ML18041A068

Forwards Revised Listed EPIPs & Rev 37 to Nine Mile Point Site Emergency Plan. Changes Are Submitted,Per 10CFR50,App E,Section V

ML18041A068
Person / Time
Site:	Nine Mile Point
Issue date:	01/12/1999
From:	Terry C NIAGARA MOHAWK POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML17059C479	List: ML18041A068
References
NUDOCS 9901220355
Download: ML18041A068 (541)

Text

h CATEGORY 1 REGULA RY INFORMATION DISTRIBUTIO SYSTEM (RIDS)

ACCESSION NBR:9901220355 DOC.DATE: 99/01/12 NOTARIZED: NO DOCKET FACIL:50-220 Nine Mile Point Nuclear Station, Unit 1, Niagara Powe 05000220 50-410 Nine Mile Point Nuclear Station, Unit 2 Niagara Moha 05000410 AUTH.NAME TERRY,C.D.

AUTHOR AFFILIATION Niagara Mohawk Power Corp.

gsqS'fgy RECIP.NAME RECIPIENT AFFILIATION Records Management Branch (Document on rol Desk) ~+P

SUBJECT:

Forwards revised listed EPIPs & rev 37 to "Nine Mile Point C Site Emergency Plan." Changes are submitted,per 10CFR50,App E,Section V. A D1STR1BUT10N CODE: A045D TITLE: OR COPIES RECEIVED:LTR Submittal: Emergency Preparedness I ENCL I SIZE: 1 Plans, Implement'g Procedures, P l)G C T

p'OTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL 0 PD1-1 PD 1 1 HOOD,D 1 1 NEER~

fi INTERNAL: AEOD/HAGAN,D 1 1 2 2 NRR/DRPM/PERB 1 1 NUDOCS -ABSTRACT 1 1 EXTERNAL: NOAC 1 1 NRC PDR 1 1 NOTE TO ALL MRZDSM RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD) ON EXTENSION 415-2083 TOTAL NUMBER OF COPIES REQUIRED: LTTR 9 ENCL 9

NIACARA MOHAWK C E N E RAT I 0 N NINE MILE POINT NUCLEAR STATION/LAKEROAD, P.O. BOX 63, LYCOMING, NEW YORK 13093/TELEPHONE (315) 349-7263 FAX (315) 349<753 BUSINESS CROUP CARL D. TERRY Vice President January 12, 1999 Nuclear Safety Assessment and Support United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 RE: Nine Mile Point Unit 1 Nine Mile Point Unit 2 Docket No. 50-220 Docket No. 50-410 DPR-63 NPF-6 Gentlemen:

Enclosed please find copies of the following cmcrgcncy procedure revisions for Niagara Mohawk Power Corporation's Nine Mile Point Nuclear Station:

~ EPIP-EPP-01, Revision 08, "Classification of Emergency Conditions at Unit 1"

~ EPIP-EPP-08, Revision 09, "Off-Site Dose Assessment and Protective Action Rccoinincndation"

~ EPIP-EPP-12, Revision 05, "Rc-Entry Procedure"

~ EPIP-EPP-14, Revision 02, "Emergency Access Control"

~ EPIP-EPP-23, Revision 08, "Emergency Personnel Action Procedures"

~ EPIP-EPP-24, Revision 02, "Nuclear Transportation Accidents"

~ EPIP-EPP-31, Revision 00, "Control Room Support Functions from the TSC"

~ EPMP-EPP-02, Revision 15, "Emergency Equipment Inventories and Checklists"

~ EPMP-EPP-06, Revision 04, "Emcrgcncy Response Organization Notification Maintenance and Surveillance"

~ EPMP-EPP-0101, Revision 02, "Unit 1 Emcrgcncy Classification Tcchnical Bases"

~ EPMP-EPP-0102, Revision 03, "Unit 2 Emcrgcncy Classification Technical Bases"

~ Nine Mile Point Site Emergency Plan, Revision 37 Thcsc procedure changes are being submitted as rcquircd by Section V to Appendix E of 10 CFR Part 50. Should you have any questions, please fccl free to contact Mr. James D. Jones, Director of Emergency Preparedness at (315) 349-4486.

Very ul yours, Carl D. Terry Vice President Nuclear Safety Asscssmcnt and lcld Support Enclosure

~c:

Mr. H.J. Miller, Regional Administrator, Region I (2 copies)

Mr. G. K. Hunegs, Senior Resident Inspector (1 copy)

Mr. D.S. Hood, Senior Project Manager, NRR (I copy)

Mr. S.S. Bajwa, Director, Project Dircctoratc, I-l, NRR (letter only)

Records Management i90<220355 ee01~2 PDR ADOCK 05000220 F PDR

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NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT NUCLEARSTATION EHERGENCY PLAN IMPLEMENTING PROCEDURE P P- P-0 R S ON 07

,C SS F CA 0 OF RG CY CONDITIONS~AT UNIT TECHNICAL SPECIFICATION RE(UIRED Approved by:

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N. L. Rademacher Plant Hanag 'r - Unit I Date

. THIS IS A FULL REVISION PER ODIC REVIEW, 02/08/98, NO CHANGE ffective Date: 06/26/97 PERIODIC REVIEW DUE DATE

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LIST OF EFFECTIV PAGES N

Coversheet .

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2 .

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5 .

January 1997 Page i EP I P-EPP-01 Rev 07

I TAB OF CONT NTS SECTION PAGE 1.0 PURPOSE ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ 1 2.0 PRIMARY RESPONSIBILITIES 1 3.0 PROCEDURE . 2 3.1 Station Shift Supervisor/Site Emergency Director 2 4.0 DEFINITIONS . 3

5.0 REFERENCES

AND COMMITMENTS 4 6.0 RECORD REVIEM AND DISPOSITION . 4 January 1997 Page ii EP I P-EPP-01 Rev 07

I 1.0 PURPOSE Provide the Control Room staff and Site Emergency Director with the criteria and method for classifying abnormal conditions into one of the four emergency classifications.

2.0 P INAR SPONSIBILITIES

2. 1 Station Shift Su ervisor (SSS)

~ Haintains awareness of any abnormal plant conditions or occurrences and evaluates the need to classify the condition in accordance with this procedure.

~ Upon initial declaration of an emergency, assumes the role of Site Emergency Director (SED) and functions as the SED until relieved of those duties by the on-call SED, other SRO or the emergency is terminated.

~ Declares any subsequent emergency classifications based on available information until relieved of SED, or other active SRO duties or the emergency is terminated.

~ For conditions classified as an Unusual Event, terminating the emergency in accordance with EPIP-EPP-25 "Reclassification and Recovery".

2.2 Site Emer enc Director (SED)

~ Upon activation of the TSC, relieves the SSS of the SED duties in accordance with EPIP-EPP-18 "Activation and Direction of the Emergency Plans".

Haintains awareness of any abnormal plant conditions or occurrences and evaluates the need to re-classify the condition in accordance with this procedure and in concurrence with the Corporate Emergency Director.

~ For conditions classified as an Alert or higher, terminating the emergency in accordance with EPIP-EPP-25 "Reclassification and Recovery".

2.3 Cor orate Emer enc Director CED For conditions classified as an Alert or higher, concur with the emergency classification or reclassification as determined by the SED.

January 1997 Page 1 EPIP-EPP-01 Rev 07

3.0 PROCEDURE NOTES: l. Entry into o-n emergency classification is not expected for planned outages of systems or equipment in which compensatory measures have been taken.

2. The SSS/SED should not delay actions that would mitigate or prevent an emergency or off-normal conditions, to classify an event. However, all events should be classified in accordance with this procedure no later than 15 minutes after indications are available in the Control Room that an EAL has been exceeded.

3.1 Station Shift Su ervisor Site Emer enc Director and Site Emer enc Director TSC 3.1.1 Continually monitor and evaluate plant conditions to determine if one or more emergency action level thresholds have been met or exceeded -Emergency Action Level Hatrix (Attachment 1).

3.1.2 While performing the following steps:

a. IF: An abnormal condition exists which meets or exceeds an emergency action level for a classification higher than is currently declared THEN: Go to Step 3.1.3 of this procedure

b. IF: It is determined that the emergency has been over classified OR that the current emergency classification is no longer warranted THEN: Enter EPIP-EPP-25 "Emergency Reclassification and

~

Recovery" and execute it concurrently with this procedure.

C. IF: . An EAL has been met or exceeded, but the EAL threshold or emergency condition no longer exists grior to making the emergency declaration (transitory event),

THEN: 1. Classify current conditions and declare the emergency, if necessary. for the declared

2. Hake notifications required emergency in accordance with EPIP-EPP-20.

3. Notify State, County and NRC of transitory event (even if no emergency is declared).

NOTE: The same Part 1 Notification Fact Sheet and NRC Notification Morksheet can be used to execute Steps 3.1.2.c.2 and 3.1.2.c.3.

January 1997 Page 2 EP IP-EPP-Ol Rev 07

3.1.3 IF: One or more emergency action level thresholds have been met or exceeded - Emergency Action Level Matrix (Attachment I)

THEN:

a. declare the highest level emergency classification for which an EAL is currently being met or exceeded

b. Enter EPIP-EPP-18 "Activation and Direction of the Emergency Plans" and execute it concurrently with this procedure

3. 1.4 WHEN: It has been determined that an emergency condition no longer exists THEN: Enter EPIP-EPP-25 "Emergency Reclassification and Recovery" and execute it concurrently with this procedure.

3.2 If the emergency declaration is due to an initiating condition affecting both Unit 1 and Unit 2, then the Unit 1 SSS shall assume the role of SED.

4.0 DEFINITIONS Unusual Event Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected.

Alert Events are in progress or have occurred which degrade plant safety systems to the extent that increased monitoring of plant safety functions is warranted. Any releases from these events are expected to be limited to small fractions of the EPA Protective Action Gui'deline plume exposure levels outside the site boundary.

Site Area Emer enc Events are in progress or have occurred which involve actual or likely major failures of plant functions intended for protection of the public. Releases are not expected to exceed EPA Protection Action Guideline plume exposure levels outside the site boundary.

General Emer enc Events are in progress or have occurred which involve actual or imminent substantial core degradation with potential for loss of containment integrity. Any releases from these events can be reasonably expected to exceed EPA Protective Action Guideline plume exposure levels outside the site boundary.

Transitor Event An event in which an emergency action level has been exceeded but the condition no longer warrants classification at that level pr',or to makin" the emergency declaration.

January 1997 Page 3 EPIP-EPP-01 Rev 07

4.0 (Cont)

Classification - Categorization of plant conditions or events into the appropriate emergency classification level.

Declaration - Announcement in the Control Room or TSC that an EAL has been met and an emergency classification level has been entered.

5.0 C S AND COMMITMENTS 5.1 icensee Documentation Unit I UFSAR, Chapter XIII 5.2 Standards Re ulations Codes

~ NUHARC NESP-007, Methodology for the Development of Emergency Action Levels 5.3 Policies 'Pro rams and Procedures

~ EPMP-EPP-0101, Unit 1 Emergency Classification Technical Bases

~ EPIP-EPP-18, Activation of the Emergency Plan EPIP-EPP-.25, Emergency Reclassification and Recovery

~ NRC Emergency Preparedness Position (EPPOS) f2, "Timeliness of Classification of Emergency Conditions" 5.4 Coamitments None 6.0 RECORD REYIEM AND DISPOSITION The following records generated by this procedure as a result of actual declared emergency at the Nine Mile Point Nuclear Station shall be maintained by Nuclear Records Hanagement for the Permanent Plant File in accordance with NIP-RHG-OI.

Not Applicable The following records generated by this procedure as a result of EP Drills/Exercises are not required for retention in the Permanent Plant File.

Not Applicable January 1997 Page 4 EPI P-EPP-01 Rev 07

This page represents the Emergency Action Level matrix/Unit 1 which is too large to fit in this document.

January 1997 Page 5 EP IP-EPP-Ol Rev 07

NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT NUCLEAR STATION EMERGENCY PLAN IMPLEMENTING PROCEDURE EPIP-EPP-08 REVISION 08 OFF-SITE DOSE ASSESSMENT AND PROTECTIVE ACTION RECOMMENDATION TECHNICAL SPECIFICATION REQUIRED Approved By: oYz2 R. G. Smith Plant Manager - Unit 1 Oate Approved By:

K. A. Dahlberg Plant Manager - Unit 2 Date THIS IS A FULL REVISION Effective Date: 04/01/98 PERIODIC REVIEW DUE DATE:

l LIST OF FF CTIVE PAG S

~N. ~ ~N" N ~ ~PN . ~NN ~PN . ~NN N Coversheet . 22 .

23 .

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25 .

2 ~ ~ 26 .

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March 1998 Page- i EP IP-EPP-08 Rev 08

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TABLE OF CONTENTS TION PAGE 1.0 PURPOSE . ~ ~ 1 2.0 PRIMARY RESPONSIBILITIES 3.0 PROCEDURE .

3. 1 Dose Assessment and Protective Action from the Control R oom . . 1 3.2 Dose Assessment and Protective Actions from the EOF 3 4.0 DEFINITIONS . 5

5.0 REFERENCES

/COMMITMENTS 5 6.0 RECORDS REVIEW AND DISPOSITION 7 ATTACHMENT 1: INITIAL DOSE ASSESSMENT AND PROTECTIVE ACTIONS . 8 ATTACHMENT 2: USE OF THE EDAMS COMPUTER 12 ATTACHMENT 3: METEOROLOGICAL DATA ACQUISITION 14 TACHMENT 4: RELEASE RATE DETERMINATION . 22 ACHMENT 5: REFINED DOSE ASSESSMENT AND PROTECTIVE ACTIONS 28 ATTACHMENT 5.1: CHRONOLOGICAL RELEASE RATE LOG . 33 ATTACHMENT 5.2: EDAMS DATA ENTRY FORM 34 March 1998 Page ii EP IP-EPP-08 Rev 08

1.0 PURPOSE To provide the methods for determining meteorology data, release rates, dose assessment and protective actions during accident conditions at Nine Mile Point.

~

~

2.0 PRIMARY RESPONSIBILITIES 2.1 The Station Shift Supervisor/Site Emergency Director (SSS/SED):

2.1.1 Ensures meteorological data acquisition, release rate determination, and dose assessment are performed during the initial stages of an emergency to support development of Protective Action Recommendations (PARs)

2. 1.2 Approves PARs and ensures their timely issue to the State and County 2.2 The Corporate Emergency Director (CED) approves PARs prior to their transmittal to the State and County, following EOF activation.

2.3 The Radiation Assessment Manager (RAM) is responsible to the SED for managing the onsite radiological monitoring and assessment aspects of the station during an emergency, following TSC activation.

Chemistry Technicians perform release rate assessments, obtain meteorological data, and develop PARs, prior to EOF activation.

2.5 The Offsite Dose Assessment Manager (ODAM) manages the offsite dose aspects of an emergency in order to assess the radiological consequences to the public, following EOF activation.

2.6 The Radiological Assessment Staff is responsible to the ODAM for obtaining meteorological data, determining source term, performing dose assessment, and developing PARs, following EOF activation.

3.0 PROCEDURE I

3.1 Dose Assessment and Protective Action from the Control Room CAUTION Calculation involving the determination of release rates and/or protection action shall be self-checked for accuracy.

• * * * * * * * * * * * * * * * * * *

• 4' * * * * * * * * * * * *

• March 1998 Page 1 EPIP-EPP-08

08

'ev

3.1.1 Chemistr Technician Actions a~ Consult the SSS/SED on plant conditions and possible release paths. If a General Emergency has been declared, assist SSS/SED in making Protective Action Recommendations based on plant conditions using Attachment l.

b. Access EDAHS computer using Attachment 2 C. Obtain meteorological data using Attachment 3.

d. Assess effluent monitor readings and conditions.

e. Determine release rate using Attachment 4. Combine multiple release points as follows:

1. Sum all release points from the same elevation (ground or elevated).

2. Calculate the total release rate from combined ground and elevated sources using the workspace on Attachment 1.

Use Attachment 1 flowchart and advise SSS/SED of any PARs recommended by the flowchart.

g. IF an unmonitored a'tmospheric release is suspected or known to be in progress, assist the SSS/SED in the following actions:

1. Advise the SSS/SED to expedite the dispatch of Radiation Protection (RP) Technician. Request assistance of the unaffected Unit or J.A.

Fitzpatrick if needed.

2. The RP Technician should be dispatched to potential plume centerline (wind direction (degrees) + 180'lume centerline), as close to the site boundary as practicable. See Attachment 1, Figure 1.4 for Site boundary location.

3. IF readings indicate > 1 rem/hr based on field survey perform the actions indicated in Attachment l.

h. Assist Communications Aide in completing. the meteorological data and release rate sections of the Part 1 Notification Fact Sheet.

Continue to monitor meteorological data, changes in effluent conditions or conditions that might lead to abnormal radiological effluents.

Parch 1998 Page 2 EPIP-EPP-08 Rev 08

3.1.1 (Cont)

j. Mhen requested, turn over all duties to the EOF.

3.1.2 SSS Actions

a. Verify that the Chemistry Technician is performing dose assessment and protective action development in a timely fashion and in accordance with Attachment 1.

b. Assess any release rates provided by the Chemistry Technician against the Emergency Action Levels (EAL).

c. Review AND approve PARs recorded on the Notification Fact Sheet Part 1, as required. Use ERPA map in Attachment 1 if desired.

3.2 Dose Assessment and Protective Actions from the EOF 3.2.1 Offsite Dose Assessment Mana er ODAM Actions NOTE: IF at any time the initiating conditions listed in Attachment 1 are met, THEN perform the actions listed in that attachment.

a ~ Perform actions as indicated in EPIP-EPP-23.

b. Verify Environmental Survey Sample Team Coordinator has been assigned and is:

1. Preparing for the dispatch of downwind survey teams.

2. Is aware of meteorological advisor status.

C. Perform or have performed the following:

1. Obtain meteorology data using Attachment 3 of this procedure.

2. Obtain effluent monitor readings and calculate release rate using Attachment 4 of this procedure.

3. Perform dose assessment calculation using Attachment 5 of this procedure.

d. Determine PARs using Attachment 5 of this procedure.

e. Interface with State and County representatives in the EOF.

l. 1. Keep State/County representatives confirmed data and results.

informed of March 1998 Page 3 EPIP-EPP-08 Rev 08

3.2.1 (Cont)

Complete Part 2 Notification Fact Sheet when ANY of the following conditions exist or are met:

1. Rad release that exceeds Tech Specification limits.

2. Significant changes in meteorological OR rad release conditions.

3. Every 30 minutes.

g. With each significant change in meteorological, actual release rate, and dose assessment data, OR every 30 minutes.

h. Constantly reassess effluent monitors (release rate) and meteorological data for changes. Perform new dose assessment as needed. Develop new PARs and/or verify the adequacy of PARs already made.

As Downwind Survey Team (DST) becomes available, utilize it to verify release rates. If these refined release rates differ significantly from those calculated from effluent monitor readings, reperform dose assessment using refined release rates.

Provide data for the Part 1 Notification Fact Sheet as requested.

k. Provide CED with pertinent information as needed.

1. Changing radiological conditions that may lead to PARs.

2. Protective actions for site staff.

e. Maintain Chronological Release Rate Log (see Attachment 5.1) .

3.2.2 EOF Dose Assessment Staff

'a ~ IF at any time the initiating conditions listed in Attachment 1 are met, THEN perform the actions listed in that attachment.

b. Perform actions as indicated in EPIP-EPP-23.

'C. Perform any actions as requested by the ODAM, including:

~ Obtaining meteorological data (Attachment 3)

March 1998 Page 4 EPIP-EPP-08 Rev 08

3.2.2.c (Cont)

~ Obtaining release rate data (Attachment 4)

~ Performing dose assessment and protective action recommendations (Attachment 5) 4.0 DEFINITIONS 4.1 CDET. Committed dose equivalent to the thyroid for the child.

4.2 EDAHS. Emergency Dose Assessment Hodeling System. A PC-based computer program that calculates release rates, doses and protective actions, and obtains meteorological data for emergencies.

4.3 HHS. Heteorological Honitoring System. Consists of the dedicated computer, main, backup and inland towers and software. Stores and edits site meteorological data.

RADDOSE. A subprogram of EDAHS, it performs the dose assessment Functions during emergencies.

4.5 SHELTERING. A protective action whose benefit is to bring the public to a heightened state of awareness. No dose reduction is assumed for sheltering.

TEDE. Total Effective Dose Equivalent.

5.0 REFERENCES

CONNITHENTS 5.1 Technical S ecifications None 5.2 Licensee Documentation 5.2.1 NHP Unit 1 FSAR, Section XV

a. Table XV-32

b. Table XV-28

c. Table XV-29

d. Table XV-23

e. Table XV-29d

f. Section 1.3.1

g. Section 2.1 Harch 1998 Page 5 EP IP-EPP-08 Rev 08

5.2.2 NHP Unit 2 USAR, Section 15

a. Table 15.6-15b

b. Table 15.4-12

c. Table 15.7-11

d. Table 15.6-8

e. Table 15.7-4

f. Table 15.6-3

g. Table 16.6-19 5.2.3 SEP, NMPC Nine Mile Point Nuclear Station Site Emergency Plan 5.2.4 NHPC Correspondence 96-MET-001 (Backup Tower Wind Speed Correction Factor) 5.2.5 NHP Correspondence 96-MET-002 (Main Tower Wind Speed Correction Factor) 5.2.6 NHP Correspondence 96-HET-004 (Backup Tower Wind Direction Concerns) 5.2.7 NHP Correspondence 96-HET-003 (Discussion at DER C-95-0693) 5.2.8 NHP Correspondence 96-MET-005 (Hain Tower 30'igma Theta Concern)

NHP Correspondence 97-HET-002 (Main Tower Wind Obstructions) 5.3 Standards Re ulations and Codes NUREG-0654, FEHA-REP-l, Rev 1, Supp 3, Criteria for Protective Action Recommendations for Severe Accidents 5.4 Policies Pro rams and Procedures 5.4.1 EPIP-EPP-07, Downwind Radiological Monitoring 5.4.2 EPIP-EPP-15, Health Physics Procedure 5.4.3 EPIP-EPP-23, Emergency Personnel Action Procedures 5.5 Commitments DER C-95-0693 (for Attachment 3)

March 1998 Page 6 EPIP-EPP-08 Rev 08

RECORDS REVIEM AND DISPOSITION The following records generated by this procedure shall be maintained by Records Management for the Permanent Plant File in accordance with NIP-RHG-OI, Records Management:

NOTE: For records generated due to an actual declared emergency only.

Attachment 1, Initial Dose Assessment and Protective Actions Attachment 4, Release Rate Determination Figure 5. 1, Chronological Release Rate Log =

Figure 5.2, EDAMS Data Entry Form 6.2 The following records generated by this procedure are not required for retention in the Permanent Plant File:

NOTE: For records generated NOT due to an actual declared emergency only.

Attachment 1, Initial Dose Assessment and Protective Actions Attachment 4, Release Rate Determination Figure 5. 1, Chronological Release Rate Log Figure 5.2, EDAMS Data Entry Form March 1998 Page 7 EPIP-EPP-08 Rev 08

ATTACHMENT 1: INITIAL DOSE ASSESSMENT AND PROTECTIVE ACTIONS Sheet 1 of 4 START NO ls Tcsal Reareae sara'O NO Table 1.1 ~~

decret elan roles rl rises<<rroyyt lo Score evidence oI sn Qryn<<cltcyod rcoosso7 YES YES YES t

SIIPAs Table I~

R/hah'hM EIIPAa PNe noes ooenwns 1.2<<ESAMs~a sss oocnccn Ap Teen co Inane oecrrcao

~lrar ecr ~c<<s Achroo SSSISEO char OEtcEIIAI.

tavrO drecnaa s I OCyrI as casse ro ace roaaroary as craccrcaA PP Teen co case oooo nce reacorg ~

W revarnrO EPPAs EICEROE~ have seen rroC

~op Ip Iaae breaco Is oreserrc, TIIEIIEVACUATE

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<u Table 14 <<EOAAIS Coasarcert YES I~ Ifeeacaes Ooae ran a I ncrrrroy 0

NO Hrs rrno cvacsorr cnrngas lry hao <<roe PAIcs wore lan nosey YES Use this formula if release has a ground AND elevated source:

Ground Release Rate (Ci/s) Elevated Release Rate (Ci/s)

IF ) 1, A General Table 1.1 Ground Release Rata Table 1.1 Elevated release rate Emergency Exists (Ci/s) (Ci/s)

Harch 1998 Page 8 EPIP-EPP-08 Rev 08

ATTACHNENT 1: INITIAL DOSE ASSESSMENT AND PROTECTIVE ACTIONS Sheet 2 of 4 TABLE 1.1 - GENERAL EMERGENCY RELEASE RATES

,;. -'round. Release',(Cj/s),,'.",,<<;:,';,,:,> ElevatedlRefease, (Ci/s).

Wind Speed Stability Class Wind Speed Stability Class (mi/h) B/C EIF/G (mi/h) B/C D E/F/G 0-3 1333 213 119 38 0-3 2041 1124 3030 769 4-6 3226 286 143 48 4-6 3703 909 769 769 7-9 5556 526 250 83 7-9 5882 1515 1075 1250 10-1 3 7692 769 357 117 10-13 7692 2083 1388 1724 14-1 7 10753 1075 500 164 14-1 7 11494 2857 1818 2273 18-21 13514 1389 667 213 18-21 14286 3704 2273 2778

>21 16393 1667 833 256 >21 17241 4348 2632 3226 TABLE 1.2 - AFFECTED ERPAs Lake: Breeze>Adjusted

'.-;. Wind Direction;From' . 5'Mile:Radius.. 10.Mile Radius (5 Mile, Radius).

14 to 'l4, 2 toi .  ;<< 14, 4,7 234 to 240 14, 15, 29 241 to-254:- '::,4;...:7. 14, 15;, 29 255 to 262 4,7 14 15. 16. 17, 29 263 to"2/8'-.-'<':"-: " '-<< -':. '.4 ".7;"9 '>c 8, 1416, 16, 17, 29 279 to 292 4, 5, 7, 9 8, 14, 15, 16, 17, 18, 29 10 tor305.--.. "': .'-'1',a,5;> 7"; 9;"...10'1 ,8, 1'4, 16, 16, 1.7, 18; 29 4, 5, 7, 9, 1 ,14,1,16,1,1,'l

2. -

4,6, 7.;- , 14, 15, 1, 17, 18, 1 333 to 340 4,6,9,10,11 8,15,16,17,18,19,20,21,25 6,7,12 341 to ,,4;.5, 9 10, 1'1 8, 17, 18, 19, 20, 21, 24, 25, 6; 7, 12 349'50 to 356 5,6,9, 10, ll 8, 13, 18, 19, 20, 21, 22, 24, 25, 4,7 12 357 to 12, .,- ~ -, << ",,~ .6;,6;..;9, 10, 1,1, 13, 18; 19, 20, 21, 22, 23, 24; 25,, .4 12 1 to 6,,1,11 1,1,1,,21,22,,

12 4, 5, 4, (21 to 51,- ;5,.6,"'.10,;.1:1 13, 19,. 20, 21, 22, 23; 24,, 25, 28, '9 12 52 to 56 5,6, 11 13, 19, 20, 21, 22, 23, 24, 28, 12 10 57 to'6'l,:,:-;, 6;.. 6',, l l. 13, 19, 21, 22, 23, 24, 28, 1,2 "lO 62 to 70 6, 11 13, 19, 21, 22, 23, 24, 28, 12 10

,71 to 89" 13, 21, 22, 23, 24, 28, 12 to 95 5, 11, 1 116 to 146 28 1 47to 2 1 3c.

'>j $ 86>'; 'v>'...>>~ '" .~: p .,z..;<>'w;",,>,, w.,~u r 28,,29 TABLE 1.3 - EPA 400 Protective Action Guidelines (EPA PAGs)

'P',i '.;,~iPAR.::., "," -;~'--', ',<<,." ."~';c '..'~:" 4+,::.-; .'TEDE (rem)" CDE, (rem)

Evacuate >5 March 1998 Page 9 EPIP-EPP-08 Rev 08

ATTACHNENT 1: INITIAL OOSE ASSESSMENT.AND PROTECTIVE ACTIONS Sheet 3 of 4 FiGURE 1A - Site Boundary Map r

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ATTACHMENT 1: INITIAL DOSE ASSESSMENT AND PROTECTIVE ACTIONS Sheet 4 of 4.

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25 LEGEND 1 ERPA Number EPPA Papulatian Mar ch 1998 Page ll EPIP-EPP-08 Rev 08

ATTACHMENT 2: USE OF THE EDANS COMPUTER Sheet 1 of 2 For the Control Room, see Section 1.0 below. For the EOF, see Section 2.0 below.

1.0 CONTROL ROOM EDANS Turn the system on: Turn on the power to the EDAMS computer, monitor, and printer. After the computer boots, the EOANS log-in menu will appear (Figure 2.1), and the default log-in selection will be highlighted.

1.2 Com uter roblems

a. If at any time problems are experienced with the computer, depress the eject button on the front of the computer. This will eject the laptop computer. Continue this procedure with the laptop.

b. If the laptop should fail, have Chemistry Tech from the unaffected Unit go to the unaffected Control Room and bring the EDANS laptop back to the affected Control Room and continue with this procedure.

NOTE: In this case, meteorological data will have to be obtained manually.

2.0 EOF EDANS 2.1 Turn the system on: Turn on the power to the EDANS computer, monitor, and printer. After the computer boots, the EDANS log-in menu will appear (Figure 2. 1), and the default log-in selection will be highlighted.

2.2 Com uter roblems.

~ If at any time problems are experienced with the computer, use the duplicate EDAMS computer in the EOF.

FiGURE 2.1 - EDAMS Main Menu g':4""')<>>"<<~'~:-<'>'::<?'i,""EDAM$:.".LGGIN.'MENU<'~<',r'<<:~:> '4 "'.

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>'?SELECT'<'ONElOF,.'<THE~?FOL'L'OWING:>OPTIONS::~g<'.",.'"~),"'.,';.;:;:,',';;-"',.g.-'.,'~<,'N~~gL'og,'on;'toiNiiie:Mile!

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j~~j~g~Fi~d--",,': "Dial-"up'.Nin'e!Mile. Met data".,',.'. "p?.'"-,,'i>",.::":~"i':

March 1998 Page 12 EPIP-EPP-08 Rev 08

ATTACHMENT 2: USE OF THE EOANS COMPUTER k

Sheet 2 of 2 3.0 SYSTEM USE A brief,description of each EDAHS Hain selections. is found below.

a. Release Rate Calculation. Computerized worksheets for all methods of release rates. This option is plant specific.

b. Emergency Meteorology Report. A single screen that contains the latest fifteen minute meteorology data.

c~ Dose Assessment Model: RADDOSE IV. Emergency dose assessment model. All plant specific release rate methods are available, and the latest fifteen minute meteorology data is automatically inputted to the model.

d. Protective Action Recommendations. Determine PARs for each Emergency Response Planning Area (ERPA)

e. Other Selections:

~ Miscellaneous Meteorological Reports. (For Het Advisor use)

Contains the Emergency Heteorology Report and historical meteorological data for all three towers.

~ RASCAL. RASCAL is the NRCs dose assessment model. The most recent version is included in EDAHS for general use.

~ Field Team Calculations. Provides a method for calculating results from downwind survey teams.

f. Logoff a.o EDANS DOSE NODEL LIMITATIONS a.1 A calculational limitation of the dose assessment model occurs when an extreme wind (direction) shift takes place. The model may not calculate doses in sectors that the plume skips over entirely within a single 15 minute calculation step.

4.2 EDAHS only allows the operation of one application at a time.

4.3 Dose rates and deposition rates reported by the model are the maximum for the sector, not necessarily the dose rate or deposition rate at the, center of the sector. This avoids the situation of a narrow (stable) plume slipping between receptor points and being missed.

4,4 Deposition data reported is not intended for an environmental evaluation; its intent is to indicate areas of potentially high ground level concentrations.

Harch 1998 Page 13 EPIP-EPP-08 Rev 08

ATTACHMENT 3: METEOROLOGICAL DATA AC UISITION Sheet 1 of 8 1.0 OBTAINING METEOROLOGICAL DATA The methods of obtaining meteorological data are listed below in the order that they should be used.

~ EDAMS (see Section 3.0 of this Attachment)

~ Strip Chart Recorder (see Section 4.0 of this Attachment)

~ Manual input from alternate sources (see Section 5.0 of this Attachment) 2.0 USE OF METEOROLOGICAL DATA: GENERAL CONDITIONS NOTE: Wind speed measurements at both the main and backup towers may on occasion be less than actual observed winds. When using the main tower winds and the wind direction is between 0'nd 100'r when using the backup tower and the wind direction is between 220'nd 270 , caution should be exercised when estimating plume arrival time, its likely that the plume will arrive sooner than what the wind speed would indicate. Additionally the actual dose may be less than forecast by EDAMS.

2.1 Hierarchy of NMP meteorological data sources is shown in Table 3. 1 below.

NOTE: Heights of meteorological instrumentation is approximate.

2.2 If substitute data is to be used, consult the Meteorological Advisor (if available).

2.3 If using the 90'igma theta as a level release, substitute stability for either the following corrections should be elevated or ground made:

If the winds blow from 232'o becomes 246'nd 270'o 281', add one stability class such that a D an E, C becomes a D and so on. If the original class is a G then no changes should be made.

If the winds blow from 247'o 269 , twobecomes stability classes should.

and so on. If be added such that a D becomes an F, C an E the original class is either an E, F, or G, the class should become a G.

2.4 If no release is in progress, or release path is unknown, use elevated data (200'ain tower), or substitute as outlined in Table 3.1.

2.5 If using the 30'igma theta and the wind is blowing from 035'o 076 substitute to the next source per Table 3.1.

source (Sodar, other towers, 2.6 The Meteorological Advisor may use any characterization tables, etc.) or skills of the trade to satisfy the need for meteorological data.

March 1998 Page 14 EP IP-EPP-08 Rev 08 Pl ti

ATTACHMENT 3: METEOROLOGICAl DATA AC UISITION Sheet 2 of 8 TABLE 3. 1: HIERARCHY FOR USE OF NMP METEOROLOGICAL DATA SOURCES

<<j<<~'.."'~g,"'~L'Q'<<<<'<<~3iIi~~>'LYLY, cNSources~j>.~4<'.::.:. <<<<'-.<<<<p)>m~g~>~~'g~<<~eg":n

i4'<<,'<<<<1%'s's.~4 '.":"."<<~~a )<<5v 4.'..a."..~~~><<<<'<<<PCS'"~w<<<<'pgxyp,<g5a 9<<<<: wg5p:+<<g(

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100'ain Substitutes 90'AF Primary 200'T 30'igma Theta 30'igma Theta 200'T Substitutes 33'nland Sigma Theta Primary equals Primary Sackup for Unit 2 (2) lf using 30'igma theta, AN0 the wind is from 035'o 076', THEN substitute the next source of data in accordance Mith Table 3.1 of this attachment.

2.7 Refer to Figure 3.2 to determine if lake breeze is a possibility (EOF only).

2.8 Refer to Figure 3.3 to determine if land breeze is a possibility (EOF only).

.0 EDAMS To obtain meteorological data for the Notification Fact Sheet Part 1, select Emergency Meteorological Report from the EDAMS main menu.

Select the data as discussed in Table 3. 1.

3.2 Hit the F5 Key to print the data.

4.0 STRIP CHART RECORDER CAUTION Do not use the LED readouts associated with the strip chart recorders.

NOTE: Use this data only if the method described in Section 3.0 of this Attachment is unavailable.

4.1 Strip chart meteorological data can be found in both of the Control Rooms, and in the TSC. Utilize Table 3. 1 to determine source of data.

NOTE: bT cannot be obtained in the TSC. Utilize f78 in determining stability.

,.z Figures temperature, hT, 3.4'nd 3.5 show. sample strip chart f78, wind speed and direction data.

traces showing ambient air Harch 1998 Page 15 EPIP-EPP-08 Rev 08

ATTACHMENT 3: METEOROLOGICAL 0 TA AC UISITION Sheet 3 of 8 I

4.3 Observe the values of the vertical temperature difference, 6T, from the primary meteorological tower over the last 15 minute period. The pr eferred reading is the 30'-200'elta temperature reading, for an elevated (stack) release.

4.4 Compare the values of lD to the Stability Classification Chart (Table 3.6) and select the appropriate stability class and record.

4.5 If values of hT are not available, then observe the values of cr8, directly from the primary or backup meteorological tower recorders, over the last 15 minute period.

4.6 Compare these values of o8 to Table 3.6. Using the chart, select the appropriate stability class and record.

4;7 If both data are available, use the hT at 30'-200'levation for elevated releases: use a8 at the 30'levation for ground (vent) releases.

4.8 If values for hT and e8 are not available, then observe the wind direction trace over the last 15 minute period. Determine u8 by dividing the horizontal deviation of the wind direction trace over the last 15 minutes by 6. To make reading of the strip charts easier, you may want to advance the chart.

5.0 MANUAL INPUT FROM ALTERNATE SOURCES NOTE: Use this data only if the methods described in Section 3 and 4 of Attachment 3 are unavailable.

CAUTION Data obtained by the methods described below will not be site-specific and will likely introduce errors into dose assessments. The Meteorological advisor shall be consulted regarding the use of all substitute data. If the Meteorological advisor is not available, use the data as obtained.

5.1 National Weather Service a ~ Telephone the National Weather Service (NWS) in Buffalo at 800-462-7751 or 716-565-9001.

4

b. Request the current wind speed and direction, stability class and temperature.

c Use this data as follows:

1. Wind speed (NWS) elevated and ground wind speed

2. Wind Direction (NWS) - elevated. and ground wind direction

3. Stability Class (NWS) stability class March 1998 Page 16 EPIP-EPP-08 Rev 08

ATTACHMENT 3: METEOROLOGICAL DATA AC UISITION Sheet 4 of 8 5.1. c (Cont)

4. Temperature (NWS) ambient temperature Other Sources

a. Other sources of meteorological data that may be utilized are as follows:

1. SODAR

2. Other (non-NWS) meteor ology towers

3. Commercial weather services FIGURE 3.2 Lake Breeze/On-Shore Flow and Fumigation Flow Chart Obtain meteorological data per Section 1.0 of this Attachment.

2. Obtain lake intake water temperature from Unit 1 or 2 proces's computer or from Control Rooms.

3. Follow the flow chart answering the appropriate questions.

Is a release possible NO between dawn and dusk?

YES the intake water NO perature less than e air tern erature?

YES Is Mind Speed less than NO 22 mph?

270'horough YES Is Mind Direction out of the wind sector from NO Is Mind Speed less NO 360'o than 10 mph?

90'?

YES YES

,Pot'ent i'.al"

,;for.;:i<On-:,Shore..'..', . lPotenti'al$ .;,for<". Lake.o~;,;.,-"'.,"

"'St'op";":A'naly's:is,:.,',"::

j.>@ <.?".~".<egg r,.'~..$ /3~<. <gb)g~'<<,i~~g'"j; Use "Lake Breeze Use "Lake Breeze No Adjusted ERPAs" for Adjusted ERPAs" for modifications PARs (Table 1.2 or PARs (Table 1.2 or to PAR EDANS) EDAMS) required

• Also for sudden shift in wind direction 245 note that there through to 65'fis thea potential a lake breeze has not already formed.

to March 1998 Page 17 EP I P-EPP-08 Rev 08

ATTACHNfNT 3: NfTfOROLOGICAL DATA AC UISITION Sheet 5 of 8 FlGURE 3.3 LAND BREEZE FLOW CHART

1. Obtain Meteorological Data.

2. Obtain lake temperature.

3. Follow the flow chart answering the appropriate questions.

t is a release possible between dusk and 10 am.?

YES NO Are sky conditions neafty clear (Le., little cr no c.'oucs)?

YES NO ls the take water temperature greater than the air temperature?

YES NO Is 200 tt. Wind Speed less than 17 mph?

YES Is 200 tL Wind Dlrectfon frcm 090 NO The potential for Land Breeze stiff btrough f80'o 270'? exists. Contktue to monitor.

YES NO Stop Analysis Is 200 ft delta T stability F or G?

A Land Breeze ts not expected YES Land Breeze may already exist and 200 tL Wind Direction may not be representative ot stack height winds.

"NOTE: There is a potential for a shift in Wind Direction to 090'hrough 180'o 270't the weather tower.

Harch 1998 Page 18 EPIP-EPP-08 Rev 08

ATTACHMENT 3: NETEOROLOGICAL DATA AC UISITIOM Page 6 of 8 FIGURE 3.4 SAMPLE AIR TEMPERATURE AND STABILITY CLASS TRACE - CONTROL ROOM

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I March 1998 Page 19 EP IP-EPP-08 Rev 08

ATTACHMENT 3: METEOROLOGICAL DATA AC UISITION Sheet 7 of 8 FIGURE 3.5 SAMPLE WIND SPEED AND WIND DIRECTION TRACE - CONTROL ROOM/TSC 54@ ottiind g~n t~

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ATTACHMENT 3: METEOROLOGICAL OATA AC UISITION Sheet 8 of 8 TABLE 3.6 - STABILlTY CLASSlFlCATION CHART TEMP CHANGE r/s OEGREES TEMP CHANGE STABILITY TURBULENCE PASQUILL WITH HEIGHT, RANGE OF WITH HEIGHT, CLASSIFICATION CLASS CAT. oF/70ftre VALUES'" oF/1 68ft'"

Extremely unstable 4TIdz < -0.73 22.5 <<Jff < 4T/4Z < ~ 1.75 Moderately unstable 4.73 < 4T/4Z < -0.65 17.5 <<rf) < 22.5 ~ 1.75 < 4T/dZ < ~ 1.57 Slightly unstable <.65 < 4T/4Z < <.58 12.5 <<rff < 17.5 ~ 1.57 < 4T/dZ < -1.38 Neutral -0.58 < hT/4Z < .0.19 7.5 < off < 12.5 -1.38 < 4T/dZ < 4.46 Slightly Stable &.19 < 4T/4Z < 0.58 3.8 <<r<) < 7.5 -0.46 < dTldZ < 1.38 Moderately Stable 0.58 < 4T/4Z < 1.53 2.1 <<rff < 3.8 1.38 < dT/4Z < 3.69 Extremely Stable f) 1.53 < 4T/hZ <rf) < 2.1 3.69 < 4T/4Z Adjusted to correspond to the 4T measured batN<aan the 30.foot and 100.loot levels.

l2) Nota on symbol convention 3.8 <<rff<7.5 means that ctf is greater than or equal to 3.8 degrees but less than 7.5 degrees.

l3) Adjusted to correspond to the 4T measured between tha 30.foot and 200.foot levels.

ATMOSPHERI TABILITYCHARACTERIZATION A. (I) Mid-afternoon only, with clear skies or skies with very few thin clouds; late spring to early fall, winds usually are below 6 miles per hour.

Late morning to mid-afternoon only, with clear or partly cloudy skies; mid spring to mid-fall, winds are usually below 9 miles per hour.

C. (II) Late morning to late afternoon only, with partly cloudy skies; spring through fall, winds are usually below 11 miles per hour.

D. {III) All daytime, with overcast or partly cloudy skies or early morning and late afternoon with clear or partly cloudy skies, all night time with overcast skies or partly cloudy year around, winds are moderate to high (greater than 6 miles per hour).

