/sec and record value in Block 1 on Attachment 1. If IAMAN is inoperable, complete appropriate attachment of Procedure 5.7.16 and record the noble gas release rate value (Cilsec) in Block 1 on Attachment 1.

[ ] NOTE - The answer to the question concerning the status of the Standby Gas Treatment System has a significant impact on the resultant dose projection calculation. The answer to this question is coordinated with Radiological, Operations, and Engineering personnel, if available.

[] 6.2 Determine if SGT is in the effluent path.

[1 ] 6.2.1 If SGT is in effluent path, enter 0.01 in Block 2 on Attachment 1.

[1 ] 6.2.2 If SGT is not in effluent path, enter 1 in Block 2 on Attachment 1.

I PROCEDURE 5.7.17 l REVISION 31 l PAGE 8 OF 33

NOTE - The Iodine to Noble Gas ratio is very dependent on the answer to the core degraded question and has a significant impact on the resultant dose projection calculations. The core is considered to be degraded if any of the following listed conditions are met OR if they were met and have subsequently dropped below the condition threshold. The answer to the core degraded question is coordinated between Radiological Protection, Chemistry, Operations, and Engineering, if available.

6.3 Determine if the core is degraded (fuel cladding loss) as indicated by any of the following conditions:

[] 6.3.1 15,000 mrem/hr on SJAE monitor.

[] 6.3.2 Reactor Coolant Sample > 300 [iCi/gm Dose Equivalent 1-131.

[] 6.3.3 LOCA with DW Rad Monitor reading > 2500 REM/hr.

[ 6.3.4 Non-LOCA with DW Rad Monitor reading > 115 REM/hr.

[] 6.3.5 Main Steam Line Radiation Monitor Readings Hi-Hi Alarm Setpoint.

[] 6.3.6 Reactor water level below 0" FZ (Fuel Zone) or cannot be determined.

[] 6.3.7 If core is degraded, obtain the Iodine to Noble Gas ratio from Table 1 of Attachment 1 and enter that value in Block 3 on Attachment 1.

[] 6.3.8 If core is not degraded, enter 1.86E-07 in Block 3 on Attachment 1.

6.4 Determine the energy factor (MeV/dis) based on time since reactor shutdown in hours and Table 2 on Attachment 1, and enter value in Block 4 on Attachment 1.

6.5 Obtain the wind speed in miles per hour (mph) from PMIS and record the value in Block 5 on Attachment 1. If wind speed is not available from PMIS, call the National Weather Service (NWS) in Valley, NE and request an estimate of wind speed at CNS for the appropriate elevation. The telephone number for the NWS may be found in the Emergency Telephone Directory -

Federal TAB.

[] 6.5.1 If the release is from the ERP, use wind speed at the 100 meter level.

If 100 meter data is unavailable, use the 60 meter data. If wind speed is unavailable from PMIS, and the NWS cannot be contacted, then use the historical default wind speed value of 13 mph.

PROCEDURE 5.7.17 l REVISION 31 l PAGE 9 OF 3

[] 6.5.2 If the release is from any other source, use the wind speed at the 10 meter level. Either MET tower 10 meter level is acceptable. If 10 meter data is unavailable, use the 60 meter data. If wind speed is unavailable from PMIS, and the NWS cannot be contacted, then use the historical default wind speed value of 8 mph.

[ 6.6 Determine the wind direction (from) in degrees from PMIS and record in Block 6 on Attachment 1. If wind direction is not available from PMIS, call the National Weather Service (NWS) in Valley, NE and request an estimate of wind direction at CNS for the appropriate elevation. The telephone number for the NWS may be found in the Emergency Telephone Directory - Federal TAB.

[ ] 6.7 Determine the atmospheric stability class ("A" through "G") from PMIS and record in Block 7 on Attachment 1. If the stability class cannot be obtained from the PMIS and the NWS cannot be contacted, use "D" as the default stability class.

[] 6.7.1 If using temperatures from the NWS to develop delta-T-based stability class, request the temperatures (10 meter (M) and 100 M) in degrees Centigrade. Determine degrees Centigrade (C) delta-T and the appropriate stability class using the following formula and table:

100 M °C - 10 M °C = delta-T °C delta-T °C < -1.7 -1.7 to -1.5 -1.5 to -1.3 -1.3 to 0.451 0.45 to 1.3 1.3 to 3.6 > 3.6 StabilityClass A B C D E F G

[] 6.8 DETERMINE IF RELEASE PATHWAY IS THROUGH REACTOR BUILDING

[] 6.8.1 If the release bypasses Reactor Building (for example direct venting of the drywell or a release from the Turbine Building), then enter 1 in Block 8 on Attachment 1.

[1 ] 6.8.2 If the release is through Reactor Building, then enter 0.5 in Block 8 on Attachment 1.

[ ] 6.9 Obtain TEDE Noble Gas Dose Conversion Factor from Table 3 of Attachment 1 and record in Block 9 on Attachment 1.

[ ] 6.10 Obtain TEDE Iodine Dose Conversion Factor from Table 3 of Attachment 1 and record in Block 10 on Attachment 1.

[] 6.11 Obtain CDE Iodine Dose Conversion Factor from Table 3 of Attachment 1 and record in Block 11 on Attachment 1.

PROCEDURE 5.7.17 REVISION 31 l PAGE 10 OF 33 l

[ ] 6.12 Obtain the mixing factor (X/Q) for the receptor point or location.