E. (IV) Typically night time only, with thin overcast or partly cloudy skies, all year around, winds less than 10 miles per hour.

F. (IV) Typically night time only, with clear to partly cloudy skies, all year around, winds less than 7 miles per hour.

G. {IV) Typically night time only, with clear skies or very few thin clouds all year around, winds less than 5 miles per hour.

Hatch 1998 Page 21 EPIP-EPP-08 Rev 08

ATTACHMENT 4: RELEASE RATE DETERMINATION Sheet 1 of 6 1.0 METHOD Access the EDAMS Computer using Attachment 2 of this procedure.

IF Unit 1 was selected, go to Section 2.0 of this Attachment.

IF Unit 2 was selected, go to Section 3.0 of this Attachment.

2.0 UNIT 1 METHODS 2.1 OGESMS a ~ Select monitor (7, 8, 10a or 10b)

NOTE: Monitor 7 indicator 112-07A Monitor 8 = indicator 112-08A Monitor 10a = indicator RNlOA Monitor 10b = indicator RN10B

b. Enter time that reading was obtained (using 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> format) c~ Enter monitor reading (cpm for monitors 7 or 8, cps for monitors 10a or 10b). Use J panel readings or the following computer points:

monitor 7, use E334 monitor 8, use E335 monitor 10a, use E488 monitor 10b, use E489

d. Enter calibration factor. If unavailable, use default values below:

~ 4.4E-8 for 7 or 8

~ 4.4E-7 for 10a or 10b

e. Enter Stack Flow (kcfm). Use computer point C320 or calculate from Table 4. 1.

Hit the "F9" key.

g Print results.

2.2 RAGEMS a ~ Enter the time that the reading was obtained (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> format).

Enter the monitor reading (cps). Use J panel reading or computer point C321.

c Enter calibration factor (use posted value).

d. Enter dilution factor as follows:

~ 1

= IE3 if if 6 liter chamber is used 30 cc chamber- is used 2E5 if 30 cc chamber plus first stage dilution is used. Use Total Dilution Ratio (TDR) x1000 as the dilution factor, if TDR is known.

4E7 if 30 cc chamber plus first and second stage dilution is used. Use TDR xl000 as the dilution factor, if TDR is, known.

March 1998 Page 22 EPIP-EPP-08 Rev 08

ATTACHMENT 4:, RELEASE RATE DETERMINATION Sheet 2 of 6 2.2 (Cont)

e. Enter Total Stack Flow (kcfm). Use computer point C320 or calculate from Table 4. 1.

Hit the "F9" key.

Print results.

2.3 Stack Teletector a~ Enter the time that the reading was obtained (24-hour format).

b. Enter the monitor reading (mrem/hr).

C. Enter the calibration factor. If unavailable, use default value of 0.5.

d. Enter Total Stack Flow (kcfm). Use computer point C320 or calculate from Table 4.1.

e. Hit the "F9" key.

Print the results.

2.4 Grab Sam le Noble Gas CAUTION In using grab samples to determine release rate, the results may be invalid if significant changes in source terms have occurred since the sample was taken.

a ~ Enter the time that the reading was obtained (24-hour format).

b. Enter total Noble Gas concentration (pCi/cc).

C. Enter Total Stack Flow (kcfm). Use computer point C320 or calculate from Table 4.1.

d. Hi t the "F9" Key.

e. Print the results.

2.5 Back Calculation Use back calculation of downwind survey team data to determine release rate when no other method is available, AND to verify calculated release rates.

'a ~ Enter the time that the reading was obtained (24-hour format).

Enter the wind speed (mi/hr).. Use the method described in Attachment 3:

c'. Enter "E" for elevated/stack or "G" for ground/vent release.

March 1998 Page 23 EP IP-EPP-08 Rev 08

ATTACHMENT 4: RELEASE RATE DETERMINATION Sheet 3 of 6 2.5 (Cont)

d. Enter the stability class (A-G).

Enter the three foot closed window reading from the ion chamber (mrem/hr). If readings are in CPM, then convert using 3500 CPM 1 mrem/hr.

Enter the downwind distance that the above reading was obtained.

g, Hit the "F9" key.

h. Print the results.

2.6 FSAR NOTE: Input from the Control Room or TSC staff is necessary to select the FSAR accident type that most closely describes the conditions being experienced.

a~ Select the accident being experienced or projected (Use Attachment 5, Table 5. 1) .

b. Print results.

2.7 Containment Hi h Ran e Monitor NOTE: This method is only valid if the monitor is able to "see" the release.

Therefore, consult Operations personnel on the validity of monitor readings.

a~ Enter the monitor ID or number.

b. Enter the time that the reading was obtained (24-hour format).

c ~ Enter the date that the reading was obtained.

d. Enter the time of reactor shutdown (24-hour format).

e. Enter the date that the reactor was shutdown.

Enter the monitor reading (rem/hr). Use computer point E467 or E468.

g. Enter the expected flow rate (kcfm) to the environment. Consult with Operations personnel if needed.

h. Hit the "F9" key.

Print results.

2.8 For 1 i uid releases consult Nl-CSP-M204 March 1998 Page 24 EP IP-EPP-08 .'.

Rev 08

II ATTACHNENT 4: RELEASE RATE DETERMINATION Sheet 4 of 6 3.0 UNIT 2 NETHODS GENS

a. Enter the time that the reading was obtained (24-hour format).

b. Enter "S" if this is a stack reading or "V" if it is a vent reading.

C. Enter monitor reading {pCi/s). Use GEMS readings from SPOS display or the 882 panel. If offscale, use GEMS computer.

d. Hit the "F9" key.

e. Print results.

3.2 Grab Sam le Noble Gas CAUTION In using grab samples to determine release rate, the results may be invalid if significant changes in source terms have occurred since sample was taken.

a. Enter the time that the reading was obtained (24-hour format)

b. Enter total Noble Gas reading (pCi/cc).

c. Enter total stack or vent flow (kcfm). Calculate from Figure 4.2 or '4.3.

d. Hit the "F9" Key.

e. Print the results.

3.3 Back Calculation Use Section 2.5 of this Attachment.

3.4 USAR Use Section 2.6 of this Attachment.

3.5 Containment Hi h Ran e Nonitor Use Section 2.7 of this Attachment. Monitor readings are available on the ORMS system (RMSla,b,c or d), the SPOS display or the 880 panel.

3.6 For li uid releases consult N2-CSP-LWS-N203 i.,

March 1998 Page 25 EPIP-EPP-08'ev 08

ATTACHMENT 4: RELEASE RATE DETERMINATION Sheet 5 of 6 TABLE 4.1 FLOW RATES CORRESPONDING TO FAN CONFIGURATIONS FOR UNIT 1 Drywall Vent, Purge, and Fill Line (10.00 KCFM) KCFM Turbine Building High Speed Fans (170.00 KCFM) KCFM Turbine Building Low Speed Fans (120.00 KCFM) KCFM Reactor Building High Speed Fans (70.00 KCFM) KCFM Reactor Building Low Spood Fans (35.00 KCFM) KCFM Waste Building (8.00 KCFM) KCFM Waste Building Extension (5.30 KCFM) KCFM Offgas Building (6.00 KCFM) KCFM Reactor Building Emergency Vent. (1.60 KCFM) KCFM RSSB Extension (10.25 KCFM) KCFM

+"":-"'.-'":<""":~"""~> "Fg>"<>"':~+-"'~Total>Stack'Flo'w~ '.i""-~">': KCFM

">"~)<'YH'>&A<8%A sM:Ai.

TABLE 4.2 FlOW RATES CORRESPONDING TO FAN CONFIGURATIONS FOR UNIT 2 STACK CST Room 1 Stack Turbine Building Turbine Building Standby Gas Nominal Nominal Fan (2200 Substructure 1 Fan (40,000 2 Fans (80,000 Treatment SCFM cm /sec SCFM) 1 Fan (1400 SCFM) SCFM) (4,000 SCFM) 5CFIIA) 4,000 1.89 E6 40,000 1.89 E7 44,035 2.08 E7 80,000 3.78 E7 84,000 3.96 E7 1400 6.61 E5 2200 1.04 E6 March 1998 Page 26 EPIP-EPP-08 Rev 08

ATTACHMENT 4: RELEASE RATE DETERMINATION Sheet 6 of 6 TABLE 4.3 LOW RATES CORRESPONDING TO FAN CONFIGURATIONS FOR UNIT 2 VENT 250adwaste Red waste Radwaste Building Radwaste on Rm 1 Fan Uner 1 Fsn Tanks 1 Fan 1 Fan (47,800 Building Aux Boiler Refueling Floor Refueling Floor 300 SCFM) (800 SCFM) (4910 SCFM) SCFM) 2 Fans (95,600 (23,000 Above (70,000 Below (70,000 Nominal Nominal SCFM) SCFM) SCFM) SCFM) SCFM cm~/sec 47,800 2.256 E7

~Pw4+~"~-"'"L+~ 95,600 4.512 E7 70,800 3.341 E7 118,600 5.597 E7 Pj~@~<~~graggggg+p~

@%8? c'<4p xhkki';@AN?c 187.800 8.864 E7 g.":O'Bdk;%4 235,600 1.112 E8

'kY>~44~~~~448 210,800 9.948 E7

~Y4LxÃ44 258.600 1.22 E8 P,".g'-;>mgg 4910 2.317 E6 800 3.775 E5 3300 1.557 E6 March 1998 Page 27 EPIP-EPP-08 Rev 08

ATTACHMENT 5: REFINED DOSE ASSESSMENT AND PROTECTIYE ACTIONS Sheet 1 of 5 1.0 DOSE ASSESSMENT General Considerations 1.1. 1 The dose assessment program is called RADDOSE.

1.1.2 Meteorological data is automatically sent to RADDOSE by the Meteorological Monitoring System (MMS). The user can use this data or manually input data.

1.1.3 Source term and release rate determination is identical to that described in Attachment 4.

1.2 Dose- Assessment Procedure NOTE: The dose assessment model has many capabilities beyond those used in this procedure. Use the "EDAMS Operators Hanual" (available in the EOF) for further reference.

1.2.1 Log on to EDAHS computer using Attachment 2.

1.2.2 Select "Dose Assessment Model: RADDOSE IV" from the EDAMS main menu.

1.2.3 Utilize "EDAHS Data Entry Form", Figure 5.2, or equivalent.

1.2.4 Select the affected unit.

.1.2.5 Select "Begin New Incident" at the "Start-UP Henu" screen.

Hit the "delete" key if prompted.

1.2.6 Enter the following at the Accident Scenario Definition screen:

a. Trip Date. This is the date that the reactor scrammed or was manually tripped. IF the reactor is not shut down, enter tomorrow's date.

b. Trip Time (24-hour format). This is the time that the reactor scrammed or was manually tripped.

c. Release Date. This is the date that the release to the atmosphere began, or is projected to begin.

d. Release Time (24-hour format). This is the time that release to atmosphere began or is projected to begin.

e. Enter the lake temperature (deg F). If unknown, hit "Enter" and historical data will be entered.

f. Enter the initials of the user (two or three initials).

g. Hit "F9" to accept all entries or "Esc" to back-up and correct.

March, 1998.- Page 28 EPIP-EPP-08 Rev 08.

ATTACHMENT 5: REFINED DOSE ASSESSMENT AND PROTECTIVE ACTIONS Sheet 2 of 5 1.2.7 Select "Enter Source Term Data" from the EDAHS main menu.

NOTES: 1. Use Attachment 4 to obtain the information needed to complete this section.

2. The pr'eferred source of release rate data is the actual isotopic distribution, if available.

a. Select "Accident Type" by hitting the "F2" key, and choosing the accident that most suits'-the current conditions. Use Table 5. 1 in making the choice.

b. Select "Y" for elevated" releases OR "N" for ground releases when asked, Is, this release Elevated'".

NOTE: "Elevated" releases are releases from the stack. "Ground" releases are 'from any other release point.

c. Select the "Method" used to determine the release rate by hitting the "F2" key and selecting. Enter the "Flowrate" and "Monitor Reading" if required.

d. Select the Iodine release rate "Method" by hitting the "F2" key. Enter the "Monitor Reading" and "Release Rate" if required.

e. Up to three Accident Types (and therefore three release paths) can be entered. To enter additional release paths, repeat Steps a - d above. When all applicable accident types have been entered, proceed to the next step.

f. Upon completion of this screen, hit. the "F9" key to accept or "Esc" to back-up.

1.2.8 The user will be queried only'or the meteorological data required. Enter meteorological data as required:

a. When queried for "Enter/Edit Meteorological Data", hit "Enter".

b. If the HHS is available, the data will be automatically displayed for the current time step. Hit "F9" to accept, or a

A

~ Hit "F4" to update'he screen

~ Hit " Ins" to insert data C. If the is unavailable, enter met data obtained from HMS alternate sources, as outlined in Attachment 3 of this procedure.

March 1998 Page 29 EPIP-EPP-08 Rev 08

ATTACHMENT 5: REFINED DOSE ASSESSMENT AND PROTECTIVE ACTIONS Sheet 3 of 5 1.2.9 Select "Perform Calculations" from the EDAHS main menu.

CAUTION Any calculations performed on actual data shall be verified. The ODAH may act as the checker for calculations performed by the Rad Assessment Staff.

a. The map of the 10 mile Emergency Planning Zone (EPZ) will appear with centerline dose rates when the calculation is complete.

b. Hit any key to go to the output menu.

c. Select "Continue Calculations" from the output menu.

d. Select "Perform Forecast" from the RADDOSE main menu.

e. Verify meteorology and source term data as required.

f. Enter "Forecast Period" (i.e. - release duration). Use 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> as a default value.

g. Hit the "F9" key.

h. After the forecast map appears hit any key to go to the output menu.

Select "Go to Report Henu".

J. Select "Print Complete 10-Mile ERPA Hap".

k. Select "Print Complete Dose/Dose Rate Report".

Attach results of Step 1.2.9.j and k to EDAHS Data Entry Form, Attachment 5.2 or equival,ent.

m. Verify that any results are supported by radiological and plant conditions. Consider:

~ Core damage

~ Drywell high range monitor readings

~ Effluent monitor readings

~ Inplant radiological conditions

~ Containment hydrogen monitor readings

n. Document the verification of the calculation using the signature lines on Figure 5.2 or equivalent.

Harch 1998 Page 30 EP IP-EPP-08 Rev 08 ILS

ATTACHMENT 5: REFINED DOSE ASSESSNENT AND PROTECTIVE ACTIONS Sheet 4 of 5 2.0 EFINED PROTECTIVE ACTIONS These actions are initiated for the purpose of verifying the adequacy of PARs made using Attachment 1 of this procedure OR to develop PARs using projected doses obtained from Attachment 5, Step 1.2.9 of this

'.2 procedure.

In determining PARs based on dose assessment, carefully consider factors such as release duration and Evacuation Travel Time Estimates (ETTE). (For example, puff releases may yield doses in excess of Protective Action Guidelines for an evacuation, but the plume will pass before an evacuation could be completed). ETTEs are available in the EOF.

NOTE: County and State PARs take many factors into account that NNP procedures do not (i.e. - road conditions, special population needs, evacuation scenarios, and shelter vs evacuation doses). Therefore, differences in PARs may occur. The ODAH must account for differences in PARs, when those differences exist. This can be accomplished via consultation with County and State representatives in the EOF as to the assumptions used in their dose calculations and PAR development.

2.3 Obtain dose projection for each ERPA.

2.3.1 PARs are listed on the 10 mile ERPA map obtained per Attachment 5, Step 1.2.9. j.

2.3.2 The following criteria are used in determining the PAR for each ERPA.

PAR TEDE rem CDE rem Evacuate >5 2.3.3 Record the PAR for each ERPA on the Part 1 Notification Form and give to the CED for approval.

2.3.4 PARs that have been made previously must be accounted for when PARs are revised. For example, if a PAR to evacuate an ERPA was previously made to the State/County and that PAR does not appear on a revised map from 1.2.9.j, that PAR must still be included on the revised recommendation to the State/County.

2.3.5 If projected doses exceed values listed in Attachment 5 Step 2.3.2 for distances greater than 10 miles, PARs shall be made using convenient geographic boundaries (such as townships).

March 1998 Page 31 EPIP-EPP-08 Rev 08

ATTACHNENT 5: REFINED DOSE ASSESSNENT AND PROTECTIVE ACTIONS Sheet 5 of 5 TABLE 5.1 - FSAR/USAR ACCIDENT TYPE

':.,., -'<.',i.-",A'cc(de'nt'Type",::;-.:;:i:-;.-.',";-)~/~I lRel'ease';Rate>,(Ci/s)':,";:Release: Ratej(CIls);;

~".<>~((P~q~:,"<j>%:;? ';,;<.,""c..",'+r?.:;(g i>8e%gV4(> 4Q?ypvg~AN

?'. 'ga qA o.") ~ p~q'ge'v "Y +1'.:?'0'g'?4~ ~' ~y gi bg~i$ 5'ji~ivvgc?qik Unit 1:

DBA Loss of Coolant 5.50E+ 0 4.53E-3 Elevated Control Rod Drop 2.51E+ 1 6.03E-5 Elevated Refueling Accident 3.78E-2 3.84E-5 Elevated Steam Line Break 6.36E+ 0 4.86E+ 1 Ground Loss of Coolant (Realistic) 1.79E-3 1.00E-6 Elevated Unit 2:

DBA Loss of Coolant 1.03E+ 1 2.03E-1 Elevated Control Rod Drop 4.22E-2 4.70E-4 Ground Refueling Accident 1.77E+1 1.65E-1 Ground Steam Line Break 3.64E+ 0 1.22E+2 Ground Rad Gas Waste System Leak 4.06E+ 0 0.00 Ground Instrument Line Failure 0.00 2.17E-2 Ground Fuel Cask Drop 2.06E+ 0 2.68E-3 Ground Loss of Coolant (Realistic) 1.05E-2 2.38E-5 Elevated Parch 1998 Page 32 EP IP-EPP-08 Rev 08

,4 l,

RA:..'.,',.,M':: "'"-", =."';"'".:;..".'." .'j '-:::.":::,';'".-..-'":"" .',"iri'O'i'i'n'-

0"Q~Q".".~:~'~j'"",'- .<~g',"%x', j"'."::.:::';.~) f..'~!;: ?.-, 'tK crm: urvey ccaucn: cmp ete y:

a'te; Re ease Environmental Sampling Data Reioase Lcg Effluent Monitor Data i ~ ~ ~ ~ ~ ~ Assigned Time Assigned Interval Gamma Dose Wind Transit Est Time of Release r '?. Release Time cf Releaso Duration Ul Rato Distance Speed Time Release Rate Rats Mcnitc Monitor Rate of Survey O (mR/hr) (Mi) (mph) (min) tram Site (Ci/soc) (Ci/sec) 'D Roadin System (Ci/soc) Release Timo Location S~~

iy yri

? X?

'?.?

r

?N

??a pg y?

'ctes:

Transit Time (min) ~ (Distance/Wind Speed) x 60 min/hr Est. Time of Release ~ Survey Time - Transit Time March 1998 Page 33 EPIP-EPP-08 Rev 08

ATTACH)lENT 5.2: EDANS DATA ENTRY FOR Cl "What lf" Cl Actual Data (Checker Required l)

Release:

Rx Trip: Date: Time: Release Duration (Hr):

CI Unit 1 Q Unit 2 A ciden T e Relea e Poin Circle On Containment DBA Elev/Grd Flow Rate Control Rod Drop Elev/Grd Refueling Accident Elev/Grd Method Steam Line Break Elev/Grd Loss of Coolant Elev/6rd Monitor Reading Rad Gas Waste System Elev/Grd tnst. Line Failure Elev/Grd iodine Method Fuel Cask Drop Elev/Grd Severe Accident Elev/Grd iodine Monitor Met Data: Cl Automatic C3 Manual (Belowj Lake Temp 'F or Default Elevated Ground Wind 'Speed (mi/hr)

Wind Direction (from - degrees)

Stability (A-G)

Temperature ('F)

Precipitation (in/15 min)

Attach: ~ Map from color printer

~ "Complete Dose/Dose Rate" report Misc:

Calculations Performed By:

Calculations Verified By:

Narch 1998 Page 34 EPEP-EPP-08 Rev 08

NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT NUCLEAR STATION EMERGENCY PLAN IMPLEMENTING PROCEDURE EP IP-EPP-12 REV SION 04 RE-ENTRY PROCEDURE TECHNICAL SPECIFICATION RE(UIRED Approved by: / /f R.~ B.~ Abbott Plant Manager - Unit 1 Date Approved by:

K. A. Dahlberg Plant Manager - Unit 2 Date Effective Date:

PERIODIC REVIEW DUE DATE

I IST OF EFFECTIVE PAGES N

NNN . ~Ch N PacaeNo. Chan e No. ~PN . ~Ch N Coversheet .

ll

~ ~

2 ~ ~ ~ ~

3 \ ~

4 5 0 ~ ~ ~

6 .

January 1998 Page i EP I P-EPP-12 Rev 04

0 T~LEN

~SCTI ON PAG 1.0 PURPOSE . 1

2. 0 RESPONSIBILITIES 1 3.0 PROCEDURE . 1 3.1 Site Re-entry 1 3.2 Site Boundary Survey . . 2 3.3 Site Survey 3 3.4: Station Re-entry Survey ~ ~ ~ 3 3.5 Unaffected Unit Re-entry Survey 4 4.0 DEFINITIONS .

5.0 REFERENCES

AND COMMITMENTS 4 6.0 RECORDS REVIEW AND DISPOSITION 5 ATTACHMENT 1: DOWNWIND/RE-ENTRY SURVEY DATA SHEET 6 January 1998 Page ii EPIP-EPP-12 Rev 04

'I PURPOSE To evaluate hazards to personnel approaching NHPNS from Off Site, and outline the method used to re-enter potential radioactively contaminated areas caused by a radiation emergency at Nine Mile Point Nuclear Station (NHPNS).

2.0 ESPONSIBILITIES 2.1 Site Emer enc Director SED Provides technical and administrative direction to the Radiological Assessment Hanager (RAH).

2.2 Radiolo ical Assessment Mana er RAM 2.2.1 Hanages the radiological monitoring and assessment features of a re-entry operation.

2.2.2 Provides technical and administrative direction to the Environmental Survey/Sample Team Coordinator and staff during re-entry operations.

2.3 Environmental Surve Sam le Team Coordinator Provides technical and administrative direction to re-entry survey teams.

2.4 Re-entr Surve Teams Perform comprehensive re-entry radiation surveys of NHPNS and assessment of radiological problem areas.

3.0 PROCEDURE NOTES: 1. Radiological Survey equipment and supplies are available at the EOF.

2. Re-entry teams shall conduct surveys in pairs, as a minimum.

3. Exposure limits shall be in accordance with EPIP-EPP-15.

3.1 Site Re-entr 3.1.1 Before leaving, the re-entry survey team should ensure they:

a. Obtain a briefing from the Environmental Survey/Sample Team Coordinator or designee on the following as a minimum:

January 1998 Page 1 EP IP-EPP-12 Rev 04

3.1.1. a (Cont)

1. Required monitoring locations.

2. Anticipated radiological conditions both on and off site.

3. Suggested routes.

4. Exposure limits.

5. Radiological conditions warranting mission cancellation or reevaluation. (i.e. back-off dose rates)

6. Review of any available survey data.

b. Obtain necessary survey and sampling equipment.

c. Don protective clothing and dosimetry if appropriate.

d. Verify communications equipment is operational.

3.1.2 Perform the surveys expeditiously for the purpose of:

a. Facilitating the relocation of emergency operations to the Technical Support Center (TSC) or Emergency Operations Facility (EOF), if necessary.

b. Assessing radiological problems possibly encountered by subsequent emergency teams (i.e., damage control.,

search and rescue, etc.).

3.2 Site Boundar Surve 3.2.1 Perform surveys at the site boundary, at assigned survey areas, and as appropriate for conditions encountered. (i.e.

moving surveys traversing the access road may be prudent)

NOTE: Sample analysis may be performed at the Station Laboratory at the New York Power Authority Fitzpatrick Station, or Environmental Laboratory at the Volney EOF, as directed by the Environmental Survey/Sample Team Coordinator.

3.2.2 Record results on Inplant/Downwind/Re-entry Survey Data Sheet (Attachment 1) and transmit the data to the Environmental Survey/Sample Team Coordinator.

January 1998 Page 2 EP IP-EPP-12 Rev 04

3.2.3 Survey Evaluation

'a ~ As survey results are received from the re-entry survey teams, the Environmental Survey/Sample Team Coordinator or designee shall calculate dose rates and airborne concentrations using the methodology and figures described in EPIP-EPP-07, Downwind Radiological Monitoring.

b. The RAM evaluates the results and makes appropriate recommendations to the Site Emergency Director.

3.3 3.3.1 If radiation levels are less than predetermined back-off levels, obtain additional surveys at the following locations:

a. NLC

b. Security West (Unit 1)

c. Security East (Unit 2)

d. New York Power Authority driveway 3.3.2 If radiation levels are encountered greater than pre-determined back-off levels, contact the ESSTC or designee for guidance before proceeding.

3.4 Station Re-entr Surve 3.4.1 The re-entry survey team should:

a. Enter via the Unit 1 Admin Building initially if possible.

b. Inform the Control Room of arrival and readiness to survey the following Emergency Response Facilities:

l. Unit 1 Admin Building, El. 277'.

Technical Support Center (TSC)

3. Operations Support Center (OSC) areas and Admin Building, El. 261'.

Unit 1 Chem Lab (Unit 1 Turbine Bldg., El. 261')

c~ Perform radiological evaluations, record results on Inplant/Downwind/Re-entry Survey Data Sheet (Attachment 1), and transmit to the Environmental Survey/Sample Team Coordinator.

January 1998 Page 3 EP IP-EPP-12 Rev 04

3.4.2 The Environmental Survey/Sample Team Coordinator shall evaluate survey results using methodology described in EPIP-,

EPP-07, Downwind Radiological Honitoring and EPIP-EPP-08, Off-Site Dose Assessment and Protective Action Recommendation.

3.4.3 The RAM shall provide appropriate recommendations to the SED concerning movement of emergency personnel back to the on-site Emergency Response Facilities.

3.5 U a fected Unit Re-entr Surve When an evaluation of areas at the unit experiencing the emergency is .

completed, the RAN should provide direction to the Re-entry Survey Team(s) to evaluate radiological conditions at the unaffected unit.

4.0 DEFINITIONS None 5.0 EFERENCES AND COMMITMENTS 5.1 Technical S ecifications None 5.2 Standards Re ulations and Codes 10CFR20, Standards for Protection Against Radiation 5.3 Policies Pro rams and Procedures 5.3.1 EPIP-EPP-06, Inplant Emergency Surveys 5.3.2 EPIP-EPP-07, Downwind Radiological Monitoring 5.3.3 EPIP-EPP-08, Off-Site Dose Assessment and Protective Action Recommendation 5.3.4 EPIP-EPP-15, Health Physics Procedure 5.3.5 S-RPIP-3.3, Contamination Surveys 5.3.6 S-RPIP-3.4, Airborne Radioactivity Surveys 5.4 Commitments None January 1998 Page 4 EPIP-EPP-12 Rev 04

6.0 ECORDS REVIEW AND DISPOSITION 6.1 The following records generated by this procedure that are the result of an actual emergency shall be maintained by Nuclear Records Management for the Permanent Plant File in accordance with NIP-RMG-Ol:

Attachment 1: INPLANT/DOWNWIND/RE-ENTRY SURVEY DATA SHEET 6.2 The following records generated by this procedure that are not the result of an actual emergency are not required for retention in the Permanent Plant File.

Attachment 1: INPLANT/DOWNMIND/RE-ENTRY SURVEY DATA SHEET January 1998 Page 5 EPIP-EPP-12 Rev 04

ATTACHMENT 1: DOWNWIND RE-ENTRY SURVE DATA SHEET Survey Meter Modal P SR I 0 Downwind Swvay OA OB OC Count Rata Meter Model g SR P 0 Ao-entry Survey Air Sampler Mod lt SRt HI>>SR st Stssd M I M d lt SR>>

Dlrocthns for Survey Teems: report General Area Radiation Data Air Sample Data Survey Team readings In shaded bhcks from loft to right Exposure Data

@O.W;;: Re'idlnrg'",'>>. .,'Sapiple', Co'unt', ~~

@Satan'>> Team Exposure Cumulative Survey g(mrs m/hr.,or,'cpm) n.: ;~@(mnim/hr) r'; Simple".

Start Stop Durathri;  : Rsto': Bkgd'>>

~",-Asti (cpm) >:;"':

Swvoyor's DatelTimo 'ID g'. ,: Members Aecolved Exposwe

,Contict <'qualm': >> : Contact >>itrri .

.'-. Factor .'

>> Time Time ,. (min) '(Cfm) (Cpm) . XM" Initials Initial~ (mrom) (mrem)

Part, Iodine'f, 5R+ qr >:"P I>> s't Plume Tracking Survey Data Moving Swvey Data Near Edge Centerline Fer Edge Tkno Locathn (Street Names, Building. Etc.) Rsdiathn levels (mrem/hr o.w. or cpm)

Odometer Radiation Levels Odometer Radiation Levels Odometer Radiation levels Reading mrom/hr o.w. or cpm Reading mrernlhr o.w. or cpm Road hg mrem/hr o.w.or cpm y 1998 EPIP ~

2 Rev s

NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT NUCLEAR STATION EMERGENCY PLAN IMPLEMENTING PROCEDURE EP I P-EPP-14 REVISION 01 EMERGENCY ACCESS CONTROL TECHNICAL SPECIFICATION RE(UIRED Approved By:

N. L. Rademacher Plant Manager Unit 1 Dat Approved By:

J. T. Conway Pla Manager Unit 2

(

Da e li THIS PROCEDURE PARTIALLY SUPERSEDES S-EPP-14 PERIODIC RPJIEtl, 07/17/97, NO CHAtlGE 12/20/96 Effective Date:

PERIODIC REVIEW DUE DATE "U'-Y 1998

LIST OF EFFECTIVE PAGES

~Pa e No. Chan e No. ~PN . ~hN Coversheet .

1 1

~ ~ ~

2 ~ ~

3 ~ 0 ~ ~

4 ~ ~

'0 5 ~ ~ ~

6 ~ 0 ~

ao 7 0 ~ ~

toN 8 0 ~ ~

9 ~ ~ ~

10 .

I ~ ~

12 .

13 .

14 .

15 .

June 1996 Page i EPIP-EPP-14 Rev Ol

TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE 1

2.0 REFERENCES

AND COMMITMENTS . 1

3. 0 DEFINITIONS 2 4.0 RESPONSIBILITIES . 2

5. 0 PRECAUTIONS 3 6.0 LIMITATIONS AND ACTIONS 4

7. 0 PREREOUI SITES 4

8.0 PROCEDURE 5 8.1 Oswego County Emergency Identification Cards . 5 8.2 . 10 Mile Emergency Planning Zone Access . . . . 8 8.3 8.4 8.5 NMPNS Site=:Aqcess, Protected Area Access Joint News Center Access . 8 9

9 8.6 Emergency Operations Facility (EOF) Access . . 9 8.7 Equipment Access . 10 9.0 ACCEPTANCE CRITERIA ll 10.0 RECORD REVIEW AND DISPOSITION ll ATTACHMENTS . ll FIGURE 1: AUTHORIZED ACCESS CONTROL IDENTIFICATION CARDS . 12 FIGURE 2: EMERGENCY STATUS BOARD . 14 FIGURE 3: AUTHORIZATION FORM FOR ISSUANCE OF THE OSWEGO COUNTY EMERGENCY IDENTIFICATION CARD 15 June 1996 Page ii EPIP-EPP-14 Rev 01

1.0 PURPOSE To provide guidance to personnel assigned to the Emergency Response Organization and Personnel with Oswego County Emergency Identification Cards; on accessing the site and protected area during emergencies.

1.2 To provide guidance to personnel assigned emergency duties when accessing various secured areas or equipment.

2.0 REFERENCES

AND COMHITHENTS 2.1 Technical S ecifications I

None ~

2.2 Standards Re lations and Codes ANSI/ANS-3.3-1982, Security for Nuclear Power Plants

~ ~> ~ >~'>>;e 2.3 Policies ro ams and Procedures

> P 2.3 1 OI-13, Termination of Access to the Protected Area 2.'.3;2 =

EPHP-EPP=92,.Emergency Equipment Inventories and Checklists 2.3-.3 ' EPHP-EPP-Ol,'aintenance of Emergency Preparedness

~ > ~

2.3.4 S-SAP-11.0, Emergency Plan Duties

~ .> ~ ee . ~ . ~ ~

2.3.5 S-SAP-11. 1, Emergency Access Control Point Duties r>

2.3e6 S-SAP-11.2, Emergency Duties EOF Security Officer NDD-EPP, Emergency Preparedness I'.3.7 IA 2.4 Commitments Sequence NCTS Number /umber Descri tion P

3197-14 Change the method for issuing Oswego County Emergency Identification cards during an emergency.

June 1996 Page 1 EPIP-EPP-14 Rev Ol

3. 0 DEFINITIONS 3 1 10 Nile Emer enc Plannin Zone EPZ A designated area approximately 10-miles in radius around NHPNS used to facilitate off-site emergency planning. Access to the 10 Nile EPZ may be controlled during a radiological emergency at NHPNS by police and military control points.

3.2 ccess Control Points J

Checkpoints for incoming traffic to be stopped and identification verified. These points are established by Niagara Mohawk Power

~

Corporation (NHPC) Nuclear Security at the Alert, or higher emergency classification, or as directed by the Site and Corporate Emergency Director. The Access Control Points are predesignated at two

.locations:.-~

3.2. 1 The intersection of County Route 29 and Private Road 3.2.2 The intersection .of 'County Route lA and Private Road 3.3 Oswe o Co nt .Emer enc - Ide tification Cardsn:.-

ID cards issued to individuals assigned emergency duties which allow admittance beyond the Access Control Points, and into the 10 Mile EPZ.

4.0 RESPONSIBILITIES 4.1 Cor orate E er enc Director Recover Hang er 4:1.1 Maintains-overall responsibility for the actual operation and control of emergency response activities.

/

Authorizes -access to Niagara Mohawk Power Corporation "(NHPC) facilities in response to an emergency event.

4.2 Nuclear Securit Director Provides overall direction for security and traffic control at affected NHPC facilities.

4.3 Nuclear Securit Coo dinator 4.3.1 Ensures plant security is maintained and institutes appropriate measures (such as initiation and maintenance of roadblocks) in accordance with the Security and Safeguards Contingency Plan, or as directed by the Site Emergency Director.

0 June 1996 Page 2 E PIP-EPP-14 Rev 01

4.3.2 Provides access and traffic control check points at the Emergency Operations Facility (EOF).

4.3.3 Authorizes the issuance of Oswego County Emergency Identification Cards;. =

4.4 Emer enc Pre aredness Trainin Lead Instructor Haintains a system for authorization, control, use, and collection of Oswego County Emergency Identification Cards.

4.4.2 Issuance of temporary ID -cards at the EOF; located on County Route 176.

4.5 Director Emer enc Pre aredness Authorizes the issuance .of Oswego County Emergency. Identification Cards.

k I' 4.6 e v'sor Technic l Trainin,,~:.-."

Authorizes the issuance of Oswego County Emergency Identification

'fy,', " '

Cards. ~ ~~ 1 f'L 1

~ p> ~ i

" 1 tg $ J df ~ ~ I 4.7 Joint News Center JNC "Director,.

Determines whether or not individuals are permitted access to the Joint, News Center:(JNC).."...". =r. l!

4.8 Emer enc Res onse Perso el ...-. \

4.8.1 Upon assignment of. emergency response .duties, obtain an Oswego County Emergency Identification Card after appropriate training..is .completed.

4.8.2 Display Oswego County Emergency Identification Card for Site access during an emergency.

4.8.3 Shall maintain Oswego County Emergency Identification Card readily available at'all,times when offsite (except as allowed in Section 7.2).

5.0 R CAUTIONS None June 1996 Page 3 EP IP-EPP-14 Rev Ol

6.0 LIMITATIONS AND ACTIONS 6.1 Personnel issued Oswego County Emergency Identification Cards shall not use the card for any other purpose (for example, non-site, non-nuclear related emergency access).

6.2 Permanent Oswego County Emergency Identification Cards are issued routinely to individuals reporting to emergency positions.

6.3 Temporary Oswego County Emergency Identification Cards are obtained by reporting to the EOF.

6.4 Other authorized governmental control cards are supplied to personnel by respective agencies.

7.0 PRERE UISITES When an evacuation is declared:

7.1 Access to the following areas is limited:

7.1.1 The Oswego County Emergency Operations Center 7.1.2 The 10 Mile Emergency Planning Zone 7.1.3 The Joint News Center 7.1.4 Nine Mile Point Nuclear Station (NHPNS)

7. 1.5 NHPNS Protected Area

7. 1.6 The Emergency Response Facilities 7.2 An individual requiring access shall provide one of the following to Nuclear Security personnel, law enforcement, or. military officials:

7.2.1 An Oswego County Emergency Identification card; +0 7.2.2 Another authorized government control identification card (for example, a New York State Police ID Card, or Federal Emergency Management Agency (FEMA) ID Card). Refer to Authorized Access Control Identification Cards (Figure 1).

June 1996 Page 4 EPIP-EPP-14 Rev 01

8.0 PROCEDURE 8.1 Oswe o Count Emer enc Identification Cards 8.1.1 Authorization, control, and use of Oswego County Emergency Identification Cards

a. During an emergency at the Nine Mile Point Nuclear Station (NMPNS) or James A. Fitzpatrick (JAF) Nuclear Station, an Oswego County Emergency Identification Card allows access:

1. Through military and/or police control points throughout the Oswego County 10 Mile Emergency Planning Zone (EPZ) 2 - Into the Joint News Center (JNC) or NMPNS

3. Into the Protected Area 4.- - Into emergency" response facilities

b. Individuals requiring an Oswego County Emergency Identification Card include:

l. gualified personnel staffing emergency positions

2. NMPC Nuclear. Security personnel:,':

3. Other personnel, as determined by the Director Emergency Preparedness'-

of

'.'ssuance 8.1.2 Permanent ID Cards

a. The following steps are to be completed before issuance of an Oswego County Emergency Identification Card:

1. The individual requesting an ID card should complete Part 1 of the Authorization Form for Issuance of the Oswego County Emergency Identification Card (Figure 3), or equivalent.

2. Upon completion of Part 1, the individual should:

a) If training is required:

1) Forward the form directly to the Nuclear Training Center.

2) Upon completion of emergency response training, ensure the Training Supervisor, or designee, signs Part II and forwards the form to the Director Emergency Preparedness.

June 1996 Page 5 EPIP-EPP-14 Rev 01

8.1.2.a.2 (Cont) b) If training is not required, forward the form directly to the Director Emergency Preparedness.

c) Obtain photographs and an identification number at the location specified by the Director Emergency Preparedness.

d) Return the.,form to the Emergency Preparedness Department for review and disposition.

b. Replacement of the Oswego County Emergency Identification Card due to loss or damage is made by:

II

1. 'ssuing a new identification card number

c. The Oswego County Emergency Identification Card may be

-" collected by Emergency -Preparedness in the -following cases:

l. .,Upon

~ ~ ~

termi.nation of the employee from NHPC

2. When a need no longer exists for the individual to I Er ~ s possess the ID card

8. 1.3 Issuance of temporary. ID.cards=

(NCTS 1) a;-' The site contact should acquire the following

=

information from an individual requiring a temporary Oswego County Emergency Identification Card:

II r ~

1. Name..of individual

2. Agency 3..'ocial-.Security number

4. Location of planned access

5. Purpose for access June 1996 Page 6 EPI P-EPP-14 Rev Ol

8.1.3 (Cont)

(NCTS 1)

b. The site contact should forward the information obtained in Step 8.1.3.a to one of the following for authorization:

1. Corporate Emergency Director/Recovery Manager

2. Nuclear Security Director

3. Nuclear Security Coordinator

4. Director Emergency Preparedness

5. Emergency Preparedness Department staff NOTE: During an emergency, this list may be expanded as required in consultation with Oswego County officials.

C. One of the authorized individuals in Step 8. 1.3.b should:

1

1. Contact. personnel at the EOF, by calling 593-5735.

2. Provide the required information.

d. The individual requiring the Oswego County Emergency Identification Card should report to:

The EOF located on County Route 176 at:

New York Power Authority/Niagara Mohawk Emergency Operations Facility 656A Airport Road (R.R.¹2)

Fulton, N.Y. 13069

e. Upon arrival at the EOF, personnel will request the following information from each individual requiring a temporary Oswego County Emergency Identification Card:

1. Name of individual

2. Agency

3. Social Security number

4. Location of planned access June 1996 Page 7 EPIP-EPP-14 Rev 01

8.1.3 (Cont)

(NCTS 1)

f. If the information provided corresponds with previously authorized information, personnel will issue the temporary Oswego County Emergency Identification Card and provide the ingress route to follow, if necessary.

8.2 10 Nile Emer enc Plannin Zone Access 8.2.1 During a declared emergency, an area within an approximate 10 mile radius from the station may be secured by roadblocks manned by law enforcement or military officials.

8.2.2 Individuals requiring access into the 10 Nile EPZ shall, upon arriving at a roadblock, provide law enforcement officials with the appropriate identification described in Section 7.2 of this procedure.

8.2.3 Individuals without the appropriate identification may be permitted access after obtaining a temporary identification card per Section 8.1.3.

8.2.4 When permitted access, follow instructions provided by law enforcement officials (such as avoiding certain roads, etc.)

8.3 Joint News Center Access 8.3.1 Before entry is permitted, individuals requiring access to the JNC shall display the appropriate identification described in Section 7.2 of this procedure or provide acceptable proof of media/press affiliation.

8.3.2 Individuals without the appropriate identification may be permitted access after obtaining a temporary identification card.

8.3.3 The following individuals may authorize the issuance of a temporary ID to access the JNC:

a. Emergency Director of Public Information (EDPI), or designee; OR

b. JNC Director 8.3.4 When permitted access, following instructions provided by Nuclear Security personnel (such as avoiding specific areas within the building).

June 1996 Page 8 EPIP-EPP-14 Rev 01

8.4 NHPNS Site Access 'I 8.4.1 Upon arrival at site roadblocks, personnel requiring access to the site during an emergency shall provide Security personnel with the appropriate identification described in Section 7.2 of this procedure.

tl ~ ~

8.4.2 If an individual with emergency response functions does not possess appropriate identification, security personnel at the roadblock shall refer the matter to the EOF Security Director for proper resolution.

~ .I-8.4.3 When permitted access, personnel shall follow instructions provided by Security personnel (such as specific routes to follow, etc.); >

8.5 Protected Area ccess 8.5. 1 Nuclear Security personnel will notify incoming personnel of

--;..=-'.thewtatus of the emergency by posting an Emergency Status Board (Figure 2) within the Unit 1 and Unit 2 Security Building entrances.

8.5.2 During an emergency, individuals seeking access to the

~ ~

""Protected>Area should:

~ a';-': Proceed'to"a<Nuclear~Security

/

Building." -

"" ' . b Provide':the appropriate identification as described in

'~Section 7.2 of this procedure to obtain protected area ACAD/ID Card.

) I'.

"-Follow rlbrmaT'ecurity access procedures.

NOTE: Any visitors -present during the implementation of this procedure (for example, a telephone repair person) may be authorized to remain in the emergency I

facility.

I

)

8.5.3 During an Alert or higher emergency classification, only personnel possessing authorized emergency identification should access the Protected Area.

8.6. ~ ~~'-"Emer enc 0 erations Fac'lit EOF Access Personnel requiring access to the Emergency Operations Facility (EOF) should:

8.6.1 Enter the primary access door at the EOF.

8.6.2 Frisk upon entering, using the portal monitor, if required.

June 1996 Page 9 EP IP-EPP-14 Rev 01

8.6.3 Proceed to the registration desk.

8.6.4 NMPC Personnel - display an Oswego County Authorized Access Control ID Card.

8.6.5 Display an Oswego County Authorized Access Control ID Card or another authorized government control ID card.

8.6.6 .,Individuals without an appropriate ID provide sufficient

-cause or permission by one of the following individuals:

a. Nuclear Security Director or Coordinator

b. = Corporate Emergency Director/Recovery Manager

c. Director Emergency Preparedness, or designee 8.7 E ui ment Access 8.7 L. .Personnel .requi.ring-4he use of emergency equipment- should:- --.

a.'btain emergency equipment (for example, radiological,

..resc'u'e, operations supplies, vehicles, etc.) at various fac1lities'-"Snd locations .'on:-site-and off site.: ~

b. Maintain appropriate keys to emergency facilities,

,equipment,"and locations by following the guidance regarding control and inventory of keys outlined in Section 8.7.2 of this procedure.'"Emergency

- ='.7.2 keys used for access shall be maintained and distributed by the Manager Emergency Preparedness, or designee. Changes in personne1'hou1d be promptly reported to-emergency .response personnel so 'keys may 'be re-assigned.

a. A set of emergency keys are:

Maintained in the Nine Mile Point Control Rooms under control of the Station Shift Supervisors (SSS).

2. Made available to responding emergency personnel requiring access to emergency facilities or equipment.

b. Selected personnel assigned from Nine Mile Point Nuclear Station (NMPNS) staff shall be provided a set of emergency keys for ease of access during an emergency situation. A complete list is contained in EPMP-EPP-02, Emergency Equipment Inventories and Checklists.

June 1996 Page 10 EP IP-EPP-14 Rev Ol

8.7.2 (Cont)

c. Several emergency kits contain keys, inventoried quarterly, as part of emergency equipment inventories.

d. Emergency vehicle keys are maintained in the Operations Support Center (OSC) key locker cabinet.

e. Inventories and checklists

1. A quarterly inventory is performed by the Manager Emergency Preparedness, or designee, to ensure availability of emergency keys.

2. A list of emergency keys requiring maintenance is contained in EPMP-EPP-02, Emergency Equipment Inventories and Checklists.

e 9.0 CCEPTANCE CRITERIA None," .

10.0 RECORD REVIEW AND DISPOSITION The following records generated by this procedure as-the 'result of an actual emergency declared at Nine Mile Point, shall be maintained by Nuclear Records Management for the 'f'ermanent Plant File in accordance with NIP-RMG-01:

Figure 3, "Authorization Form for Issuance of 'th> Oswego County Emergency Identification Card" I U ~ t) ~

The- Al'lowing 're'cords generated by this procedure as the result of emergency drills and exercises are not required for retention in the Permanent"PTant Fj3e: *+ '

~ l~ ' t ~ Ir s,,ll Figure 2, "Emergency Status Board" Figure 3, "Authorization Form for Issuance 'of;the .Oswego County Emer g enc Identification Card" STIACHIIENTS Figure 1: Authorized Access Control Identification Cards Figure 2: Emergency Status Board Figure 3: Authorization Form for Issuance of the Oswego County Emergency Identification Card June 1996 Page ll E PIP-EPP-14 Rev 01

FIGURE 1 Page 1 of 2 ACC SS CON OL ID NTI ICATION CARDS OSWEGO COUNTY HEW YORK.EXECQTJVE LAW K.AKROKHCY MAHAQEQ6ÃfOFHCE ART.2-$

Tha ss ra Carr&

This cant vviil he displayed at LEOMELO P CQSTZLLO ail times Photo

~

where paataqrran Ncf slgfN!N Inn<<t ho@os nQ'r hajji~~

~ rnltgonaY i':i&ill ~ N 5f3j

.thrauqn MIUTARfaa thrauynaur O~lKlt~

POLtC- CQNTROI County nk r5 ISQlÃD 4XPfRA ncH 4A rK MPC

$ 4PrsIQPf... Ja EH c'r 'cc AQTHCRQSQ SY Black on Light Green

--..The~aie~mf, thizjpnfchaH hzi e Tenipiicarj'~esniij enam zi a

r splrhfl~Ap gn8-/ai - IQE~iF CAT(

'h DOUCE CQNTRCK+ Qf5fTS thou fr iu'aego Caunty

~l>> No 0200 empcr~iy IdentificWcn.

Cclcr;vill he announce as iclenttfief lcr hufividual eterne cr event

'HAY 'ORK STATE PQL!CE A

~ ~

~ 4

~ ~

T Rr Ls ur ~ dree aCC ~mare appear Acme ls e ue W ~ Topi Scexc toPacc.

Black cn White - 'Pulpit insignia June 1996 Page 12 EP IP-EPP-14 Rev 01

~ I ~ ~ ~

FIGURE 2 EMERGENCY STATUS BOARD EMERGENCY STATUS BOARD IS THIS Cl IS NOT A DRILL NINE MILE NUCL'EAR. STATION UNIT

'4 4 IS EXPERIENCING A(N).O UNUSUALEYENT CI ALERT.

2? SITE AREA EMERGENCY C3 GENERAL EMERGENCY ALLEMERGENCY VlORKERS HAVE YOUR EMERGENCY 1DENTtFlCATION READY.

CI REPORTTO YOUR NORMALWORK LOCATlON..

CI REPOFlTTO YOUR EVACUATlON ASSEMBLY AREA OR ASS)GNED EMERGENCY RESPONSE FAClUTY.

Page 14 EPIP-EPP-14 June 1996 Rev 01

FIGURE 3 Section 1 Authorization ls for: Emergency Role 0 New Card 0 Replacement Card, Reason:

Name (Last, First, Mo Social Security No.

Department Work Phone No.

Work Address Home Phone No.

Section 2 The above designatedindividual has completed all pertinent Emergency Plan Training, and I request that an Identification Card beissued for hislher use.

raining Supervisor or Designee Date Section 3 The above individualis authorized to receive an Identification Card.

irector EP/Designee Date "This formis to be returned to NMPNS Emergency Preparedness Training" June 1996 Page 15 EP IP-EPP-14 Rev 01

NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT NUCLEAR STATION EMERGENCY PLAN IHPLEHENTING PROCEDURE EPIP-EPP-23 REVISION 07 EMERGENCY PERSONNEL ACTION PROCEDURES TECHNICAL SPECIFICATION REQUIRED Approved By:

R. B. Abbott Date Approved By:

K. A. Dahlberg Plant Manager Unit Date PER/ODXC REVIEW 3.0/2R/98, NO CHANGE Effective Date: 12 31 97 PERIODIC REVIEW DUE DATE

f

j~'IAGARA MOHAWK POWER CORPORATION NINE MILE POINT NUCLEAR STATION EMERGENCY PLAN IHPLEHENTING PROCEDURE EP IP-EPP-23 REVISION 07 EHERGENCY PERSONNEL ACTION PROCEDURES TECHNICAL SPECIFICATION REQUIRED Approved By:

R. B. Abbott Date Approved By:

K. A. Dahlberg Plant Manager Unit Date Effective Date: 12 31 97 PERIODIC REVIEW DUE DATE

LIST OF EFFECTI P G S

/

~PN. ~NI I. ~PN . ~NI ~PN . ~IN .

Coversheet . 22 . 47 ~ ~ ~ ~

1 ~ ~ ~ ~ 23 48 .

1 1 ~ ~ ~ ~ 24 . 49 ~ o ~ ~

111 o ~ o ~ 25 ~ ~ ~ ~ 50 .

1 ~ ~ ~ ~ 26 .

2 ~ ~ ~ ~ 27 .

3 ~ ~ ~ ~ 28 .

4 ~ ~ ~ ~ 29 5 ~ ~ ~ ~ 30 .

6 ~ ~ ~ ~ 31 7 ~ ~ ~ ~ 32 8 ~ ~ ~ ~ 33 9 ~ ~ ~ ~ 34 10 . 35 .

36 .

12 ~ ~ ~ ~ 37 .

13 . 38 .

14 ~ ~ ~ ~ 39 15 40 .

16 . 41 17 . 42 .

18 . 43 19 . 44 .

20 . 45 .

21 46 .

December 1997 Page i EPIP-EPP-23 Rev 07

TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE . ~ ~ ~ 1 2.0 PRIMARY RESPONSIBILITIES ~ ~ ~ 1 3.0 PROCEDURE . ~ ~ ~ 1 4.0 DEFINITIONS . ~ ~ ~ 1

5.0 REFERENCES

AND COMMITMENTS ~ ~ ~ 1 6.0 RECORD REVIEW AND DISPOSITION . ~ ~ ~ 3 ATTACHMENT 1: ERF GENERAL ACTIONS 4 ATTACHMENT 2: SITE EMERGENCY DIRECTOR 5 ATTACHMENT 3: TECHNICAL DATA COORDINATOR . ~ ~ ~ 6 ATTACHMENT 4: REACTOR ANALYST COORDINATOR 8 ATTACHMENT 5: MAINTENANCE COORDINATOR 9 ATTACHMENT 6: RADIOLOGICAL ASSESSMENT MANAGER 10 ATTACHMENT 7: RAD SUPPORT STAFF 15 ATTACHMENT 8: DOSE ASSESSMENT ADVISOR 17 ATTACHMENT 9: SECURITY LIAISON . 18 ATTACHMENT 10: TSC/EOF/CR LIAISON . 20 ATTACHMENT ll: TSC/NED COORDINATOR 21 ATTACHMENT 12: OPERATIONS SUPPORT CENTER COORDINATOR 22 ATTACHMENT 13: OPERATIONS SUPPORT CENTER COMMUNICATOR . 23 ATTACHMENT 14: PERSONNEL ACCOUNTABILITY COORDINATOR . 24 ATTACHMENT 15: RADIATION PROTECTION TEAM COORDINATOR 25 ATTACHMENT 16: DAMAGE CONTROL TEAM COORDINATOR 26 ATTACHMENT 17: STOC SECURITY COORDINATOR 27 ATTACHMENT 18: CORPORATE EMERGENCY DIRECTOR/RECOVERY MANAGER 28 ATTACHMENT 19: TECHNICAL LIAISON AND ADVISORY MANAGER . 33 December 1997 Page ii EPIP-EPP-23 Rev 07

TABLE OF CONTENTS SECTION PAGE ATTACHMENT 20: ADMINISTRATIVE/LOGISTICS MANAGER . 35 ATTACHMENT 21: SECURITY DIRECTOR 38 ATTACHMENT 22: EOF ADMINISTRATOR 40 ATTACHMENT 23: OFF-SITE DOSE ASSESSMENT MANAGER . 43 ATTACHMENT 24: JOINT NEWS CENTER DIRECTOR . 45 ATTACHMENT 25: EOF-JNC LIAISON 47 ATTACHMENT 26: ENVIRONMENTAL SURVEY/SAMPLE TEAM COORDINATOR . 48 ATTACHMENT 27: CONTROL ROOM INFORMATION LIAISON . 50 December 1997 Page iii EPIP-EPP-23 Rev 07

1.0 PURPOSE The attachments to this procedure list tasks that should be completed by emergency personnel at the Emergency Response Facilities depending on the nature and severity of the emergency situation.

2. 0 PRIMARY RESPONSIBILITIES Each individual assigned an emergency response position is responsible for implementing the guidance found in the respective Attachment of this procedure.

3.0 PROCEDURE Each individual for which attachments are provided should use the appropriate attachment for that emergency position to perform the unique actions.

4. 0 DEFINITIONS None

5.0 REFERENCES

AND COMMITMENTS 5.1 Licensee Documentation None 5.2 Standards Re ulations Codes None 5.3 References 5.3.1 EPIP-EPP-03, Search and Rescue 5.3.2 EPIP-EPP-04, Personnel Injury or Illness 5.3.3 EPIP-EPP-05, Station Evacuation 5.3.4 EPIP-EPP-06, In-Plant Emergency Surveys 5.3.5 EPIP-EPP-07, Downwind Radiological Monitoring 5.3.6 EPIP-EPP-OS, Off-Site Dose Assessment and Protective Action Recommendations 5.3.7 EPIP-EPP-13, Emergency Response Facilities Activation and Operation December 1997 Page 1 EP IP-EPP-23 Rev 07

5.3.8 EPIP-EPP-15, Health Physics Procedure 5.3.9 EPIP-EPP-16, Environmental monitoring 5.3. 10 EPIP-EPP-17, Emergency Communications Procedures 5.3. 11 EPIP-EPP-18, Activation and Direction of the Emergency Plan 5.3.12 EPIP-EPP-19, Site Evacuation Procedure 5.3. 13 EPIP-EPP-20, Emergency Notifications 5.3. 14 EPIP-EPP-22, Damage Control 5.4 Commitments Section/Step ,

Commitment b Number Descri tion NCTS 003093-14 OSC Coordinator should assure exterior doors are closed.

NCTS 003093-04 Personnel Accountability Coordinator should keep OSC Coordinator informed.

NCTS 003093-04 OSC Coordinator should keep Maintenance Coordinator informed of accountability activities.

NCTS 003170-14 Technical Data Coordinator should review status boards for accuracy.

NCTS 003152-02 Assure that the HPN Hotline is continuously manned by a technically qualified member of the Radiological or Dose Assessment Group. Decide whether the HPN Hotline is to be manned from the TSC or the EOF.

NCTS 503911-00 Change emergency procedures to accommodate increased Control Room dose during a LOCA due to increased HSIV Leakage.

December 1997 Page 2 EPIP-EPP-23 Rev 07

6.0 RECORD REVIBf ND DISPOSITIO The following records generated by this procedure as a result of actual declared emergency at the Nine Mile Point Nuclear Station shall be maintained by Nuclear Records Management for the Permanent Plant File in accordance with NIP-RMG-01, "Records Management".

Any records, logs or notes-The following records generated by this procedure as a result of EP Drills/Exercises are not required for retention in the Permanent Plant File.

Any records, logs or notes December 1997 Page 3 EPIP-EPP-23 Rev 07

ATTACHllENT 1 RF GENE L ACTIONS

~RE I II IRI Tl I All Emergency Response Personnel responding to an emergency are responsible for implementing the applicable actions of this attachment when reporting to an Emergency Response Facility.

2.0 ~AC IONS

2. 1 Observe and adhere to frisking requirements as required.

2.2 If responding within five hours of alcohol consumption (NIP-FFD-Ol, 3.7), inform the Security Director and cooperate with Security for Fitness for Duty determination.

2.3 Upon arrival at the ERF, or upon hearing the announcement for accountability card in at the accountability card reader. (Card in one time only for accountability).

2.4 Adhere to posted requirements for eating/drinking restrictions.

2.5 Assist in the activation of the facility if needed.

2.6 Perform respective duties per the Emergency Plan Implementing Procedures.

2.7 Sign in on the ERF staffing board.

2.8 Give/Receive complete turnover of emergency situation before being relieved or assuming ERO duties.

2.9 Maintain a log of activities performed for the emergency.

2.10 As necessary, update personnel within your area of responsibility on changing plant conditions.

2.11 Ensure personnel actively assigned to you are accounted for at all times.

2. 12 As necessary, determine need for additional equipment, supplies and/or personnel.

2. 13 Ensure travel restrictions due to safety or radiological conditions are provided to responding personnel.

2.14 Inform Security Director if responding personnel required identification to gain access to NMPNS.

do not have

2. 15 Upon termination of the emergency or at shift change:

a. Sign out at registration log or card out at accountability card reader.

b. Turn in dosimetry.

2.16 Retain for inclusion in the Permanent Plant File all records generated as a result of an actual declared emergency.

December 1997 Page 4 EPIP-EPP-23 Rev 07

ATTACHMENT 2 SITE EMERGENCY DIRECTOR 1.0 RESPONSIBILITIES The Site Emergency Director responsibilities are listed in EPIP-EPP-18.

2.0 CTIONS

2. 1 The SED shall implement actions (as required) of Attachment 1, ERF General Actions.

2. 1 The SED shall implement all actions required of the Site Emergency Director as contained in EPIP-EPP-18 "Activation and Direction of the Emergency Plan".

2.2 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 5 EPIP-EPP-23 Rev 07

ATTACHMENT 3 TECHNICAL DATA COORDINATOR Sheet 1 of 2 I.B ~EE P II IBI III E The Technical Data Coordinatot is responsible for making the TSC operational and directing and coordinating Technical Department personnel in the analysis of emergency conditions and the development of plans and procedures in support of station operations personnel.

2.0 ACTIONS 2.1 Activate the TSC as necessary per EPIP-EPP-13.

2.2 Refer to Attachment 1, ERF General Actions.

2.3 Verify that sufficient numbers of secondary responders are available and are reporting to the emergency facility by reviewing the fax from Community Alert Network (CAN) located at the CAN designated fax.

2.4 Coordinate with the Technical Liaison and Advisory Manager in the EOF and enter information onto the INPO Nuclear Network System.

2.5 Determine need for and request additional equipment, supplies and manpower.

2.6 Obtain briefing from Site Emergency Director on plant status, corrective actions in progress, and identified or anticipated needs from the technical group.

2' Verify sufficient personnel are present to assist in the following duties:

~ Reactor Analyst Coordination

~ ENS Communications (Emergency Notification System)

~ Support Staff

~ TSC/EOF/CR Liaison

~ Control Room Information Liaison 2.8 Assign individuals to act as aides to the Site Emergency Director and to act as data loggers for status boards (Plant Status and Emergency Events).

2.9 Brief staff on plant status, corrective action in progress, and identified or anticipated technical needs.

2. 10 Assign a member of your staff to staff the Tech Info Line, as the TSC-EOF/CR Liaison.

2. 11 Assign a plant qualified member of your staff to man the NRC ENS Hotline necessary.

and perform duties per EPIP-EPP-20 Section 3.4.2 if December 1997 Page 6 EPIP-EPP-23 Rev 07

ATTACHNENT 3 CHNIC DATA COO IN TOR Sheet 2 of 2

2. 12 Verify the NRC Event Notification Worksheet is completed as required per EPIP-EPP-20, Section 3.4.3.

2.13 Brief the TSC/EOF/CR Liaison periodically on TSC activities (e.g., engineering assessment, planned on-going activities, PARs).

(C4) 2.14 Ensure all relevant data received is posted on the appropri ate status board.

2. 15 Assess plant conditions against the EALs and recommend emergency classifications to the SED.

2.16 Direct and coordinate the efforts of the assigned technical staff in analyzing problems and developing solutions, guidance, and emergency operating procedures for operations personnel.

2. 17 Provide the interface between the Site Emergency Director on technical problems, analyses and resolutions.

2. 18 Periodically brief the Site Emergency Director on actions/assessments/status/results.

2. 19 Continuously analyze plant conditions and recommend r e-prioritization of emergency response activities as necessary.

2.20 Assist the SED in developing termination and/or recovery criteria per EPIP-EPP-25; 2.21 Develop long term staffing plans for Technical Support as appropriate.

2.22 Recover technical data developed during the emergency for later use.

2.23 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 7 EPIP-EPP-23 Rev 07

ATTACHNENT 4 EACTOR ANAL ST COORDINATOR 1.0 ESPO SIB I IES The Reactor Analyst Coot dinator is responsible for analyzing and resolving reactor physics related problems, assisting in the development of emergency operating procedures for conducting emergency operations and performing core damage estimates per EPIP-EPP-09.

2.0 ACTIONS 2.1 Refer to Attachment 1, ERF General Actions.

2.2 At the direction of the Site Emergency Director or the Technical Data Coordinator, and in consultation with the Shift Technical Advisor (STA), analyze problems, determine alternate solutions, and design and coordinate the installation of short term modifications.

2.3 Operate Control Room cameras as necessary for determining plant status.

2.4 Monitor trends in plant parameters for early detection of core damage.

2.5 Perform core damage estimates and calculations per EPIP-EPP-09, and provide to Technical Data Coordinator.

2.6 As necessary, consult fuel vendor on issues regarding failed fuel.

2.7 Develop long term action plan for core monitoring and continued assessment (as necessary).

2.8 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 8 EPIP-EPP-23 Rev 07

ATTACHNENT 5 NINTENANC COORD INATO 1.0 RESPONSIBILITIES The Naintenance Coordinator is responsible for the management of all maintenance efforts to provide technical and administrative direction to Damage Control Teams through the OSC Damage Control Team Coordinator and/or the Operations Support Center Coordinator.

2.0 ~CTIOMS 2.1 Activate the TSC as necessary per EPIP-EPP-13.

2.2 Refer to Attachment 1, ERF General Actions.

2.3 Ensure coordination with the Damage Control Team Coordinator.

2.4 Upon activation of the Operations Support Center ensure that the following positions are staffed:

~ OSC Coordinator

~ OSC Communicator

~ Damage Control Team Coordinator 2.5 Establish communications with the OSC Coordinator and keep Site Emergency Director informed relative to OSC activities such as:

Activation status Manpower status Habitability status of OSC areas Damage Control Activities 2.6 Determine the need for damage inspection and repair activities in accordance with EPIP-EPP-22.

2.7 Assist in the installation of special structures, systems, and components as required or in the coordination of contamination control activities as the need arises.

2.8 If a "Site Evacuation" is ordered, coordinate the use of maintenance personnel for the decontamination of evacuating vehicles with the Radiological Assessment Manager.

2.9 Keep Site Emergency Director and Technical Data Coordinator apprised of information received from Damage Control Teams.

2.10 Develop long term staffing plan for maintenance support as appropriate.

2. 11 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 9 EPIP-EPP-23 Rev 07

ATTACHMENT 6 RADIOLOGICAL ASSESSMENT NNGER Sheet 1 of 5 1.0 RESPONSI ITIES The Radiological Assessment Nanager is responsible for managing the on-site dose assessment aspects of an emergency to determine radiological consequences and hazards to station personnel.

2.0 ACTIONS CAUTION IF notified that a LOCA has occurred, THEN go to Step 2.7.

2. 1 Activate the TSC as necessary in accordance with EPIP-EPP-13.

2.2 Refer to Attachment 1, ERF General Actions.

2.3 Verify that sufficient numbers of secondary r esponders are available and are reporting to the emergency facility by reviewing the fax from Community Alert Network (CAN) located at the CAN designated fax.

2.4 Request that the TSC/EOF/Control Room Liaison ask the Control Room if a LOCA has occurred. IF a LOCA has occurred, THEN go to (C6) Step 2.8.

2.5 Ensure that the HPN Hotline is continuously staffed in accordance with EPIP-EPP-20, if necessary.

2.6 Ensure exposure control is in accordance with EPIP-EPP-15.

2.7 Obtain briefing from Site Emergency Director on plant status, corrective actions in progress, identified or anticipated survey/sample needs, and dose assessment requirements.

CAUTION Step 2.8 pertains ONLY to Unit 2 in the event of a LOCA.

2.8 Perform the following:

2.8.1 Evaluate the air intake pathway (either the East or West side of the Control Building) to the Unit 2 Control Room to determine the least contaminated air intake to the Control Room Special Filter Train. The higher potentially contaminated pathway should be isolated. Evaluation should include consideration of:

December 1997 Page 10 EPIP-EPP-23 Rev 07

ATTACHNENT 6 IOLOGICAL ASSESS E NNAGE Sheet 2 of 5 2.8.1 (Cont)

~ release point(s)

~ wind direction 2.8.2 Hake recommendation to the SED on appropriate control room actions based upon this evaluation.

2.8.3 IF unable to determine the higher potentially contaminated pathway, THEN recommend isolation of the East intake.

2.8.4 Advise the OSC Radiation Protection Team Coordinator to direct Control Room personnel AND those reporting to the Control Room to don protective clothing and eyewear (C6) for the purpose of reducing beta dose.

2.9 Verify personnel are present to fill the following positions:

~ Radiation Protection Team Coordinator

~ Off-Site Dose Assessment Manager

~ Rad Support Staff (as needed)

(C5) ~ HPN Communicator 2.10 Request additional personnel as needed from the OSC (preferably Chemistry and Radiation Protection Department personnel) to assist in performing the following activities:

Radiological control activities On-site dose projections Communications (radio and dedicated lines)

Habitability surveys of emergency response facilities Source Term Assessment Post Accident Chemistry Samples 2.11 Designate an individual to coordinate the issuance of dosimetry to non-site personnel if and when appropriate.

2. 12 Brief RP Team Coordinator and ODAM on plant status, corrective action in progress, and identified or anticipated survey/sample needs. Discuss survey/sample strategy and develop plans.

2. 13 Contact on-call Chemistry Supervisor support is required.

if additional chemistry

2. 14 Before dispatch of emergency teams ensure that appropriate measures are implemented to adequately monitor and control personnel exposures. (Refer to EPIP-EPP-15)

2. 15 Ensure on-site protective actions (shelter or site evacuation) are being evaluated and implemented.

December 1997 Page ll EPIP-EPP-23 Rev 07

ATTACHMENT 6 RAD OLOGIC L SSESS E GE Sheet 3 of 5 2.16 If it is determined that safety or radiological hazards exist offsite or onsite:

a. Consult with ODAM regarding best possible ingress and egress routes.

b. Determine the need for a site evacuation using EPIP-EPP-15.

c. Coordinate with the SED the implementation of onsite protective actions.

2.17 If site evacuation is to be implemented, determine best route to leave site.

2.18 Assign priorities using Table 6.1 as a guide.

2.19 If radiological conditions warrant, ensure a general announcement is made prohibiting smoking, eating and drinking when deemed appropriate.

2.20 Ensure TSC habitability surveys are performed using EPIP-EPP-13.

2.21 If radiological conditions warrant, ensure step off pads and monitors are set up at the entrances to TSC.

2.22 To ensure TSC habitability for 30 days following a Loss of Coolant Accident (LOCA), Direct an air sample to be taken for I-131 concentration following TSC emergency ventilation system initiation. If the LOCA occurs at Unit 2 you may compare the I-131 results with EPIP-EPP-13, Determination of TSC Habitability following a Design Basis Accident (DBA).

2.23 Perform on-site dose assessment activities outlined in EPIP-EPP-15.

2.24 Consult with activities.

ODAN, if necessary, on results of assessment 2.25 Consult with the Environmental Survey/Sample Team Coordinator (ESSTC), as necessary, on on-site and off-site environmental monitoring results.

2.26 Ensure on-site dose rates and protective actions are posted.

2.27 Assist Environmental Survey/Sample Team Coordinator in selecting proper monitoring locations and assessing radiological conditions expected in the field.

2.28 Assist Rad Support Staff in selecting proper monitoring and sample collection points, data required, and the assessment of radiological conditions at those points.

December 1997 Page 12 EPIP-EPP-23 Rev 07

ATTACHMENT 6 IOLOGIC L SSESS ENT GER Sheet 4 of 5 2.29 Consult with Chemistry Supervisor to assess the release rate and required sampling.

2.30 Haintain interface with the Rad Support Staff in the following matters:

Required survey/sample activities Disposition of results (including disposition of various samples)

Requests for outside assistance, (such as JAF, Ginna, INPO, FRHAP) are to be made through the SED interfacing with these groups.

2.31 Implement use of RWPs for on-site activities through the Rad Support Staff and additional staff in TSC (i.e., repair and damage control, assessment activities, operations, etc.).

2.32 Provide technical and administrative direction to the ESSTC during re-entry operations in accordance with EPIP-EPP-12.

2.33 Assist the SED in developing termination and/or recovery criteria per EPIP-EPP-25.

2.34 Develop a long term staffing plan for Radiological Protection support as appropriate. Utilize JAF personnel as appropriate.

2.35 Collect Radiological Protection data developed during the emergency for later review and analysis.

2.36 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 13 EPIP-EPP-23 Rev 07

ATTACHMENT 6 0 OGIC SSESS ENT MANGER Sheet 5 of 5 TABLE 6.1 RADIOLOGIC L ASSESSMENT MANGER CTIVITY PRIORITIES Priori t T s Procedure to Im lement Search and Rescue and First Aid:

Lifesavin Onl EPIP-EPP-03, EPIP-EPP-04, EPI P-EPP-15 Initial On-site protective actions EP IP-EPP-15 In-Plant Surveys EPIP-EPP-06 Provide Personnel to Accompany Damage Control Team EPIP-EPP-06, EPIP-EPP-22 Provide Personnel to Monitor at Accountability Areas for Radiation/

Contamination EPIP-EPP-05 Emergency First Aid and Decontamination:

not Lifesaving EPIP-EPP-04, EPIP-EPP-15 Provide Personnel to Accompany Follow-Up Re-entry Teams EPIP-EPP-22 Personnel Exposure Control (Routine Dosimetry Issuance and Completion of Special Radiation Mork Permits) EPIP-EPP-15, EPIP-EPP-22 Follow-Up In-Plant/On-Site Monitoring and C Sample Collection EP IP-EPP-06, EP IP-EPP-07 10 Sample Analysis EPIP-EPP-15 Minor First Aid and Decontamination EPIP-EPP-04, EPIP-EPP-15 12 Personnel Re-entry to Site EPIP-EPP-12 This list of activity priorities is sequenced in a "likely order" for a fast breaking radiological emergency when personnel resources may be limited. Personnel assignments should be made as needed by the specific plant and personnel requirements.

December 1997 Page 14 EPIP-EPP-23 Rev 07

ATTACNEHT 7 SUPPORT STAFF Sheet 1 of 0

1. 0 RESPONSIBILITIES The Rad Support Staff is responsible for providing technical and administrative direction to In-Plant monitoring and sampling/survey teams, and post accident sampling team(s).

2.0 ACTIONS 2.1 Activate the TSC as necessary in accordance with EPIP-EPP-13.

2.2 Refer to Attachment 1, ERF General Actions.

2.3 Determine need for and request additional equipment, supplies and staff.

2.4 Ensure exposure control is in accordance with EPIP-EPP-15.

2.5 Obtain briefing from Radiological Assessment Hanager on plant status and corrective actions in progress.

2.6 Assess plant status and communicate these conditions to appropriate personnel.

2.7 Establish communications with the Radiation Protection Team Coordinator in the OSC.

2.8 Request the Rad Protection Team Coordinator assign personnel to perform In-Plant monitoring as directed by the Radiological Assessment Hanager. Priorities for assignment will depend on plant conditions; the following order of tasks is provided as a guide:

Support of source term calculations needed for initial dose projection when radiation monitors are inoperable In-Plant surveys Accompany initial Damage Control Teams (EPIP-EPP-06, 22)

Accompany subsequent Damage Control Teams (EPIP-EPP-06, 22)

In-Plant sample collection (EPIP-EPP-06, 15)

Sample analysis (EPIP-EPP-15)

Other missions as required 2.9 Provide radiological control for the facility in accordance with standing radiological procedures.

2. 10 Provide Rad Protection Team Coordinator with appropriate precautions on expected or potential hazards, protective clothing requirements, and exposure control (in accordance with EPIP-EPP-06 and EPIP-EPP-15).

December 1997 Page 15 EPIP-EPP-23 Rev 07

ATTACfNENT 7 RAD SUPPO S Sheet 2 of 2

2. 11 Keep Radiological Assessment Manager apprised of all data received.

2. 12 Ensure a radiation protection technician is dispatched with any emergency team to provide radiation protection coverage. Arrange for this through the Radiation Protection Team Coordinator in the OSC.

2. 13 In the event of a station evacuation, request Rad Protection Team Coordinator dispatch a survey team to monitor Assembly Areas (see EPIP-EPP-05, Station Evacuation).

2. 14 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 16 EP IP-EPP-23 Rev 07

ATTACHNENT 8 DOSE ASS SS E VISO

l. ~BEBB BBIB Tl The Dose Assessment Advisor is responsible for providing on a regular basis:

~ Heteorological data

~ Determining effluent release rate

~ Off-site radiological assessment

~ Protective Action Recommendations for SED approval.

2.0 ACTIONS 2.1 Report to the control room when notified of an emergency.

2.2 Notify the on call Chemistry Supervisor is required.

if additional assistance 2.3 Implement dose assessment activities in accordance with EPIP-EPP-08.

2B4 Provide meteorological information as requested by the SSS/SED.

2.5 Perform Dose Assessment activities and PARs per EPIP-EPP-08 until relieved by the ODAN.

2.6 Assist the SSS/SED in the control room as directed.

2.7 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 17 EP IP-EPP-23 Rev 07

ATTACNENT 9 SECU ITY I ISON Sheet 1 of 2 1.0 RESPONSIBI ITIES The Security Liaison is responsible for maintaining:

~ Communications link between Site disciplines

~ Security Tactical Operations Center (STOC)

~ Updating the SED and staff on current, on-going security events

~ Communicating command directives from the SED to the Security Coordinator in the STOC (when staffed).

2.0 ~CT IONS 2.1 Activate the TSC as necessary in accordance with EPIP-EPP-13.

2.2 Refer to Attachment 1, ERF General Actions.

2.3 Determine need for and request additional equipment, supplies and personnel.

2.4 Obtain briefing by Site Emergency Director or his designee on emergency status and any security needs.

2.5 Contact the Security Coordinator in the Security Tactical Operations Center (STOC) located in the Security Building to determine status of station security and update the SED of the status of applicable security and contingency procedures.

2.6 Ensure that requests for assistance are provided to the Personnel Accountability Coordinator in accounting for station personnel in accordance with EPIP-EPP-05, "Station Evacuation", EPIP-EPP-19, "Site Evacuation", and security procedures, if appropriate.

2.7 Ensure that requests for access and traffic control for Off-Site Niagara Mohawk Power Corporation (NMPC) ERF locations are communicated to the Security Director.

2.8 Consult with the Radiological Assessment Manager on protective measures that should be taken by security department personnel, as appropriate.

2.9 Communicate with the Security Director the need to provide personnel to allow NRC personnel access to the Learning Center roof or the Loomis Corners radio tower so that they may install portable radio equipment as needed.

2. 10 Maintain liaison with the Security Director.

2.11 Ensure that the NMPC helicopter is secured upon request.

December 1997 Page 18 EPIP-EPP-23 Rev 07

ATTACHNENT 9 S CU SON Sheet 2 of 2 2.12 Communicate, in a timely manner, all SED directions for the use of security personnel on site to the Security Coordinator.

2. 13 Coordinate the assignment of security personnel to Damage Control Teams as directed/requested.

2.14 Assist the SED in developing termination and/or recovery criteria as needed.

2. 15 Develop long term staffing plan for security in conjunction with the Security Coordinator, as needed.

2.16 Collect all paperwork developed during the emergency for later review and analysis.

2.17 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 19 EPIP-EPP-23 Rev 07

ATTACHMENT 10 SC 0 SON 1.0 RESPONSIBI ITI S The TSC/EOF/CR Liaison is responsible for maintaining liaison with the Control Room Information Liaison Technical Assistant located in the EOF Technical Assessment Room and providing the technical interface between the EOF, TSC and the Control Rooms.

2.0 ~CTIONS 2.1 Refer to Attachment 1, ERF General Actions.

2.2 Determine and request additional support as needed from the Technical Data Coordinator.

2.3 Obtain the names of individuals filling the emergency positions

.in the Control Room and provide this information.to the Technical Data Coordinator for posting.

2.4 Receive briefing from the Site Emergency Director or his designee on plant status and corrective actions in progress.

2.5 Obtain information from the Control Room Information Liaison and keep the technical briefers in the EOF Technical Assessment Room 2.6 informed of on-site developments.

Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

0 December 1997 Page 20 EPIP-EPP-23 Rev 07

ATTACHMENT 11 TSC HED COORDINATOR RESPONSIBILITIES The TSC/NED Coordinator is responsible for coordinating Nuclear Engineering Department support and Licensing.

2.0 ACTIONS

2. 1 Refer to Attachment 1, ERF General Actions.

2.2 Determine need for and obtain additional equipment, supplies and personnel.

2.3 Obtain a briefing from the Site Emergency Director on plant status, corrective actions in progress, and identified or anticipated problem areas.

2.4 Establish and maintain contact with the Technical Liaison and Advisory Manager in EOF, and brief on current situation and corrective actions in progress.

2.5 Analyze mechanical, electrical, structural, instrumentation and control and radiological problems; determine alternate solutions; design and assist in the coordination of short-term modifications.

2.6 Analyze thermohydraulic and thermodynamic problems and develop problem resolutions.

2.7 Assist in the development of Emergency Operating Procedures, Operating Procedures, etc. as necessary for conducting emergency operations.

2.8 Analyze conditions and develop guidance for the Site Emergency Director and operations personnel for protection of the reactor core.

2.9 Develop long term staffing plan for engineering support as needed.

2. 10 Collect paperwork developed during the emergency for later review and analysis.

2.11 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 21 EPIP-EPP-23 Rev 07

ATTACHMENT 12 OPERATIONS SUPPORT CENTER COORDINATOR

1. 0 RESPONSIBILITIES The Operations Support Center Coordinator is responsible for making the OSC operationa't, coordinating and supervising the overall emergency response operations of the OSC.

2.0 ACTIONS 2.1 Activate the OSC in accordance- with EPIP-EPP-13.

2.2 Refer to Attachment 1, ERF General Actions.

2.3 Ensure proper use of communications equipment in accordance with EP IP-EPP-17.

2.4 Establish communications with Technical Support Center (TSC)

(normal hours) or Control Room (off-hours) and request snformation on plant status and corrective actions in progress.

2.5 If a radioactive release has occurred, or is in progress, ensure a general announcement is made prohibiting smoking, eating, and drinking until habitability surveys have been completed and found to be satisfactory.

2.6 Direct Radiation Protection to survey the facility and provide radiological control in accordance with standing radiological procedures. Notify Site Emergency Director immediately of results.

2.7 Place sign on door to the Unit 1 Administration Building Lobby directing all personnel to enter via the employee entrance.

(Cl) 2.8 Ensure all exterior doors to the Unit 1 Administration Building are closed during a radiological emergency.

2.9 When sufficient numbers of personnel are available to support emergency functions, notify the Maintenance Coordinator thy OSC is operational.

2. 10 Keep Site Emergency Director informed of all available information concerning repairs, staff, surveys, etc.

2. 11 Provide appropriate announcements in OSC to keep personnel informed.

(C3) 2. 12 Obtain information from the Personnel Accountability Coordinator on the status of the efforts to find missing people and provide this information to the Maintenance Coordinator in the TSC. If necessary, implement EPIP-EPP-03.

2. 13 In conjunction with the Maintenance Coordinator, develop long term staffing plans for maintenance support.

2. 14 Collect all paperwork developed during the emergency for later review and analysis.

2. 15 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 22 EPIP-EPP-23 Rev 07

ATTACHMENT 13 OPERATIONS SUPPORT CENTER CONNUNICATOR

l. ~RiEPIINEIBtLI Y The OSC Comunicator is responsible for maintaining communications with the Control Rooms, Technical Support Center (TSC) and Personnel Accountability areas.

2.0 ACTIONS

2. 1 Refer to Attachment 1, ERF General Actions.

2.2 Ensure proper use of communications equipment in accordance with EP IP-EPP-17.

2.3 Establish and maintain communications with the TSC (normal hours) or Control Room (off-hours), as appropriate.

2.4 Install additional phones as necessary and test. Test backup radio.

2.5 Frequently request Emergency Status updates from TSC and provide information to OSC Coordinator for disbursement to OSC staff.

2.6 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 23 EPIP-EPP-23 Rev 07

ATTACHNENT 14 PERSO NEL CCOUNTABI COO INATOR

~PEP I I IE The Personnel Accountability Coordinator is responsible for the accounting of all site personnel, visitors and contractors.

2. 0 ACTIONS

2. 1 Refer to Attachment 1, ERF General Actions.

2.2 Inform the OSC Coordinator that the Personnel Accountability Coordinator position is staffed and ready to perform accountability when requested.

2.3 Establish communications with personnel accountability assembly areas (as required) and carry out actions required in accordance with EPIP-EPP-05 and EPIP-EPP-19.

2.4 Establish contact with Security Liaison located in TSC to coordinate the computerized accountability process.

(C2) 2.5 Keep the Security Liaison in TSC and the OSC Coordinator informed of accountability activities, including the status of finding missing people.

2.6 Coordinate with the OSC Coordinator and implement search and recuse actions of EPIP-EPP-03 as necessary.

2.7 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 24 EPIP-EPP-23 Rev 07

ATTACHNENT 15 RADI ION PROTECTION TEAN COORDINATOR 1.0 ~RE PRNEIEIL TIEE The Radiation Protection Team Coordinator is responsible for providing technical and administrative direction to survey/sample teams and determining OSC habitability.

2. 0 ACTIONS 2.1 Activate the OSC in accordance with EPIP-EPP-13 as necessary.

2.2 Refer to Attachment 1, ERF General Actions.

2.3 Verify that sufficient numbers of secondary responders are available and are reporting to the emergency facility by reviewing the fax from Community Alert Retwork (CAN) located at the CAN designated fax.

2.4 Provide radiological control for facilities in accordance with standing radiological procedures.

2.5 Ensure exposure control is in accordance with EPIP-EPP-15.

2.6 Contact Radiological Assessment Manager or the Rad Support Staff in the Technical Support Center (TSC) and receive briefing and instructions.

2.7 Assign radiation protection technicians to the following tasks as appropriate and log the assignments:

Downwind Survey, Team A Downwind Survey Team B Downwind Survey Team C In-Plant Survey Teams 1-6 Repair/Damage Control Team 1 Repair/Damage Control Team 2 Fire/Rescue/Medical Brigade 1 2.8 Direct survey teams to prepare for dispatch and inform when ready. Advise OSC Coordinator when teams have been dispatched and their destination.

2.9 Report OSC radiation survey and air sample results to the OSC Coordinator.

(COMM 1) 2.10 If radiological conditions warrant, set up step-off pads and monitors by the employee and lobby entrances to the Unit 1 Administration Building and the Unit 1 entrance to the bridge connecting Unit 1 and 2.

2. 11 Inform the OSC Coordinator when these areas are established.

NOTE: Step 2.12 is only for Unit 2.

2.12 In the event that a LOCA has occurred, or as directed by the RAN, Control Room personnel and others who may report to the Control Room shall be directed to don protective clothing and eyewear for (C6) the purpose of reducing beta dose.

2. 13 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 25 EP IP-EPP-23 Rev 07

ATTACHNENT 16 DAMAGE CONTROL T COO OR 1.0 RESPONSIBILITIES The Damage Control Team Coordinator is responsible for providing technical and administrative direction to Damage Control Teams, providing an assessment of any damaged equipment and necessary personnel or equipment needs to effect emergency repairs, keeping OSC personnel appraised of Damage Control and Repair activities, and assuring that Damage Repair Team leaders maintain accountability of their team members at all times.

2.0 ACTIONS 2.1 Activate the OSC in accordance with EPIP-EPP-13 as needed.

2.2 Refer to Attachment 1, ERF General Actions.

2.3 Verify that sufficient numbers of secondary responders are available and are reporting to the emergency facility by reviewing the fax from Community Alert Network (CAN) located at the CAN designated fax.