[ ] 6.12.1 Record location or receptor point ID at the top of Attachment 1.

[ ] 6.12.2 Obtain the proper X/Q isopleth overlay based on stability class and release point.

[] 6.12.2.1 Overlays are available in the TSC or EOF for both elevated and ground level releases for each stability class.

Use ground level isopleths for all releases which are not from the ERP.

[ ] 6.12.3 Place the isopleth overlay on an Emergency Planning Zone map scaled to 1" per mile. The preferred map is the "Cooper Nuclear Station 20 Mile Plume Exposure" map with sectors, radii, and wind direction labeled. One is posted in the TSC and EOF.

[] 6.12.4 Orient the isopleth overlay so the centerline of the isopleth is over the wind direction radius, the open end of the isopleth is downwind, and the asterisk is over CNS.

[ ] 6.12.5 Lightly mark the desired receptor location on the isopleth with a pencil.

[] NOTE - All X/Qs have negative exponents.

[ ] 6.12.6 Using the legend in the lower right hand corner of the isopleth overlay, linearly interpolating as necessary, determine a X/Q value for the receptor site.

[ ] 6.12.7 Record the X/Q value in Block 12 on Attachment 1.

[ 6.13 Compute TEDE dose rate (REM/hr) and record in Block 13 on Attachment 1.

[(Block 1)(Block 4)(Block 9)1+r(Block 1)(Block 2)(Block 3)(Block 8)(Block 10)1 x (Block 12)

(Block 5)

[ ] 6.14 Estimate the duration of the release (consult with Operations and/or Engineering for this time estimate) in hours and record the value in Block 14 on Attachment 1. If the estimated duration of release cannot be determined, use 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months as a default value.

[ ] 6.15 Compute the integrated TEDE dose (REM) and record in Block 15 on Attachment 1.

(Block 13) x (Block 14)

PROCEDURE 5.7.17 l REVISION 31 l PAGE 11 OF 33l

[ ] 6.16 Compute CDE dose rate (REM/hr) and record in Block 16 on Attachment 1.

(Block 1)(Block 2)(Block 3)(Block 8)(Block 11) x (Block 12)

(Block 5) 6.17 Compute CDE dose (REM) and record in Block 17 on Attachment 1.

(Block 14) x (Block 16) 6.18 Record name, time, and date at the bottom of Attachment 1.

7. CORRELATING OFF-SITE SAMPLE RESULTS WITH DOSE PROJECTIONSO NOTE 1 - This section describes the methodology to be used to correlate CNS-DOSE results (estimated gross iodine concentrations) with gross iodine concentrations sampled in the field.

NOTE 2 - This section is to be used by dose assessment personnel in the EOF once field teams have been dispatched and sample results become available.

NOTE 3 - Initial dose projections (computer and hand-calculated) are based upon assumed radionuclide concentrations until actual concentrations have been measured. Off-site sample results are used to determine a dose correction factor which may be applied to adjust the CNS-DOSE Program.

7.1 FIELD TEAM SAMPLE TO CNS-DOSE COMPARISON

[1 ] 7.1.1 Radiological Assessment Supervisor shall:

[] NOTE 1 - Prior to comparing field team air sample results, ensure that the time of the field team air sample and the time of "CNS-DOSE" dose assessment are comparable.

[] NOTE 2 - If the field team air sample is reported from a distance other than 1, 2, 5, or 10 miles, use the appropriate stability class/release point isopleth to determine what CNS-DOSE predicted iodine air sample results would be at that distance prior to performing the field team sample comparison.

[1 ] 7.1.1.1 Compare the field team iodine air sample concentrations with the predicted CNS-Dose iodine air sample concentrations using the decision tree in Step 7.1.2.

[] 7.1.1.2 Radiological Control Manager shall review the field team corrected dose assessment results and communicate any change in PARs or Classification to the Emergency Director.

I PROCEDURE 5.7.17 REVISION 31 PAGE 12 OF 33

[] 7.1.2 N 1

.\ (Validate Inputs to the

_- CNS-Dose Model Y

Provide dose assessment results and atL 4 to Radiological Control END Manager

[] 7.2 APPLYING FIELD TEAM CORRECTION TO CNS-DOSE

[] 7.2.1 Apply the correction to CNS-Dose using the "Field Adjust" OPTION of CNS-DOSE.

[] 7.2.1.1 At the MAIN CNS-DOSE screen, select option "Field Adjust".

PROCEDURE 5.7.17 l REVISION 31 l PAGE 13 OF 33

[ ] 7.2.1.2 Enter the radius distance from CNS in miles at the prompt (1, 2, 5, and 10 are the only options).

[ ] 7.2.1.3 Enter the gross iodine concentration (in i[Ci/cc) obtained from the field at the prompt.

[] 7.2.1.4 After obtaining new Results from CNS-DOSE, compare new PARs to any PARs previously transmitted to off-site authorities.

8. CORE DAMAGE ESTIMATE USING IN-CONTAINMENT HI-RANGE RADIATION MONITORS NOTE 1 - Attachment 7 is only used for core damage estimates where the in-containment radiation monitors are exposed to coolant or steam (i.e., only for primary containment LOCA situations). For other accident sequences, a Reactor Coolant System (RCS) sample and Core Damage Assessment Program (CORDAM) must be used. The Post-Accident Sampling System (PASS) may be used, as required, to obtain the RCS sample.