2.4 Ensure proper use of communications equipment in accordance with EP I P-EPP-17.

2.5 Contact Maintenance Coordinator in TSC for briefing and any instructions.

2.6 In consultation with Maintenance Coordinator, determine any preparations necessary for damage control teams in accordance with EPIP-EPP-22 and advise the Damage Control Teams as appropriate.

2.7. Assign Maintenance personnel to standby as teams for any necessary repair/damage control activities.

2.8 Advise Maintenance Coordinator and OSC Coordinator of team assignments.

2.9 If it is determined that On-Site security is needed for assistance with access control or personnel protection, request assistance through the Security Liaison in the TSC.

2.10 Obtain system engineering support for specific damage teams, as needed.

2. 11 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 26 EPIP-EPP-23 Rev 07

ATTACNEHT 17 STOC SECU I COORDINATO RESPONSIBILITIES The STOC Security Coordinator is responsible for maintaining plant security and instituting appropriate measures per the Site Security Plan or as directed by SED in assisting the Personnel Accountability Coordinator in search and rescue activities to account for missing personnel.

2. 0 ACTIONS 2.1 On a continuing basis, inform and update Security Liaison in TSC, and the Security Director in EOF of current security events.

2.2 Ensure that all personnel actively assigned to you are accounted for at all times.

2.3 Maintain a log of Security related activities.

2.4 Determine need for and request additional equipment, supplies and personnel 2.5 Assist the Personnel Accountability Coordinator in search and rescue efforts.

2.6 Develop long term staffing plans for security as needed 2.7 Collect paperwork developed during the emergency for later review and analysis.

2.8 Provide access and traffic control check points at EOF and AEOF and coordinating on-Site security emergency activities.

2.9 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 27 EP IP-EPP-23 Rev 07

ATTACHMENT 18: CO ORATE EMERGENCY DIRECTOR RECOVERY MANAGER Sheet 1 of 3 1.0 ESPONS IB I L ITY The Corporate Emergency Director/Recovery Manager is responsible for managing all aspects of the NMPC response to an emergency at NMPNS.

2.0 ~CTIONS

2. 1 Refer to Attachment 1, ERF General Actions.

2.2 Call for 'information from the following as appropriate:

~ Technical Support Center

~ Unit 1 Control Room

~ Unit 2 Control Room 2.3 Notify the Chief Nuclear Officer, President, and Chairman of the Board (Chief Executive Officer) of the situation at the NMPNS and actions to be taken. (See Emergency Events Phone List) 2.4 Establish and maintain communications with the Site Emergency Director in the TSC, and obtain plant status updates.

2.5 Ensure communications with State and Oswego County are transferred to the EOF in accordance with EPIP-EPP-20.

2.6 Verify with EOF Administrator that a NMP Technical Representative has been assigned to report to the State and County EOCs.

2.7 Verify that a NMP Technical Representative has been assigned to report to the State and County EOC's.

2.8 Brief EOF staff on initial accident conditions. Attachment 18, Figure 1, "Ingredients for a Good Update" may be utilized for ~

this.

2.9 Direct EOF managers to evaluate resource needs.

2. 10 When sufficient numbers of personnel are available in the EOF to support emergency functions, assume overall direction, control and authority of Niagara Mohawk's emergency response activities.

2. 11 Transfer responsibility from the Site Emergency Director to the CED/RM.

2.12 Direct the SED in the TSC to make the announcement to TSC emergency personnel.

December 1997 Page 28 EP IP-EPP-23 Rev 07

ATTACHMENT 18 CO PORATE EMERGENCY DI ECTOR RECOVERY MANAGER Sheet 2 of 3

2. 13 Make announcement in the EOF (see below for an example of the EOF announcement).

EOF ANNOUNCEMENT Transfer of Emer enc Direction and Control From the TSC to the EOF "Attention. This is/is not a drill. This is (name), Corporate Emergency Director. As of hrs, I have relieved the Site Emergency Director, (name) of overall direction and control of the emergency." (Provide brief status of the emergency situation) "This is/is not a drill."

2. 14 Advise State and County Emergency Operations Centers of this formal transfer.

2.15 Review and approve NMPC Protective Action Recommendations (PARs).

The CED/RM shall not delegate the approval of notifications or protective actions to off-site agencies.

2.16 Interface with the J. A. FitzPatrick Nuclear Power Plant Liaison to obtain support as necessary.

2.17 Provide periodic briefings to the EOF staff regarding emergency status and progress. Attachment 18, Figur'e 1, "Ingredients for a Good Update" may be utilized for this.

2. 18 Meet with Federal, State and County officials to discuss plant status, the prognosis of the emergency, and protective action recommendations, if appropriate. Utilize Attachment 18, Figure 2, "CED Guidelines for NRC and Offsite Agency Interface".

2.19 Review and approve all press releases.

2.20 Periodically brief NMPC corporate officer(s) 2.21 Assi st the Site Emergency Director in continued assessment of emergency conditions and in determining and directing actions per the Site Emergency Plan and Procedures.

2.22 Interface as needed directly or through the Technical Liaison and Advisory Manager, with representatives of the Legal, Claims and Risk Management Departments.

December 1997 Page 29 EPIP-EPP-23 Rev 07

ATTACNENT 18 CO PO GEC CO ECOEY AGE Sheet 3 of 3 2.23 Establish communications with INPO and/or other vendor organizations as conditions warrant and request their assistance, if deemed necessary.

2.24 Coordinate SORC/SRAB review as appropriate, of any emergency actions, procedures, modifications, etc.

2.25 Approve all outside technical and vendor contracts.

2.26 Authorize purchases of necessary equipment and supplies, as appropriate.

2.27 When appropriate, implement actions in accordance with EPIP-EPP-25 for reclassification, termination and/or recovery.

2.28 Coordinate with the Work Control groups to schedule recovery meetings and prepare agenda.

2.29 If outside groups are to conduct investigations (e.g., NRC, Congressional Subcommittees, etc.) 'coordinate through the Technical Liaison and Advisory Manager to arrange for legal and technical interface as necessary. Also, determine the advisability of conducting an independent and parallel in-house investigation and direct same as appropriate.

2.30 If required, request D.O.E. assistance through FRMAP (Federal Radiological Monitoring and Assessment Plan) via the TLAH.

2.31 Ensure the initiation of the development of environmental impact studies.

2.32 Ensure an evaluation of a release is performed in accordance with 10CFR140.84, Radiological Criteria for Extraordinary Nuclear Occurrence per EPIP-EPP-16, Environmental Monitoring.

2.33 Ensure an estimate of the total population dose is made per EPIP-EPP-16, Environmental Monitoring.

2.34 Develop long term staffing plans for CED/RM positions and review staffing plans for other ERF's.

2.35 Ensure collection of paperwork developed during the emergency for later review and analysis.

2.36 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 30 EP IP-EPP-23 Rev 07

ATTACHNENT 18 FIGURE 1 INGREDIENTS FOR A GOOD UPDATE 0 "Attention in the EOF; This (is/is not) a drill; This is an Update."

0 Emergency Classification 0 Plant Status

~ Briefly - Where we'e been....

~ Where we are

~ Where we are going......time frame if known 0 Release information 0 Protective Action status...Clarify NHPC PARs versus County Actions 0 Outside involvement...NRC, INPO, GE, Others?

0 "What other information or corrections does anyone have that relate to our status or plan?"

0 "Any questions?"

0 "End of update" December 1997 Page 31 EP IP-EPP-23 Rev 07

ATTACHNENT 18 FIGURE 2 CED GUIDELINES FO NRC AND OFFSITE AGENCY INTERFACE NOTE: This guideline refers to NRC, County or State (hereafter referred to as NRC/Offsite) emergency response personnel.

IF an additional CED is available, direct them to complete the actions contained in this guideline.

2. Introduce yourself to arriving NRC/offsite personnel.

3. Direct EOF Administrator to show above personnel to their respective EOF rooms.

Assign Nine Nile Point ERO personnel as contacts in each of the following areas:

~ dose assessment (request persons name from the ODAM)

~ plant assessment (request persons name from the EOF Administrator)

~ command/control (assign this person yourself)

5. Announce the following over the EOF PA system:

"Attention in the EOF. The following persons have been assigned as primary contacts for the NRC, State and County EOF responders (state the name of each contact person and their area of responsibility). I would request that all NRC, State and County personnel direct all questions to those individuals. Thank you.

6. Periodically update NRC/Offsite personnel regarding plant and radiological conditions, as well as intended protective actions for onsite and offsite.

NOTE: The assignment of contact personnel does NOT preclude the NRC/Offsite personnel from talking with other NHP EOF staff.

December 1997 Page 32 EPIP-EPP-23 Rev 07

ATTACHMENT 19 TECHNIC L LIAISO D VISORY NAGER Sheet 1 of 2

l. ~RE P II IBI IYY The Technical Liaison and Advi sory Manager is responsible for advising the CED/RM on technical/engineering matters and coordinate an advisory group comprised of technical and managerial personnel from government, contract and consultant support organizations.

2.0 ACTIONS 2.1 Activate the EOF per EPIP-EPP-13 as needed.

2.2 Refer to Attachment 1, ERF General Actions.

2.3 Obtain initial briefing from the TSC/NED Coordinator.

2.4 Contact representatives of the Legal Department and advise the individual contacted of the emergency situation. If necessary, request that an Attorney and a Claims Department representative be dispatched to the EOF.

NOTE: Provide proper travel direction (to avoid radioactive plume) as appropriate. Also determine if individuals have an Oswego County Access Control ID card.

coordinate obtaining these cards through the If not, EOF Security Director.

2.5 Contact the American Nuclear Insurers (ANI) and provide a technical briefing on the accident situation. Provide the names and phone numbers of Risk Management personnel.

2.6 Inform the Communications Coordinator in the EOF that you have taken over the notifications to ANI.

2.7 Interface with onsite G.E. representative.

2.8 Contact a representative of the Risk Management Department and advise the individual contacted of the emergency situation and of your conversation with ANI.

2.9 Contact a representative of the guality Assurance Department and advise the individual contacted of the emergency situation.

December 1997 Page 33 EPIP-EPP-23 Rev 07

ATTACHMENT 19 TECHNIC LI ISON D SORY MAN GE Sheet 2 of 2

2. 10 When contacted by the INPO Liaison, make arrangements for entry into the EOF.

2. 11 Interface with the INPO Liaison on matters relating to assistance requests made to INPO and/or the industry.

2. 12 Contact the EOF/JNC Liaison and coordinate release of information to public.

2. 13 Establish an advisory group of engineers and technicians (including outside consultants, Legal and Claims personnel) to provide assistance to the Corporate Emergency Director/Recovery Manager.

2. 14 Ensure that necessary plant modifications, designs, etc. are appropriately reviewed by the guality Assurance Department.

2. 15 Authorize purchases of necessary equipment and supplies, as appropriate.

2. 16 Ensure all engineering-related activities and support are properly initiated and carried out.

2. 17 Review and approve all changes to emergency procedures and ensure appropriate review of all necessary plant modifications, designs, etc. Interface with the SORC and SRAB, as applicable.

2.18 Periodically interface with the Work Control groups to assure appropriate scheduling and prioritization of activities.

2.19 After the emergency condition has subsided, assist the CED/RM in the development of termination and/or recovery criteria in accordance with EPIP-EPP-25.

2.20 If outside groups are to conduct investigations (e.g., NRC, Congressional Subcommittees, etc.) coordinate with the Corporate Emergency Director/Recovery Manager, Legal Department, and others as necessary to arrange for legal and technical interface.

2.21 Determine the advisability of conducting an independent and parallel in-house investigation, and direct same as appropriate.

2.22 Develop long term staffing plans for support organizations as needed.

2.23 Collect paperwork developed during the emergency for later review and analysis.

2.24 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 34 EPIP-EPP-23 Rev 07

ATTACHMENT 20 e RESPONSIB LITY The AD I IST IVE LOGIST CS MANAGER Sheet Administrative/Logistics Nanager is responsible for administrative 1 of 3 and logistic functions required to support the entire off-site and on-site emergency organizations. The types of support services could include:

~ General Administration

~ Transportation of materials, personnel, etc.

~ Personnel administration and accommodations

~

Purchasing

~ Petty Cash

~ Outside plant support

~ Commissary

~ Safety

~ Sanitation

~ Human Resources

~ Communications

~ Non-technical staffing 2.0 ACTIONS 2.1 Activate the EOF in accordance with EPIP-EPP-13.

2.2 Refer to Attachment 1, ERF General Actions.

2.3 Verify that sufficient numbers of secondary responders are available and are reporting to the emergency facility by reviewing the fax from Community Alert Network (CAN) located at the CAN designated fax.

2.4 Obtain a briefing from the CED/RM or the TLAM and determine administrative/logistics needs.

2.5 Contact each of the following groups and advise the contact of the situation and relate any current or anticipated assistance that may be needed:

~ NHPNS Admin. Support/Services

~ NHPC Purchasing

~ NMPC Transportation

~ NMPC Treasury

~ NHPC Materials Management

~ NHPC Network Management NOTE: Provide proper travel direction (to avoid radioactive plume) as appropriate. Also determine if individuals contacted have an Oswego County Access Control ID card. If not, coordinate obtaining these cards through the EOF Security Director.

December 1997 Page 35 EPIP-EPP-23 Rev 07

ATTACHNENT 20 2.6 ADHINISTRATIVE L06ISTICS AGER Establish general administrative activities, as required or Sheet 2 requested, for all emergency response/recovery centers, including the following:

of 3 0

~ Typing services

~ Xerox services

~ Stenographic support

~ Facsimile services Audio/visual aids, graphics, printing and photography

~ Communications services

~ Office furniture 2.7 Establish a commissary (if appropriate) and arrange for food service and water supply support for personnel at each emergency response/recovery facility.

2.8 Establish areas for handling transportation and housing functions, and evaluate their needs daily.

2.9 Secure use of the NHPC helicopter upon request via the Transportation Coordinator.

NOTE: Consult with the Environmental Sample/Survey Team Coordinator before requesting the helicopter so that radiological conditions at and in route to the helipad may be evaluated.

2.10 Arrange for office facilities as necessary which may include the following:

~ Additional trailers (including power supplies, HVAC, etc.)

~ General maintenance, housekeeping and janitorial services

~ Lavatory and sanitation facilities

~ Trash removal

~ Hail delivery

~ Communications

~ Repair of office equipment

2. 11 Periodically review human resources and needs, including the following:

~ Work schedules

~ Staff replacement

~ Payroll and petty cash 2.12 Arrange for miscellaneous resources, including the following:

~ Laboratory supplies

~ Additional dosimetry and radiation equipment

~ Additional Staff

2. 13 Arrange for the coordination and supply of materials and equipment from the NHPNS stores facilities, as appropriate.

~ i December 1997 Page 36 EPIP-EPP-23 Rev 07

ATTACHMENT 20 ADNINIS RATIVE LOG S ICS NNAGER Sheet 3 of 3 2.14 Coordinate with the Work Control groups in developing work schedules and prioritizing administrative/logistics activities

2. 15 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 37 EPIP-EPP-23 Rev 07

ATTACQIENT 21 SECU C 0 Sheet 1 of 2 1.0 RESPONSIBILITY The Security Director is responsible for providing overall direction for security and traffic control at the NHPC facilities, providing additional security personnel (as required), and coordinating with the Security Coordinator the off-site security and police forces involved in the emergency.

2.0 ACTIONS 2.1 Refer to Attachment 1, ERF General Actions.

2.2 Perform, or arrange for performance of, breath analysis of individuals declaring alcohol consumption within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of reporting for duty by qualified breathalyzer technique.

2.3 Notify appropriate Security personnel of the situation at NHPNS.

2.4 Obtain briefing from the CED/RH or TLAH of plant status and Security needs..

2.5 Ensure the EOF registration desk is manned as necessary.

2.6 2.7 Call in (or put on standby) additional security personnel to establish/maintain security (site, Ensure provisions EOF, JNC, for security at the Joint etc.).

News Center have been

~ IJ initiated.

2.8 Establish and maintain a security communications center with NHPC security personnel and appropriate off-site law enforcement groups.

2.9 Coordinate on-site security efforts via the Security Coordinator located in the Security Tactical Operations Center (STOC).

Provide direction to the Security Coordinator as necessary.

2. 10 Coordinate off-site security efforts with appropriate off-site security and law enforcement groups.

NOTE: Requests for any outside law enforcement assistance must be coordinated through the Oswego County Sheriff.

2. 11 Ensure that appropriate security measures (including badging) have been established and maintained at all emergency response/recovery facilities (e.g., EOF, JNC, Site).

December 1997 Page 38 EPIP-EPP-23 Rev 07

ATTACHMENT 21 SECURITY DIRECTOR Sheet 2 of 2

2. 13 Ensure that the Oswego County Sheriff's Office has established road blocks at both ends of private road (i.e., site entries).

Supplement with and maintain NMPC security at all roadblocks, if appropriate.

2.14 Establish and maintain traffic-control patterns (flow) at all NHPC facilities involved in the emergency response/recovery.

2. 15 Consult with the ODAH on protective measures to be taken by Security Department personnel.

2. 16 Coordinate security activities with the Corporate Emergency Director/Recovery Hanager.

2. 17 Interface with the Work Control groups as appropriate to schedule and prioritize security-related activities and requirements.

2. 18 Interface with the Legal Department representative, as appropriate, on legal implications of and authorities in security-related activities.

2.19 If necessary, request and coordinate contractual off-site security assistance.

2.20 Periodically consult with Manager System Security, security personnel at Corporate Headquarters, and outside law enforcement agencies to determine and arrange for any additional security resulting from the emergency situation (e.g., potential protest demonstrations, telephoned security threats, etc.).

2.21 Upon request, secure the Administrative/Logistics NMPC helicopter Manager if the is not available to carry out this responsibility.

NOTE: Consult with the Environmental Sample/Survey Team Coordinator before requesting the helicopter so that radiological conditions at and in route to the helipad may be evaluated.

2.22 Upon request, provide personnel to allow NRC personnel access to the Loomis Center radio tower so that they may install portable radio equipment.

2.23 Assist the CED/RM as necessary in developing termination and/or recovery criteria as needed.

2.24 In conjunction with the Security Coordinator, develop long term staffing plans as necessary.

2.25 Collect paperwork developed during the emergency for later review and analysis.

2.26 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 39 EP IP-EPP-23 Rev 07

ATTACHNENT 22: EOF ADMINISTRATOR Sheet 1 of 2 1.0 RESPONSIBILITY The EOF Administrator is responsible for EOF setup, staffing, operations and equipment and coordinates these activities with the Administrative Logistics Manager (ALM).

2.0 ~CT IONS 2.1 Activate the EOF in accordance with EPIP-EPP-13.

2.2 Refer to Attachment 1, ERF General Actions.

2.3 Verify that sufficient numbers of secondary responders are available and are reporting to the emergency facility by reviewing the fax from Community Alert Network (CAN) located at the CAN designated fax.

2.4 Hake an announcement using the EOF PA system requesting that all EOF staff ensure they have registered at the EOF Registration Desk.

2.5 Assign EOF Technical Assistants to staff the EOF Technical Assessment Room:

~ To act as an assistant to the CED/RH

~ To staff the EOF/TSC/JNC communications

~ To post information 2.6 Call the NLC Receptionist (x2080) and inform them to complete Attachment 10 of EPIP-EPP-13.

2.7 Assign Technical Liaison to the Oswego County and New York State Emergency Operations Centers. Any NHPC employee that has plant specific knowledge of the affected unit may be used. A job aid with directions to both facilities is available at the EOF Administrators desk.

2.8 Ensure communication notifications with outside agencies are transferred to the EOF and maintained as per EPIP-EPP-20.

2.9 Periodically evaluate status boards for technical accuracy.

2.10 Supply personnel to act as Technical Briefers to the Joint News Center.

2. 11 When members of the NRC arrive during an emergency situation, notify the Corporate Emergency Director/Recovery Manager and escort the NRC Team to a conference room for a briefing. Utilize Attachment 22, Figure 1, "EOP Administrator Guidelines for NRC and Offsite Agency Interface".

December 1997 Page 40 EPIP-EPP-23 Rev 07

ATTACHMENT 22 EOF ADMI S 0 Page 2 of 2 2.12 Obtain support from computer support personnel for equipment problems.

2.13 Collect paperwork developed during the emergency for later review and analysis.

2. 14 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 41 EPIP-EPP-23 Rev 07

~

ATTACHMENT 22

~GU EOF ADNINISTRATOR GU LINES FOR N C 0 SITE AGENC I ER CE NOTE: This guideline refers to NRC, County or State (hereafter referred to as NRC/Offsite) emergency response personnel.

1. When directed by the CED, assign a contact person to meet the needs of NRC/Offsite personnel responding to the EOF.

2. Assign that contact person to complete the remainder of this guideline.

NOTE: The remainder of this guideline is to be completed by the Technical Assessment offsite contact person.

3. Introduce yourself and the EOF Plant/Technical Assessment Team to NRC/Offsite personnel.

4. Request that any questions or concerns be directed to you.

NOTE: It is acceptable for the NRC/Offsite personnel to ask questions of the tech assessment staff. Tech assessment staff may answer any questions they feel appropriate.

5. Respond to any questions, requests for information or other needs as requested by NRC/Offsite.

6. Verify that NRC/Offsite personnel are aware of emergency classification changes and significant changes in plant conditions.

December 1997 Page 42 EPIP-EPP-23 Rev 07

ATTACHHENT 23: OFF-SITE DOSE ASSESSHENT HANAGER T.B BEEP N IBILITT The Off-Site Dose Assessment Nanager (ODAN) is responsible for managing the off-site dose assessment aspects of an emergency to determine radiological consequences and hazards to the general public for the purpose of protective action recommendations.

2. 0 ~CTI 0NS 2.1 Activate the EOF as necessary in accordance with EPIP-EPP-13.

2.2 Refer to Attachment 1, ERF General Actions.

2.3 Verify that sufficient numbers of secondary responders are available and are reporting to the emergency facility by reviewing the fax from Community Alert Network (CAN) located at the CAN designated fax.

2.4 Ensure the following positions are filled:

~ One Radiological Assessment staff member

~ Heteorological Advisor

~ Environmental Survey Sample Team Coordinator (ESSTC) 2.5 Obtain a briefing from the CED/RH, RAH, and the control room chemistry technician.

2.6 Coordinate the staffing of the HPN line with the RAH.

2.7 Implement EPIP-EPP-08.

2.8 Direct the Dose Assessment staff to maintain radiologically status boards as needed.

2.9 Direct an RP Tech to perform periodic radiological surveys as necessary.

2. 10 Continually update the CED/RH on dose assessment activities and protective action recommendations.

2.11 Provide updated Part II Notification Fact sheets at approximately 30 minute intervals to the Communications Coordinator.

2. 12 Coordinate dose projection activities with New York State and Oswego County representatives in the EOF.

2. 13 Interface with offsite agency personnel as directed by the CED using Attachment 23, Figure 1, "ODAH Guidelines for NRC and Offsite Agency Interface", as a guide.

2. 14 Haintain hard copies of status board updates, dose calculations, meteorological data and downwind survey team results for later review and analysis.

2.15 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 43 EP IP-EPP-23 Rev 07

ATTACHMENT 23 FIGURE 1 ODAN GUIDELINES FOR NRC AND OF SITE AGENCY INTERFACE NOTE: This guideline refers to NRC, County or State (hereafter referred to as NRC/Offsite) emergency response personnel.

1. When directed by the CED, assign a contact person to meet the needs of NRC/Offsite personnel responding to the EOF.

2. Assign that contact person to complete the remainder of this guideline.

NOTE: The remainder of this guideline is to be completed by the Assessment offsite contact person.

3. Introduce yourself and the EOF Dose Assessment Team to NRC/Offsite personnel.

4. Request that any questions or concerns be directed to you.

NOTE: It is acceptable for the NRC/Offsite personnel to ask questions of the dose assessment staff. Dose assessment staff may answer any questions they feel appropriate.

5. Respond to any questions, requests for information or other needs as requested by NRC/Offsite.

6. Resolve differences in NRC/Offsite dose projections or protective actions.

7. Verify that NHPC dose projections, downwind survey team results, meteorology forecasts and source term data are provided to NRC, County and State.

December 1997 Page 44 EPIP-EPP-23 Rev 07

ATTACHHENT 24 JOINT NEWS CENTE I ECTO Sheet 1 of 2

1. 0 RESPONSIBILITY The Joint News Center Director (JNC Director) is responsible for preparing news releases, coordinating all outgoing public information, ensuring news releases are reviewed and approved by the CED/RH or SED as appropriate, and ensuring news releases are provided timely and accurate to public officials, the press and the general public.

2.0 ACTIONS

2. 1 Notify appropriate personnel within your department of the situation at NHPNS and any actions to be taken. (Use PACC On-Call schedule).

2.2 Report to the Joint News Center (JNC) when notified.

2.3 Verify that sufficient numbers of secondary responders are available and are reporting to the emergency facility by reviewing the fax from Community Alert Network (CAN) located at the CAN designated fax.

2.4 Activate the JNC in accordance with EPIP-EPP-27 2.5 Refer to Attachment 1, ERF General Actions as appropriate.

2.6 Contact the Site Emergency Director or TSC/EOF Liaison in the TSC and receive a briefing on initial accident conditions.

2.7 Establish and maintain communications with the Vice President PACC Department and keep him informed on the status of the emergency.

2.8 Establish and maintain coordination with the Corporate Emergency Director/Recovery Hanager directly or through the EOF-JNC Liaison and ensure that all press releases are reviewed and approved.

2.9 Haintain coordination with the EOF-JNC Liaison located in the EOF.

2. 10 Assist in the preparation of news releases.

2. 11 Ensure a copy of every news release is sent to the PACC offices in Syracuse.

2. 12 Implement actions in accordance with EPIP-EPP-27.

2. 13 Ensure that the Joint News Center, Hedia Response and Rumor Control Programs are being activated for an Alert, Site Area Emergency or General Emergency.

December 1997 Page 45 EPIP-EPP-23 Rev 07

ATTACHNENT 24 JOINT NEMS CENTER D EC OR Sheet 2 of 2

2. 14 Establish contact and coordinate activities with both State and local Public Information Officers (PIOs).

2. 15 Develop, as soon as possible, a schedule for press briefings.

2. 16 Contact EOF administrator and request Technical Briefer be sent to the Joint News Center.

2. 17 Contact ODAM and request a Rad Briefer be sent to the Joint News Center.

2. 18 Ensure legal department representative is available for providing consultation regarding public information.

2.19 If possible, periodically arrange for a knowledgeable senior company official to attend press conferences (e.g., Corporate Emergency Director/Recovery Manager).

2.20 Develop long term staffing plans as necessary for the JNC staff.

2.21 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 46 EPIP-EPP-23 Rev 07

ATTACHNENT 25 EOF-JNC LIAISO

1. 0 RESPONSIBILITY The EOF-DNC Liaison is responsible for coordinating all outgoing information and ensuring news releases are provided to the CED/RH or SED (as appropriate).

2.0 ACTIONS

2. 1 Refer to Attachment 1, ERF General Actions.

2.2 Contact the TLAH or EOF Technical Assistants and receive a briefing on initial accident conditions.

2.3 Establish and maintain communications with the JNC staff and keep them informed of the status of the emergency.

2.4 Establish and maintain coordination with the Corporate Emergency Director/Recovery Hanager to ensure review and approval of all press releases.

2.5 Assist in the preparation of news releases:

a. Ensure information to be released to the public has been reviewed by the TLAH and is both technically accurate and easily understandable.

b. If possible, ensure a representative of the Legal Department reviews all news releases to guard against legal or insurance problems, as necessary.

2.6 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 47 EP IP-EPP-23 Rev 07

ATTACHMENT 26 ENVIRON E L SURVEY SAMP COORD INTO Sheet 1 of 2 1.0 RESPONSIBILITY The Environmental Survey/Sample Team Coordinator is responsible for providing technical and administrative direction to environmental monitoring teams during a declared emergency, and assisting in the evaluation of on-site and off-site dose assessment aspects of an emergency to determine potential or actual radiological impacts to site personnel and the general public based on environmental measurements.

2.0 ACTIONS 2.1 As necessary, activate the EOF in accordance with EPIP-EPP-13.

2.2 Refer to Attachment 1, ERF General Actions.

2.3 Obtain a briefing as to plant conditions, radiological data and other information as appropriate.

2.4 Ensure proper use of communications equipment in accordance with EPIP-EPP-17.

2.5 Ensure exposure control is in accordance with EPIP-EPP-15.

2.6 Interface with the ODAM for corrective actions in progress and for projected off-site doses to the public based on the type of accident; 2.7 Interface with the ODAM to discuss a survey strategy that would verify projected off-site doses.

2.8 Assign personnel to perform environmental monitoring as directed by Radiological Assessment Manager per guidance provided in EPIP-EPP-07. Priorities for assignment will depend on plant conditions; the following order of tasks is provided as a guide:

Dose Rate Confirmation - EPIP-EPP-07 Off-Site Monitoring EPIP-EPP-07 and EPIP-EPP-16 Monitoring of Evacuating Vehicles and Per sonnel-EP IP-EPP-19 2.9 Establish communications with environmental (downwind) survey teams. Assess their availability and location. indicate survey team locations on maps provided.

2. 10 Provide appropriate precautions and directions on expected or potential hazards, protective clothing requirements, and exposure control (per EPIP-EPP-15, "Health Physics Procedure" ).

December 1997 Page 48 EP IP-EPP-23 Rev 07

ATTACHNENT 26 ENVIRONNENT L SURVEY SANPLE TEAN COORDINATOR Sheet 2 of 2 2.11 Provide data to the ODAH for dose projections. Ensure all data received is logged on status boards.

2. 12 Ensure survey teams are briefed periodically on plant conditions (use discretion so as not to alarm the public).

2. 13 Notify downwind teams as soon as you know that a release has occurred.

2. 14 Coordinate environmental monitoring activities with local, state and federal agencies.

2. 15 Ensure that the EOF radio operator is recording all data reported by the survey teams on the Survey Team Report form.

2. 16 Ensure that data received from the survey teams is being transmitted to the TSC.

2.17 Provide copies of survey team report data logged on the status board sheet to county, state and federal personnel located in the EOF as well as the ODAN and public information personnel.

2. 18 Periodically update instructions to the survey teams as new information becomes available.

2. 19 Ensure that meteorological data is being posted on status boards and survey maps. Ensure forecasts are being obtained.

2.20 Provide administrative and technical direction to the re-entry teams in accordance with EPIP-EPP-12.

2.21 Retain for inclusion in the Permanent Plant File records generated as a result of an actual declared emergency.

December 1997 Page 49 EPIP-EPP-23 Rev 07

ATTACKNENT 27 CONT OL ROON INFORSL ION LIAISON 1.0 RESPONSIBILITIES The Control Room Information Liaison is responsible for providing the Emergency Response Facilities (ERF) with plant conditions/events, systems status, and operator responses and actions. This position reports to the TSC Technical Data Coordinator.

2.0 ~ctions NOTE: The purpose of the Control Room Information Liaison is for the transmission of technical data only. This position should not be used for "command and control" activities, requests for action or Communications Aide activities.

2.1 Enter affected control Room and inform the SSS that the Control Room Information Liaison position is now staffed.

2.2 Inform the Technical Data Coordinator in the TSC that the Control Room Communicator position is now staffed. (This will normally be done through the EOF/TSC/CR Liaison) 2.3 Establish and maintain communications with the following, using the Tech Information Line or telephone:

~ TSC (EOF/TSC/CR Liaison)

~ EOF (Tech Assistant) 2.4 Provide plant status/events, systems status, alarms, and operator responses/actions to all ERFs as they occur or as requested.

2.5 Respond to any requests for information from the ERFs.

2.6 Retain for inclusion into the Permanent Plant File records generated as a result of an actual declared emergency..

December 1997 Page 50 EPIP-EPP-23 Rev 07

NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT NUCLEAR STATION EMERGENCY PLAN IMPLEMENTING PROCEDURE PIP-EPP-24 REVISION 01 NUC EAR TRANSPORTATION ACCIDENTS TECHNICAL SPECIFICATION REQUIRED Approved by N. L. Rademacher Plant Manager - Unit I Date Approved by:

J.~ T. Conway

~ Pl Manager - Unit 2 Date PERIODIC REVIEM, 07/17/97, NO CHANGE PERIODIC REVIEW,07/22/96,NO CHANGE Effective Date: 04/0 PERIODIC REVIEW DUE DATE

LIST OF EFFECTIVE PAGES

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March 1996 Page i EPIP-EPP-24 Rev Ol

A TAB E OF CONTENTS SECTION PAGE 1.0 PURPOSE . 1

2.0 REFERENCES

AND COMMITMENTS . . . . . . . . . . . . . . . . . . . 1 3.0 DEFINITIONS........................... t 2

4.0 RESPONSIBILITIES . . . . . . . . . . . . . . . . . . . . . . . . 7 5.0 PRECAUTIONS........................... 8 6.0 LIMITATIONS AND ACTIONS...................... 9 7.0 PREREQUISITES.......................... 9 8.0 PROCEDURE . ~ ~ ~ 9

8. 1 Accident Notification ~ ~ ~ 9 8.2 Accident Response 10 8.3 PACC . ~ ~ ~ ~ ~ 16 8.4 Response Outside NMPC Geographica1 Boundary 16 9.0 ACCEPTANCE CRITERIA . . . . . . . . . . . . . . . . . . . . . . . 19 10.0 RECORD REVIEW AND DISPOSITION.................. 19 ATTACHMENT 1: VOLUNTARY ASSISTANCE AGREEMENT BY AND AMONG ELECTRICAL UTILITIES INVOLVED IN TRANSPORTATION OF NUCLEAR MATERIALS 21 FIGURE 1: NUCLEAR TRANSPORTATION ACCIDENT REPORT FORM ......... 29 FIGURE 2: NUCLEAR TRANSPORTATION ACCIDENT (SSS OR CSO CHECKLIST).... 30 FIGURE 3:

NUCLEAR TRANSPORTATION ACCIDENT CONTROL ROOM COMMUNICATIONS CHECKLIST AID 31 FIGURE 4: NUCLEAR TRANSPORTATION ACCIDENT SECURITY CHECKLIST...... 33 ENCLOSURE 1: INPO ADDRESS AND TELEPHONE . . . . . . . . . . . . . . . . 34 March 1996 Page ii EPIP-EPP-24 Rev Ol

1.0 PURPOSE This procedure provides guidelines for NHPNS personnel response to an off-site nuclear transportation accident.

2.0 REFERENCES

AND COMHITHENTS 2.1 Technical S ecifications None 2.2 Standards Re ulations and Codes 2.2.1 10CFR20, Standards for Protection Against Radiation 2.2.2 10CFR30, Rules of General Applicability to Domestic Licensing of Byproduct Haterial 2.2.3 10CFR40, Domestic Licensing of Source Haterial 2.2.4 10CFR50, Domestic Licensing of Production and Utilization Facilities 2.2.5 10CFR71, Packaging and Transportation of Radioactive Haterial 2.2.6 49CFR171, General Information, Regulations, and Definitions Hazardous Haterials Regulations.

2.2.7 49CFR173, Shippers-General Requirements for Shipments and Packages 2.2.8 40CFR302, Designation-Reportable quantities and Notification 2.2.9 6-NYCRR, Rules and Regulations for Protection and Control of Environmental Pollution by Radioactive Haterial 2.3 Policies Pro rams and Procedures 2.3.1 Fire Command Lesson Plan, NHP Nuclear Training Department 2.3.2 Oswego County Radiological Emergency Response Plan and Procedures 2.3.3 New York State Radiological Emergency Preparedness Plan 2.3.4 EPIP-EPP-27, Emergency Public Information Procedure 2.3.5 EPIP-EPP-04, Personnel Injury or Illness 2.3.6 EPIP-EPP-07, Downwind Radiological Honitoring Harch 1996 Page 1 EPIP-EPP-24 Rev Ol

2.3.7 EPIP-EPP-16, Environmental Monitoring 2.3.8 EPIP-EPP-20, Emergency Notifications 2.4 Technical Information 2.4. 1 "A review of the Department of Transportation (DOT)

Regulations for Transportation of Radioactive Materials",

USDOT, August, 1976 2.4.2 INPO, "Voluntary Assistance Agreement By and Among Electrical Utilities Involved in Transportation of Nuclear Material", September 14, 1982 2.5 Commitments Sequence Commitment Number Number Descri tion NCTS 3122-12 Responders may report directly to the accident scene.

NCTS 3122-9 Consider the use of tarps to control contamination spread.

NCTS 3122-11 Incorporate applicable Incident Command System information.

NCTS 3122-8 Incorporate applicable NY State procedure information.

NCTS 3122-13 Describe what non-NMP organizations expect of NMP responders 6 NCTS 3122-7 It may be helpful to converse with the truck driver.

NCTS 3122-10 Verify all packages are accounted for.

NCTS 3122-16 Provide more detailed guidance for PACC members at the accident scene.

3.0 DEFINITIONS 3.1 A

-1 The maximum activity of special form radioactive material permitted in a Type A package.

March 1996 Page 2 EPIP-EPP-24 Rev Ol

3.2 The maximum activity of radioactive material, other than special form or low specific activity radioactive material, permitted in a Type A package.

3.3 A reement State Any state with which the Nuclear Regulatory Commission has entered into an effective agreement regarding licensing. "Non-agreement State" means any other state.

3.4 Airborne Radioactive Material Any radioactive material dispersed in the air in the form of dusts, fumes, mists, vapors or gases.

3.5 B roduct Material Any radioactive material (except Special Nuclear Material) made radioactive by exposure to the radiation incident to or yielded in the process of producing or utilizing Special Nuclear Material (SNM). The majority of the radwaste produced by nuclear power utilities falls within this category.

3.6 Carrier A person engaged in the transportation of passengers or property by land or water as a common, contract, or private carrier or by civil aircraft.

3.7 Closed Trans ort Vehicle A transport vehicle equipped with a securely attached exterior enclosure that during normal transportation restricts the access of unauthorized persons to the cargo space containing the radioactive materials. The enclosure may be either temporary or permanent and in the case of packaged materials may be of the "see-through" type and must limit access from top, sides, and ends.

3.8 Curie That amount of radioactive material which disintegrates at the rate of 37 billion atoms per second.

March 1996 Page 3 EPIP-EPP-24 Rev 01

3.9 Exclusive Use Also referred to as "sole use" or "full-load" The sole use of a conveyance by a single consignor, for which all initial, intermediate, and final loading and unloading are carried out in accordance with the direction of the consignor or consignee. Specific instructions for maintenance of exclusive use shipment controls must be issued in writing and included with the shipping paper information provided to the carrier by the consignor.

3.10 Fissile Mate i l Any material (except natural or depleted uranium) consisting of or containing one or more of the fissile radionuclides. Fissile radionuclides are plutonium-238, plutonium-239, plutonium-241, uranium-233 or uranium-235.

3.11 Hi h Inte rit Container HIC A container designed to prevent the egress of its contents in a land burial environment for approximately 300 years.

3.12 Hi hwa Route Controlled uant't A single package quantity which exceeds 3,000 times the A, value (special form), 3,000 times the Az value (normal form), or 30,000 curies, whichever is least. A, Az Tables, 49CFR173.435.

3.13 Licensed Material Source material, special nuclear material, or by-produce material received, possessed, used, or transferred under a general or specific license issued by the U. S. Nuclear Regulatory Commission.

3.14 Limited uantit A quantity of radioactive material not exceeding the materials package limits of 49 CFR 173.423 and which conform with the requirements in 49CRF173.421.

3.15 Low S ecific Activit LSA The definition is broad in scope and should be referred to as stated in 49CFR173.403(n).

March 1996 Page 4 EP IP-EPP-24 Rev Ol

3. 16 NMPC Geo ra hic l Bo nd ries The figure below shows typical NHPC Nuclear Transportation Accident Response boundaries. The bold dark line represents these boundaries.

The area within the bold lines is the area to which NMPC will respond.

C 3.17 Non-fixed Radioactive Contamination Radioactive contamination that can be readily removed from a surface by wiping with an absorbent material. See 49CFR173.433 for limits.

3.18 Normal Form Radioactive material which has not been demonstrated to qualify as "special form radioactive material".

3.19 ~Packa e The packaging together with its radioactive contents as presented for transport.

3.20 Packacaing For radioactive materials, the assembly of components necessary to ensure compliance with packaging requirements.

3.21 Radioactive Article Any manufactured instruments and articles such as an instrument, clock, electronic tube or apparatus, or similar instrument and article having radioactive material as a component part.

March 1996 Page 5 EP IP-EPP-24 Rev 01

3.22 Radioactive Mater'al Any material having a specific activity greater than 0.002pCi/g.

3.23 Source Material (1) Uranium or thorium, or any combination thereof, in any physical or chemical form, or (2) ores which contain by weight one-twentieth of one percent (0.05 percent) or more of (a) uranium, (b) thorium, or (c) any combination thereof. Source material does not include special nuclear material (SNM).

3.24 S ecial Form A radioactive material that would present a direct radiation hazard but its little hazard package.

of contamination or radiotoxicity if released from Special form is usually a solid piece or a sealed encapsulation and must meet the criteria established in 49CFR173.469.

3.25 S ecial Nuclear Material SNM (1) Plutonium, uranium-233, uranium-enriched in the isotope-233 or the isotope-235, or (2) any material artificially enriched by any of the foregoing. This definition does not include source material.

3.26 Trans ort Index The dimensionless number (rounded up to the first decimal place) placed on the label of a package to designate the degree of control to be exercised by the carrier during transportation. The transport index is to be determined as follows:

The number expressing the maximum radiation level in millirem per hour at one meter from the external surface of the package. For example, at one meter 1.04 mrem/hr would become a Transport Index of 1.1.

3.27 T e A Packa in Packaging designed to retain the integrity of containment and shielding under normal conditions of transport as demonstrated by tests set forth in 49CFR173.465, 173.466, and 49CFR173.412.

3.28 T e B Packa in Packaging that meets the requirements of Type A packaging and additionally meets hypothetical accident test conditions on 10CFR71.

March 1996 Page 6 EPIP-EPP-24 Rev Ol

4.0 RESPONSIBILITIES 4.1 Station Shift Su ervisor SSS - Unit 4.1.1 Assumes overall responsibility for all off-site emergencies.

4.1.2 Notifies station emergency response teams.

4.1.3 Maintains knowledge regarding the current status of off-site emergency response actions taken by station personnel.

4.2 Station Shift Su ervisor SSS Chief Shift 0 erator CSO Unit 2 4.2.1 Any notification of an Offsite Nuclear Transportation Accident shall be immediately referred to the. Unit 1 SSS or CSO.

4.3 Chief Shift 0 erator CSO - Unit 1 4.3.1 Assumes overall responsibility for completion of the Nuclear Transportation Accident Form (Figure 1).

4.3.2 Forwards the Nuclear Transportation Accident Form (Figure 1) to the Unit 1 SSS.

4.4 Su ervisor Radwaste 0 erations 4.4.1 Coordinates the safe and efficient conduct of station related radioactive waste operations.

4.4.2 Schedules, coordinates, and supervises radioactive waste shipments.

4.4.3 Directs and supervises work of station radwaste operators.

4.4.4 Assists Unit Supervisor RP in determining material content and assessing the radiological consequences and actions pertaining to a nuclear transportation accident.

4.4.5 Provides advice and assistance regarding contact of off-site organizations associated with the packaging and shipment of the radioactive material.

4.4.6 Assists Manager RP in accident recovery activities as necessary,incl.uding cleanup, repackaging, relabeling, shipping, and agency notification.

4.5 Mana er Radiation Protection RP 4.5.1 Dispatches a station team to the accident site.

4.5.2 Evaluates the radiological consequences of an off-site accident and the effect on the off-site general population.

March 1996 Page.7 EPIP.-EPP-24 Rev Ol

4.5.3 Provides advice and support to medical personnel regarding contaminated wounds.

4.5.4 Provides advice and support for radiological activities as requested by off-site authorities.

4.5.5 Dispatches representatives to the hospital to provide hospital personnel with pertinent information regarding contamination, if necessary.

4.5.6 Assists in accident recovery activities as necessary, including cleanup, repackaging, relabeling, and restoration of the accident scene to preaccident status.

4.6 Radiation Protection Technicians 4.6.1 Respond as directed by the Unit 1 SSS or the Hanager RP, or designee.

4.6.2 Provide radiation protection advice and support as directed by the team leader.

4.7 INPO Dut Officer 4.7.1 Coordinates requests for assistance in locating emergency manpower and equipment among signatories of the INPO, "Voluntary Assistance Agreement By and Among Electrical Utilities Involved in Transport of Nuclear Haterials".

4.7.2 Disseminates information to agreement signatories concerning the incident as applicable to operation.

4.6.3 Organizes industry experts and advises signatories of agreement.

5.0 PRECAUTIONS 5.1 Hazardous substances other than radioactive material may be present at the accident scene including flammable solids, liquids or gasses; poisons; corrosives; compressed gases; reactive or toxic chemicals; irritants and biological agents (biohazards).

5.2 These other substances may pose a hazard to emergency response personnel, either through direct exposure or by their interactions with each other or with packages of radioactive materials.

5.3 A substance may be both radioactive and corrosive, flammable or toxic.

5.4 Other classes of hazardous material present problems that are of more immediate concern and danger than radioactive materials.

5.5 Be alert to the presence of hazardous substances and take extra precautions during response operations.

Harch 1996 Page 8 EPIP-EPP-24 Rev. Ol

6.0 IM TA IONS N ACTIONS 6.1 The Unit 1 Control Room staff shall take the lead in all notifications, requests, etc. relating to a nuclear transportation accident.

6.2 All news media inquiries should be referred to:

6.2.1 Public Affairs and Corporate Communication Department (PACC) personnel present at the accident site; OR 6.2.2 PACC at the Syracuse office of Niagara Mohawk Power Corporation.

6.3 It is not necessary to execute steps or actions in the order listed to successfully perform this procedure.

7.0 PRERE UISITES None 8.0 PROCEDURE 8.1 Accident Notification 8.1.1 Any notification of an off-site nuclear transportation accident will be immediately referred to either the Unit 1 Station Shift Supervisor (SSS) or the Unit 1 Chief Shift Operator (CSO).

NOTE: If accident is outside NHPC's geographical boundaries, refer to Section 8.4 of this procedure for special considerations.

8.1.2 The Unit 1 CSO or SSS should:

a. Obtain as much information as possible.

b. Record pertinent information on Part B of the Nuclear Transportation Accident Report Form (Figure 1).

c. Record the time Part B of Figure 1 was completed.

d. Sign Part B of Figure l.

e. Upon completion of Part B, forward Figure 1 to the SSS.

March 1996 Page,9 EPIP-EPP-24 Rev Ol

8.1.3 The Unit 1 SSS should:

a. Complete the Nuclear Transportation Accident SSS or CSO Checklist (Figure 2).

b. Assess the nature of the accident and the request for assistance.

c. Provide the caller with information regarding station response.

d. Contact the Supervisor, Radwaste Operations and the Manager RP.

1. During off-hours, use the on-call list for the Radiation Protection Department.

2. Transmit accident information.

3. Request the dispatch of a response team, with appropriate equipment, to the accident scene.

8.1.4 The response team should consist of a MINIMUM of:

a. One Supervisor RP

b. One Supervisor Radwaste Operations

c. A minimum of One RP Technician 8.1.5 During off-hours, the Supervisor RP should call in (COMM 1) additional off-duty station personnel (i.e., RP technicians) to either the station or the accident scene directly, as necessary.

8.2 Accident Res onse 8.2.1 The Manager RP and Supervisor Radwaste Operations should:

'a ~ Designate the response team leader.

b. Arrange for station response team(s) to meet.

8.2.2 When possible, station response personnel should:

(COMM 1)

'a ~ Meet at the station.

b. Use the On-Call Emergency Response Vehicle to proceed to the accident location.

March 1996 Page 10 EPIP-EPP-24 Rev Ol

8.2.3 The response team leader should:

a~ Provide direct supervision to the station response team.

b. Act as liaison with non-station emergency response personnel.

C. Identify station response personnel to non-station emergency response personnel.

d. Coordinate station personnel assistance.

e. Ensure station personnel response is conducted safely.

'eport accident status information to the Unit I SSS as soon as possible and as often as necessary.

Discuss station personnel emergency response activities with the Unit I SSS.

8.2.4 The response team should:

a ~ Ensure equipment is available in the On-Call Emergency Response Vehicle, including as a minimum the supplies and equipment necessary to evaluate and monitor the situation and to provide adequate personnel protection for team members.

b. Obtain additional equipment, as necessary, from normal station or off-site emergency facilities.

1. Consider information known about the accident, such as material involved, material packaging, weather, etc.

(COMM 2) 2. Consider equipment which might be useful in mitigating the consequences of the accident, such as:

Pail s Shovels Plastic sheeting Tarps Plastic bags Protective equipment March 1996 Page ll EPIP-EPP-24 Rev Ol

8.2.4 (Cont)

c. Maintain an incident log recording significant response related activities.

CAUTION Upon approaching the accident scene, special attention

~US be given to special circumstances such as fire, chemical, or multiple hazards.

d. Proceed to the accident location safely, obeying all traffic regulations.

8.2.5 When first to arrive at the accident scene, the response team should:

a~ Provide emergency medical first aid, as necessary.

b. Secure the scene of the accident.

C. Ensure traffic will not present a further hazard.

d. Rescue of the injured and life-saving first aid take precedence over radiological hazards.

1. Accomplish rescue safely following common-sense guidelines (i.e., time, distance, shielding).

2. Perform rescue and first aid based on individual qualifications (i.e., knowledge of first aid techniques) of responding personnel.

e. If there is a fire or a high probability of a fire starting:

l. Establish a "control zone" of at least TWO HUNDRED (200) FEET from debris.

2. Clear personnel at least 2000 feet from the downwind direction.

If movement can be accomplished without undue risk to personnel, move undamaged packages of radioactive material from fire or corrosive acid areas.

March 1996 Page 12 E PIP-EPP-24 Rev Ol

8.2.5 (Cont)

• • • • • • • • • • • • • • • • • • • • • • • • • • CAUTION Be alert for fumes, smoke, and irritating or noxious odors.

g. When possible, identify the presence of hazardous material cargo:

l. Observe package labels.

2. Observe United Nations identification number and vehicle placards.

NOTE: Do NOT rely entirely on vehicle placards since certain materials do not require placarding. Also, two or more classes of hazardous materials (other than radioactive) may be identified by a single "DANGEROUS" placard instead of by individual placards for each hazard class.

3. Use radiological survey instruments.

4. Review all available shipping documents.

8.2.6 Incident Command System (CONN 3)

a. System commonly used by US emergency response organizations (i.e., police, fire companies, nuclear plant emergency response personnel).

b. Purpose is the efficient, effective mitigation of emergency consequences.

c. Facilitates cooperative emergency response effort by establishing a universally accepted system for:

1. Communication

2. Command hierarchy

3. Response organization

d. Basic approach that can be applied to any type of emergency.

March 1996 Page 13 EPIP-EPP-24 Rev 01

8.2.6 (Cont)

e. Terminology

l. Incident Commander a) Individual in charge of the accident site and accident related activities.

b) Normally the highest ranking local fire official present.

(COHH 4) c) The County Director of Emergency Hanagement has the lead responsibility for directing incident mitigation unless the governor declares a state of emergency.

2. Command Post a) Location determined by Incident Commander b) Hay be located at accident scene or remotely c) Generally contains communication abilities

3. Sector NOTE: Incidents are subdivided into sectors to more easily manage mitigation activities.

a) Distinctive aspect of the incident b) Examples of sectors

~ Hedical assistance

~ Traffic control

~ Accounting for shipment packages

~ Specific geographic area

4. Sector Commander a.. Designated by Incident Commander

b. Responsible for designated sector and informing Incident Commander of sector related status

-Harch 1996 Page 14 EPIP-EPP-24 Rev Ol

8.2.7 On arrival at an accident scene already being controlled by (COMM 5) local or government officials the team should:

a. Report to the Incident Commander

b. Provide technical assistance as requested

c. Perform environmental radiological monitoring or sampling in accordance with EPIP-EPP-07, Downwind Radiological Monitoring and EPIP-EPP-16, Environmental Monitoring, as appropriate 8.2.8 The Supervisor Environmental Protection may direct environmental monitoring activities.

8.2.9 The response team should:

a 0 Obtain as much information as possible. When appropriate:

1. Discuss the situation with the individual directing emergency response at the scene.

(COMM 6) 2. Discuss the event with the driver of the vehicle transporting radioactive material.

3. Review the shipping documentation carried by the driver.

(COMM 7) 4. Account for packages listed in the shipping documentation.

5. Observe integrity of the packages.

b. Attempt to determine:

1. Name/address/phone number of the "shipper" and "carrier"

2. Origin of shipment

3. Weight of entire assembly or load

4. Nature and quantity of material

5. Destination of shipment C. Record information from Step 8.2.9.b on the incident log.

March 1996 Page 15 EPIP-EPP-24 Rev Ol

8.2. 10 The Unit 1 SSS or CSO should:

a. Complete Figure 1, Part B.

b. Attempt to contact the shipper (material sender) and the carrier (material Transport Company):

1. For notification of the incident and actions taken

2. To obtain more complete information concerning the potential hazard from the material

c. Immediately report appropriate information to the response team leader at the accident scene.

d. Record significant actions of station personnel in the Station Log.

8.2.11 When the response team leader determines that the accident scene is properly controlled by responsible off-site authorities, station personnel may return to the station.

8.3 PACC (COHH 8)

Public Affairs and Corporate Communication (PACC) personnel at the accident site should:

8.3.1 Obtain as much information as possible about the accident including current status, consequences, and mitigation.

8.3.2 Forward information to the PACC Department in Syracuse.

8.3.3 Handle new media requests as directed by the Hanager, Nuclear Communications and Public Affairs or designee.

8.4 Res onse Outside NHPC Geo ra hical Boundar 8.4.1 Direct station response in accordance with the following steps in addition to those tasks already described in this procedure.

8.4.2 These steps apply to station response as governed by INPO's Voluntary Assistance Agreement and are independent of accident material ownership.

8.4.3 Special consideration will be given to material originating from the station. Station assistance will generally not be provided outside NHPC's geographical boundary for non-station material.

8.4.4 Station Shift Supervisors (Unit 1 and Unit 2), the Hanager RP, and the Supervisor Radwaste Operations make the decision to have station emergency personnel respond to the accident.

Harch 1996 Page 16 EPIP-EPP-24 Rev Ol

8.4.5 Call the INPO Duty Officer for assistance.

NOTE: All legal and financial requirements will be satisfied in accordance with INPO's "Voluntary Assistance Agreement By and Among Electrical Utilities Involved in Transportation of Nuclear Materials", dated September 14, 1982, and any other agreements entered into by NMPC.

8.4.6 The Unit 1 SSS and CSO shall question the INPO Duty Officer providing the accident notification to determine:

a ~ If the accident location is within the geographical boundary of one of the signatories of the INPO "Voluntary Assistance Agreement By and Among Electrical Utilities Involved in Transportation of Nuclear Material".

b. If the shipper is one of the signatories of the Voluntary Assistance Agreement (refer to Figure 1, Part B).

8.4.7 When a Voluntary Assistance Agreement signatory is involved:

a ~ The organization requesting assistance will be referred to as the "Requesting Company".

b. The organization providing assistance will be referred to as the "Responding Company".

C. The Unit 1 SSS or CSO shall obtain from the INPO Duty Officer the names and phone numbers of signatory contacts and provide this information to the Unit 1 Manager RP.

d. The SSS, the Manager RP, and the Supervisor Radwaste Operations shall discuss the accident details and determine tentative station response (or assistance request) recommendations.

e. The Manager RP, or designee, shall contact INPO and discuss accident details.

f. Assistance rendered shall be voluntary.

g. Determine emergency response assistance, either to or from the signatory and record on Figure 1, Part B.

h. INPO shall coordinate emergency response.

March 1996 ,Page 17 EPIP-EPP-24 Rev Ol

8.4.7 (Cont)

i. The "Requesting Company" through an INPO Duty Officer should:

Provide the "Responding Company" with a description of assistance requested and the anticipated duration for which such assistance is desired.

2. Provide general direction regarding actions to be taken by the "Responding Company".

3. Be responsible for making any report to governmental authorities and news media.

Notify the "Responding Company" when its assistance is no longer needed.

5. Inform the "Responding Company" of any specific equipment which may be required in the particular situation.

The "Responding Company" should:

Be responsible for determining the procedures to be followed in furnishing assistance.

2. Furnish the requested emergency personnel and equipment at its direction.

3. Have the right, at any time and in its sole judgment and discretion, to withdraw personnel and equipment furnished to the "Requesting Company" and return such personnel and equipment to their working base.

4. Give notice, through an INPO Duty Officer, to the "Requesting Company" of the withdrawal of

,. personnel or equipment furnished.

5. Make all arrangements for the transportation of its personnel and equipment from and to their working base or home.

6. Equip personnel with such normal working and protective equipment as shall be compatible with the circumstances under which said personnel shall function.

7. Keep time sheets and work records pertaining to "Responding Company" personnel and equipment.

March 1996 Page 18 EPIP-EPP-24 Rev 01

8.4.7.j (Cont)

8. Furnish the "Requesting Company" with a detailed statement of costs and expenses paid or incurred by the "Responding Company" in connection with the furnishing of assistance to the "Requesting Company".

k. Direct requests for assistance to INPO. Refer to INPO Address and Telephone (Enclosure 1).

INPO contacts "Responding Companies" to locate sources of emergency manpower and equipment on behalf of the "Requesting Company".

m. INPO shall not be responsible for implementing, enforcing, or interpreting any of the above guidelines.

8.4.8 The Manager RP or designee shall log significant (major) activities associated with emergency response.

9.0 CCEPTANCE CRITERIA None 10.0 RECORD REVIEW AND DISPOSITION The following records generated by this procedure as a result of an actual emergency reported at the Nine Mile Point Nuclear Station shall be maintained by Nuclear Records Management for the Permanent Plant File in accordance with NIP-RMG-Ol, "Identification, Maintenance, Storage, and Transfer of Nuclear Division Records":

Attachment 1, Voluntary Assistance Agreement By and Among Electric Utilities Involved in Transportation of Nuclear Materials, Counterpart Signature Page Attachment 1, Voluntary Assistance Agreement By and Among Electric Utilities Involved in Transportation of Nuclear Materials, Letter Confirming Requested Assistance Figure 1, Nuclear Transportation Accident Report Form Figure 2, Nuclear Transportation Accident SSS or CSO Checklist Figure 3, Nuclear Transportation Accident Control Room Communications Aid Checklist Figure 4, Nuclear Transportation Accident Security Checklist March 1996 Page 19 EPIP-EPP-24 Rev Ol

10.0 (Cont)

The following records generated by this procedure during Emergency drills or exercises are not required for retention in the Permanent Plant File:

Attachment 1, Voluntary Assistance Agreement By and Among Electric Utilities Involved in Transportation of Nuclear

~

Materials, Counterpart Signature Page Attachment 1, Voluntary Assistance Agreement By and Among Electric Utilities Involved in Transportation of Nuclear Materials, Letter Confirming Requested Assistance Figure 1, Nuclear Transportation Accident Report Form Figure 2, Nuclear Transportation Accident SSS or CSO Checklist Figure 3, Nuclear Transportation Accident Control Room Communications Aid Checklist Figure 4, Nuclear Transportation Accident Security Checklist March 1996 Page.20 EPIP-EPP-24 Rev Ol

ATTACHMENT 1 Page 1 of 8 VOLUNTARY ASSISTANCE AGREEMENT BY AND AMONG ELECTRICAL UTILITIES INVOLVED IN TRANSPORTATION OF NUCLEAR MATERIALS This Voluntary Assistance Agreement (hereinafter "Agreement" ) has been entered into by and among electric utilities involved in transportation of source material, special nuclear material, or byproduct material received, possessed, used or transferred under a general or specific license issued by the U.S. Nuclear Regulatory Commission pursuant to Title 10 of the Code of Federal Regulations (hereinafter "nuclear materials" ) and which have subscribed counterpart signature pages in the form attached hereto (hereinafter "Parties" ).

The Parties wish to set forth herein their understanding and agreement with respect to their mutual undertaking to each other in the situation wherein an emergency arises during the transportation of nuclear materials shipped by or on behalf of a Party and a request for assistance is issued to another Party in respect to such emergency and such assistance is provided.

This Agreement is intended only to define the terms and conditions under which such assistance is provided. This Agreement is intended only to define the terms and conditions under which such assistance, if volunteered, will be rendered and received. It is understood that this Agreement does not impose any obligation on any Party to render or continue to render any such assistance but this Agreement does record the understanding of the Parties with respect to the rights and obligations which will be incurred in responding to requests for assistance.

NOW, THEREFORE, it is agreed, that:

1. Assistance rendered by a Party as described hereunder shall be entirely voluntary and, when given in response to a request by any Party for help during an emergency arising by reason of the transportation of nuclear materials shall be rendered in accordance with the terms and conditions herein.

2. The Party that requests assistance shall be known as the "Requesting Company" and the Party furnishing assistance shall be known as the "Responding Company". Attachment A is a suggested letter confirming an agreement whereby assistance will be furnished pursuant to this Agreement.

3. (a) Requesting Company shall notify Responding Company of the type of assistance requested and the anticipated duration during which such assistance is desired. Requesting Company may also provide general direction as to the actions to be taken by Responding Company. Responding Company shall furnish such assistance as it may decide. Except as such companies may agree otherwise, Responding Company shall be responsible for determining the procedures to be followed in furnishing such assistance and for supervising wor k at the site of the emergency. Requesting Company, in cooperation with Responding Company, shall make any report to governmental authorities and the news media. Requesting Company will notify Responding Company when its assistance is no longer needed.

March 1996 Page 21 EP IP-EPP-24 Rev 01

ATTACHMENT I (Cont) Page 2 of 8 VOLUNTARY ASSISTANCE AGREEMENT BY AND AMONG ELECTRICAL UTILITIES INVOLVED IN TRANSPORT TION 0 UC A A ALS

3. (b) The furnishing of assistance hereunder shall be deemed to have commenced when personnel of the Responding Company are assigned to other than normal duties or transportation of equipment commences pursuant to a determination by the Responding Company to provide assistance to a Requesting Company under this Agreement and shall be deemed to have terminated when the transportation of such personnel or equipment back to their working base, or home (for personnel returning at other than regular working hours), is completed.

(c) The Responding Company shall make all arrangements for the transportation of its personnel and equipment from and to their working base or home.

4. (a) Employees of Responding Company shall at all times continue to be employees of and remain under the supervision and control of the Responding Company, including work procedures and/or safety rules, shall be those of the Responding Company.

(c) All personnel of the Responding Company shall be equipped by the Responding Company with such normal working and protective equipment as shall be compatible with the circumstances under which said personnel, shall function hereunder; Requesting Company shall inform Responding Company of any specific equipment which may be required in a particular situation.

5. (a) Responding Company shall furnish the requested personnel and equipment to the extent that the Responding Company may determine to do so at its sole judgement and discretion.

(b) Responding Company shall have the right, at any time and in its sole judgement and discretion, to withdraw personnel and equipment furnished to the Requesting Company and return such personnel and equipment to their working base. Responding Company shall give written notice at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in advance to Requesting Company of the permanent withdrawal of personnel or equipment furnished. Responding Company's withdrawal of personnel or equipment shall not affect any obligations which may have been incurred hereunder prior to such withdrawal or which may arise out of events occurring prior to such withdrawal.

6. All time sheets and work records pertaining to Responding Company personnel and equipment shall be kept by the Responding Company. The Responding Company shall furnish the Requesting Company with a detailed statement of all costs and expenses paid or incurred by the Responding Company in connection with the furnishing of assistance to the Requesting Company, which statement shall be paid by Requesting Company within thirty (30) days after receipt.

March 1996 Page 22 EPIP-EPP-24 Rev Ol

ATTACHMENT 1 (Cont) Page 3 of 8-VOLUNTARY ASSISTANCE AGREEMENT BY AND AMONG ELECTRICAL UTILITIES INVOLVED IN TRANSPORTATION OF NUCLEAR MATERIALS

7. The Requesting Company shall reimburse Responding Company for all direct and indirect costs and expenses, not including a profit, incurred by Responding Company in giving assistance pursuant to this Agreement, including but not limited to costs and expenses related to or resulting from compliance with governmental requirements such as Title 10 of the Code of Federal Regulations Part 20. Such costs and expenses shall be computed in accordance with Responding Company's standard rates and accounting practices including such overheads as are determined by Responding Company to be applicable to such direct and indirect costs and expenses incurred by Responding Company.

Requesting Company shall have the right to audit the records of Responding Company relative to work performed pursuant to this Agreement.

8. (a) In addition, and subject to the provisions of paragraph 8(b) hereof, Requesting Company shall indemnify and hold Responding Company, its officers, directors and employees, jointly and severally, harmless from and against any and all liability or loss, damage, cost or expense which any of them may incur by reason of bodily injury, including but not limited to death, to any person or persons, or by reason of damage to or destruction of any property, including but not limited to the loss of use thereof, which results from furnishing assistance pursuant to this Agreement, whether due in whole or in part to any act, omission, or negligence of Responding Company,,its officers directors or employees.

(b) Where payments are made by Responding Company or its insurers to Responding Company's officers, directors or employees of their beneficiaries for bodily injury or death resulting from furnishing assistance pursuant to this Agreement, including but not limited to workers'ompensation, disability, pension plan, medical and hospitalization, or other such payments, Requesting Company shall make reimbursement to Responding Company to the extent such payments increase the Responding Company's employee-related costs, whether such increase in costs occurs in the form of an increase in premiums or contributions, a reduction in dividends or premium refunds, or otherwise.

Requesting Company shall also reimburse Responding Company for any deductible amounts or for any amounts paid by Responding Company as a self-insurer.

Responding Company will request its insurer to waive any right of subrogation it may have against Requesting Company as a result of any payment described in this paragraph 8(b) which such insurer may make on behalf of Responding Company because of Responding Company's furnishing of assistance pursuant to this Agreement.

March 1996 Page 23 EPIP-EPP-24 Rev 01

ATTACHMENT 1 (Cont) Page 4 of 8

,VOLUNTARY ASSISTANCE AGREEMENT BY AND AMONG

-ELECTRICAL UTILITIES INVOLVED IN TRANSPORTATION OF NUCLEAR ATERIALS

8. (c) In the event any claim or demand is made or suit, action or proceeding is filed against Responding Company, its officers, directors or employees, jointly or severally, alleging liability for which Requesting Company shall indemnify and hold harmless Responding Company, its officers, directors and employees under paragraph 8(a) hereof, Responding Company shall promptly notify Requesting Company thereof, and Requesting Company at its sole cost and expense, shall settle, compromise or defend the same in such manner as it in its sole discretion deems necessary or prudent. Responding Company shall cooperate with Requesting Company in the resolution of any such matter.

(d) Each party to this Agreement agrees to carry the amount of financial protection required by the Atomic Energy Act of 1954, as amended, and self-insurance or comprehensive liability insurance, including contractual liability coverage covering the indemnification and defense obligations set forth herein, subject to such types and amounts of self-insurance, retentions or deductibles as are consistent with good business practice in the industry.

(e) In the event a Responding Company provides assistance pursuant to this Agreement through an affiliate or subsidiary, the indemnification provided in this paragraph 8 to the officers, directors and employees of that Responding Company shall apply with equal force to the officers, directors and employees of that affiliate or subsidiary.

9. Each Party shall provide the Institute of Nuclear Power Operations (hereinafter "INPO") with an executed counterpart signature page to this Agreement and to any amendments hereto. This Agreement shall become effective when counterpart signature pages executed by at least two Parties shall have been received by INPO. This Agreement shall remain in effect as to any Party until such Party has withdrawn from the Agreement as provided below.

Any electric utility involved in the transportation of nuclear materials may become a Party upon execution of the Agreement.

10. (a) INPO may provide certain administrative and emergency response support services in furtherance of this Agreement, such as maintaining and distributing to the Parties a roster of the signatories to this Agreement; providing copies of the Agreement and any amendments thereto to all Parties; and preparing and distributing to the Parties other documents, such as a list of sources of emergency manpower and equipment. INPO may provide such other services as may be requested of INPO from time to time by the Parties. The Parties recognize that INPO shall not be responsible for implementing, enforcing or interpreting this Agreement.

March 1996 Page 24 EPIP-EPP-24 Rev 01

ATTACHMENT 1 (Cont) Page 5 of 8 VOLUNTARY ASSISTANCE AGREEMENT BY AND AMONG ELECTRICAL UTILITIES INVOLVED IN TRANSPORTATION OF NUCLEAR MATERIALS

10. (b) The Parties shall defend, indemnify and hold harmless INPO, its officers, directors and employees, jointly and severally, from and against any and all liability or loss, damage, cost, or expense which results from performance of INPO's functions described in paragraph 10(a) of this Agreement, except as may result from the sole negligence or willful misconduct of INPO, its officers, directors or employees. Each Party hereby expressly waives any right it may have to assert any claim against INPO, its officers, directors, or employees arising out of its or their performance of the duties described in paragraph 10(a), except as may result from the sole negligence or willful misconduct of INPO, its officers, directors or employees.

(c) Following the occurrence of an emergency involving the transportation of nuclear materials INPO may, if asked to do so by a Requesting Company, help to locate sources of emergency manpower and equipment with which the Requesting Company may contract for assistance. If INPO does furnish such assistance and unless otherwise agreed by INPO and the Requesting Company, the Requesting Company and INPO shall have the same rights and obligations as if INPO were a Responding Company (including but not limited to the Requesting Company s obligations to INPO, its officers, directors and employees under paragraph 8 hereof), except that paragraphs 6 and 7 shall not apply either to Requesting Company or INPO and paragraph 8(d) shall not apply to INPO.

ll. This Agreement will not create any rights or defenses in favor of any entity or person not a signatory to this Agreement except to the extent provided in this paragraph and in paragraphs 8 and 10 of this Agreement. This Agreement shall be binding upon and inure to the benefit of each signatory to this Agreement and the subsidiaries and affiliates of each such signatory.

12. Except as otherwise provided in paragraph 13, any Party may withdraw from this Agreement upon at least thirty (30) days prior written notice to INPO with a copy to all of the other Parties. Notice of withdrawal shall not affect any obligations which may have been incurred hereunder prior to the effective date of such notice or which may arise out of events occurring prior to that date. No party may withdraw from this Agreement while it is receiving assistance pursuant to this Agreement.

March 1996 Page 25 EP IP-EPP-24 Rev Ol

ATTACHMENT 1 (Cont) Page 6 of 8 VOLUNTARY ASSISTANCE AGREEMENT BY AND AMONG ELECTRICAL UTILITIES INVOLVED IN RANSPO TATION OF NUCL AR MATERIALS

13. This Agreement may be amended by agreement of a majority of the Parties hereto. Such amendment shall be effective and binding upon all Parties thirty (30) days after INPO has received counterpart signature pages for the amendment executed by at least a majority of the Parties to the Agreement. INPO shall notify all Parties when at least a majority of the Parties have executed an amendment to the Agreement. No amendment shall affect any obligation which may have been incurred hereunder prior to the effective date of such amendment or which arises out of events occurring prior to that date. Notwithstanding the first sentence of paragraph 12, any Party may withdraw from this Agreement by submitting written notice to INPO at any time during the thirty (30) day period prior to the effective date of such amendment with a copy to all of the other Parties.

14. If any provision of this Agreement is determined to be invalid or unenforceable as to any Party or otherwise, such determination shall not affect the validity or enforceability of the other provisions of this Agreement as to that Party or otherwise.

March 1996 Page 26 EPIP-EPP-24 Rev Ol

ATTACHMENT 1 (Cont) Page 7 of 8 VOLUNTARY ASSISTANCE AGREEMENT BY AND AMONG ELECTRICAL UTILITIES INVOLVED IN TRANSPORTATION OF NUCLEAR MATERIALS COUNTERPART SIGNATURE PAGE The undersigned company hereby agrees to become a Party to the Voluntary Assistance Agreement By And Among Electric Utilities Involved in Transportation of Nuclear Materials dated

~21 1 1 2.

Date Company By see roster of si natories Corporate Officer Signature The roster of the signatories of the Transportation Agreement is provided below. (Signatures maintained on file in the NMPC Emergency Preparedness Department).

0, Alabama Power Company

2. Arkansas Power Light 22.

23.

Northern States Power Company Pacific Gas & Electric

& Company

• 3 Cincinnati Gas Electric

& 24. Pennsylvania Power & Light

• 4 Cleveland Electric Illuminating Company 25. Philadelphia Electric Company

5. Commonwealth Edison 26. Portland General Electric Company

6. Consumers Power Company 27. Public Service Company of Colorado

7. Detroit Edison Company 28. Public Service Company of Indiana

8. Duke Power Company 29. Rochester Gas & Electric Corporation

9. Florida Power & Light Company 30. South Carolina Electric & Gas Company

10. Gulf States Utilities Company 31. Southern California Edison Company

11. Illinois Power Company 32. Tennessee Valley Authority

12. Indiana & Michigan Electric Company 33. Texas Utilities Generating Company

13. Iowa Electric Light & Power Company 34. Toledo Edison Company

14. Jersey Central Power & Light Company 35. Union Electric Company

• 15 Kansas Gas & Electric Company 36. Vermont Yankee Nuclear Power Company

16. Long Island Lighting Company 37. Virginia Electric and Power Company

17. Maine Yankee Atomic Power Company 38. Washington Public Power Supply System

18. Hetropolitan Edison 39. Wisconsin Electric Power Company

19. Mississippi Power & Light Company 40. Wisconsin Public Service Company

20. Niagara Mohawk Power Corporation 41. Yankee Atomic Power Company

21. Northeast Utilities

• New Members March 1996 Page 27 EPIP-EPP-24 Rev Ol

ATTACHHENT 1 (Cont)

VOLUNTARY ASSISTANCE AGREEHENT BY AND AHONG ELECTRICAL UTILITIES INVOLVED IN RANSPORTATION OF NUCLEAR HATERIALS Re uestin Com an etterhead Date 19 (Name and Address of Responding Company)

This letter confirms the telephone conversation on insert date and

~time between eur and your in which our company requested assistance pursuant to the terms of the Voluntary Assistance Agreement By and Among Electric Utilities Involved in Transportation of Nuclear Haterials dated , 1982 and your company agreed to provide assistance pursuant to that Agreement.

Please acknowledge your agreement to the foregoing by signing and returning to me the enclosed copy of this letter.

Requesting Company Name and Address Corporate Officer Signature Responding Company Name and Address Corporate Officer Signature and Date Harch 1996 Page 28 EP IP-EPP-24 Rev Ol

(EPIMPP-24+140)

Instructions: When making notifications of a Nuclear Trans. Accident, read the following statement.

"This is the Nine Mile Point Nuclear Station. This (isfis not) a drill. My name is This is to report that we have received information of a transportation accident involving a vehide carrying radioactive materials.'Provide additional information as requested from the items listed in Part B below)

Instructions: Complete the following items upon notification of a Nuclear Transportation Accident.

Location'ate/Time Name of Caller Title/Organization Accident Return Phone No.

~

Nearest Aitpott Time of Accident Nature of Accident t7 Highway(ie.truck,car,etc.) C7 RaiAvay < Airplane < Other(Explafn):

Special Accident Details (Mark all appropriate boxes)

> Fire < Injun'es(Explain):

C7 Radioactive Material Involved (Type ifinformation available):

Cl Other Hazardous Materials pjpes):

FederaVState/Local Authorities Notified or On-Scene Nuclear Station Assistance Requested Name of Shipper: Shipment No, Phone No, ( )

Address: Contacted: Cl Yes Cl No Special instructions or Information from Shipper.

Name of Canier. Phone No.: ( )

Address: Contacted: + Yes <No Special Instructions or Information from Carrier.

Instructions Given to Caller Briefly Record Add'I Important Info.

Call Received by (Unit 1 SSS or CSO) Time SSS NotiTied

~ yes ~No

• If acddent is located outside NMPC's geographical boundaiy, ref. to Sect.8.4 of this proc. for special considerations such as requesting INPO assistance.

• • If the shipper Is a signatory of the INPO Voluntaiy Assistance Agreement, refer to Section 8.4 of this procedure for spedal considerations regarding emergency response either to or from that signatory.

March 1996 Page 29 EPIP-EPP-24 Rev Ol

tEPIIKPP-24+240)

Name Cl SSS Cl CSO Date Initia) I Time

1. Complete the Nuclear Transportation Accident Report Form (Figure 1).-

2. Provide caller with the nature of station response and initial response instructions:.

3. Designate a Control Room Communications Aide to perform notifications and complete the Control Room Communications Aide Checklist (Figure 3).

4. Record all significant accident related activities in the Station Log.

5. If possible, notify the shipper. Obtain information concerning the shipment and record on Figure 1..

6. If appropriate, report shipper information to the Unit Supervisor RP at the accident scene..

7. If necessary, contact the INPO Duty Officer to determine if the shipper is an INPO Transportation Agreement signatory. Refer to Section 8.4 of this procedure if the shipper is a signatory.. C7 Yes Cl No

8. If known, notify the carrier. Obtain information concerning the shipment and record on Figure 1..

9. If appropriate, report carrier information to the Unit Supervisor RP or the Supervisor Radwaste Operations..

10. Provide the Unit Supervisor RP at the accident scene with instructions concerning accident response and recovery operations.

March 1996 Page 30 EPIP-EPP-24 Rev Ol

7 N,UC,LEAR!0 NlAGARA,:.:;::::::-.,::.:.:;::::.::,':::::::::',::;:,'-'!.':~:

N 0MOHAWK j',;:;.CONTROL"'.':;.ROOKIlYCO5ll!NVNICATlONBl'Alo,',CHECKLlST4j";::

IEPiPZPP-24k 3a00)

Name Date Initial I Time Inform the Supervisor Radwaste Operations of the accident by providing the information given in the Nuclear Transportation Accident Report Form (Figure 1, Part A)....

NorE:

If the shipment originated from NMPC, contact the Supervisor Radwaste Operations of the unit where the shipment originated, if known. (This will facilitate getting information on the the shipment.) Otherwise contact ONE of the following:

Unit 1 General Supervisor Radwaste Office: 349-2543 Jack Torbitt, Jr. Home: 593-2713 Beeper. 876-1282 Unit 2 General Supervisor Radwaste Office: 3494231 Ron Cole Home: 3434045

2. Inform Chemistry and Radiation Protection Management using the information given in Figure 1,.

Part A,through use of the department onmll schedules.

3. Contact the Manager RP to request the Supeivisor Radwaste Operations and a minimum of one RP Technician report immediately to the accident scene, if appropriate..

4. Instruct the Supervisor RP to contact the Unit 1 SSS from the accident scene and provide all pertinent information concerning the incident.

5. Provide the Nuclear Security Department with the information given on Figure 1, Part A.

Request Security complete notiTications in accordance with the Nuclear Transportation Accident Security Checklist (Figure 4)..

Nuclear Security Department 349-2401 Or Gaitronics

6. Inform the Oswego County Emergency Management Office (OCEMO) of the accident, providing the information given in Figure 1, Part A.

Normal Hours Oswego County Emergency Management Office 598-1191 5934912 5984678 (Telecopy)

Or Radio Off Hours Oswego County Warning Point 343-1313 March 1996 Page 31 EPIP-EPP-24 Rev Ol

NUCLEAR TRANSPORTATION ACCIDENT FIGURE 3 CONTROL ROOM COMMUNICATIONSAID CHECKLIST

~ Continued ~

Name Date Initial / Time

7. Inform the New York State Emergency Management Office (SEMO) of the accident, providing the information given in Figure 1 NY State Emergency Management Office 518-457-2200 51 8-457-6811 518-457-9942 (lelecopy)

8. Inform the Unit 2 Control Room SSS of the incident, providing the information give in Figure 1, Part A ~...... ~... ~..... ~......

~

Unit 2 SSS 349-2170 342-1929 342-3059 349-21 68 (CSO)

9. When instructed by the SSS, inform the NRC of the accident, providing the information given in Figure 1, Part A, ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

NRC 301-816-5100 (MainJ 301-951-0550 (Back-upJ 301-415-0550 (Second Backup) 301-816-5151 (TelecopyJ

10. Inform the NRC Senior Resident Inspector of the accident, providing the information givenin Figure 1, Part A ..

NRC Resident Inspector Office: 349-2529 Office: 342-4041 Pager 716/528-0925 March 1996 Page 32 EP IP-EPP-24 Rev 01

j".'..:'?<$'j;<?g):,q>jjgg';"<'.@$g@:. @5N ?:.A?4; .??'??>'.< 0':<:rYWAN??:N>'4"??4F4:<:<?%774?F6jW ?N< Swy?. <:<??>Kg'?qspgg??Ã$ %??@<<+)?:;?.:?'.?~??

FIGURE 4 Name Date Initial / Time

1. Upon being notified of a nuclear transportation accident, complete a copy of the Nuclear Transportation Accident Report Form (Figure 1).................

2. Inform the Supervisor Environmental Protection of the accident by providing the information given in Figure 1, Part A.............................

Carey Merritt Office: 349-4200 Home: 298-7490 Beeper: 876-3169 OR See Environmental Protection Department On-call Schedule

3. Inform the Manager Nuclear Communications and Public Affairs of the accident by providing the information given in Figure 1, Part A................ ~...

Robert Burtch Office: 349-7601 Home: 342-2271 Beeper: 876-1124 OR See PACC Department On-call Schedule 4 'C'otification complete................................,......

Hatch 1996 Page 33 EPIP-EPP-24 Rev Ol

ENCLOSURE 1 INP ADDR A TELEPH NE INSTXHJTE OF NUCLFAR POWER OPERATIONS 1100 Circle 75 Parkway Suite 1500 Atlanta, 6eorgia 30339 (770) 644-8000 Emergency Phone No.: (800)32't-06'14 Narch 1996 Page 34 EPIP-EPP-24 Rev 01

NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT NUCLEAR STATION EMERGENCY PLAN MAINTENANCE PROCEDURE, EPMP- PP-02 REVISION 14 EMERGENCY E UIPMENT INVENTORIES AND CHECKLISTS TECHNICAL SPECIFICATION REQUIRED Approved by:

R. G. Smith Plant Manager Unit I Date 1

Approved by:

K. A. Dahlberg Plant anager Undec 2 Date Effective Date:

~pl"

~

TABLE OF CONTENTS S CTION PAG 1.0 .

PURPOSE 1

2. 0 PRIMARY RESPONSIBILITIES 1 3.0 PROCEDURE . 2

3. 1 Performing Inventory . . . . . . . . . . . . . . . . . . . . 2 4.0 DEFINITIONS ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~, ~ ~ ~ ~ ~ 3

5.0 REFERENCES

AND COMMITMENTS . . . . . . . . . . . . . . . . . . . 4 6.0 RECORD REVIEW AND DISPOSITION.................. 5 ATTACHMENT 1: FIRE CABINET INVENTORY.................. 7 ATTACHMENT 2: MEDICAL/RESCUE EQUIPMENT................. 8 ATTACHMENT 3: STOKES BASKET/BACKBOARDS - UNIT 1 . . . . . . . . , . . . 9 ATTACHMENT 4: STOKES BASKET/BACKBOARDS - UNIT 2 . . . . . . . . . . . . 10

. ACHMENT 5: RESCUE CABINET INVENTORY . . . . . . . . . . . . . . . . . 11 ATTACHMENT 5A: CONFINED SPACE RESCUE EQUIPMENT CABINET INVfNTORY.... 12 ATTACHMENT 6: SECURITY BUILDING INVENTORY: AMBULANCE AND FIRE KIT UNIT - 2 o ~ ~ ~ ~ ~ ~ ~ ~ * ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o o 13 ATTACHMENT 7: RADIATION PROTECTION SUPPLIES AND EQUIPMENT OSC / TSC / ONSITE / DOWNWIND .............. 14 ATTACHMfNT 8: RADIOLOGICAL MONITORING EQUIPMENT

/ TSC / ONSITE / .............. 15 OSC DOWNWIND ATTACHMENT 8a: MISC. R.P. EQUIPMENT 16 ATTACHMENT 9: RADIATION PROTECTION SUPPLIES AND EQUIPMENT EOF ..... 17 ATTACHMENT 10: RADIOLOGICAL MONITORING EQUIPMENT EOF.......... 18 ATTACHMENT 11:

OFFSITE ASSEMBLY AREA..................

RADIATION PROTECTION SUPPLIES AND EQUIPMENT .

19 ATTACHMENT 12: DELETED . . . . . . . . . . . . . . . . . . . . . . . . . 20 ATTACHMENT 13: OSWEGO HOSPITAL NUCLEAR EMERGENCY CABINET INVENTORY . . . 21 TTACHMENT 14: PERSONNEL DECONTAMINATION ROOM SUPPLIES INVENTORY.... 23 ATTACHMENT 15: DELfTED ~ ~ ~ t a ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 24 February 1998 Page i EPMP-fPP-02 Rev 14

~ gJ

. ~

TABL OF CONTENTS (Cont)

S CTION PAG ATTACHMENT 16: TECHNICAL SUPPORT CENTER . . . . . . . . . . . . . . . . 25 ATTACHMENT 17: EOF (EMERGENCY OPERATION FACILITY) . ~......... 28 ATTACHMfNT 18: EMERGENCY VENTILATION FILTER LOG ............ 30 ATTACHMENT 19: OPERATIONS SUPPORT CENTfR (OSC)............. 31 ATTACHMENT 20: JOINT NEWS CENTER JNC . . . . . . . . . . . . . . . . . . 32 ATTACHMENT 21A: DAMAGE CONTROL TOOL BOX INVENTORY (MECHANICAL) . . . . . 35 ATTACHMENT 218: DAMAGE CONTROL TOOL BOX INVENTORY (I&C) . . . . . . . . 37 ATTACHMENT 22: ELECTRIC DAMAGE REPAIR EQUIPMENT INVENTORY ....... 39 23:

ATTACHMENT INVENTORY........................