NOTE 2 - The release from the core may bypass the containment, be retained in the primary system, or not be uniformly mixed. Therefore, a low containment radiation reading does not guarantee a lack of core damage. The levels of damage indicated by the value in Attachment 7 are considered minimum levels unless there are inconsistent monitor readings.

NOTE 3 - Inconsistent monitor readings may be due to the uneven mixing in containment (e.g., steam rising to the top of the dome). It may take hours for uniform mixing.

[ ] 8.1 The Chem/RP Coordinator or designee shall perform following steps to determine an estimate of core damage, if decisions must be made which are based on core conditions and PASS results are not available.

[] 8.1.1 Obtain highest in-containment hi-range radiation monitor reading from RMA-RM-40A(B), DRYWELL RAD MONITOR, and record in Block 1 on Attachment 7.

[1 ] 8.1.2 Complete the calculations on Attachment 7.

[1 ] 8.1.3 Report results to the TSC Director.

I PROCEDURE 5.7.17 l REVISION 31 l PAGE 14 OF 33 l

(

ATTACHMENT 1 HAND-CALCULATED DOSE PROJECTION (NON-CENTERLINE)

Location or Receptor ID:

[(1) Noble Gas Release (2) Release Path (3) Iodine/Noble Gas (4) Energy Factor (5) Wind Speed (mph) (6) Wind (7) Stability (8) Release through Rate from KAMAN or through SBGT? Ratio Direction Class Reactor Building?

5.7.16 (pCi/Sec) No= 1; Yes= 0.5 l Yes = 0.01; No =1 (from Table 1) (from Table 2) ERP = 13; Other = 8 (° from) Default D For Columns 5, 6, and 7, use PMIS, NWS, or Defaults.

Mixing Factor Conversion Factors (from Table 3) (from Isopleths)

TEDE Noble Gas (9) (12)

TEDE odine (10)

CDE Iodine (1 1).

[.TEDE Dose Rate (13): (REMIhr)

[ Duration (Hours)

Default = 4 hrs TEDE Dose (REM)

(Block 13) x (Block 14)

. [(Block 1)(Block 4)(Block 9)1+f(Block 1)(Block 2)(Block 3)(Block 8)(Block 10)l (Bl k 12) (14) (15)

L ~~~~~~~~~(Block

)x(Bok1)- )

CDE Dose Rate (16): (REM/hr) 11 ll CDE Dose (REM)

(Block 14) x (Block 16)

[(Block 1)(Block 2)(Block 3)(Block 8)(Block 11)1 x (Block 12) (17)

Name/Time/Date: / I PROCEDURE 5.7.17 REVISION 31 PAGE 15 OF 33

ATTACHMENT 1 HAND-CALCULATED DOSE PROJECTION (NON-CENTERLINE)

TABLE 1 - IODINE TO NOBLE GAS RATIO VS. TABLE 2 - ENERGY FACTORS TIME SINCE SHUTDOWN TIME SINCE IODINE/NOBLE GAS RATIO TIME SINCE ENERGY SHUTDOWN NON-DEGRADED DEGRADED SHUTDOWN FACTOR (hrs) CORE CORE (hrs) (MeV/dis) t<1 1.86 E-7 2.71 E-1 t<1 0.75 1<t <2 1.86 E-7 3.57 E-1 1*< t < 2 0.60 2 < t<4 1.86 E-7 3.41 E-1 2 <t<4 0.40 4 < t < 10 1.86 E-7 2.81 E-1 4 < t < 10 0.25 10 < t < 30 1.86 E-7 2.30 E-1 10 < t < 30 0.15 30 < t< 100 1.86 E-7 1.65 E-1 30 < t< 100 0.09 100 t 1.86 E-7 1.40 E-1 100 t 0.07 TABLE 3 - DOSE CONVERSION FACTORS NON-DEGRADED CORE DEGRADED CORE TEDE Noble Gas 1.48 E-3 9.19 E-4 TEDE Iodine 8.77 E-2 2.98 E-2 CDE Iodine 2.04 E 0 4.96 E-1 PROCEDURE 5.7.17 REVISION 31 l PAGE 16 OF 33

ATTACHMENT 2 TRANSIT TIMES AND EFFECTIVE AGES OF NOBLE GASES AT RECEPTOR SITES

1. Effective Age is defined as time elapsed (hrs) since shutdown. For off-site locations, the effective age of the isotopic mixture may be obtained through summarizing following components:

[] 1.1 The effective aye at the time of release onset.

[] 1.2 The transit time from the release point to the receptor site (refer to Section 2 below).

2. CALCULATION OF TRANSIT TIME FROM THE RELEASE POINT TO THE RECEPTOR LOCATION

[ ] 2.1 Estimate the downwind distance (miles) to the receptor location.

[ ] 2.2 Divide the distance in miles by the 100m meter level wind speed (mph) to determine the plume transit time.

(1) RECEPTOR SITE (2) 100 METER LEVEL (3) PLUME TRANSIT DOWNWIND DISTANCE WIND SPEED TIME (hrs)

(miles) (mph) (1) (2)

+

4 +

3. DETERMINATION OF EFFECTIVE AGES AT RECEPTOR SITES 4.

(3) EFFECTIVE AGE OF (1) EFFECTIVE AGE OF (2) TRANSIT TIME FROM ISOTOPIC MIXTURE AT MIXTURE AT TIME RELEASE POINT TO RECEPTOR LOCATION OF RELEASE ONSET RECEPTOR LOCATION (hrs)

(hrs) (hrs) (1) + (2)

Name/Time/Date: l l I PROCEDURE 5.7.17 l REVISION 31 l PAGE 17 OF 33

C (

ATTACHMENT 3 HAND-CALCULATED DOSE PROJECTION (CENTERLINE)

(1) Noble Gas Release Rate (2) Release Path (3) Iodine/Noble Gas (4) Energy Factor (5) Wind Speed (mph) (6) Stability (7) Release through from KAMAN or through SBGT? Ratio (MeV/dis) Class Reactor Building?

5.7.16 (Ci/Sec) No = 1; Yes 0.5 Yes = 0.01; No = 1 (from Table 1) (from Table 2) Defaults ERP = 13; Other = 8 Default =D For Columns 5 and 6, use PMIS, NWS, rDefaults.

Conversion Factors (from Table 3) TEDE Sub-Calculation (11):

TEDE Noble Gas (8)

TEDEIodin (9) [(Block 1)(Block 4)(Block 8)1+f(Block 1)(Block (Block 5) 2)(Block 3)(Block 7)(Block 9)l CDE Iodine (0 TEDE RATE (REM/hr) Duration (hours) TEDE Dose (REM)

Mixing Factors (from Table 4) (Block 11 x Block 12) Default = 4 hrs (Block 13 x Block 14) 1 Mile (12) 1 Mile (13) (14) 1 Mile (15) 2 Mile (12) 2 Mile (13) 2 Mile (15) 5 Mile (12) 5 Mile (13) 5 Mile (15) 10 Mile (12) 10 Mile (13) 10 Mile 1 (15)

CDE Sub-Calculation (16): CDE Rate (REM/hr) CDE Dose (REM)

(Block 16 x Block 12) (Block 14 x Block 17) 1 Mile (17) 1 Mile (18)

[(Block 1)(Block 2)(3lock 3)(Block 7)(Block 10)1 2 Mile (17) 2 Mile (18)

(Block 5) 5 Mile (17) 5 Mile (18) 10 Mile (17) 10 Mile l (18)

Name/Time/Date: I /

I - -PROCEDURE 5.7.17 REVISION 31 PAGE 18 OF 33

ATTACHMENT 3 HAND-CALCULATED DOSE PROJECTION (CENTERLINE)

TABLE 1 - IODINE TO NOBLE GAS RATIO VS. TABLE 2 - ENERGY TIME SINCE SHUTDOWN FACTORS TIME SINCE IODINE/NOBLE GAS RATIO TIME SINCE ENERGY SHUTDOWN NON-DEGRADED DEGRADED SHUTDOWN FACTOR (hrs) CORE CORE (hrs) (MeV/dis) t<1 1.86 E-7 2.71 E-1 t<1 0.75 1*t<2 1.86E-7 3.57 E-1 1*t<2 0.60 2 t<4 1.86 E-7 3.41 E-1 2

t<4 0.40 4 t < 10 1.86 E-7 2.81 E-1 4*t<10 0.25 10 t < 30 1.86 E-7 2.30 E-1 10 t < 30 0.15 30 t < 100 1.86 E-7 1.65 E-1 30 t < 100 0.09 100 t 1.86 E-7 1.40 E-1 100 t 0.07 TABLE 3 - DOSE CONVERSION FACTORS NON-DEGRADED CORE DEGRADED CORE TEDE Noble Gas 1.48 E-3 9.19 E-4 TEDE Iodine 8.77 E-2 2.98 E-2 CDE Iodine 2.04 E 0 4.96 E-1 TABLE 4 - PLUME CENTERLINE X/Q'S (MIXING FACTORS)

I RELEASE JSTABILITY B I POINT lCLASS l A l B l C l D l E l F l G 1 MILE 2.87E-6 6.04E-6 1.17E-5 8.35E-6 1.03E-6 2.35E-11 1.31E-23 ERP 2 MILE 7.94E-7 1.78E-6 4.55E-6 8.21E-6 4.98E-6 8.12E-8 5.62E-13 (ELEVATED) 5 MILE 1.50E-7 3.42E-7 1.18E-6 3.77E-6 4.66E-6 1.09E-6 5.67E-9 10 MILE 4.51E-8 1.03E-7 4.58E-7 1.82E-6 3.13E-6 1.44E-6 4.OOE-8 OTHER 1 MILE 3.01E-6 6.90E-6 1.73E-5 5.10E-5 i.O9E-4 3.07E-4 7.67E-4 THAN ERP 2 MILE 8.03E-7 1.84E-6 5.15E-6 1.78E-5 3.86E-5 1.09E-4 2.71E-4 (GROUND 5 MILE 1.50E-7 3.44E-7 1.21E-6 4.98E-6 1.25E-5 3.52E-5 8.81E-5 LEVEL) 10 MILE I 4.51E-8 I 1.03E-7 4.63E-7 I 2.07E-6 6.43E-6 1.81E-5 4.52E-5jl PROCEDURE 5.7.17 REVISION 31 PAGE 19 OF 33

ATTACHMENT 4 CORRELATING OFF-SITE SAMPLE RESULTS WITH DOSE 1 PROJECTIONS l

1. CORRECTION FACTOR DETERMINATIONS USING OFF-SITE SAMPLING DATA (3) (4)

FIELD GROSS CNS-DOSE (5)