TEMPORARY RESTORATION OF POWER FOR POST ACCIDENT SAMPLING 41 ATTACHMENT 24: EMERGENCY RESPONSE FACILITY COMMUNICATIONS SURVEILLANCE . 42 25A: EMERGENCY RESPONSE FACILITY COMMUNICATIONS SURVEILLANCE ACHMENT RADIOLOGICAL EMERGENCY COMMUNICATIONS SYSTEM (RfCS)

TESTING (MONTHLY) 44 ATTACHMfNT 25B: EMERGENCY FACILITY COMMUNICATIONS SURVEILLANCE COMMERCIAL TELEPHONE TESTING (MONTHLY)......... 46 ATTACHMENT 25C: EMERGENCY FACILITY COMMUNICATIONS SURVEILLANCE EMERGENCY NOTIFICATION SYSTEM (ENS) TESTING (MONTHLY) . 48 ATTACHMENT 25D: EMERGENCY FACILITY COMMUNICATIONS SURVEILLANCE DEDICATED TELEPHONE TESTING (ANNUALLY)......... 51 ATTACHMENT 25E: fMERGENCY FACILITY COMMUNICATIONS SURVEILLANCE RADIO CONSOLE TESTING (ANNUALLY)............ 53 ATTACHMENT 25F: EMERGENCY FACILITY COMMUNICATIONS SURVEILLANCf RADIO TESTING (ANNUALLY)................ 55 ATTACHMENT 25G: PORTABLE RADIO BATTERY EXCHANGE (QUARTERLY) . . . . . . 57 ATTACHMENT 26A: RESPIRATORY EQUIPMENT MONTHLY INSPECTION........ 58 ATTACHMENT 26B: RESPIRATORY EQUIPMENT MONTHLY INSPECTION........ 59 TTACHMENT 26C: RESPIRATORY EQUIPMENT MONTHLY INSPECTION SCOTT PAK... 60 ATTACHMENT 27: HAZARDOUS WASTE AND EMERGENCY SPILL RESPONSE KIT INVENTORY 61 February 1998 Page ii EPMP-EPP-02 Rev 14

ra

\ ~

-0

TABLE OF CONTENTS (Cont)

SECTION PAGE ATTACHMENT 28: ALTERNATE POWER SUPPLIES FOR PORTABLE AIR SAMPLERS . . . 62 ATTACHMENT 29: N2-EOP-6 TOOL BOX FOR BY-PASS OF STAND-BY GAS (N2-PM-QOO&) 63 ATTACHMENT 30: EMERGENCY FACILITIES TLD LISTING 64 ATTACHMENT 31: EMERGENCY TLD ISSUE SHEET . ~ ~ ~ 65 ATTACHMENT 32: NINE MILE POINT NUCLEAR STATION PROCESS RAD MONITORING BOARD UNIT 1 . 66 ATTACHMENT 33: NINE MILE POINT NUCLEAR STATION PROCESS RAD MONITORING BOARD UNIT 2 . 67 ATTACHMENT 34: NINE MILE POINT NUCLEAR STATION INPLANT SURVEY/SAMPLE STATUS BOARD 68 ATTACHMENT 35: NINE MILE POINT NUCLEAR STATION DOWNWIND SURVEY/SAMPLE STATUS BOARD . 69 ATTACHMENT 36: NINE MILE POINT NUCLEAR STATION EMERGENCY EVENTS STATUS BOARD . 70 ATTACHMENT 37: NINE MILE POINT NUCLEAR STATION EQUIPMENT SURVEY/SAMPLE STATUS BOARD 71 ATTACHMENT 38: PLANT STATUS TRENDING BOARD . 72 ATTACHMENT 39: NINE MILE POINT NUCLEAR STATION AREA RAD MONITORS - UNIT 1 73 ATTACHMENT 40: NINE MILE POINT NUCLEAR STATION AREA RAD MONITORS UNIT 2 74 ATTACHMENT 41: EMERGENCY PROCEDURES TELEPHONE NUMBERS QUARTERLY PHONE CHECKS 75 ATTACHMENT 42: EMERGENCY KEY INVENTORY (QUARTERLY) 80 February 1998 Page iii EPMP-EPP-02 Rev 14

~ 1 0

I. PURPOSE To provide a mechanism for ensuring that emergency equipment necessary to implement the Site Emergency Plan is maintained by all responsible departments.

2.0 PRINARY RESPONSIBILITIES FREQLKNCY RESPOIS I BLE N=fiontfa ly INVENTINIT/SINIVETLLANCE BRANCH Q=Qmsrterly MANAGER AR=As Required IIRWot R ired Fire Cabinet Inventor perations U- I Medical/Rescue Equipment perations U-I Stokes Basket/Backboards - Unit 1 perations U-I Stokes Basket/Backboards - Unit 2 oerations U-l Rescue/Confined Space Rescue Equipment Inventory perations U-I Securit Bid Inventor : Ambulance/Fire Kit - Unit 2 Rad protection Radiation Protection Supplies and Equipment Rad Protection OSC/TSC/Onsite/Oovnwind Radiological Monitoring Equipment Rad Protection OSC/TSC/Onsite/Oownuind Ba Misc Rad Protection Eauicment Rad Protection Rad Protection Supplies and Eauicment EOF Licensin 10 Radiological Monitoring Equipment EOF Licensing Rad Protection Supplies and Fcuicment OAA Licensin 12 Oeleted 13 Oswego Hosp>tal Nuclear Emergenc Cabinet Inventor Licensing Personnel Oecontamination Room Supplies Invento Rad Protection 15 Oeleted 16 TSC Inventor Trainin 17 EOF Inventory raining 18 Emergency Ventilation Filter Log Training 19 OSC inventor v Trainin 20 JNC Inventory Training 21 Oama e Control Tool Box Invento aintenance/IEC Electric Oamage Repair Equipment Inventory aintenance Tereorarv Restoration of Poser for PASS tnvento aintenance Emergency Response Facility Caenanication Training AR Surveillance 25 Emergency cammnications Surveillance Sheets Training AR 26 Respiratory Protection Monthly Inspections Licensing/

perations/

ad Protection 27 ~ Hazardous Maste and Emergency Spill Response Kit perations U-1 Inventory 28 Alternate PoMer Su lies for Portable Air Samolers aintenance 29 NZ-EOP-6 Tool Box for B ss at Stancby Cas perations U-2 30 Emergency Facilities TLO Listing /A NR 31 Emer enc TLO Issue Sheet N/A AR 32-40 Emergency Facility Status Boards N/A NR 41 Quarterl Phone Checks Trainin 42 Emergency Key Inventory I'raining 2.1 De artment Su ervisor Signs the inventory or surveillance for final approval to indicate satisfactory completion and resolution of any identified abnormalities.

February 1998 Page 1 EPMP-EPP-02 Rev 14

2.2 Oirector er enc Pre a ed e Responsible for ensuring completion and documentation of required inventories and checklists.

3.0 PffO~CQUUi.

3.1 Per formin Inventor

~NOT: Inventories or checklists performed by the New York Power Authority, that are determined to be equivalent to NMPC requirements by the Oirector Emergency Preparedness, shall provide acceptable proof of completion for those equivalent forms found in this procedure. Ouplication of effort by NMPC is

, not required in these cases.

3.1.1 The Emergency Preparedness Oepartment shall ensure emergency equipment "inventory checklists are completed by assigned persons and, where required, retained for documentation of the surveillance.

3.1eZ Inventories, unless otherwise, specified, should be performed at least ance ~darts each quarter and aster each use.

a. Post use inventories may be used to satisfy routine inventory requirements and should clearly indicate this on the form as applicable.

b. Equivalent forms may be used for inventories.

3.1.3 "UNSAT" Oiscrepancies should be corrected, or action initiated by the responsible party to correct them within 3 working days.

Resolution of the "UNSAT" discrepancies shall be noted on the checklist.

~NOT : A discrepancy or "UNSAT" condition should not preclude the completion of the checklist.

a. In the case of a discrepancy or an unsatisfactory condition, a note shall be made on the checklist indicating the corrective action taken and date completed.

b. In the case of discrepancies that can not be corrected on the spot (i .e. equipment not in stock and must be ordered) a copy of the completed inventory checklist identifying the discrepancy (where practical) should be included with that Emergency Equipment until such time as the deficiency is resolved or corrected.

c A second copy of the as-completed inventory checklist (with discrepancies identified) should be sent to the Emergency Preparedness Oepartment.

d. Upon resolution/correction of the discrepancies, the original completed inventory/surveillance form should be sent to Emergency Preparedness in accordance with Step 3.1.7.

e. If N/A (Not Applicable) or N/R (Not Required) is used in this procedure, provide an explanatory note to document the reason.

February 1998 Page 2 'EPMP-EPP-02 Rev 14

3. 1.4 A complete inventory and inspection should be performed on sealed supplies at least once per year.

3.1.5 Contents of supplies need not be inventoried if:

a. Seal is not broken (except in case of step 3.1.4 above).

b. Opened only to remove equipment for testing, source check, one for one changeouts, etc.

c. Opened to verify specific equipment availability.

d. Used for training and has been restored to pre-class condition.

3.1.6 The entire Emergency Communications System is subject to periodic testing. This shall be accomplished using the instructions in Attachments 24 and 25.

3.1.7 Department Supervisor or designee shall:

a. Ensure corrective actions are initiated promptly and appr opri ately (See 3.1.3) .

b. Ensure discrepancies are resolved satisfactorily.

c. Ensure that any items that may be expiring are ordered or available from stores as needed.

d. Sign the completed surveillance or inventory indicating satisfactory completion and resolution of discrepancies.

e. Forward signed, completed form to the Emergency Preparedness Department within ten working days from the date of Supervisor approval.

3.1.8 The Director Emergency Preparedness or designee shall:

a. Hake a determination of the effect discrepancies have on the Site Emergency Plan and ensure appropriate priorities have been assigned to resolution.

b. Initial each "corrective action" for an "Unsat" and add notes as appropriate, prior to signing the form for final approval.

4.0 DEFINITIONS

'Sat" - Satisfactory means an item is available in at least the minimum quantity specified and capable of performing its intended function.

February 1998 Page 3 EPHP-EPP-02 Rev 14

4.0 (Cont)

"Unsat" Unsatisfactory means an item is not available in at least its minimum quantity, or it is not capable of performing its intended function.

5.0 REFERENCES

AND COMMITMENTS 5.1 Technical S ecifications None 5.2 Licensee Documentation None 5.3 Standards Re ulations and Codes 5.3.1 NUREG 0654, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants 5.3.2 10CFR50 Appendix E Emergency Planning and Preparedness for Production and Utilization Facilities 5.3.3 NRC-IE Information Notice 86-97 Emergency Communication System 5.3.4 NRC-IE Information Notice 85-44, Emergency Communication System Monthly Test 5.3.5 NRC Memorandum dated Sept. 18, 1984, RE: Emergency C'ommunication Systems at Licensee Sites 5.4 Policies Pro rams and Procedures 5.4. 1 NDD-EPP, Emergency Preparedness 5.4.2 NIP-RMG-Ol, Records Hanagement 5.4.3 EPMP-EPP-Ol, Maintenance of Emergency Preparedness 5.4.4 S-RRI-9, Issuing Emergency Kit Dosimetry 5.4.5 S-RPIP-4.4, Maintenance, Inspection, and Testing of Respiratory Protection Equipment 5.4.6 N2-COMP-GEN-W001, Weekly Preventive Maintenance Checklist 5.4.7 NIP-CHE-Ol, Chemical Control Program 5.5 Commitments Sequence NCTS

/umber Number Descri tion None February 1998 Page 4 EPMP-Epp-02 Rev 14

0 RECORD REVIEW AND DISPOSITION 6.1 The following records generated by this procedure shall be maintained by Records Management for the Permanent Plant File in accordance with NIP-RMG-01, Records Management:

~ All Inventories, Surveillances, or lists containing signatures indicating completion ATTACHMENT 1: FIRE CABINET INVENTORY ATTACHMENT 2: MEDICAL/RESCUE EQUIPMENT ATTACHMENT 3: STOKES BASKET/BACKBOARDS UNIT 1 ATTACHMENT 4: STOKES BASKET/BACKBOARDS UNIT 2 ATTACHMENT 5: RESCUE CABINET INVENTORY ATTACHMENT 5A: CONFINED SPACE RESCUE EQUIPMENT CABINET INVENTORY ATTACHMENT 6: SECURITY BUILDING INVENTORY: AMBULANCE AND FIRE KIT UNIT-2 ATTACHMENT 7: RADIATION PROTECTION SUPPLIES AND EQUIPMENT OSC/TSC/ONS ITE/DOWNWIND ATTACHMENT 8: RADIOLOGICAL MONITORING EQUIPMENT OSC/TSC/ON SITE/DOWNWIND ATTACHMENT 8a: MISC. R.P. EQUIPMENT ATTACHMENT 9: RADIATION PROTECTION SUPPLIES AND EQUIPMENT EOF ATTACHMENT 10: RADIOLOGICAL MONITORING EQUIPMENT EOF ATTACHMENT 11: RADIATION PROTECTION SUPPLIES AND EQUIPMENT OAA ATTACHMENT 13: OSWEGO HOSPITAL NUCLEAR EMERGENCY CABINET INVENTORY ATTACHMENT 14: PERSONNEL DECONTAMINATION ROOM SUPPLIES INVENTORY ATTACHMENT 16: TECHNICAL SUPPORT CENTER ATTACHMENT 17: EMERGENCY OPERATIONS FACILITY (EOF)

ATTACHMENT 19: OPERATIONS SUPPORT CENTER (OSC)

ATTACHMENT 20: JOINT NEWS CENTER (JNC)

ATTACHMENT 21A: DAMAGE CONTROL TOOL BOX INVENTORY (MECHANICAL)

ATTACHMENT 21B: DAMAGE CONTROL TOOL BOX INVENTORY (IBLC)

ATTACHMENT 22: ELECTRIC DAMAGE REPAIR EQUIPMENT INVENTORY ATTACHMENT 23: TEMPORARY RESTORATION OF POWER FOR POST ACCIDENT SAMPLING INVENTORY ATTACHMENT 25A: EMERGENCY RESPONSE FACILITY COMMUNICATIONS SURVEILLANCE RADIOLOGICAL EMERGENCY COMMUNICATIONS SYSTEM (RECS) TESTING (MONTHLY)

ATTACHMENT 25B: EMERGENCY FACILITY COMMUNICATIONS SURVEILLANCE COMMERCIAL TELEPHONE TESTING (MONTHLY)

ATTACHMENT 25C: EMERGENCY FACILITY COMMUNICATIONS SURVEILLANCE EMERGENCY NOTIFICATION SYSTEM (ENS) TESTING (MONTHLY)

ATTACHMENT 25D: EMERGENCY FACILITY COMMUNICATIONS SURVEILLANCE DEDICATED TELEPHONE TESTING (ANNUALLY)

ATTACHMENT 25E: EMERGENCY FACILITY COMMUNICATIONS SURVEILLANCE RADIO CONSOLE TESTING (ANNUALLY)

ATTACHMENT 25F: EMERGENCY FACILITY COMMUNICATIONS SURVEILLANCE RADIO TESTING (ANNUALLY)

ATTACHMENT 25G: PORTABLE RADIO BATTERY EXCHANGE ATTACHMENT 26A'ESPIRATORY EQUIPMENT MONTHLY INSPECTION ATTACHMENT 26B: RESPIRATORY EQUIPMENT MONTHLY INSPECTION ATTACHMENT 26C: RESPIRATORY EQUIPMENT MONTHLY INSPECTION February 1998 Page 5 EPMP-EPP-02 Rev 14

6.1 (Cont)

ATTACHMENT 27: HAZARDOUS WASTf AND EMERGENCY SPILL RESPONSE KIT INVENTORY ATTACHMENT 28: ALTERNATE POWER SUPPLIES FOR PORTABLf AIR SAMPLERS ATTACHMENT 29: N2-EOP-6 TOOL BOX FOR BY-PASS OF STAND-BY GAS (N2-PM-Q008)

ATTACHMENT 31: EMERGENCY TLD ISSUE SHEET ATTACHMENT 32: NINE MILE POINT NUCLEAR STATION PROCESS RAD MONITORING BOARD UNIT 1 ATTACHMENT 33: NINE MILE POINT NUCLEAR STATION PROCESS RAD MONITORING BOARD UNIT 2 ATTACHMENT 34: NINE MILE POINT NUCLEAR STATION INPLANT SURVEY/SAMPLE STATUS BOARD ATTACHMENT 35: NINE MILE POINT NUCLEAR STATION DOWNWIND SURVEY/SAMPLE STATUS BOARD ATTACHMENT 36: NINE MILE POINT NUCLEAR STATION EMERGENCY EVENTS STATUS BOARD ATTACHMENT 37: NINE MILE POINT NUCLEAR STATION EQUIPMENT SURVEY/SAMPLE STATUS BOARD ATTACHMENT 38: PLANT STATUS TRENDING BOARD ATTACHMENT 39: NINE MILE POINT NUCLEAR STATION AREA RAD MONITORS-UNIT 1 ATTACHMfNT 40: NINE MILE POINT NUCLEAR STATION AREA RAD MONITORS UNIT 2 ATTACHMENT 41: EMERGENCY PROCEDURES TELfPHONE NUMBERS QUARTERLY PHONE CHECKS ATTACHMENT 42: EMERGENCY KEY INVENTORY 6.2 The following records generated by this procedure are not required for retention in the Permanent Plant File:

ATTACHMENT 18: EMERGENCY VENTILATION FILTER LOG

~ The following status boards when generated for any other reason than an actual emergency event (i.e., drill, training):

ATTACHMENT 31: EMERGENCY TLD ISSUE SHEET ATTACHMENT 32: NINE MILE POINT NUCLEAR STATION PROCESS RAD MONITORING BOARD - UNIT 1 ATTACHMENT 33: NINE MILE POINT NUCLEAR STATION PROCESS RAD MONITORING BOARD UNIT 2 ATTACHMENT 34: NINE MILE POINT NUCLEAR STATION INPLANT SURVEY/SAMPLE STATUS BOARD ATTACHMENT 35: NINE MILE POINT NUCLEAR STATION DOWNWIND SURVEY/SAMPLE STATUS BOARD ATTACHMENT 36: NINE MILE POINT NUCLEAR STATION EMERGENCY EVENTS STATUS BOARD ATTACHMENT 37: NINE MILE POINT NUCLEAR STATION EQUIPMENT SURVEY/SAMPLE STATUS BOARD ATTACHMENT 38: PLANT STATUS TRENDING BOARD ATTACHMENT 39: NINE MILE POINT NUCLEAR STATION ARfA RAD MONITORS UNIT 1 ATTACHMENT 40: NINE MILE POINT NUCLEAR STATION AREA RAD MONITORS UNIT 2 LAST PAGE February 1998 Page 6 EPMP-EPP-02 Rev 14

ATTACHMENT 1: F CABI NV NTORY ation: Unit 1 Turbine Bldg. El. 261, 1st 8 Bridge CI guarterlyt I2 3 4 Unit 1 Screenhouse El. 261, SM Corner circle one Unit 1 Admin. Bldg. El. 261, Vestibule (year)

Unit 2 AP Hall El. 261, East CI Post Drill/

Unit 2 Turbine Bldg. El. 250, South East Exercise/Emergency Unit 2 Screenwell Bldg. El. 261 (date)

CI Other Item/Equipment Min. Qty Sat Unset Corrective Action Date Resolved inventory Sealed 0 CI

1. Fire Axe (1) 0 0

2. Wrecking Bar (1) 0 0

3. Portable Hand Light (5) 0 Cl

4. Extension Cord (1) CI 0

5. Forcible Entry Tool (1) CI 0

6. Bolt Cutters (1) 0 0

7. Rescue Belts (2) 0 0

8. Life Lines (2) 0 0

9. White Turn-out Coat (1) 0 0

10. Yellow Turn-out Coat (4) 0 0

11. Rre Fighters Gloves (5) 0 Cl

12. Boots (5) CI CI

13. Fire Helmet (5) CI CI

14. Spare SCBA Bottles (10) CI 0

5. Scott Air Packs (5) CI 0 is.E i men

1. Exhaust Fan (1) 0 0

2. Duct Tubing (1) CI 0 Change Batteries Every 24 Months Last Battery Change Date NOTE: If batteries will expire before the next inventory then order or obtain replacements.

Performed by Date Supervisor Approval Date E.P. Review Date February 1998 Page 7 EPMP-EPP-02 Rev 14

ATTACHMENT 2: HEOECA RESCUE E UIPMENT Unit 1 Turbine Bldg. El. 261 , 1st (tt Bridge 0 Quarterly: 1 2 3 4 Unit 1 Screenhouse El. 2 61, SW Comer circle one Unit 1 First-Aid,Room El .2 61 (year)

Unit 2 AP Hall El. 261, E ast 0 Post Oriil/

Unit 2 Turbine Bldg. EL 250 , Sou th East Exercise/Emergency Unit 2 Screenwell Bldg. El. 261 (date)

Item/Equipment Mkt. Qty Sat unset Ccrrs cdve Actions Oats Itsso(wd lnwntory Sealed 0 0 Cabinet

l. Olspcsablo Blenkots (3) 0 0 24 Oispcsablo Bootios/Gloves (1 Bag) 0 0

3. Padded Board Splint IGt (1) 0 0

4. Here Traction Splint (1) 0 0

5. Free-Pack (1), 0 0 6 ~ Mast Pants (1) a 0 7 ~ Triage Kit (1) 0 0

a. Heed Irnmcbilizor (1) 0 0

9. Ivied. Cervical Collar (1) 0 0

10. Smo Cervical Collar (1) 0 0

11. Lg. Cervical Collar (1) a 0

12. Straps (3) 0 0

13. K.E.O. Board (1) 0 0

14. Oxygen Iot (1) 0 a gr Beni ~ rs Reqularer irervrtebraathar Mask Bao Valve Mask

15. Infection Control Kit (4) 0 0 16.'7.

Stair Chair (1) a 0 Trauma Kit (1) 0 0 Blood Prossure Cuff (1) 0 0 Stothosccpo (1) 0 0 Kling 6" x 5" Yards (2) 0 0 Kling 4 x 5" Yards (2) 0 0 Kling 2" x 5" Yards (2) a 0 Pen Ught (2) 0 0 EMT Scissors (1) 0 0 And-Bacterial Ointmont (5) 0 Q Instant Glucose (1) 0 0 Ammonia (nba(ants (6) 0 0 Cotton Tipped Applicators (4) 0 0 Oval Eye Pads (4) 0 0 Telfa Sterile Pad (5) 0 0 2x2 Gauzo Pad (5) 0 0 3x3 Gauze Pad (5) 0 0 4x4 Gauze Pad (5) 0 0 Triangular Bandage (3) 0 0 Tape 1 (2) 0 0 Tape 2" (1) 0 0 Tapo 3 (Cloth) (1) 0 0 Vaseline Gauze (2) 0 0 Ace Bandago (1) 0 0 Surgi Pad (4) 0 0 Trauma Oressing (2) 0 0 Storilo Bum Sheets (2) 0 a Ico Packs (2) 0 0 PCR'5 (2) ~

0 a Safety Pins (2) 0 0 Pen (1) 0 a Stop Watch (1) Q 0 Extra Latex Gloves (6 pairs) 0 0 Butte rflys (5) 0 0 Band-Aids (1o) 0 0 Band-Aids extra largo (5) 0 0 Alcohol Preps (5) 0 0 Betadino Props (5) ~ 0 0 Performed by Date Supervisor Approval Date E.P. Review Date

• Items not required at Ul Screenhouse, U2 Screenwell, and U2, Turb. Bldg.

250'ebruary 1998 Page 8 EPHP-EPP-02 Rev 14

ATTACHMENT 3: STOK S BASK BACKBOAROS UNIT Cl guarterlyt 1 2 3 4 circle one (year) 0 Post Drill/

Exercise/Emergency (date)

Cl Other Item/Equipment Min. Qty Sat Unsat Corrective Actions Date Resolved Inventory Sealed CI 0 Turbine 261'y Elevator Stokes Basket Backboard, l.ong

2. Screenhouse Basket Q Q 261'tokes Backboard, Long Q Q

3. A'dmin 261'irst Aid Room Stokes Basket Q Q Backboard, Long Q Q TE: A satisfactory verification of equipment shall include:

Stokes Basket - Good Condition, Bridle Backboard - Good Condition, Straps and Immobilizer Performed by Date Supervisor Approval Date E.P. Review Date February 1998 Page 9 EPMP-EPP-02 Rev 14

ATTACHMENT 4: STOKES BASK T BACKBOAROS "- UNIT 2 CI quarterly: 1 2 3 4 circle one (year)

Cl Post Drill/

Exercise/Emergency (date)

CI Other Item/Equipment Min. Qty Sat Unset Corrective Actions Date Resolved Inventory Sealed AP 261'tokes Basket (1)

Backboard, Long 0 0 Backboard, Short (1) 0 0

2. Screenwell Basket 261'tokes (1) 0 0 Backboard, Long (1) 0 0 Backboard, Short (1) 0 0

3. Turbine Basket 250'tokes 0 0 Backboard, Long 0 0

4. Emergency Response Vehicle Stokes Basket (1) 0 0 Backboard, Long (1) 0 0 Backboard, Short (1) 0 0

5. Turbine 306'W Stokes Basket (1)

(1)

Backboard Basket Rigged for Crena 0 0 a a

'erformed by Oate Supervisor Approval Oate E.P. Review Oate February 1998 Page 10 EPMP-EPP-02 Rev 14

ATTACHMENT O'ESCUE CABINET INVENTORY CI quarterly: 1 2 3 4 circle one (year)

CI Post Drill/

Exercise/Emergency (date)

CI Other

~pic: Unit 1 G Bldg EI. 261'estibule Item/Equipment Min. Qty Sat Unsat Corrective Actions Date Resolved Inventory Sealed CI Cl Crow Bars (2) Q Q

2. Boltcutter (1) Q Cl

3. Hacksaw (2) Cl Cl

4. Burning Torch (1) Cl Cl Come-"Along (1) Cl Cl

6. Cable Sling, (1) Q Cl

7. Sling, 3'able (1) a a

8. Jack, 6'ydraulic 1 Cl CI Ton Hydraulic Jack, 5 Q a Ton 10..Sledgehammer, 6¹ a Cl Sledgehammer, Cl Q 12¹

12. Rope 1/2" x (2) Q Q 100'ife

13. Lines 'l00'4.

(2) Cl a Forcible Entry Tool (1) Q Q

15. Wrecking Bar (5') (1) cl a

16. Box Small Clevis (1) Q Q Pins Performed by Date Supervisor Approval Date E.P. Review Date February 1998 Page ll EPMP-EPP-02 Rev 14

ATTACHMENT 5A: CONFIN D SPAC R SCUE UIPHENT CABINET INVENTORY CI quarterly: 1 2 3 4 circle one (year)

Cl Post Drill/

Exercise/Emergency (date)

Cl Other

~@i~n: Unit 2 Service Bldg. H. 261 Foam Room itemlEquip ment Min. Qty Sat Unset Corrective Actions Date Resolved Inventory Sealed

1. Tripod 0 0

2. Winch (1) 0 0

3. 4 Point Harness (2) 0 0

4. Shock Absorbing 0 0 Lanyard

~A" 100'1) (2)

5. Rope, x 100'.

(2)

Life Lines, (2)

Performed by Date Supervisor Approval Date E.P. Review Date February 1998 Page 12 EPHP-EPP-02 t Rev 14

ATTACHMENT 6: SECURITY BUILDING INV NTORY: AMBULANC AND FIRE KIT UNIT-Cl guarterlyt 1 2 3 4 circle one (year)

Cl Post Drill/

Exercise/Emergency (date)

Cl Other ggg@i~n: Security Unit 2 item/Equipment Min. Qty Sat Unset Corrective Actions Date Resolved inventory Sealed Cl 0 TLDs with controls (50) 0 0 and issue sheets

2. Finger Rings with (6) controls

3. Masking Tape 2" (2 rolls) 0 0 Sealed Sets of PCs (3) 0 0

5. Disposable Gloves (1 box) 0 0

6. Full Face (3)

Respirator with Canister Spare Canisters (3) 0 0

8. Bandage Scissors (2) Cl Cl

9. Herculite Green 0 0

10. Herculite Yellow or (21 0 Cl White

11. Clip Board, Pencils CJ CI

12. Paper Pads (1) 0 0

'l 3. Plastic Bags (4) 0 Q (assorted)

I erformed by Date Supervisor Approval Date E.P. Review Date February 1998 Page 13 EPMP-EPP-02 Rev 14

I l

I ATTACHMENT 7: RADIATION PROTECTION SUPP I S AND UIPM T OSC TSC ON SIT OOWNW IND 0 quarterly: 1 2 3 4 circle one (year)

Cl Post Drill/

Exercise/Emergency (date)

CI Other Location: OSC Storeroom - Unit 1 - Elevation Min. Qty 261'temlEquipment Sat Unset Correcdve Actions Date Resolved Inventory Sealed 0 0 OTECllVE EQUIPMENT Protective Clothing fcomplere sealed (40 sets) 0 0 pecks gef

• Full Face Rospirator with voice Amplifier 0 0 and Canister

3. Sparo Canisters 140 lodinel40 HEPA1 (80) 0 0

$4 Flashlights (30) 0 0

'5. Extra D-Cell Batteries (50) 0 0

'6. Kl Tablots fborrfesf (12) 0 0 Duo Date inventory Sealed SUPP UES

1. PA-235 koys for Post Accidont Sampling (2) 0 0

2. "P-5" keys to Environmental Stations (3) 0 0

3. Key to Softball Field (1) 0 0 4,

5.

6.

7.

8.

New York State Rood Map Rolls of Tapo Misc. Plasdc Bags Disc Smears Moslin Cloth (20)

('IO bx)

(10 pkg)

(3) 0 0

0 Q

Q 0

0 0

0 0

. ~

9. Extension Cord (6) 0 0

10. Latex Gloves (10 bx) 0 0 Rubber Boots (6 pr) 0

12. Rain Suits (6) 0 0

13. Gym Bags (10) 0 0

14. Rad Rope fsr least 100'f 0 0

15. Step off Pads (4) 0 0

16. Radiadon Material Tags fpeperl (40) 0 0

17. Rodiotion Signs and inserts l3) 0 0

18. Plastic Booties (40 pr) 0 0

19. 112 Amp Fuso for VAMP (1) 0 0 Chango battories evory 24 months, lest battery change date:

aNOTE: If botteries or Kl tablets will expire before next inventory then order or obtain replacoments.

Performed by Date Supervisor Approval Date E.P. Review Date

. ~

February 1998 Page 14 EPMP-EPP-02 Rev 14

ATTACHMENT 8: RADIOLOGICAL MONITORING UIPM NT OSC TSC ONS ITE DOWNWIND Q Quarterly: 1 2 3 4 circ1e one (year)

Q Post Drill/

Exercise/Emergency (date)

Q Other

~L~n: OSC Storeroom - Unit 1 - Elevation Min. Qty Sat Unset Corrective Actions Date Resolved 261'temJEquipment inventory Sealed

+l IIPMENT

1. Count Rate Meter (7) CI Cl

2. Dose Rate Meter (0-5R/hrJ (4) 0 0

3. Dose Rate Meter (0-50RlhrJ (6) 0 0

4. High Range Dose Rate Meter (6) 0 Cl (0-1000FVhrJ

5. Sealed Silver Zeolite (15) 0 Q Air Sample Packs 1 Patri Oish 1 Particulate Filter 2 Collection Envelopes

6. Sealed Charcoal (20) Cl Cl Air Sample Packs 1 Petri Dish 1 Particulate Filter 2 Collection Envelopes

7. Radeco AC Air Sampler with (10) 0 0 Spare Fuse

8. Radeco DC Air Sampler (3) 0 0 9, Head for Air Sampler (10) C3 CI

10. GasTech Meter (1) Cl Cl DOSIMETRY - Located in Box in Unit 1 RP Office Box Sealed Cl Cl

1. TLDs (50) C3 0

2. Finger Rings (40 pr) 0 0

3. Dosimeters (0-5RJ (20) Cl 0 4, Dosimeters (0-50RJ (20) Cl C3

5. Dosimeters (0-200RJ (5) 0 0

6. Dosimetry Issue Sheets (2) Cl 0 Performed by Date Supervisor Approval Date E.P. Review Date February 1998 Page 15 EPMP-EPP-02 Rev 14

ATTACHMENT Sa: MISC. R.P. E UIPMENT 0 quarterly: 1 2 3 4 circle one (year)

CI Post Drill/

Exercise/Emergency (date)

CI Other Item/Equipment Min. Sat Unset Corrective Date Qty Actions Resolved

1. Hand and Foot Monitor (TSCJ (2) 0 0 Serial ¹:

Cal. Due:

Serial ¹:

Cal. Due:

2. PING (TSCJ (1) 0 0 Serial ¹:

Cal. Due:

3. VAMP (TSC Rad Assessmenr RoomJ (1) 0 0 Serial ¹:

Cal. Due:

4. VAMP (OSC CofeJ (1) 0 0 Serial ¹:

Cal. Due:

Performed by Date Supervisor Approval Date E.P. Review Date Febr uary 1998 Page 16 EPMP-EPP-02 Rev 14

ATTACHMENT 9: RADIATION PROTECTION SUPPLIES AND E UIPMENT

/OF CI quarterly: 1 2 3 4 circle one (year)

CI Post Drill/

Exercise/Emergency (date)

Cl Other Location: fOF Dock and Storage Area Item/Equipment Min. Qty Sat Unset Corrective Actions Date Resolved PROTECTIVE EQUIP MENT

1. Protoctivo Clothing (comp/ete see/ed 10 sots Q Q peckege/

Inventory Sealed (1W) 0 0

~

1 ~ Rashlights 4 0 0

'2. Extra 0-Call Batto ries 8 0 0

3. Kl Tablots /bott/es/ 12 0 0 Oue Octo:

4, Sealed Silvor Zeolite Air Sample Packs (6) 1 Petri Dish 1 Perticula'te Filter 2 Collection Envelopes

5. Sealed Charcoal Air Sample Packs (6) 0 0 1 Petri Dish 1 Paruculato Filtor 2 Collection Envelopes Bo ots (3 Pair) 0 0 SUPPUES Inventory Sealed ('I-17l 0 0

1. Key to Softball Field (1) 0 0 2 Now York Stato Road Mop (1) 0 0

3. Rolls of Tapo /2'J (4) 0 0 4, Adhesive Labols (10) 0 0

5. Tie Labels (10) Q 0

6. Plastic Bag Ties (10) 0 0

7. Tepo Measure /100 ft./ (1) 0 0

8. Water Sample Container /1 ga/.J (12) 0 0

9. Grass Clippers (1) 0 0

10. Pruning Shears (1) 0 0

11. Mallet (1) 0 0

12. Magnetic Pocket Compass (1) 0 0

13. Twine (3 rolls) 0 0

14. Garden Trowel (1) 0 0

15. Rod Roroscant Tape (1) 0 0 ~

16. Stakes (20) 0 0

17. "P-5 keys to Environmental Stations (1) 0 0 e18 Shovels (2) 0 0 e18. Roinsuits (4) Q 0

~change batteries ovary 24 months, Last battery change date:

• Located outside of sealed kits NOTE: If batteries or KI tablets will expire before the next inventory, then order or obtain replacements.

Performed by Date Supervisor Approval Date E.P. Review Date February 1998 Page 17 EPMP-EPP-02 Rev 14

ATTACHMENT 10: D IO OGI CA MONITORING UIPM NT

/OF CI quarterly: 1 2 3 4 circle one (year )

Q Post Drill/

Exercise/Emergency (date)

C1 Other I,gg~ijnn: EOF Dock and Storage Area Item/Equipment Min. Qty Sat Unsat Correcthre Date Actions Resolved ggglPMENT Count Rate Meter (4)

Cal Oue Date SN:

SN'N:

SN:

2. Dose Rate Meter (3) 0 0 Cal Due Date SN:

SN'N:

3. Sealed Silver Zeolite Air Sample Packs (6) 1 Petri Dish 1 Particulate Filter 2 CoSection Envelopes

4. Sealed Charcoal Air Sample Packs (6) 1 Petri Dish 1 Psrticulete Filter 2 Collection Envelopes

5. Radeco AC Air Sampler with Spare Fuse (2)

Cal Due Date SN:

SN:

6. Radeco OC Air Sampler 0 Cl Cal Due Date SN:

7. Head for Air Sampler (21 Cl Cl

8. Check Source (for merersJ CI Q

9. High Range Dose Rate Meter (0-1000RlhrJ Cl 0 Cal Due Date SN:

10. Dosimeter Charger 0 0 DOSIMETRY - Located ln one box:

Box Sealed 0 0 TLDs (100) 0 0

2. Dosimeters (0-500mrJ (4) 0 0

3. Oosimeters (0-5RJ (4) Cl 0

4. Dosimeters (0-50RJ (4) 0 0

5. Dosimetry Issue Sheets 0 0 Change batteries every 24 months, Last battery change date:

~NOT  : If batteries will expire before the next inventory then order or obtain r'eplacements.

Performed by Date Supervisor Approval Date E.P. Review Date February 1998 Page 18 EPHP-EPP-02 Rev 14

ATTACHHENT ll: RADIATION PROT CTION SUPPLIES AND E VIP'T OFFSIT ASS HB Y ARE CI guarterlyt 1 2 3 4 circle one

{year)

CI Post Drill/

Exercise/Emergency

{date)

CI Other Qgggijnn: Offsite Assembly Area - Volney Service Center PPLtE and PR TE IV IPMENT: Located in sealed drums and footfockers in line crew warehouse its mlEquip mant Min. Qty Sat Unset Correct(ve Acdons Oats Resolved SUPPUES: in footlocker inventory Sealed

l. Misc. Plastic Bags (10) 0 0 24 Oisc Smears (3 bx) 0 0

3. Muslin Cloth (3 pkg) 0 0

4. Extension Cord (1) 0 0

5. Surgical Gloves (3 bx) 0 0

6. Cotton Uners (12 pr) 0 0

7. Gym Bags (3) 0 0 S. Rad Rope lat least 50'J (50') 0 0

9. Rad Matorial Tags (6) 0 0

10. Cotton Tip Swabs (1 pkg) Q 0

11. Surgical Scrub Brushes (5) 0 0

12. Step off Pads (4) Q 0 Bandage Scissors (2) 0 0 Soap bars (2) 0 0 Shampoo (1) 0 0

16. Pocket Watch (3) 0 0

17. Masking Tape (5 Rollsl 0 0

16. Material IO Togs (10) 0 0 Empty Yellow Rod Orums (3)

PROTECTIVE EQUIP.:In 66 gal drum inventory Sealed 0 0

1. Disposable Coveralls (1 box) 0 0

2. Papor Bath Towols (25) 0 0

3. Paper Hand Towels (2 pkg) Q 0

4. Plastic Shoo Covers (10) 0 0 S. Shovels (2) 0 0 Outside footlocker Performed by Date Supervisor Approval Date E.P. Review Date February 1998 Page 19 EPHP-EPP-02 Rev 14

ATTACHMENT 12 ATTACHMENT 12 DELETED.

PAGE LEFT INTENTIONALLY BLANK.

February 1998 Page 20 EPHP-EPP-02 Rev 14

ATTACHMENT 13: OSWEGO HOSP TA NUCL AR M RG NCY CABIN INVENTORY 0 quarterly: 1 2 3 4 circle one (year )

Cl Post Drill/

Exercise/Emergency (date)

Cl Other

~gjgn: Hallway Adjacent to X-Ray Department or closet next to Conference/Radiation Treatment Rootn Item/Equip mant Min. Qty Sat Unset Corrscdve Acdons Oats Rasolwd

~ut Green Herculite (1) 0 0 Stop-Otf Pads (2) 0 0

3. Mosking Tape (10) 0 0 4, 'adiation Signs (10) Cl 0

5. Yellow & Magenta Rope (3) Ci 0 Magnets (6) Q 0 Yollow Trash Bags (15) 0 0 Dosimeter Charger fl bettery & 1 ACl (2) 0 0

9. RMC Samplo Taking Kit finventory (1) 0 0 contents /AWAtt. Gin Hospital Pion/

10. RMC Decontamination Kit (inventory contents fAVYAtt. Gin Hospital Plenl

11. RMC Accident Proc. Poster (1) y12. Seolod Protective Clothing Kits (10) oe TLD badge Oue Date:

b. AMOOmR/ Dosimeter Oue Date:

c fo-20RJ Dosimeter Ouo Date:

0

13. RMC Decontamination Table Top (1) 0 0

14. Hose and Nonle for Oecontaminadon (2) 0 0 Table Top

15. Yellow Wator Receptacles (2) 0 0

16. Yellow Trash Receptacles (2) 0 0 Movablo Base for Trash Rocoptaclos

'7.

(2) 0 0

18. Lead Pig (1) C! 0

19. White Horculite Matting (2) 0 0

20. Portable Stanchion (1) 0 0

21. Radiadon Tags /tie/ - misc. (10) 0 0

22. Rodiation Togs fedhesive/ - misc. (10) 0 0

23. Disc Smears (50) 0 0

24. Atomic Wipes (50) 0

~change battaries avery 24 months, Last battery change dato:

0 t If battorios will expiro before the next inventory thon order or obtain replacomonts.

February 1998 Page 21 EPMP-EPP-02 Rev 14

II ATIACHMENT 13 (Cont)

OSW GO HOSPITAL NUC R HERGENCY CABIN INV NTORY Item/Equip mont INIn. Qty Sat Vnoat Corroctiw Actions Date Roeoiwd

26. Count Rata Motor /NYPAI (1) 0 0 Oue Doter SN:

28. Dose Rate Metor /NYPAI (1) 0 0 Due Date: SNr

27. MS-2 w/HP 210 Probe (NYPAJ snd spare (1) 0 0 fusee

28. Extension Cord /for count rare meterJ 0 0

29. Count Rata Mater /NMPCI 0 0 Ouo Dote: SN:

30. Doso Rato Metor (NMPCJ (1) 0 0 Ouo Onto: SN:

31. NMPC Check Source (1) 0 0 Number:

32. Dosimoters /0-1.5RJ(NMPCJ 0 0

33. EAP-2, Porsonnal Injury /JAPJ 0 0 RdV

'4.

RP-OP&02&1, Personnel (1) 0 0 Oecontarninsdon" (JAPJ Rav.:

38. RP-OPS<2.01, Att. 2 Oocontsrninstion (10) 0 0 Incident Report" (JAR Rev.:

38. RP-INST&2.09 PAR (1) 0 0 Rev.:

37. Invontory Checklists .

(1) 0 0

~ SAP-2 (JAP) (1), 0 0 Rov.:

~ EPMP-EPP-02 (NMPC)

Rev,:

38. Control TLO (NMPC) (2) 0 0 Ouo Osto:

39. Dosimetry Issuo Log snd /NMPCJ Cross 0 Q Refarenco to Kit S

40. Tho Oswego Hospital PIsn for tho (1) 0 0 Decontamination snd Treatment of tho Rodioactivaly Contaminated Psdont

/(ocsted et nurses'srsrionJ Performed by Oate Supervisor Approval Oate E.P. Review Oate February 1998 Page 22 EPNP-EPP-02 Rev 14

ATTACHMENT 14: P RSONN 0 CONTAMINATION ROOM SUPP S INV NTORY Q quarterly: 1 2 3 4 circle one (year)

Cl Post Orill/

Exercise/Emergency (date)

CI Other Q~Zg~ion: 0 Unit 1 OSC /StoreroomJ Cl Unit 2 El. 261'CB Item/Equipment Min. Gty Sat Unset Corrective Actions Date Resolved Inventory Sealed 0 0 Coveralls (6) 0 Cl

2. Paper Bath Towels (6) 0 0

3. Paper Hand Towels (6) CI 0 4, Disposable Gloves (1 box) 0 Cl

5. Assorted Plastic Bags (6) CI 0

6. 4 x 4 Steri Pads (1) 0 0

7. Scissors /Bandage Type/ (1) 0 0

8. Shampoo (4) Cl 0

9. Shaving Cream (2) 0 0

0. Disposable Razors (1) 0 Cl Cotton Swabs (1 box) Cl 0

12. Surgical Scrub Brushes (10) Cl 0

13. Masking Tape (2) 0 0

14. Sample Envelopes (6) 0 0

15. Assorted Radiation/ (6) 0 0 Contamination Tags

16. Soap (10) Cl CI Performed by Date Supervisor Approval Date E.P. Review Date Febr uary 1998 Page 23 EPMP-EPP-02 Rev 14

ATTACHM NT 15 ATTACHMENT 15 OELETED.