(1) (2) IODINE IODINE CORRECTION SAMPLE SAMPLE CONCENTRATION CONCENTRATION FACTOR (CF)

LOCATION TIME @Ci/cc) (,uCi/cc) (3) (4) 4 + 4- 4 I + 4- 4 I 1- 4- 4 Name/Time/Date: l l

3. Route completed form to Emergency Preparedness Department.

PROCEDURE 5.7.17 l REVISION 31 l PAGE 20 OF 33

ATTACHMENT 5 METEOROLOGICAL AND RADIOLOGICAL DATA SOURCES FOR CNS-DOSE NOTE 1 - When the normal source of meteorological data (PMIS MET screen) is not available or is "unhealthy", attempt to obtain the data by PMIS point ID. If PMIS is not available, call the National Weather Service (NWS) in Valley, NE to obtain the data. The telephone number is contained in the Emergency Telephone Directory - Federal TAB. If the NWS cannot be contacted, use default values.

NOTE 2 - If the user is not familiar with the use of PMIS, Attachment 6 provides an overview and instructions on access and selected use of PMIS.

NOTE 3 - The Turn-On-Code "VALUE" is used to display single point values and qualities.

NOTE 4 - The Turn-On-Code "MET" is used to display most meteorological point values and stability classes.

PMIS POINT ID DESCRIPTION MET001 IOOM LVL SIGMA THETA (15 MIN AVE)

MET004 100M LVL TEMPERATURE MET005 DELTA TEMPERATURE (lOOM-lOM)

MET006 100M LVL WIND DIR. (15 MIN AVE)

MET007 100M LVL WIND SPEED (15 MIN AVE)

MET009 60M LVL SIGMA THETA (15 MIN AVE)

MET012 60M LVL TEMPERATURE MET013 DELTA TEMPERATURE (10OM-60M)

MET014 60M LVL WIND DIR. (15 MIN AVE)

MET015 60M LVL WIND SPEED (15 MIN AVE)

MET017 10M LVL SIGMA THETA (15 MIN AVE)

MET020 10M LVL TEMPERATURE MET021 DELTA TEMPERATURE (60M-1OM)

MET023 IOM LVL WIND DIR. (15 MIN AVE)

MET024 10M LVL WIND SPEED (15 MIN AVE)

MET027 PRECIPITATION (15 MIN PERIOD)

MET028 10M TWR SIGMA THETA (15 MIN AVE)

MET029 10M TWR TEMPERATURE MET030 IOM TWR WIND DIR. (15 MIN AVE)

MET031 !OM TWR WIND SPEED (15 MINAVE)

N8000 RX BLDG EFFLUENT FLOW AVE N8001 TURB BLDG EFF HI RAD MON AVE N8002 TURB BLDG EFF NORM RAD MON AVE N8003 TURB BLDG FLOW AVE N8004 AOG & RW EFF HI RAD MON AVE N8005 AOG & RW EFF NORM RAD MON AVE N8006 RX BLDG EFF RAD MON AVE N8007 AOG & RW BLDG EFF FLOW AVE N8010 ERP HI RAD MON AVE N8011 ERP NORMAL RAD MON AVE N8012 ERP FLOW AVE N8013 SGT FLOW TO ERP AVE PROCEDURE 5.7.17 REVISION 31 PAGE 21 OF 33

I ATTACHMENT 6 PMIS SYSTEM ACCESS AND USE

1. PLANT MANAGEMENT INFORMATION SYSTEM (PMIS) 1.1 The PMIS System (PMIS) is a set of programs and hardware provided by NPPD that make use of VMS functions and additional peripherals (Data Concentrators) which provides access to plant parameters.

2. PMIS COMPUTERS 2.1 PMIS computers share a common set of peripherals (disk drives, tape drives, terminals, etc.) and software.

3. VMS OPERATING SYSTEM 3.1 The VMS Operating System (VMS) is the host operating system for the PMIS computers. It is a set of programs that interface with the computer hardware and peripherals, and allows the computers to recognize and process commands.

4. PMIS MODES 4.1 PMIS has three operational modes, Primary, Primary/Backup, and Backup, and will operate on either computer in one of the three modes. A computer with PMIS operating in either the Primary or Primary/Backup Mode is referred to as the Primary System and the one with PMIS operating in the Backup Mode is referred to as the Backup System.

4.2 The Primary and Primary/Backup Modes provide full PMIS capabilities, consisting (in part) of data acquisition and conversion, data display, data archiving, alarm processing, self monitoring, and many other functions that perform specialized calculations and displays.

4.3 The Backup Mode monitors the Primary System, transfers information necessary to keep the Backup System files and tables up-to-date, and automatically changes to the Primary Mode when a loss of the Primary System is detected (referred to as a FAILOVER). Although many functions are available on the Backup System, their use is discouraged because the lack of real-time data results in the display of inaccurate information (CNS-DOSE is an exception).

5. PMIS ACCESS 5.1 Access to PMIS is gained through various video display terminals, printer/plotters, and printers, including color graphic Information Display Terminals (IDTs) dedicated exclusively for PMIS access in the Control Room, TSC, and EOF.

I PROCEDURE 5.7.17 l REVISION 31 l PAGE 22 OF 33 l

I ATTACHMENT 6 PMIS SYSTEM ACCESS AND USE I 5.2 The IDTs and printers are selectively connected to either computer through a switching device controlled by PMIS. At system start or during a FAILOVER, all terminals and printers are switched to the Primary System. However, the SWITCH position may be changed at any time after that.