PAGE LEFT INTENTIONALLY BLANK.

February 1998 Page 24 EPMP-EPP-02 Rev 14

ATTACHMENT 16: T CHNICA SUPPORT C NTER Cl quarterly: 1 2 3 4 circ1e one

{year) 0 Post Dri11/

Exercise/Emergency

{date)

CI Other NOTE: These are suggested locations for those items: however, the material may be found in other areas within the facility.

All computer equipment is checked by l&C Department Computer Technicians on a monthly basis. See completed Preventative Maintenance Checklist.

Its mlEqulpment Min. Oty Sat Unset Conactlve Actions Date Resolved TSC. COMMUNICATIONS ROOM Communicator Headset 0 0 Telecopier 0 0 TSC, RADIOLOGICALASSESSMENT ROOM Maps l20 mile radius or largerl 0 0 Printers:

GE TermiNet 200 0 0 Genicom 200 0 0

3. Digital DecWriter III 0 0 CONFERENCE ROOM DiagrsmslDra wings:

Electrical Diagrams, Unit 1 (1 set) 0 0 Electrical Diagrams, Unit 2 (1 set) 0 0 Isometrics, Unit 1 (1 set) 0 0 Mechanical Diagrams, Unit 2 (1 set) 0 0 P&IDs, Unit 1 (1 set) 0 0 P&IDs, Unit 2 (1 set) 0 0 TSC, UBRARY fo(JTSIDE COREl

1. Aperture Cards Units 1 & 2 (1 set) 0 0 TSC, TECHNICAL ASSESSMENT ROOM

1. Closed Circuit TV 0 0

2. Computer Printer Paper (1 pkg) 0 0

3. GE Terminet 200 Printer 0 0

4. Honeywell Monitors 0 0

5. Pump Curve Book, Unit 1 0 0

6. Telecopier 0 0

7. Telecopier Paper 0 0 S. Terminet Printer funder Q 0 Honeywell Monirorsi February 1998 Page 25 EPMP-EPP-02 Rev 14

ATTACHMENT 16 (Cont)

TECHNICA SUPPORT CENTER Item/Equipment Mln. Qty Sst Unset Corrective Actions Oats Resolved TSC, CORE Clock 0 20 Compose Rosa (SYr x 11 I 0

3. Oiagrams/Orawings:

Control and Instrument Powor figuro IX-2 Eloctrical Foeds, Unit 1 Area Rod Monitors Hectrical Feeds. Unit 1 Process Rad Monitors Hectrical Powor Oistnbudcn Oisgram Emergoncy Oporaticn Prccodure /EOPJ Rcw Charts, Unit 2 (1 sot)

Emergency Operation Procedure (EOPJ ficw Charts, Unit 1 (1 set) 0 0 Gonoralizod Station Orawing, Unit 1 (1) 0 0 Genorelized Station Drawing, Unit 2 (1) 0 0 Roactcr Vessel Orawing, Unit 1 (1) 0 0 Reactor Vessel Orawing, Unit 2 (1) 0 0 Stoticn Power Oistributicn figur IX-1 (1) 0 0 Emergency Action Levels (EAU, Unit 1 (1) 0 0 Emorgency Action lovols (FAU, Unit 2 (1) 0 0 4, Eating/Orinking/Smoking Is/Is Nct Authcrizod Sign (1) 0 0

5. Ernergoncy Classifications Signs: (1 each) 0 0 Emergency Class Unusual Event Alert Site Aroa Emergency Qenoral Emergency Forms Cabinet Procedure/Occumentst Chemistry Surveillanco Prccodures /CSPJ, Unit 2 (1) 0 0 Coro Operating Limits Roport /COLRJ (1) 0 0 Oamage Repair Procedures, (DRPJ, Unit 1 (1) 0 0 Emergency Chemistry Prccodures /ECPJ. Unit 1 (1) 0 0 Emergency Preparedness Implementing Procedure s (1) 0 0 (EPIPJ Emergency Preparedness Maintonence Procedures (EP(MPJ final Safety Analysis Report (FSARJ, Unit 1 0 Final Safety Analysis Ropcrt Appendices I/a 0 Supplements, Unit 1 Fuel Handling Prcceduros (FHPJ, Unit 1 (1) 0 0 Fuel Handling Proceduros (FHPJ, Unit 2 (1) 0 0 Generation Administrativo Procedures /GAPJ (1) 0 0 INPO Emergency/Resources Manual (1) 0 0 Now York State Radiological Emergency Ran (1) 0 0 NMPC Users Quido Equipmont History Si Users l1) 0 0 Manual Nucloar Interfacing Prccodures /Iy(PJ Oswogo County Radlaticn Ernorgency/Responso Plan 0 Occupational Safety Sa Health Manual (SFTJ 0 Redieticn Protection Administrative Prccodures (S-RAP)

Rsdieticn Protection Technical 4 Analytical Procedures (RTPJ, Unit 1 0 Redieticn.Protection Technical Ea Analytical 0 Procedures (RTPJ, Unit 2 Radiation Protection implementing Procedures (RPIPJ, 2 books Emergency Action Level Rofarenco Manual February 1998 Page 26 EPHP-EPP-02 Rev 14

ATTACHMENT 16 {Cont)

T CHN C SUPPORT C NT Item/Equip mont Min. Qty Sat Unsat Corrscdw Actions Data Rose(vad C, CORE

7. Proceduro/Documents (Cont)

Reaotor Engineering Procedures /REPJ, Unit 2 (1) 0 0 Reactor Engineering Surveillance Procoduros (RESP/, Unit 2 (1) 0 0 Site Chemical Analysis Procedure (CAP/ (1) 0 0 Site Emergency Plan (SEPJ (1) 0 0 Site Radiation Protection Technical 6L Anoiytical Procedures /RTPJ (1) 0 0 Special Operating Procedure /SOPJ, Unit 1 (1) 0 0 Technical Specification Amendment Letters, Unit 1 (1) 0 0 Technical Specification Amendment Letters, Unit 2 (1) 0 0 Technical Speci((cations. Unit 1 (1) 0 0 Technical Specificsdons. Unit 2 (1) 0 0 Technical Support Administrative Procedures (1) 0 0

/TDPJ Updated Safoty Analysis Report (USARJ, Unit 2 (1) 0 0 Waste Handling Procedures (WHP) (1) 0 0 0 0

8. Release ls/is Not in Progress Sign 0 0

9. Status Boards:

Area Rsd Monitor Board, Unit 1 (1) 0 0 Ares Rod Monitor Board. Unit 2 (1) 0 0 Downwind Survey Sample Status Board (1) 0 0 Ernergoncy Events Status Board (1) 0 0 Equipment/Team Status Board (1) 0 0 Inplant Survey Board (1) 0 0 Plant Status Board, Unit 1 (1) 0 0 Plant Status Board, Unit 2 (1) 0 0 Plant Trending Board (1) 0 0

, Process Monitor Status Board, Unit 1 (1) 0 0 Procoss Monitor Status Board, Unit 2 li) 0 0

10. 10 Milo Radius Maps:

Map ¹1 (1) 0 0 Map ¹2 (1) 0 0 IVlap ¹3 (1) 0 0 Map ¹4 (1) 0 0 Map ¹5 (1) 0 0 Map ¹6 (1) 0 0 Map ¹7 (1) 0 0 Map ¹8 (1) 0 0 Drafting Table TSC, PROTECTIVE EQUIPMENT ROOM/SUPPLY CABINETS INVENTORY

1. Calculators (1) 0 0

2. Cassotto Tapes (2) 0 0

3. Rashlight (2) 0 0

4. Liquid Cleaner for Status Boards (1) 0 0

5. 'ortable Cassotto Recorder (1) 0 0

6. Sleeping Cote /Collspsib(e/ (12) 0 0

7. Batteries (6 oach) 0 0 AA Cell C Cell D Call Chango battorios ovory.24 months, Last battory chango dato:

NOTE: If battories will oxpiro before the next inventory then ordor or obtain replacomonts.

Performed by Date Supervisor Approval Date E.P. Review Date February 1998 Page 27 EPMP-EPP-02 Rev 14

ATTACHMENT 17: OF RG NCY OP RATION FACI ITY Q quarterly: 1 2 3 4 circle one (year)

CI Post Drill/

Exercise/Emergency (date)

Cl Other NOTE: These are suggestod lccsdcns for these items; however, tho material may bo found in other areas within tho facility.

All computer equipment is checked by INC Department Computer Technicians on s monthly basis. See completed Preventadvo Maintenenco Checklist.

Item/Equipment N(ln. Qty Sst Unset Corrscthis Actions Dsts Resolved Disqrams/Drawings:

Emerqancy Action Levels /EAQ. Unit 1 0 0 Emergoncy Action Levels /EALJ, Unit 2 0 0

2. Status Boards Downwind Survoy/Samplo Status Board (1) 0 0 Ernorgancy Evont Status Board (1) 0 0 Plant Status Board Unit 1 (1) 0 0 Plant Status Board Unit 2 (1) 0 0 P(ant Trending Board (1) 0 0

3. Prccodures/Documents: (CART)

Emergoncy Preparodness Irnple'menting Prccoduros 0 0

/EPIPJ Emerqency Preparedness Maintenance Procedures 0 0 (EPMP/

Site Emergency Plan (2) 0 0

/SEPJ PLANT ASSESSMENT ROOM Diagrams/Drawings:

Emorgency Operadcn Prccaduro (EOP/ Flow (1 set)

Charts, Unit 1 Emorqoncy Oporaticn Procedure (EOPJ Flow Charts, Unit 2 (1 set) 0 0 Reactor Vessel Drawings, Unit 1 (1) 0 0 Reactor Vessol Drawings, Unit 2 (1) 0 0 Emorgency Action Levels /EAQ, Unit 1 (1) 0 0 Emergoncy Action Levels /EALJ, Unit 2 (1) 0 0

2. Procedures/Documents: (BOOKSHELF)

Coro Operatinq Limits Report /COLRJ, Unit 2 (1) 0 0 Emerqoncy Operation Procedures, Unit 1 (1) 0 0 Emergoncy Operadcn Procedures, Unit 2 (1) 0 0 Emergency Preparodness Implementing Procedures

/EP/PJ 0 0 Emergency Proparodness Maintenance Prccoduros (EPMPJ 0 0 Final Safoty'Analysis Report /FSARJ, Unit 1 0 Q Final Safoty Analysis Ropcrt /PSARJ Supplements with Technical Supplements and Amendments (1) 0 0 INPO Roscurcos Manual (1) 0 0 Site Emorgoncy Plan /SEPJ (1) 0 0 Special Operating Procedures /SOPJ. Unit 1 (1) 0 0 Technical Spocificaticn Amendment Letters, Unit 1 (1) 0 0 Tochnicai Spocificaticns, Unit 1 (1) 0 0 Technical SpociTicaticns, Unit 2 (1) 0 0 Updated Safety Analysis Report (USARJ, Unit 2. (1) 0 0

4. Forms Cabinet 0 0 February 1998 Page 28 EPMP-EPP-02 Rev 14

ATTACHMENT 17 (Cont)

EOF EMERGENCY OPERATIONS FACILITY ftem/Equip mont Mln. Cty Sat Unset Correcdve Actfons Oste Resolved OOSE ASSESSMENT ROOM Maps with Overlays (1) 0 0 10 mile radius (1) 0 0 50 milo radius (1) 0 0 Map Arimuth Indicator (1) 0 0 Procedures/Oocum enter Emergency Preperodness lmplomenting Procoduros

/EiP/P/ 0 0 Emergency Preparedness Maintenance Procedures

/EPNP/ 0 0 Environmental Protectfon Manual cf Protective Action Guides and Protective Actions for Nucfoar Incidents /EPA~/ 0 0 Evacuadcn Travel Time Estimato 0 0 New York Stato Radiological Emergoncy Proparednoss Plan and Procedures 0 0 Oswego County Radiological Emergency Proparodness Plan Ca Procedures 0 0 Performed by Date Supervisor Approval Date E.P. Review Date February 1998 Page 29 EPMP-EPP-02 Rev 14

ATTACHMENT 18: M RG CY V NTILATION FILTER OG 1.0 ~ROCIRURE I.I Oetermine the time that the emergency ventilation ran during the past quarter.

1.2 Record the time (in hours this quarter) below. Send the sheet to:

John Oriscoll Unit I Technical Support 2.0 Complete the following:

quarter (Circle) I 2 3 4 Oate Checked (DD/HH/YY)

Checked by: Total Run Time Hours February 1998 Page 30 EPMP-EPP-02 Rev 14

ATTACHMENT 19: OP T ONS SUPPORT C NT R OSC Cl quarterly: 1 2 3 4 circle one (year)

Cl Post Drill/

Exercise/Emergency (date)

CI Other Corrective Date Item/Equipment Min. Qty Sat Unsat Actions Resolved Clocks (1) 0 0

2. Drawings/Diagrams:

Mechanical PAID Diagrams (1 set) 0 0

3. Forms Cabinet (1) 0 0

4. Procedures/Documents:

Damage Repair Procedures /DRPJ 0 0 Emergency Preparedness Implementing Procedures /EPIP) Q Q Emergency Preparedness Maintenance Procedures /EPM/ J 0 0 Site Emergency Plan (SEP) 0 0

5. Emergency Events Status Board 0 0

6. Telephones:

Outside Line (1) 0 0 TSC-Damage Control 5 Repairs (1) 0 0 TSC-Chem 5. Rad Mgt. (1) 0 0 TSC-OSC PA Speaker (1) 0 0 Performed by Date Supervisor Approval Date E.P. Reviell/ Date February 1998 Page 31 EPMP-EPP-02 Rev 14

ATTACHMENT 20: JOINT NEMS CENTER JNC CI quarterly: 1 2 3 4 circle one (year)

CI Post Drill/

Exercise/Emergency (date)

CI Other

~OTg: Nese are suggested kicadons for these items: however, the material may be found In other areas witgn the fecNlty.

AN computer equipment is checked by IdaC Department Computer Technicbins on a monthly basis, See completed Preventative Maiinenance Checklist.

Item/Equip ment Min. Qty Sat Unset Corracdw Aedons Data Resolved E4%IEFINQ AREA

1. Poster printers {2) 0 0 24 Poster printar paper (1) 0 0 COUNTY/STATE ROOM

1. 60-second clock- 0 0 24 V(dea Monitor/TV 0 0 UTIUTY ROOM

1. Clock 0 0

2. Camputor{s) 0 0

3. Emorgency Classification Signs:

~ Unusual Event (1) 0 0

~ Alort (1) 0 0

~ Sita Area Emergency (1) 0 0

~ General Ernorgency (1) 0 0 4, Printors {1)

5. Procedures/Documents:

~ Emergency Plan Implementing Procedures (EPIP) 0 a

~ Emergency Plan Maintenance Procoduras (EPMP) (1) 0 0

~ Sito Emergency Plan (SEP) (1) 0 0

~ Emergency Actloii Level Roferanco Manual (1) 0 0

6. Typewriter {1) 0 0

7. Video Monitor/TV ('I) 0 0 S. Dosk-top copier (1) a a

9. Diskottas (10) 0 0

10. Sign-off rubber stamp (1) 0 Ci STORAGE AREA Batteries

~ AA {8) Ci 0

~ C {8) Q 0

~ D (6) 0 0

~ 9V (6) 0 0 Forms:

~ Plant Status pastor {S 1/2 x 11) (50) 0 0

3. Mise. Office supplies:

~ Bulbs (EN X) (2) 0 0

~ Diskettes {10) 0 0 February 1998 Page 32 EPMP-EPP-02 Rev 14

~

~

- ATTACHMENT 20 JOINT N MS CENT R JNC (Cont)

Item/Equipment Min. Gty Sat Unset Correctfve Acdone Date Resolved STORAGE AREA (Continued)

~ Printer cartndges 0 0

~ Typewriter ribbons 0 0

4. Rubber stamper

~ Oriii 0 0

~ Exercise Only 0 0

~ Reviewed by Q 0

5. Telephone headsets 0 0 COPY ROOM (Supplios may bo in storage area)

1. Copy Machines (1) 0 0 20 Toner (1) 0 0 30 Ory ink cartridge (1) 0 0 4, Copier paper (1) 0 0

5. Tolocopy rubbor stamp (1) 0 0 B. Telocopy machinos (1) 0 0 70 Telecopier paper (1) 0 0 NRC/FEMA ROOM

1. Cock

2. Typewriter RUMOR CONTROL Forms

~ Media Response Log (50)

~ Ruiner Control Log (50)

2. Rumor Control Machine (1)

3. Cartridges for Rumor Control Machina (2)

4. Video cassette recorder/monitor (1)

MEDIA MONITORING Farms

~ lUledia Monitoring Log (50) 0 0

2. Video Cassetto recorders (1) 0 0 30 Video monitors (1) Q 0

4. Head phonos (1) 0 0

5. Radios (1) 0 0 AUDIO VISUAL AREA

1. Power Supply/Charger (1) 0 0 2e Tripod (1) 0 0

3. Audio cassottes (25) 0 0

4. V(deo cassattos (25) 0 0 February 1998 Page 33 EPHP-EPP-02 Rev 14

ATTACHMENT 20 (Cont)

JO NT N S C NT R JNC Item/Equipment Min. Cty Sat Unset Corrsctfve Actions Oats Resolved

5. Overhead projector (1) 0 0

6. Slide projector (1) 0 0 70 Projection screen (1) 0 0

8. RCA color video camera (1) 0 0 TV BOOTH AREA Audio dlstribudon amp (1) 0 0 2 Audio mixer (1) 0 0 30 Balt pack transmitter (1) 0 0

4. Camera remote control (1) Q 0

5. Diversity receiver (1) 0 0

6. Microphones (1) 0 0

7. Muid+ox (1) 0 0

8. Power empliRar (1) 0 0

8. Tripod (1) 0 0

10. VHS video recorder (1) 0 0

'I 1. Video/audio dlstribudon amp (1) 0 0

12. Video camera (1) 0 0

13. Video cassette recorders (3) 0 0

14. Video dataltima generator (1) 0 0

15. Video monitor (1) 0 0

16. Video switcher (1) Q 0 REQtSTRATION AREA Registration Logs:

~ Blue (so) 0 0

~ Pink (50) 0 0

~ Yellow (so) 0 0 Zo Badge Holders (200) 0 0

3. Badges

~ Blue (100) 0 0

~ Pink (100) 0 0

~ Yellow (100) 0 0

4. Press Kits:

~ Nine Mile 1 (10) 0 0

~ Nina Mila 2 (10) 0 0

~ NYPA (10) 0 0 Performed by Date Supervisor Approval Date E.P. Review Date February 1998 Page 34 EPMP-EPP-02 Rev 14

l ~, ~

ATTACHMENT 21A: DAMAG CONTROL TOOL BOX TNV NTORY MECHAN CA C3 quarterly: 1 2 3 4 circle one (year)

CI Post Drill/

Exercise/Emergency (date)

Cl Other

~gjgn: Unit 1 Screenhouse ttem/Equipment Mln. Qty Sat Unset Correct(w Actions Oste Reso(ved inventory Sealed MECHANlCAL TOOL UST)NC

1. Hack Saws (2) 0 0

2. 2'evel (1) 0 0

3. Wrecking Bars (2) Q 0 4, Crow Bar (1) 0 0

6. 1/2" Black Ea Decker Drill (1) 0 0

6. 1/4 Black if< Decker Driil (1) 0 0

7. 6" C.Clamps (2) 0 0

8. 6'ooden Rules (2) 0 0

9. 2 lb. Slugging Hammer (1) 0 0

10. Large Rubber Hammers (2) 0 0 12 oz. Machinist Hammers (2) 0 0

12. 16 oz. Machinist Hammers (2) 0 0

13. 50'xtension Cord (1) 0 0

14. 25'xtension Cord (1) 0 0

15. Low Voltage Lead Ught (1) 0 0

16. Fluorescent Ughts (2) 0 0

17. 3/4" Socket Sat 3/4 to 2" (1) 0 0

18. "1/16 to 1/2 by 1/64 Drill Indexes (2) 0 0

19. 18" Adjustable Wrench (2) 0 0

20. 12 Adjustable Wrench (4) 0 0

21. 10 Adjustable Wrench (4) 0 0

22. 7 Visa Grip Pliers (1) 0 0

23. 10 Vise Grip Pliers (1) 0 0

24. 1/2 Ton to 3/4 Ton Chain Fall (1) 0 0

25. 50'ength 1/2 Rope (1) 0 0

26. 6" Adjustable Wrench (4) 0 0

27. Duckbill Snips (2) 0 0

28. Straight Snips (2) 0 0

29. Regular Standard Pliers (2) 0 0 30 "

Largo Channel Lock Pliers (2) 0 0

31. Torpedo Levels (2) 0 0

32. 100'teel Tape (1) 0 0

33. 10 lb. Slugging Hammer (1) 0 Q

34. Screwdriver Set /Ror ond Phillips/ (1) 0 0

36. 1/2 Socket Sat 3/8" to 1 1/4" (1) 0 0

38. 1/4 Shackles (2) 0 0

37. 3/8 Shackles (2) 0 0

38. 1/2" Shackles (2) 0 0

39. Allan Wrench Sat (1) 0 0

40. 10 Pipe Wrarich (1) 0 0

41. 14 Pipe Wrench (1) 0 0

42. 18" Pipe Wrench (1) 0 0

43. Inspection Mirror (1) 0 0

44. Grey Tape (2) 0 0

45. Tape

'asking (2) 0 0 e46. Nuclear Grade Pipe Sealant (2) 0 0 Exp. Date:

Pairs Work Gloves (4) 0 0 Baling Wire (1) 0 0 February 1998 -Page 35 EPMP-EPP-OZ Rev 14

ATTACHHENT 21A (Cont)

DAMAGE CONTROL TOO BOX INV NTORY N CHANICA item/Equip mant . Min. Oty Sat Unset Conecthre Acdons Oats Resolved lNECHANICALTOOL USTlNQ (Cont)

48. Large Wire Brushes (2l 0 0

50. Smei( Wire Brushes (2l 0 0

51. Pair Esr Plugs (B) 0 0

52. G.F.I. (1) 0 0

53. 1" Putty Knife (1) 0 0

54. 2 Putty Knife (1) 0 0

56. 24 Pfpo Wrench (1) 0 0

56. Porte Bend Sow (1) 0 0

87. - 5/8" Shocklos (2l 0 0

58. 3/4" Shackles (1) 0 0

59. 3B" Pipe Wronch (1) 0 0

60. Nose Bag (1) 0 0

61. Res blight (2) 0 0

62. Navar-Seat (1) 0 0 B3. RTV ¹106 or equivalent (1) 0 0

~Chango batteries ovary 24 months, Last battery chango data:

NOTE: IF batteries or pipo sealant will"axbira before tha next inventory, thon ordor or obtain replacements.

Performed by Date Supervisor Approval Date E.P. Review Date February 1998 Page 36 EPHP-EPP-02 Rev 14

ATTACHMENT 21B: DAMAGE CONTROL TOOL BOX INV NTORY IjjtC 0 quarterly: 1 2 3 4 circle one (year)

CI Post Drill/

Exercise/Emergency (date) 0 Other

~L)~ca ittn: Unit 1 Screenhouse Item/Equipment Min. Qty Sat Unset Corrective Actions Date Resolved Inventory Sealed/Locked INSTRUMENTATION AND CONTROL LISTING

1. Hand Tool Box (21 0 0 I2. Digital DMM (1) 0 0

~

3 Test Gauge 0-30 PSI O.l Subd. (1) 0 0

~4 Test Gauge 0-100 PSI 0.5 Subd. (1) 0 0

~5 Digital Pressure Calibrator or equivalents (1) 0 0

~

6 Piuke Temperature Probe (1) 0 0

7. Current Source/Test Set (1) 0 0

8. Air Regulators (0-30 psig,0-100 psig,0-300 (3) 0 0 psig)

9. Meter Test Lead Set (1) 0 0

10. Soldering Gun (1) 0 0

11. Tubing Cutter (1) 0 0

12. Tubing Cutter-Spare Wheel (1) 0 0 1/4" Tubing Bender (1) 0 0

14. Pipe Wrench 6" l1) 0 0

15. Pipe Wrench 10" (11 0 0

16. Open/Box End Wrench Set ¹K-25 (1) 0 0

17. Nut/Screw Driver Roll Set (1) 0 0

18. Adjustable Wrench 4" (1) 0 0

19. Adjustable Wrench 6" (1) 0 0

20. Adjustable Wrench 8" (2) 0 0

21. Adjustable Wrench 10" (1) 0 0

22. Vise Grip Plier 7" (1) 0 0

23. Channel Loc Plier 7" (1) 0 0

24. Channel Loc Plier 10" (1) 0 0

25. Wire Stripper/Crimper (1) 0 0

26. Needle Nose-Stgt. 5 1/2" (1) 0 0

27. Needle Nose.Stgt. 6" (1) 0 0

28. Needle Nose.offset 5 1/2" (1) 0 0

29. Needle Nose-Offset 6" I'I) 0 0

30. Diag. Cutter - 4" (2) 0 0

31. Diag. Cutter - 5 (1) 0 0

32. Plier/Cutter Combination (1) 0 0

33. Holding Tweezers l1) 0 0

34. Allen Koy Set (1) 0 0

35. Hex Socket Driver Set (1) 0 0

36. Socket Set - 1/4" Drive (1) 0 0

37. Screwdriver-Standard 6 (1) 0 0

38. Screwdriver-Standard 4" (1) 0 0

39. Screwdriver.Phillips 6" (1) 0 0 Screwdriver-Phillips 4" (1) 0 0

41. Screwdriver-Phillips 3" (1) 0 0

42. Screwdriver-Pocket 2" I'I) 0 0

43. Screwdriver-Holding 3 (1) 0 0 Scrowdriver-Holding 4" (1) 0 0 Screwdriver. Holding 6" (1) 0 0 Screwdriver-Holding Combo (1) 0 0 February 1998 Page 37 EPMP-EPP-02 Rev 14

ATTACHMENT 218 (Cont)

DAMAG CONTRO TOO BOX INV NTORY IKC Item/Equipment Mln. CLty Sat Unset Carrecthre Acdons Oats Resolved INSTRUMENTATION AND CONTROL USTING (Cant)

47. Pocket Rule 6" (1) Q

48. Examinadon Mirror 1 (1) Q

48. Gauge Pointer Puller (1) 0

50. Alignment Teal (na~anductivo screw (1) 0 drivo/)

~'51. Hoctranic Grade Sil. Rubber, 1 Tubo Expiradan Oats:

52. "Snoop Leak Detector (1) 0 0

53. Black Hectrical Tapo (1) 0 0

54. 8 Ty.Wraps with Label (5) 0 0

55. 1/4" Copper Tubing (50') 0 0

56. 1/4 Tygan Tubing (50') 0 0

57. Oispasablo Surgeons Glavos (2) 0 0

58. White Masslin Wipes (2) 0 0

58. Surface Prop Cleaner (1) 0 0

60. 1/4" Whitoy Volvo SS-IVS4 (1) 0 0

61. 1/4 Whitey Valve 8-IVS4 (1) 0 0

62. Pens, Pencil Ik Papor Pad 0 0

63. Miscellaneous Fittings:

Nuts (1/4" S wage(ok) (20) Q 0 Innor Ferrulos (1/4" Swagelak) (20) 0 0 Outer Forrulos (1/4 S wagelak) (20) 0 0 l/4" NPT Male x 1/4" Swegelak Union (12) 1/4" NPT Male x 3/8" Swagelak Union (3) 1/4" NPT Male x 1/2 S wage(ok Union (3) 1/4 Swagelak Tee's (3) 1/8" NPT Female x 1/4 Swagelak Elbow 1/8 NPT Female x 1/4" Swagelak Union Q 0

'64. Nitrogen Tank with Cart 0 0 Hydra Test Oats:

65. Nitrogen Tank Accessories lin tool bax/

a. Throad Sealant Expiration Octo:

b. Regulator: Victor 443781 (1) 0 0

c. Tubing (1) 0 0

d. Adapter Fittings (1) 0 0

e. Instructions (1) 0 0 Thermamoter 504F - 2504F (1) 0 0 Safety Glasses (1) 0 0 Test Equipment Power Cord GF)

'Hydrastadc Tosting required at least every 5 years.

'NOTE: These instruments are nat maintained in this kit but are available fram the Unit 1 Meter and Test issue room.

"Ifthis item will expire bofaro the next inventory, then ardor or obtain replacements.

'I Performed by Oate Supervisor Approval Date E.P. Review Oate February 1998 Page 38 EPMP-EPP-02 Rev 14

ATTACHMENT 22: ECTRIC OAMAGE REPA R UIPM NT INVENTORY CI Quarterly: 1 2 3 4 circle one (year)

C3 Post Orill/

Exercise/Emergency (date)

Cl Other Location: Unit 1 Storeroom Item/Equipment Sat Vnsat Correct(ve Actfons Date Reso(ved Inventory Sealed 0 0 500 Ft Triplex 4/0 Cu 5 KV Insulated Cable with 1/0 Cu. 5KV Insulated Ground

2. 1000 Ft Triplex ¹2 AWG Cu, 600V Insulated Cable 0 0

3. 20 Ft 1 Conductor ¹10 SIS Wire C) 0

4. 20 Ft 1 Conductor ¹12 SIS Wire 0 a

~

5 600 Ft 1 Conductor ¹4/0 0 C3

~

6 600 Ft 1 Conductor ¹2 AWG 0 0

7. T35 Tape (min. 12) 0 0

8. T95 Tape (min. 12) C3 0

9. 3M 88 Tape (min. 12) C) 0

10. 2 Kellems Cable Support Grips Model No. RR250-HE or equivalent 0 0 2 Kellems Cable Support Grips Model No. RR150-HE or equivalent 0 C3 i2. 8 Burndy Hyline No. YS28, ¹4/0 Splices or equivalent C3 C)

13. 2 Burndy Hyline No. YS2C, ¹2 Splices or equivalent 0 Q

'I 4. 1 Burndy Hylink No. YSM27, Parallel Splices or equivalent C3 C!

15. 1 Burndy Hylink No. YSM25, Parallel Splices or equivalent 0 0

16. 3 Burndy Hylug No. YA28-2N 4/0 Terminal or equivalent 0 0

17. 1 Burndy Hylug No. YA25-2N 1/0 Terminal or equivalent 0 0

18. 8 Burndy Hylug No. YAZC-2N ¹2 Terminal or equivalent 0 0

19. 2 Burndy Reducing Adaptor No.

Y2825R or equivalent (4/0 to 1/0) 0 C}

20. 2 Burndy Reducing Adaptor No.

Y2826R or equivalent (4/0 to 2/0) 0 0

21. 4 Burndy Hylug Ring - Tongue Terminals - No. YAV10-T3 or equivalent 0 0

22. 2 Fuse 6 Amp (for Powerboard 171 Control Circuit) Cl C3

23. 1 Burndy Hytool Crimping tool MY28 or equivalent 0 Q

24. 1 Burndy Crimping Tool MY29-3 or equivalent 0 0 Breaker Elevator Hand Crank (GE for

. Magnet Blast Circuit Breaker/ 0 0 February 1998 Page 39 EPMP-EPP-02 Rev 14

ATTACHMENT 22 (Cont)

ECTRIC DAMAG R PAIR UIPMENT INVENTORY item/Equipment Sat Unset Corrective Actions Date Resolved

26. Hacksaw and 20 extra blades Ci 0

27. 5/8" Ratchet Wrench (for Breaker Closing Spring Charging/ 0 0

28. 2 sets - Wrenches and Screwdrivers to Cable and Wire Disconnection 0 0

29. 2 sets - Cable Cutting and Splicing.

Tools a 0

30. 2 Insulated Fuse Pullers 0 0 31 ~ 3 Sets - 8us Grounding Cables a a (Materia/ for 3 setsJ

32. Fire Retardant Putty 0 - 0 4 ¹12 AWG Ring-Tongue Terminals 0 0 4 Portable Compressed Air Cylinders 0 0

35. 1/2 x 3/4 NPT 8ushing 0 0

36. 3/4 NPT Street E11 Ci 0 Air Regulator Assembly 0 0

38. 10 Ft High Pressure Air Hose with Swivel Fitting 0 0

'39. Cable Quad ¹4 and ¹6 0 0

40. Cable Lugs ¹4 and ¹6 Cl 0

41. Safety Switch, 600 Volt/200 Amp 0 0

'42. Portable 60 KW Generator (located at Bui(ding 008 in Level B StorageJ 0 0 High Pressure Hose (Jijmper R915 and R925 Air Samples J 0 CI NOTE: ' unsealed inventory. All other equipment is in sealed tool box.

Performed by Date Supervisor Approval Date E.P. Review Date February 1998 Page 40 EPMP-EPP-02 Rev 14

A1TACHHENT 23: T MPORARY R STORATTON OF POM R FOR POST ACCIDENT SAHPL NG

~INV NTORY CI quarterly: 1 2 3 4 circ1e one (year )

0 Post Ori11/

Exerci se/Emergency (date)

C1 Other Lgg@ fgn: Unit 2 Control Building ltern/quupment Unset Corrective Actions Date Resolved inventory Sealed E3 0

$9~ Jumpers are 1/C, No. 12 AWG lNJN-59) 1.

237'at (SR) Nominal 4 feet length with lugs

(¹10 stud) 6 Jumpers, stored inside Panel 2CES PNL564, East Wall, Oiv. I, Cable Spreading Area, Ei.

2. Test Box Jumper, per E061A in accordance with OWG. EE%03X Rev. 01, located in North East Comer of Control Room

3. Located in SSS Office

a. Key ¹CAT60 - for 2VBS PNL102A, a a 302A, AND 2LAC-PNLU03 a ci

b. Key ¹11-CH751 - for 2CES PNL564 Performed by Oate Supervisor Approval Oate E.P. Review Oate February 1998 Page 41 EPMP-EPP-02 Rev 14

~ ~ ~ ~

ATTACHMENT 24: RG CY R SPO S FAC TY OMMUN CATIONS SURV I NC 1.0 G E GUIDE INES 1.1 Determine the required testing using the matrix in Section 2.0.

1.2 Perform the'esting of each communications system in accordance with the associated attachment.

1.3 The surveillance is considered successful checked.

if all "Sat" boxes are 1.4 Initiate corrective actions on all "Unsat" entries in accordance with Step 3.0.

a. Record details of failure and initiated corrective actions in appropriate "Remarks" section.

b.. After repair/correction, perform surveillance (only with agency that was "Unsat") and record on new attachment.

2.0 RE VIREO TESTING FRE UENCY

' + +ykS g

Unit 1 Control Raam

.'AECS

~y a%gs'-;;

Telephone~~;;Tetephan'e:

R'iiV. a iS.~2~ 'o+ 'k~~<:&:) '.".'A<<sH A

a< '>~~ ~HO~'a+<

Unit 2 Control Roam A EOF OSC TSC JNC M = Monthly A Annual(y PERFORMED 8Y NYPA 3.0 REPORTING'ROBLEMS 3.1 Radiolo ical Emer enc Communication S stem REGS Failure Report all .failures to 518-457-2200 during the hours of 9 am to 4 pm.

3.2 Radi o Failures Contact the Central Region Communications Group at 460-2378 or 460-2379.

February 1998 Page 42 EPMP-EPP-02 Rev 14

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ATTACHMENT 24 (Cont)

G NCY R SPONS FAC TY COMMUN CATIONS SURV I ANC 3.3 Commercial Tele hone and Dedicated ines Complete a "Telephone Request Form" and fax to Facilities in accordance with the instructions on the form.

~OT : Mith a Dedicated Line, use the "Circuit Number" in place at the "Extension" number on the "Telephone Request Form".

3A ~th

a. Immediately report any "Unsat" results as follows:

g:~gags Q -v: $ A'io g +'c gj,jjcA'~~<

Control Room, Unit 1 Unit 1 SSS Control Room, Unit 2 Unit 2 SSS Both TSC ENS Phones Unit 1 SSS

b. Report failure to NRC Operations Center at one of the following numbers.

~ (301) 816-5100

~ (301) 951-0550

c. IF requested by the NRC Operations Center, call NYNEX, (315) 479-2161, for assistance.

February 199S Page 43 EPHP-EPP-OZ Rev 14

ATTACHMENT 25A: H RG CY R SPONS FACI ITY COMMUNICATIONS SURV I NC RAOIOLOGICAL EMERGENCY COMMUNICATIONS SYSTEM (RECS) TESTING (MONTHLY) 1.0 ~ROC D~UR 1.1 Pick up the handset and dial A>>.

~NOT : Oepress push to talk switch in the handset to talk.

1.2 After. about 15 seconds state the following:

"This is a test. This is the Nine Mile point (location) calling all stations for a RECS test. Stand by for roll call."

1.3 State each agencies name as they appear on the RECS Testing Sheet. As each agency responds, check "Sat" or "Unsat".

~NOT : "Sat" agency responded without comment "Unsat" anything beside "Sat" response 1.4 Repeat Step 1.3 for any agency not answering roll call.

1.5 When roll call is completed, state:

"This concludes the test. Thank you."

1.6 Should an agency fail to answer, contact them by telephone, and if necessary, repeat Steps 1.1 through 1.3 for the problem agency only.

February 1998 Page 44 EPHP-Epp-02 Rev 14

ATTACHME>, A (Cont)

EMERGENCY RESPO SE FACILIT C ICATIONS SURVEILLANCE RECS TESTING SHEET Month Year

,'Tempted PY Prom Agency itI'.

C'elephori"e (".:'~ rj'pemitks .

Unit )'CR Unit 2 CR 349-2480 N/A 0 Sat Nine Mile Point Unit 1 CR 0 Unsat 349-2170 0 Sat N/A Nine Mile Point Unit 2 CR 0 Unsat 349-6666 0 Sat 0 Sat Fitzpatrick CR 0 Unsat 0 Unsat 911 0 Sat 0 Sat Oswego County 911 Center 0 Unsat 0 Unsat 598-1191 0 Sat 0 Sat Oswego County EOC 0 Unset 0 Unsat NYS Warning Point I 51 Sl 457-2200 0 Sat 0 Sat 0 Unset 0 Unset 593-5735 0 Sat 0 Sat EOF 0 Unsat 0 Unsat Tested by:

Initials/Date Supervisor Approval Date E.P Review Date Febt uaty 1998 Page 45 EPMP-EPP-02 Rev 14

ATTACHMENT 258: G NCY F C ITY COMMUN CA 0 S SURV NC COMM RC T HON T ST G MON 1.0 ~PROCEDUR 1.1 For each "Location" listed, test the telephone. by placing and receiving a call to any other telephone.

1.2 Check to "Sat" or "Unsat" box on the "Commercial Telephone Testing Sheet".

r ~

~OT: "Sat" satisfactory transmission and reception "Unsat" anything but "Sat" response Page 46 EPMP-EPP-02 February 1998 Rev 14

ATTACtNfNl (Cont)

GENC C I Y CO U 0 S SUR LA Cf COMMERCIALTELEPHONE TESTING SHEET Month Year

~4/'oatea;by~:~~

EOF Comm Area 593-5875 0 Sat 0 Unsat TSC Comm Rm 349-2487 0 Sat 0 Unsat Offsite Assembly Area 592-0125 no test required'nit 1 Control Room no test required'nit 2 Control Room no test required'oint News Center 592-3720 0 Sat in Rumor Control) 0 Unsat

'No test Is required from the Control Rooms or Offsite Assembly Area since their telephones are used regularly.

Supervisor Approval Date E,P Review Date February 1998 Page 47 .. fPHP-f PP-02 Rev 14

~ > ~

ATTACHMENT 25C: RG CY FAC TY COMMUNICAT ONS SURV ANC RG NCY NOTIFICATION SYST M NS T STING MONTH Y 1.0 ~PR ~D~R 11 r n lRo ms

a. Solicit the time of the daily plant operations status call from the NRC Operations Center to the Control Room from the SSS.

b. Record "Sat or "Unset" on the ENS Testing Sheet.

~N: "Sat" satisfactory transmission and reception "Unsat ~ anything beside "Sat" response 1.2 ~For 7 C

a. Verify the operability at each ENS phone listed on the ENS Testing Sheet by placing and receiving a call from any other ENS phone.

b. Record "Sat" or "Unsat" on the ENS Testing Sheet.

~N: "Sat" ~ satisfactory transmission and reception "Unsat" anything besides "Sat" response February 1998 Page 48 EPMP-EPP-02 Rev 14

~ I ~

ATTACHMENT 25C (Cont)

MERGENCY FACILITY COMMUNICATIONS SURVEILLANC ENS TELEPHONE TESTING SHEET Month Year CoNTRoL RooM UNIT 1 Daily Operations Status Call: Date Time (24 HourJ 0 Sat. 0 Unset CotnaoL RooM UNtt' Daily Operations Status Call: Date Time (24 HourJ 0 Sat 0 Unsat TSC

'Fnona ~,','.'.'k.Phonii" Na;::," '..

ENS 700-371-5324 NRC Room ENS 700-371-5324 Tech Assessment Room 0 HPN 700-371-6329 NRC Room HPN 700-371-5329 RAM Desk PMCL 700-371-5326 NRC Core SCL 700-371-5327 NRC Core MCL 700-371-6323 NRC Room 0 TESTED BY: Initials/Date

~N: EOF testing completed by NYPA.

Supervisor Approval Date E. P. Review Date Page 49 EPMP-EPP-02 February 1998 Rev 14

1 THIS PAGE INTENTIONALLY LEFT BLANK February 1998 Page 50 EPHP-EPP-02 Rev 14

ATTACHMENT 25D: G C FAC TY COMMUNIC T ONS SURV ANC Dl N ANN Y 1.0 ~PR gggg+

1.1 The dedicated line will automatically ring or flash the other end when the handset is lifted.

1.2 Verify that someone is available at the other end to test.

1.3 Verify proper operation by initiating, receiving, and transmitting from each end of each line listed'on the "Dedicated Telephone Testing Sheet".