6. SCREEN FORMAT 6.1 When a terminal is under control of PMIS (instead of VMS), the screen display will be in a standard format consisting of four areas, OCA, GGDA, SSA, and FKA.

6.2 The OCA (Operator Communication Area) consists of the top two (one and two lines on the screen. This area is generally used to prompt-for and receive user inputs and display advisory and warning messages. In addition, some displays that require only one or two lines of screen use the OCA for display. Also (though technically not part of the OCA), the current date and time (updated once a second) is displayed at the right side of the screen on lines 1 and 2.

6.3 The GGDA (General and Graphic Display Area) consists of lines 4 through 47 and is used for most displays. In addition, some displays (chiefly functions requiring significant editing) also prompt-for and receive user inputs in the GGDA.

6.4 The SSA (SPDS Status Area) consists of lines 45 through 48 and contain four boxes that represent (by color code) the status of the SPDS (Safety Parameter Display System), which is a software system that monitors selected plant parameters and determines overall plant safety status.

6.5 The FKA (Function Key Area) consists of the bottom two (50 and 51) lines of the screen. The FKA is used to indicate which of the definable function keys are enabled. It also indicates which mode PMIS is in, the Plant Mode, and whether or not a PMIS "event" has occurred.

7. SCREEN-COPY FUNCTION 7.1 The screen-copy function, which is activated by pressing the HARD COPY key, provides full screen reproduction in color on a printer located in the same general area as the terminal.

8. PRINTER 8.1 The printers are connected to a specific computer and are generally accessed when a "...PRINT..." option is selected and a "logical name" is entered.

PROCEDURE 5.7.17 l REVISION 31 l PAGE 23 OF 33

I ATTACHMENT 6 PMIS SYSTEM ACCESS AND USE

9. LOGICAL NAME 9.1 Printers and terminals are usually referenced by "logical names", in the format of TTOO, TT01, etc. (IDTs), and LAOO, LA01, etc. (printers). The "logical name" for a device can usually be found on a tag on the device.

10. RESET FUNCTION 10.1 This function, which is activated by pressing the RESET key (PC keyboard) or CONTROL-RESET keys (IDT keyboard), clears the screen, sounds the bell, and resets internal parameters to the default settings, producing the same effect as a re-boot or turning power off and on.

11. IDE FIELD 11.1 User input to PMIS Programs is through an open IDE (Interactive Data Entry) field on the terminal. An open IDE field is denoted by a yellow box that appears in the OCA or GGDA area. Anything typed on the keyboard will be echoed in the box. Erasing or back-spacing is accomplished with the DEL key.

All entries into an IDE field must be terminated by pressing the ENTER key unless the field is overfilled or a function key is pressed (the terminal automatically adds a carriage return character in those cases).

12. TURN-ON-CODE 12.1 The Turn-On-Code (TOC) is the mechanism by which commands are issued to PMIS. This is a one to eight character code which is interpreted by PMIS and a corresponding command is issued.

13. PMIS DATABASE 13.1 All plant parameters (or additional data based on plant or PMIS parameters) that are processed by PMIS SYSTEM are defined in the PMIS DATABASE, which is a file that specifies the origin of the data, the frequency at which it is processed, the type of processing to be performed, etc. Each parameter is referred to as a "point" and is identified by a one to eight character name or POINT-ID (PID).

14. PMIS DATA PROCESSING 14.1 Some PMIS points are processed by scanning plant sensors (through the Data Concentrator) while others are calculated based on the values of previously processed points or PMIS parameters. All points values are then assigned a quality code stored in the Current Value Table (CVT).

I PROCEDURE 5.7.17 l REVISION 31 l PAGE 24 OF 33 l

ATTACHMENT 6 PMIS SYSTEM ACCESS AND USE 14.2 Data in the CVT is considered to be "real-time" and representative of current plant and system conditions.

14.3 At regular intervals (and other special circumstances) point values are also stored in an Archive File, which provides - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of on-line historical information.

15. PMIS DATA ACCESS 15.1 All point values in the CVT and Archive File are accessed by the POINT-ID.

16. QUALITY CODES 16.1 The Quality Code, assigned when point values are assigned, represents the general status and "health" of the point, and determines how it is used by PMIS Programs. The following is a list of PMIS quality codes and related information.

CODE DESCRIPTION COLOR HEALTH UNK Value unknown - not yet processed White Bad DEL Processing has been disabled Magenta Bad INVL Data concentrator error Magenta Bad RDER Data concentrator error Magenta Bad OIC Data concentrator error Magenta Bad BAD Outside instrument range Magenta Bad STAG Point failed stagnation check Magenta Bad UDEF Undefined (spare) Magenta Bad REDU Fails redundant point check Magenta Bad HALM Above high alarm limit Red Good LALM Below low alarm limit Red Good HWRN Above high warning limit Yellow Good LWRN Below low warning limit Yellow Good ALM State/Change-of-State alarm Red Good SUB Value has been substituted Blue Good DALM Alarm checking has been disabled Green Good NCAL Value cannot be calculated White Good INHB Alarm inhibited by cut-out point Green Good GOOD Passes all other checks Green Good 16.2 Not listed above is quality code OSUB (Operator Substituted), which is treated the same as SUB, and indicates that the value was substituted within that program. OSUB is not used in the CVT.

PROCEDURE 5.7.17 l REVISION 31 l PAGE 25 OF 33

ATTACHMENT 6 PMIS SYSTEM ACCESS AND USE

17. PMIS LOGIN 17.1 If the current date and time is displayed in the OCA and is being updated about once a second:

17.1.1 If "ENTER PASSWORD..." is displayed on line 2, press the ENTER key.

17.1.2 If "SELECT FUNC. KEY OR TURN ON CODE..." and an open IDE field is displayed on line 2, the IDT is logged into PMIS. No further action is necessary.

17.1.3 If a display is operating, press the CANC key.

17.1.4 If terminal does not respond or does not meet any of the above criteria, press the XOFF key once. The terminal should be automatically reset (screen clears and the bell sounds) after about 30 seconds, and either the "ENTER PASSWORD..." or

"...TURN-ON-CODE..." prompt should be displayed. Refer to the applicable previous step for more instruction.

17.2 If the current date and time is NOT displayed or is displayed but is not being updated:

17.2.1 Press the RESET key (PC keyboard) or CONTROL-RESET keys (IDT keyboard), wait at least 10 seconds, and press the ENTER key. If the date and time appear and began updating, refer to the previous (date and time updating) step.

17.2.2 If a "$" is displayed at the left of the screen, enter "LO" and press the ENTER key. After the "...LOGGED OFF..." message is displayed, press the ENTER key again.

17.2.3 After "Username:" is displayed, enter "PMIS" and press the ENTER key. A welcome message followed by "PMIS LOGGED OUT..." will be displayed. Do not press any keys for 5 minutes or until the PMIS login display appears. When the "ENTER PASSWORD..." prompt is issued, refer to the previous (date and time updating) step and login to PMIS.

17.3 If neither of the above criteria is met or the specified sequence of events does not occur, contact the Nuclear Information Services (NIS) Department for assistance.

PROCEDURE 5.7.17 REVISION 31 PAGE 26 OF 33

I ATTACHMENT 6 PMIS SYSTEM ACCESS AND USE

18. ACTIVATING A TURN-ON-CODE 18.1 If a display is currently operating in the area of the screen that the desired TOC requires, press the CANC key.

18.2 When "SELECT FUNC. KEY OR TURN ON CODE..." is displayed followed by an open IDE field, enter one of following:

18.2.1 A TOC (i.e., "GROUP" -- activates the Group Display Program; the program will then prompt the user to select a menu option).

18.2.2 A TOC followed by a space and optional text (i.e., "PLOT ARMl" --

activates the Real-Time Plot Program and plots the group "ARMl" without further user input; note that optional text is recognized by only selected TOCs).

18.2.3 Press one of the programmable function keys on the right hand key pad or top row of function keys (i.e., blue "GROUP DISP" key --

functions the same as the first example).

18.3 Refer to the FKA for the function keys that are enabled and their descriptions.

Use other options as provided by each program.

18.4 To exit a program, use the specified exit option (if provided) or press the CANC function key.

19. DETERMINING TO WHICH SYSTEM A TERMINAL IS CONNECTED The PMIS System to which a terminal is connected is indicated by the "CONSOLE =..." on the bottom line of the FKA as follows:

CONSOLE = PRIMARY -- Connected to the Primary System operating in the Primary Mode.

CONSOLE = PRIM/BAC -- Connected to the Primary System operating in the Primary/Backup Mode.

CONSOLE = BACKUP -- Connected to the Backup System.

CONSOLE = UNKNOWN -- PMIS is in a transition or unknown state.

I PROCEDURE 5.7.17 l REVISION 31 l PAGE 27 OF 33 l

I ATTACHMENT 6 PMIS SYSTEM ACCESS AND USE l

20. SWITCHING A DEVICE TO THE OTHER SYSTEM 20.1 On a terminal located in the same area as the device to be switched and connected to either PMIS System, activate the TOC "SWITCH".

20.2 A list of all devices that can be switched from that terminal will be displayed.

Included will be their logical names, description, and the CPU to which the device is connected.

20.3 To switch a device, press function key Fl and then enter the logical name at the prompt.

20.4 If the device is an IDT, it will be logged off PMIS.

20.5 If the device being switched is a terminal other than the one running SWITCH, both are connected to the same system and a TOC is currently active, a message will be displayed to that effect, and the user will be asked if it is to be switched anyway. If the answer is not YES, the device is not switched.

PROCEDURE 5.7.17 l REVISION 31 PAGE 28 OF 33

I ATTACHMENT 7 CORE DAMAGE ESTIMATION NOTE - This attachment is only used for core damage estimates where the in-containment radiation monitors are exposed to coolant or steam (i.e., only for primary containment LOCA situations). For other accidents sequences, utilize the Post-Accident Sampling System (PASS) and Core Damage Assessment Program (CORDAM).

(1)

HIGHEST (5)

DRYWELL (2) (3) (4) PERCENT RAD MONITOR 100% CORE MELT PERCENT CLAD READING CORE MELT FRACTION CORE MELT FAILURE (RMA-RM-40A,B) FACTOR (1) (2) (3) x 100 (4) x 10 2.44E+6 Report the results of the core damage estimate (Blocks 4 and 5) to the TSC Director.

Name/Time/Date: l l I PROCEDURE 5.7.17 l REVISION 31 l PAGE 29 OF 33

ATTACHMENT 8 INFORMATION SHEET DISCUSSION 1.1 This procedure covers dose projection. Dose projection represents calculation of an accumulated dose at some time in the future if current conditions continue.

1.2 The CNS-DOSE Computer Program is a software application operated on the PMIS computers. It makes use of current meteorological and radiological data from PMIS and manually entered data to perform dose projection for the area surrounding CNS. CNS-DOSE is the primary method of dose projection.

1.2.1 The PMIS Computer System consists of two computers operating in a Primary and Backup Mode. Historical data may be obtained from either system; however, current data may be obtained only from the Primary System.

1.2.2 Personal unfamiliar with the operation of PMIS should reference procedures governing the operation of PMIS or refer to Attachment 6.

1.3 The manual dose projection methods in this procedure are intended to be used when CNS-DOSE is unavailable. Where possible, data used is from the same source as that used by the computer programs. The hand calculations are divided into two sections. Section 5 is intended to be used by the on-shift personnel for centerline dose projections. Section 6 is intended for dose assessment personnel in projecting non-centerline values.

1.4 The correlation methodology as described in Section 8 provides EOF dose assessment personnel with a means of correlating field team iodine concentration data with CNS-DOSE projected iodine concentration. Such a correlation is necessary to determine if initial Protective Action Recommendations (PARs) were adequate to protect the health and safety of the public.

1.5 Containment radiation level provides a measure of core damage, because it is an indication of the inventory of airborne fission products (i.e., noble gases, a fraction of the halogens, and a much smaller fraction of the particulates) released from the fuel to the containment (refer to NEDO-22215, Pages 1 and 2; NEDC 02-009).

I PROCEDURE 5.7.17 l REVISION 31 l PAGE 30 OF 33

I ATTACHMENT 8 INFORMATION SHEET

2. REFERENCES 2.1 TECHNICAL SPECIFICATIONS 2.1.1 Technical Specification Bases B.3.3.6.1(2.d), Main Steam Line Radiation - High.

2.2 CODES AND STANDARDS 2.2.1 NRC Regulatory Guide 1.109, Revision 1, October 1977, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR50, Appendix I, Iodine Inhalation Dose Factors.

2.2.2 NRC Regulatory Guide 1.111, July 1977, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors.

2.2.3 NRC Regulatory Guide 1.145, August 1979, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants.

2.2.4 Health Physics Journal, November 1981, Noble Gas Dose Rate Conversion Factors.

2.2.5 ICRP 59, Working Breathing Rate.

2.2.6 EPA 400-R-92-001, May 1992, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents.

2.3 DRAWINGS (APS) 2.3.1 NPPD Drawing CNS-MI-102, Atmospheric Dispersion Model (EPM2)

Special Receptor Points, 10 Mile Radius.

2.3.2 NPPD Drawing CNS-MI-03, Preselected Radiological Sampling and Monitoring Points in the Vicinity of Cooper Nuclear Station, 10 Mile Radius.

2.3.3 NPPD Drawing 2.2 (P3-A-45), Revision 1, Cooper Nuclear Station Site and Property Boundary, 1 Mile Radius.

2.3.4 Cooper Nuclear Station 50 Mile Emergency Planning Zone, Revision 2, 50 Mile Radius.

I PROCEDURE 5.7.17 l REVISION 31 l PAGE 31 OF 33

ATTACHMENT 8 INFORMATION SHEET 2.4 VENDOR MANUALS 2.4.1 CNS Number 0984, PMIS Operator's Manual - SAIC Document 502-85500107-72.

2.5 PROCEDURES 2.5.1 Emergency Plan Implementing Procedure 5.7.1, Emergency Classification.

2.5.2 Emergency Plan Implementing Procedure 5.7.16, Release Rate Determination.

2.5.3 Emergency Plan Implementing Procedure 5.7.20, Protective Action Recommendations.

2.6 MISCELLANEOUS 2.6.1 Engineering Evaluation EE 02-056, Elimination of Meteorological Instrumentation System Strip Chart Recorder References.

2.6.2 General Electric Corporation, NEDO-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions.

2.6.3 NEDC 99-034, Control Room, EAB, and LPZ Doses Following a CRDA.

2.6.4 NEDC 02-004, Estimation of the Steam Jet Air Ejector Radiation Monitor, RMP-RM-15OA(B), Readings Following a 1% Fuel Clad release (Degraded Core) in the Reactor Coolant System.

2.6.5 NEDC 02-009, Estimation of Primary Containment High Range Monitor, RMA-RM-40A(B), Readings Following 1% Clad Failure in the RCS under Non-LOCA Conditions.

2.6.6 NEDO-31400, Safety Evaluation for Eliminating the BWR MSIV Closure Function and Scram Function for the MSL Rad Monitors.

2.6.7 NRC Inspection Report 89-35.

2.6.8 C NRC Inspection Report 91-12, Emergency Preparedness Annual Inspection Report. Affects Section 7 and NOTE prior to Step 5.2.

I PROCEDURE 5.7.17 l REVISION 31 l PAGE 32 OF 33 l

I ATTACHMENT 8 INFORMATION SHEET l 2.6.9 NRC Inspection Report 92-14, Emergency Preparedness Annual Inspection Report.

PROCEDURE 5.7.17 l REVISION 31 _ PAGE 33 OF 33 l

NLS2003056, Emergency Plan Implementing Procedures for Cooper Nuclear Station

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