1.4 As each line is tested, mark "Sat" or "Unsat" on the Testing Sheet.

~N: "Sat" "Unsat" =

properinitiating, receiving, and transmitting from each end anything other than "Sat" February 1998 Page 51 EPMP-EPP-02 Rev 14

ATTACHMENT 25D (Cont)

MERGENCY FACI ITY COMMUNICATIONS SURV ILLANCE DEDICATED TELEPHONE TESTING SHEET Year 1 Roots E.O. Hotline 36 LCGL 199800 . 0 Sat 0 Unset CR¹1-TSC ¹63PLNT22750 . 0 Sst 0 Unset CR¹1-JAF C.R. ¹63PLNA28109 0 Ssc 0 Unset Tach Info Une 63 PLNA 37227 0 Sec 0 Unset Ramarkst TESTEO BY: Initials/Deca U Corrrn Roots CR¹2-TSC SEO 0 Sat 0 Unset E.D. Hodine 36 LCGL 199800 . 0 Sat 0 Unset CR¹2-JAF C.R. ¹83PLNA34299 0 Sat 0 Unset Tech Info Une 63 PLNA 37227 0 Set 0 Unset Remarks:

TESTED BY: Initials/Data Tach Info Une 63 PLNA 37227 0 Sat 0 Unset E.D. Hodina 36 LCGL 199800 . 0 Sat 0 Unset CED/SED Hotline 63 PLNA 37200 0 Sst 0 Unset Remarks:

TESTEO BY: Initials/Date Tach Info Une 63 PLNA 37227 0 Sat 0 Unset E.D. Hotline 36 CGL 199800 0 Sat 0 Unset TSC-EOF Security ¹63 PL-16919 0 Sat 0 Unset TSC-OSC I6tC Coord. ¹63 PL-16969 0 Set 0 Unset TSC-OSC SSST Coord. ¹63 PL-16918 0 Set 0 Unset CED/SED Kodine 63 PLNA 37200 0 Sac 0 Unset TSC-CR¹ 1 S.E.D. ¹63 PLNT 22750 0 Sat 0 Unset TSC-CR¹ 2 S.E.D. 0 Sat 0 Unset TSC-JAF/CR IU1) ¹63PLNA28109 ~ ~ \ ~ 0 ~ 0 Sec 0 Unset TSC-JAF/CR IU2) ¹63LAOA34299 .. ... 0 Sat

~ 0 Unset Remarks:

TESTED BY: Initials/Date JNC Tach Info Une 63 PLNA 37227 . 0 Sat 0 Unset Remarks:

TESTED BY: Inidels/Data OSC OSC Chem/RP - TSC ¹63 PL-16918 0 Sat 0 Unset OSC Damage Ctrl -.TSC Maint Coord. ¹63 PL-16969 0 Set 0 Unset Remarks:

TESTED BY: Inidels/Date Supervisor Approval Date E. P. Review Date February 1998 Page '52 EPMP-EPP-02 Rev 14

~ lg I i ~

ATTACHMENT 25E: EM RG NCY FAC ITY COMMUNICATIONS SURVE LANC RADIO CONSOLE TESTING (ANNUALLY) 1.0 ~RO~IRURE',

1.1 Testing from the TSC, Unit 1 or Unit 2 Control Room

a. Turn the volume knob on the Select Audio speaker to the twelve o'lock position.

b. Depress the "Volume" button on the "Rad/Teams" module until the light next to "full" is lit.

c. Utilizing a person equipped with an EP portable radio, verify the selected channel, and depress the "Transmit" button and give a short test message to the portable radio.

d. Repeat Steps a through c for all required channels as per the Radio Console Testing Sheet.

e. Record "Sat" or "Unsat" on the Testing Sheet.

NOTE: "Sat" satisfactory transmit and receive "Unsat" - anything beside "Sat" response 1.2 Testing from the EOF

a. Turn the volume knob to the twelve o'lock position.

b. Select channel to be tested using the up-arrow or down-arrow

. buttons until the desired channel number is displayed.

c. Utilizing a person equipped with an E.P. Portable Radio, on the same channel, depress the "transmit" bar on the microphone and give a short test message to the portable radio.

d. Repeat steps a through c for all required channels, as per the Radio Console Testing Sheet.

e. Record "SAT" or "UNSAT" on the Testing Sheet using the criteria in l.l.e.

February 1998 Page 53 EPMP-EPP-02 Rev 14

~ ~

~ g ATTACHMENT 25E (Cont)

RG NCY FACI ITY COMMUNICATIONS SURV I ANC RADIO CONSOLE TESTING SHEET Year Y.g>> ' ~'@5'~<++:-.Mmk ~<+Dr"+MKV+AXwV4+P'x~%~'g pC;.)'jj~+~~ '.:>;. ~~asm >e4$ ~>i, ~< 4j~~@<X w'c y>~'. 'Z~~'4('>S, '~ a~a)

Admin Rad Teams Unit Control Room 1

(one console onlyJ CI Sat 0 Sat 0 Unsat 0 Unsat Admin Rad Teams Unit 2 Control Room Cl Sat CI Sat (one console onlyJ 0 Unsat Cl Unsat Admin Rad Teams EOF (Rad Assmt Rm onlyJ 0 Sat 0 Sat 0 Unsat 0 Unsat Admin U1 Fire U2 RP U2 Maint U2 Fire TSC (Red Assmt Rm onlyJ Cl Sat Cl Sat 0 Sat CI Sat 0 Sat 0 Unsat Q Unsat Ci Unsat 0 Unsat CI Unsat Ul Fire U2 RP U2 Maint U2 Fire OSC Cl Sat Cl Sat 0 Sat 0 Sat 0 Unsat 0 Unsat 0 Unsat 0 Unsat Remarks:

Supervisor Approval Date E. P. Review Date February 1998 Page 54 EPMP-EPP-02 Rev 14

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(

ATTACHMENT 25F: G C FAC TY COMMUNICA ONS SURV ANC PORTABLE RADIO TESTING (ANNUALLY) 1.0 ~ROCEOU 1.1 Portable radios are tested by calling another radio and having another radio call back.'.2 Turn on the radios to be .tested and select any available onsite channel.

1.3 Transmit a short test message. Verify transmission on another radio.

1.4 On the other radio, transmit a short test message. Verify reception on the other radio.

1.5 Check "Sat" or "Unsat" on the Portable Radio Testing Sheet.

NOTE: "Sat" proper receive and transmit "Unsat" anything beside "Sat" response February 1998 Page 55 EPMP-EPP-02 Rev 14

ATTACHHENT 25F (Cont)

G NCY AC TY OM VNICATIONS SVRV I NC PORTABLE RADIO TESTING SHEET Year

'UNSAT OSC Core HT4 0 HT4 Cl HT4 0 HT4 0 HT-¹ 0

2. OSC Storeroom Habitability .. HT4 Cl 0 PAS Sample ~ .. HT¹ 0 0 PAS Analysis HT-¹ 0 0 Downwind B HT-¹ CI Downwind C HT-¹ 0 0 Inplant 1 HT-¹ CI 0 lnptant 2 HT-¹ 0 0 Inplant 3 HT-¹ 0 0 Inplant 4 ~ .. HT¹ 0 Cl In'plant 5 HT-¹ 0 0 OSC Spares HT4 0 Cl HT-¹ 0 Cl HT4 0 0

3. RP Fire Response Unit 1 (7B 248'J Unit 2 (RBACBJ HT-¹ HT-¹ Offsite Assembly Area Facility (OAAJ Offsite... ~ ~ ~ ~ ~ ~

HT-¹'T-¹ Offsite... 0 0 ffslte ~ ~ ~ ~ ~ ~ * ~ ~ ~ ~ ~ ~ ~ ~ HT-¹ 0 0

5. Emergency Operation Facility (EOJ=J

.0 ffsite.............. HT-¹ Offsite... HT4 0 ffslte ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ HT-¹

6. Joint News Center (JNCJ HT-¹

7. Vehicles Env. Prot ¹6-243 or 0 0 EP ¹2-1077 or Cl 0 EP ¹6-484 or 0 0 TESTED BY: Initials/Date Supervisor Approval Date E. P. Review Date February 1998 Page 56 EPMP-EPP-02 Rev 14

ATTACHMENT 25G: PORTABL RAOIO BATT RY CHANG UART RLY

~NO  : One week prior to this test, request replacement batteries From the Radio Shop in sufficient quantities to accommodate all HTs listed in Attachment 25F.

1.0 ~ROC ~UR 1.1 Remove the battery attached to the portable radio.

1.2 Obtain a replacement battery and verify the date to be less than 3 months old.

1.3 Attach the replacement battery to the portable radio.

1.4 Replace portable radio in charger.

1.5 When all batteries are replaced:.

a. Complete "Portable Radio Battery Exchange Sheet"

b. Send old batteries to Radio Shop.

Portable radio battery exchange completed for the quarter of (year)

Remarks:

Exchange Performed By: Initials/Oate Supervisor Approval Oate E.P. Review Gate February 199S Page 57 EPHP-EPP-02 Rev 14

ATTACHMENT 26A: R SPIRATO UIPM NT MONTH Y NSP CT ON Q Month 0 Post Drill/Exercise/Emergency Cl Other:

Onsite No. Resp./No Canister Voice Amp Bat Use Sat S Unset Location Canister Due Date Due Date

1. Ambulance and Fire U2 Security ACR

2. Security Building U1 Sec Gun Locker 8/8 (2/2)

Emergency

3. Security Building U2 Sec Gun Locker 8/8 Emergency

4. Control Room U2 Control Building 306'0/10 5, R.P, Supplies &

U1 Storeroom 40/80 Equipment MSA Duo-

6. Post Accident Sampling U1 Storeroom Flow 4 Systems Respirator Performed By Date Details/Items Resolved Supervisor Approval Date E. P. Review Date February 1998 Page 58 EPNP-EPP-02 Rev 14

ATI'ACHMENT 268: RESPIRATORY E UIPMENT MONTH Y INSP CTION 0 Month 0 Post Drill/Exercise/Emergency 0 Other:

Offsite No. Resp./No Canister Use Location Canister Due Date Sat Unsat

1. R.P, Supplies 5, Equipment EOF 10/20 0 0 Performed By Date Details/items Resolved Supervisor Approva1 Date E. P. Review Date February 1998 Page 59 ~ EPMP-EPP-02 Rev 14

ATTACHMENT 26C: R SPIRATORY U PM NT MONTH Y INSP CTION SCOTT PAK Cl Month 0 Post Drill/Exercise/Emergency CI Other.

Gate Inspection completed per S-RPlP-4A Verified by Date Locations inspection Completed by Date Unit 1 Control Room Name:

Pak's and 7anks) 277'Scott Signature:

2. Unit 1 Turbine Building Name:

Pak's and 7anksJ 261'Scott Signature:

3. Unit 1 Screen House Name:

Pak's and 7anksJ 261'Scott Signature:

4. Unit 1 Admin Building Name:

Pak's and TanksJ 261'Scott Signature:

6. Unit 1 Store Room Name:

(Scott Pak's and 261'Spares) 7anksJ Signature:

6. Unit 2 Control Room Name:

Pak's and 7anksJ 306'Scott Signature:

7. Unit 2 Turbine Building Name:

Pak's and 7anksJ 250'Scott Signature:

8. Unit 2 Screenwell Name:

Pak's and TanksJ 261'Scott Signature:

8.. Unit 2 Access Passage Name:

Pak's and TanksJ 261'Scott Signature:

10. Emergency Response Vehicle Name:

32-7-1 (Scott Pak's and Tanks/ Signature:

Supervisor Fire Protection Date Details/Items Resolved Supervisor Approval Date E. P. Review Date February 1998 Page 60 EPMP-EPP-02 Rev 14

ATTACHMENT 27: HAZARDOUS WASTE AND EMERGENCY SPILL R SPONSE KIT INVENTORY

~oc tion: Unit 1, T.B., 261

~ ~ Ct (quarterly: 1 2 3 4 Unit 1/2, Passageway circle one (year)

C1 Post Drill/

Exercise/Emergency (date)

Cl Other Its rnlEquip ment Min. Oty Sot Unset Correcdve Acdone Oats Resolved inventory Sealed 0 0 Garment Store e Locker

1. Chemical Splash Goggles (3) 0 0

2. Chemical Splash Shields (3) 0 0

3. Chemico) Rosistant Gloves l3pr) 0 0 Inventory Sealed ti Pu oee Safe E ui . Store e Locker Chemical Splash Suits fpeckegedJ (2) SM, (2) MEO, ftt (1) LG (S) 0 0

• Chemical Splash Goggles (S) 0 0

0

'3 0 Chemical Face Shields (5) 0 4, Chomical Resistant Gioves (5 pr) 0 0

5. Ouct Tape (2 rolls) 0 0 S. ()lank "Oenger" Signs (1O) 0 0

7. Roar Stand Signs "Oongor Chomical Spill - Keep Away" (3)

o. Roolod Barrier Tape "Caution Chemical Spill" (3) Q 0 "Caution - Oo not Enter" (3) 0 0

8. Acid Neutralization Kit (1) 0 0

10. Caustio Neutralizadon Kit (1) 0 0

11. Solvont Noutrolization Kit (1) 0 0

'2. Absorbants (conteinst pillowslblenkets/ebsorbontsl (1 Drum) 0

13. Plug Kit (1) 0

14. 2 Whee) Hand Cart (1) 0 Performed by. Date Supervisor Approval Date E.P. Review Date February 1998 Page 61 EPMP-EPP-02 Rev 14

~ ~

~

ATTACHMENT 28: ALT RNAT POW R SUPPLI S FOR PORTAB AI SAMP S Cl guarter1y: 1 2 3 4 circ1e one (year)

Cl Post Ori11/

Exercise/Emergency (date)

Cl Other EMiatomcv Vaacaa A. C. Images Vehicle Number Operation: Sat Unset A. ¹2-1077 /Emergency PreparednessJ B. ¹5-484 (Emergency PreparednessJ 0 C. Other D. Other

~NTE: Perform test with vehicle operating, using an AC-High Volume Air Sampler and run for 5 minutes.

DH'atLs/lees REsoLvEo By Date Performed By Supervisor Approva1 Date E. P. Review Date February 1998 Page 62 EPHP-EPP-02 Rev 14

~ l~ le I p ATTACHMENT 29: N2- OP-6 TOO BOX FOR BY-PASS OF STAND-BY GAS N -PM- 008 Cl quarterly: 1 2 3 4 circle one (year)

El Post Drill/

Exercise/Emergency (date)

CI Other LggQi~n: EOP Box El. 261'nder stairway off the Rx. Track Bay Item/Equipment Nlin. Qty Unset Corrective Actions Date Resolved Inventory Sealed 1" Nylon Sling 6 ft. long (1)

2. 2" Nylon Sling 8 ft. long (1)

3. 2" Nylon Sling 10 ft, (1) long

4. Two Ton - Ten foot Chain Falls (2)

5. 5/8" Shackles (2)

6. 3/8" Shackles (1)

7. 3/4" Shackle (1)

8. 3/8" Nut Drivers (2) 5/'I6" Nut Drivers (1)

10. 1/4" Nut Drivers (1)

11. 1/4" Rachet, 1/4 Drive (1)

12. T/4" 8reaker Bar, 1/4 (1)

Drive

13. 1/4" Socket, 1/4 Drive (1)

14. 7/16 Socket, 1/4 Drive (2)

15. 7/16 Deep Well Socket, (1) 1/4 Drive

16. 3/8 Socket, 1/4 Drive (1)

17. 5/16 Socket, 1/4 Drive (1)

18. 12" Extension, 1/4 Drive (1)

19. Pry Bar (1)

20. 1-13/16 Combo (2)

Wrenches

21. 1-1/2 Combo Wrenches (2) 0

22. 1-1/4 Combo Wrenches (2) 0

23. 7/8 Combo Wrenches (2) 0

24. Flanges (2) 0

25. Flexitallic Gaskets (2) 0 Performed by Date Supervisor Approval Date E.P. Review Date February 1998 Page 63 EPHP-EPP-02 Rev 14

ATTACHMENT 30: G CY FACI ITI S STING Red Monitonng Equipment (OSC/7SC/Ons/taDowmvina7 ln Box for Lh1 RP ONce:

1. Whole Body f7XDJ (50)

2. Extromity (RingsJ (40 pr} (1 pr}

3. Oosimotors (0-5RJ (20)

4. Oosimatars (0-5ORJ (20)

5. Oosimotore /0-200RJ (5)

6. Dosimetry lssuo Sheets (2)

Red Monitoring Equ)pment Emergency Operet/ons FeciVty /n Bc'cr EOF (contact environments/J

1. Whole Body /7LDJ (100)

2. Oosimatars /O-500mr) (4)

3. Dosimators (0-5RJ (4) 4, Oosimatars (0-50RJ (4)

5. Dosimetry Issue Sheets (2)

Ambulance Ea Rre Kit ln Box for Il-2 Security Whole Body ITLDJ (50) (2)

Extremity fRingsJ (I) pr) (1 pr) 34 Dosimetry lssuo Sheets (2)

Oswego Hospital ln Box for Oswego Hospital: (contact anvironnrente/J Whole Body /TLDJ (10)

2. Extremity /Rings/ (10 pr) (1 pr)

'3. Oosimators fO-500mr/ ( 10)

'4. Oosimators (0-20RJ (10)

5. Dosimators /0-/.5RJ I}. Dosimetry issue Shoots (2)

Should bo placed in plastic bags as 10 sets. Each sat contains ono of each item.

February 1998 Page 64 EPMP-EPP-02 Rev 14

ATTACHMENT 31: E 4 L ISSU S Fac)lity/Kkt Location TLD DATE ON MHOLE ISSUED RETURNED RESULT TLD NNE EXTREMITY BODY SSt SRPD8 DATE/TINE DATE/TINE mRem RENRKS NUMBER

• DO NOT 1SSUE CONTROL TLD TLD NUMBER TLD NUMBER February 1998 Page 65 EPMP-EPP-02 Rev 14

~ ~

s ATTACHMENT 32: NINE MILE POINT NUCLEAR STATION PROC SS RAD MONITORING BOARD - UNIT I Oate (HH/00/VV)

Trendm Monitors Trend Main Steam Uncs mR/hr 11 121 fnR/hr 12 112 mR/hr mR/hr Contfdnment HIgh Range - DW 263'h 11 R/hr 301'h 12 R/hr mR/hr 121 mR/hr Reactor Building PING 112 mR/hr pCi/cc 122 mR/hr cpm pCi/cc cpfn pCi/cc Drywall CAM cpm Turbine Buflding PING P CP foal pCi/cc Reactor Building Vent Radladon cpm pCi/cc Ch 11 mR/hr NG cpm pCi/cc Ch 12 mR/hr Radtvaste 261'ING Service Water cpfn pCi/cc cpm pCi/cc cpm pCi/cc Radwaste Discharge A cps B cps cpm pCI/cc cpm pCi/cc Stack ENuent cpfn . pCI/cc 03 tptsl apm pC csea 08 i08el apm pCils a Stack Flow 3a Itpel apm pCilsea KCFM 3b /10/f/ apm pCilsaa High Range Stack Effluent Offgas mR/hr Ch 11 mR/hr Ch 12 mR/hr

'rend Symbols: t ~ Increasing 4 ~ Decreasing pm a No Chango February 1998 Page 66 EPMP-EPP-02 Rev 14

ATTACHMENT 33: NINE MILE POINT NUCLEAR STATION PROC SS RAD MONITOR NG BOARD - UNIT Date (W/DD/YY)

Monitor f¹INamr//Reading Tran de Monitor (/J/NrmrJ(Reading Trendr GEMS-TB/S~tack RE 170 Station (/Menus/J Contsinm't High Rg DryweN Area El 261

1. Panfcufate pCI/sso 79-RMS1 A R/hr

2. Iodine pCI/soc 88-RMS1B R/hr

3. Noble Gas pCf/seo 80-RMS1C R/hr Stack Flow SCFM 89-RMS10 R/hr GEMS-Rx/RW Bfgd-Vent RE 180 Station ((t/(anus/J Above Suppression Pool

1. Particulate yCilsoc 27-RMS139 R/hr

2. Iodine pCilsoc Main Steam Rsd Monitor (/Ifanua/J.

3. Noble Gss pCi/soo MSS 48A mR/hr Stack Row SCFM MSS 468 mR/hr Servfce Water Mani'tora MSS 46C mR/hr 82-SW146A. yCi/ml MSS 460 mR/hr 91-SW146B pCI/ml Continuous Air Mon. (Drywo// Atmos./

Rad Waste Uquid Effluent Monitor 74-CMS10A-Ch 1 pCilcc 8-LWS206 pCi/ml Ch 2 pCilcc Cooling Tower Slowdown 83-CMS108-Ch 1 yCi/cc 70-CWS157 pCi/ml Ch 2 pCi/cc Service Water Monitors Rx Bldg Vent/Recirc Modo (SGTS On/

81-SWP23A pCi/ml 39-HVR229-Ch 1 yCi/cc 90-SWP238 pCI/ml Ch 2 pCI/cc Reactor Building Ventilation (SGTS off/

Above Auxiliary Bay Vent N.

77-HVR14A-Ch 1 pCI/cc 34-HVR237-Ch 1 pCi/cc Ch 2 pCi/cc Ch 2 yCi/cc 86-HVR14B yCi/cc AuxiTiary Bay Ven! S.

Below 35-HVR238-Ch 1 yCi/cc 78-HVR32A-Ch 1 pCi/cc Ch 2 pCi/cc Ch 2 pCi/cc Turbine Building Vent 87-HVR32B pCI/cc 65-HVT206-Ch 1 yCi/cc Standby Gas Treatment (Post Tiestm'tJ Ch 2 pCi/cc 68-GTS105 pCi/cc Rsd Waste Equipment Exhaust Offgas Monitors ((Joforo C/is icos/J 16-HVW195-Ch 1 pCi/cc 63-OFG13A pCI(cc Ch 2 yCi/cc 64-OFG13B pCi/cc Rad Waste Tank Exhaust 17-HVW196-Ch 1 pCllcc Trend Symbols: Ch 2 pCi/cc t ~ Increasing I ~ Decreasing a No Chango Rsd Waste Building Ventilation 18-HVW197-Ch I yCI/cc Ch 2 pCI/cc February 199S Page 67 EPMP-EPP-02 Rev 14

ATTACHMENT 34: NINE NILE POINT NUCLEAR STATION NP ANT SVRV Y SAMP STATUS BOARD DATE THIS 0 IS A DRILL N CI IS NOT A DRILL TINE LOCATION DATA FRON RESULTS RENARKS February 1998 Page 68 EPMP-EPP-02 Rev 14

ATTACHMENT 35: NINE MILE POINT NUCLEAR STATION DOWNWIND SURV SANP STATUS BOARD UNIT aI aZ I THIS Cl IS A DRILL 0 IS NOT A DRILL LOCATICNIERPA OATA FROAl REsvLTs REMARKS HABITABILITYSURVEY RESULTS:

CIRCLE EACH Protective Action Recommendations/Implementations AFFECTED ERPA NMPC RECOMMENDED 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 ERPAs FOR EVACUATION 16 17 18 19 20 21 22 23 24 25 26 27 28 29 TIME NMPC RECOMMENDED 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 ERPAs FOR SHELTER 'l6 17.18 19 20 21 22 OSWEGO COUNTY 1 234567891011 12131415 ACTUAL ERPAs 16 17 18 19 20 21 22 23 24 25 26 27 28 29 EVACUATED TIME OSWEGO COUNTY 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 ACTUAL ERPAs 16 17 18 19 20 21 22 SHELTERED February 1998 Page 69 EPMP-EPP-O2 Rev 14

ATTACHMENT 36: NINE NILE POINT NUCLEAR STATION ERGENCY EV NTS STATUS BOARD Date February 1998 Page 70 EPHP-EPP-02 Rev 14

ATTACHMENT 37: NINE MILE POINT NUCLEAR STATION UIPM T SURV SAMP STATUS BOARD THIS Cl IS A DRILL UNIT Cl 1 CI 2 I 0 IS NOT A DRILL TEAMS RETURNED TO CQNDION CQRRECTIVE ACTION NAME/LEADER TEAM STATUS SERVICE 0 ESTIMATED TEAM IO B STAN OCT DATE IKAOOI B SHIBIHO TIME TIME DISPATCH BI 10 0 COMPLETED IIATE 0 ON THE JOB TIMK 0 OTHGI 0 ESTIMATED TEAM B a CTMIDST OATS LEADBI K! SIUKIINII II ME II DISPATCH BI ME IO 0 COMPI.ETED 'ATE 0 CN THK JOB II ME 0 OTHGI 0 ESTIMATED TEAM IO 0 STANDBY OATS II ME TIME DISPATCHED IO 0 COMPLETED DATE 0 ON THE JOB TIME 0 OTHER 0 ESTIMATED TEAM ID 0 STANDBY DATE LlAOER Cl C IVER HO TIME IIM E II S PATCH BI I

IO 0 COMPLETED DATE 0 ON THE JOB TIME 0 OTHGl 0 ESTIMATED TEAM IO 0 STANDBY I!ATE LEAOBI B EIIIBINO II ME IIM E Ol SPA TON Bl IO 0 COMPLETED DATE 0 ON THE JOB TIME 0 OTHGl 0 KSIIMATEO TEAM IO 0 STANDBY I!ATE LEADER 0 BI8EIINO IIM TIME DISPATCHED IO 0 COMPLETED DATE 0 ON THK JOB TIMK 0 OTHGI 0 ESTIMATED TEAM IO 0 STANDBY DATE LKAOBI B SIUERNO TIME TIM E IIISPA TCNID IO 0 COMPLETED DATE 0 ON THE JOB TIME 0 OTHER Q ESTIMATED TEAM IO 0 STANDBY DATE LCAO Bl 0 8RIEPINO IlM C TIME DISPATCHED IO 0 COMPLETKD DATE 0 ON THE JOB IIME 0 OTHGI

~NOT:,"*" INOICATES SAME AS BEFORE February 1998 Page 71 EPMP-EPP-02 Rev 14

ATTACHMENT 38: P ANT STATUS TRfNOING BOARO Oate (MM/DD/YY)

PLANT STATUS BOARD TIME PARAMETERS Reactor Pressure (psip J Reactor Temperature (F J Reactor Level ~ (IN)

Drywell Pressure (psi gJ Drywall Temperature (F J Release Ftate (pCilSecJ Wind from Direction ( 4J Wind Speed (MPHJ Stability.

Class February 1998 Page 72 EPHP-EPP-02 Rev 14

ATTACHMENT 39: NINE HILE POINT NUCLEAR STATION AREA RAO.MONITORS - UNIT I

%ate (MM/DD/YY) Time (24 Hour)

Process Com uter Dis la ed Time No. Results (rrrR/hrJ Trends 1 TB 28'l'E 2 RB 318'ew Fuel Storago Area 3 TB 277'ontrol Room NW 4 TB 277'E-i6iC Rosults Shop 5 TB 300'urbin~enerstor End 8 TB 300'urbine-Feed Pump End 7 TB 261'ondensation Pump Valve Carr.

8 TB 261'eed Pump Area 9 TB 261'witchgear Area 10 TB 257'ondensation Deminerslizer Valve Aroa 1 1 TB 261'egen. Area 12 TB 261'W-MUD Ares 13 Old W.B. 225'rum Fill Op. Aisle 14 Old W.B. 229'ump Room 15 W.B. 261'adwasto Control Room 16 Old W.B. 281'Storage and Shipping Ares 17 RB 249'lP Area 29 RB 340'perator's Platform RFB RB 340'perator's Platform (Refuel Bridgel 18 RB 340'mergency Condensation Shield Wall 19 RB 198'E-RB Equipment Drain Tank Area 20 RB 298'-RB Closo Loop Coal Area 21 RB 261'E-Clean Up Pump Area 22 RB 281'E-Rx Fuel Pool Cooling System Aron 23 RB 237'W-Containment Rod Drive Mod Aroa 24 RB 261 High Lovel Laboratory 25 RB 340'-Spent Fuel Pool Area 26 TB 261'argo Equipment Decontamination Roam 27 TB 318'W-Containment Spent Heat Exhaust Area 28 RB 237'x N. Instrumentation Room 30 WB 261'W-Decontamination Sink Area

'31 WB 247'W-West Well 32 WB 229'W-South Wall 33 OGB 229'est Well 34

• Trend Symbols: t Increasing J Oecreasing No Change February 1998 Page 73 EPHP-EPP-02 Rev 14

ATTACHMENT 40: NINE MILE POINT NUCLEAR STATION REA RAO MONITORS - UNIT 2 Date (MM/DD/YY) Time (24 Hour)

(DRMS Computer Displayed Time) 0-ARM Monitor Results (md'hrJ Trend 19-RMS108 RB 289'outheast CRD Maintenance Area 21-RMS144 RB 281'RD Module Area South 22-RMS106 RB 261'ntrance Aria 23-RMS143 RB 261'RD Module Area North 24-RMS145 RB 240'ample Sink 25-RMS105 RB 240'IP Drive Mechanical Equipment Area 26-RMS2B RB 215'ecirc Pump Instrument Panel B 28-RMS2A RB 215'ecirc Pump Instrument Panel A 29-RMS101 Auxiliary Bay North 175'HS Heat. Exchange Equipment Room 31-RMS104 RB 175'quipment Drains Sumps 8L Pumps West 32-RMS103 Auxiliary Bay South 175'HS Heat Exchange Equipment Room 33-RMS102 RB 175'quipment Drains Sumps 8L Pumps East 42-RMS112 RB 354'uel Handling Platform 43-RMS111 RB 354'uel Handling Platform 59-RMS192 TB 308'as Effluent Monitor Area iViralArea Monitor) 60-RMS191 TB 306'ow-Level Count Room iVitolArse Afoniror/

69-RMS193 Main Stack 261'as Effluent Monitor Ares fVitaiAree Monitor/

71-RMS130 CB 261'emote Shutdown Panel Area

• Trend Symbols: t Increasing l- Decreasing ~ = No Change February 1998 Page 74 EPMP-EPP-02 Rev 14

ATTACHMENT 41: EMERGENCY PROCEDURES TELEPHONE NUMBERS UARTERLY PHONE CHECKS 1.0 PROCEDURE 1.1 For each person/organization listed, verify that the number(s) listed in this Attachment, are correct by contacting that person/organization.

NOTE: For multiple numbers a verbal verification from the person/

organization that other numbers are correct is "SAT",

1.2 Check "SAT" if the number is verified correct.

1.3 If the number is incorrect or no longer working, then perform the following:

a. If it is a number change, draw one line through the old number and write the new number next to it.

b. Verify the new number and check "SAT".

c. Generate an Immediate PCE to any affected EPIPs listed under Procedure Reference.

d. Generate a Future PCE to any affected EPMPs listed under Procedure Reference.

1.4 for all other discrepancies which cannot be resolved, record the discrepancy in the Remarks section and notify the Emergency Preparedness Organization.

1.5 Include a copy of Attachment 83, EPIP-EPP-30, annotated so as to indicate verification of the phone numbers listed.

February 1998 Page 75 EPMP-EPP-02 Rev 14

TTACHMENT 41 (Cont) 0 quarterly: 1 2 3 4 circle one (year) 0 Post Drill/

Exercise/Emergency (date) 0 Other PROCEDURE ER N ORGANIZATI N EAT gNSAT REFERENCE American Nuclear Insurers (860) 561-3433 0 0 EPIP-EPP-20 Ext. 304 Bell Atlantic 479-2161 0 0 EPIP-EPP-17 EPMP-EPP-02 Burtch, Robert Home: 342-2271 0 0 EPIP-EPP-24 Beeper: 876-1124 0 0 Office: 349-7601 0 0 Community Alert Network (CAN) (518) 382-0675 0 0 EPMP-EPP-06 (518) 382-8030 (Emergency) 0 0 EPIP-EPP-20 (518) 382-8042 0 0 (800) 992-2331 0 0 Control Room - Unit 1 349-2480 0 0 EPIP-EPP-20 342-3462 0 0 349-2478 0 0 349-2842 0 0 Control Room Unit 2 349-2170 0 0 EPIP-EPP-20 342-1929 0 0 342-3059 0 0 349-2168 0 0 349-1260 0 0 Control Room Communications Aide - Unit 1 349-2869 0 0 EPIP-EPP-20 Control Room Communications Aide Unit 2 349-2173 Cl 0 EPIP-EPP-20 Control Room Fax Rapid Com ¹'s - U1/U2

~ EOF ¹5 593-5961 CI 0 EPIP-EPP-20

~ TSC ¹01 349-2111 0 0

~ JNC ¹14 592-3850 0 0

~ Oswego County ¹27 698-6678 0 0 DOE (516) 344-2200 CI 0 EPIP-EPP-20 (516) 344-3424 0 0 Emergency Preparedness Vehicle Cellular Phones ~ 6-484 593-4646 0 CI EPIP-EPP-07

~ 5-487 593-4645 0 0

~ 5-243 593-4651 0 0

~ 2-1077 593-9606 CI 0 Environmental Survey Sample Team 593-5991 0 Cl EPIP-EPP-07 Coordinator (ESSTC) 693-5988 0 0 593-5987 CI 0 EOF 593-5740 0 0 EPIP-EPP-14 593-5735 0 0 EOF Communications Coordinator 693-5875 0 0 EPIP-EPP-20 EOF Security Director 593-5890 Cl 0 EPIP-EPP-14 February 1998 Page 76 EPHP-EPP-02 Rev 14

~PR CEDURE R N ANIZATI TELEPH N NO. /AT Q NEAT Qg/EREN~E EOF Technical Liaison Advisory Manager 593-5884 0 0 EPIP-EPP-20 (TLAM) 593-5818 CI 0 Fl'S T ggF

~ ENS 700-371-5324 700-371-0064 Cl 0 EPIP-EPP-17

~ HPN 700-371-5329 700-371-6299 0 0

~ PMCL .700-371-5326 700-371<062 Q 0

~ RSCL 700-371-5327 700-371-0063 0 0

~ MCL 700-371-5323 700-371-0060 0 0

~ LAN 700-371-5328 700-371-0061 0 0 General Electric (408) 971-1038 0 0 EPIP-EPP-20 JAFNPP Control Room 349-6665 0 CI EPIP-EPP-20 349-6666 0 CI 342-3840 0 0 349-6323 Fax 0 0 Merritt, Carey Home: 298-7490 0 0 EPIP-EPP-24 (Environmental) Beeper: 876-3169 0 0 Office: 349-4200 0 0 National Weather Service (800) 462-7751 0 0 EPIP-EPP-08

{716) 565-9001 0 0 New York State Warning Point (618) 457-2200 0 0 EPIP-EPP-17 (518) 457-6811 CI 0 EPIP-EPP-20 (518) 457-9942 Fax 0 0 (518) 457-8926 Fax CI 0 (518) 457-9997 0 0 New York Page (800) 753-2337 0 0 EPIP-EPP-17 NY State Emergency Mgmt. Office (518) 457-2200 0 0 EPIP-EPP-24 (518) 457-6811 0 0 (618) 467-9942 Fax Cl 0 Central Regional Communications Group ,- (315) 460-2378 EPMP-EPP-02 (Radio Shop) (315) 460-2379 0 0 EPIP-EPP-17 NLC Receptionist 349-2080 0 0 EPIP-EPP-23 NRC Emergency Operations Center (301) 816-5100 Main 0 CI E PIP-EPP-20 (301) 951-0550 Backup 0 0 EPIP-EPP-24

{301) 415-0650 Backup CI 0 EPMP-EPP-02 (301) 816-5151 Fax 0 0 EPIP-EPP-17 NRC (FTS Problems) (301) 816-6100 0 CI EPIP-EPP-17 (301) 951-0550 0 CI NRC Resident Office 349-2529 CI CI E PIP-EPP-24 342-4041 0 CI Beeper:(888) 364-4960 CI 0 Nuclear Security 349-2401 EPIP-EPP-24 February 1998 Page 77 EPHP-EPP-02 Rev 14

ATTACHMENT 41 (Cont)

PRRO EDURE PER ON OR ANIZATI N TELEPHONE N /AT gNSAT REFERENCE O'rien, David Home: 343-2484 0 CI EPIP-EPP-15 (DoctorJ Office: 343-4348 Cl 0 Oswego County Emergency Mgmt. Office 598-1191 CI 0 EPIP-EPP-24 598-1192 0 0 598-6678 Fax CI 0 Oswego County Sheriff 911 0 0 EPIP-EPP-20 343-5490 0 0 349-3409 0 0-349-3410 0 0 349-3411 CI 0 Oswego County 911 Center 911 0 0 EPIP-EPP-04 (Oswego County Warning Point) 343-1313 Cl 0 EPIP-EPP-03 349-8502 0 0 EPIP-EPP-24 349-8500 Fax 0 0 EPIP-EPP-30 EPIP-EPP-20 E PIP-EPP-28 Oswego Hospital 349-5522 0 0 EPIP-EPP-04 Page Activation Number 876-XXXX 0 CI EPIP-EPP-17 1-800-732-4365 0 CI Pager Coordinator 428-6700 OR (821)6700 0 0 EPIP-EPP-17 Personnel Accountability Coordinator 2662 0 0 EP IP-EPP-1 8 EPIP-EPP-05 Radiation Mgmt. Consultants (215) 243-2990 0 0 EPIP-EPP-15 0

Radiological Assessment Manager (RAM)

RP Team Coordinator (RPTC)

(216) 824-1300 349-1353 343-6408 349-1272 CI Cl 0

CI 0

0 CI EPIP-EPP-06 EPIP-EPP-07 EPIP-EPP-06

~

EPIP-EPP-07 REGS Line Trouble (518) 457-2200 0 0 EPMP-EPP-02 System Hydro Supervisor (315) 785-5203 0 0 EPIP-EPP-20 (316) 785-5206 0 0 (315) 785-7177 0 0 (315) 785-7186 0 0 (315) 785-7184 Fax 0 0 Taylor, Arthur (Skip) Home: 342-5337 0 0 EPIP-EPP-24 Office: 349-4982 0 0 Torbitt, Jack Home: 593-2713 CI 0 EPIP-EPP-24 Beeper: 876-1282 CI CI Office: 349-2543 0 0 iren Problem Grou Farrell, Kevin Office: 460-2378/9 0 0 EPIP-EPP-30 Home: 484-3337 0 0 Beeper: 876-3147 CI 0

'Rebeor, Anthony Office: 592-0166 EPIP-EPP-30 EPNP-EPP-02

-0 February 1998 Page 78 Rev 14

ATTACHMENT 41 (Cont)

PRIRPCED RE PER N ORGANI2ATION TELEPH NE NO. SAT ~UN AT RE~~ENIEE Attachment ¹3 N/A 0 0 EPIP-EPP-30 EPIP-EPP-30 NOTE: lt is acceptable to fax or ask verbally the individuals to verify the phone numbers of the people in their respective group as listed in Attachment 3 of EPIP-EPP-30.

Remarks:

Performed by Date Supervisor Approval Date E.P. Review Date February 1998 Page 79 EPHP-EPP-02 Rev 14

ATTACHMENT 42: EMERGENCY K Y INVENTORY UARTERLY

. CI quarterly: 1 circle 2 3 4 one r (year)

CI Post Drill/

Exercise/Emergency (date)

CI Other

.,'<.;~y~.:,'i~'<sg '~&<'<~,,>Ra<'gg'<~5<; <$~,GM;is'<~~@< ~- .1'gag'6$ $ <@+:,q@'S .8.'.;@'> Pjg~'2D25<,,<<;< ~>dj'j'",'.";.'.'";.'.",',:,:::iNC,;..':.,';;

Yehj;c I:;e's'.:;;,':-..'(Mastei:,)".'SC CI SAT CI UNSAT x' OSC CI SAT'I UNSAT JNC CI SAT CI UNSAT X OAA CI SAT CI UNSAT

'ontained in "break away" box outside facility.

Contained in key box inside Utility Room.

Remarks:

Performed by Date Supervisor Approval Date E.P. Review Date February 1998 Page 80 EPMP-EPP-02

'ev 14

NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT NUCLEAR STATION EMERGENCY PLAN MAINTENANCE PROCEDURE EPMP-EPP-06 EVISION 03 EMERGENCY RESPONSE ORGANIZATION NOTI ICATION MAINTENANCE AND SURVEILLANCE Approved By:

J. D. Jones D'toV E ency Preparedness D te Effective Date: 9/03/96

4~

~ ~

t r

LIST OF EFF CTIVE PAGES

~PN . ~NN ~PN . gC ~PN. ~CN N.

Coversheet .

~ ~

11 2 0 ~ ~ ~

3 ~ ~

5 .

6 .

August 1996 Page i EPHP-EPP-06 Rev 03

~ I TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE . ~ ~ ~ ~ ~ 1 2;0 PRIMARY RESPONSIBILITY 3.0 PROCEDURE .

3. 1 Emergency Preparedness Actions . . . . . . . . . .

3.2 ERO Initial Responder (with secondary responder responsibilities) Actions ~ ~ 2

'.0 3.3 ERO Member Notification Test Actions . . . . . . . ~ ~ 3 DEFINITIONS . 3

5.0 REFERENCES

AND COMMITMENTS ~ ~ 3 6.0 RECORD REVIEW AND DISPOSITION . ~ ~ 3 ATTACHMENT 1: COMMUNITY ALERT NETWORK (CAN) SYSTEM DESCRIPTION..... 4 ATTACHMENT 2: CAN DATABASE CHANGE FORM (EXAMPLE) 5 ATTACHMENT 3: NOTIFICATION DRILL RESPONSE FORM . 6 August 1996 Page ii EPMP-EPP-06 Rev 03

P 1.0 PURPOSE To provide guidance on the maintenance and surveillance of the methods used to notify the Emergency Response Organization (ERO) of drills, exercises and emergencies.

2.0 PRINARY RESPONSIBILITY 2.1 Director - Emer enc Pre aredness

~ Assigns the performance of maintenance and surveillance of the ERO notification systems.

~ Oversees the maintenance of secondary responder notification method maintenance.

2.2 ERO Initia1 Res onders with secondar res onder res onsibilities Assigns the performance of maintenance and surveillance of their notification systems.

3.0 P OCEDURE 3.1 Emer enc Pre atedness Actions 3.1.1 a er Surveillance a ~ Should be conducted the first Friday of each month.

b. The test will consist of activation of ERO initial responder pagers by sending a "000999" code via telephone activation.

C. A test is considered successful if a single ERO initial responder pager receives and displays the "000999" message.

d. The failure of any single pager to meet the success criteria should be resolved between the pager owner and the NMPC pager coordinator.

e. The failure of the pager system to meet the success criteria shall result in immediate corrective actions by EP.

3.1.2 Tele hone Notification S stem Maintenance NOTE: Automated telephone notification for the ERO is provided by Community Alert Network (CAN).

a ~ The CAN System configuration should be maintained in accordance with Attachment l.

August 1996 Page 1 EPMP-EPP-06 Rev 03

P 3.1.2 (Cont)

b. Review the CAN List for initial responder s quarterly.

Nake any changes needed to the CAN List so that it accurately reflects the current duty roster. Utilize Attachment 2 or equivalent form, for making changes.

c. Any other group rosters on CAN should be sent to the responsible owners quarterly, for review and modification.

3.1.3 Tele hone Noti ication S stem Surve llance and Testin a 0 The CAN System will be tested quarterly as follows:

1. Contact CAN in accordance with EPIP-EPP-20.

2. Request activation of the system and provide an appropriate emergency message.

CAUTION Selecting "Alert or higher" will result in the CAN message instructing ERO members to respond to emergency duty locations.

3. Successful activation is indicated by:

~ Activation of any ERO initial responder pager with the appropriate code.

~ Activation of the proper CAN telephone list based on the printed report from CAN.

b. Failure of any test criteria shall result in immediate corrective actions by EP.

3.2 ERO Initial Res onder with secondar res onder res onsibi1ities Actions 3.2.1 IF a CAN group roster exists, EP will send the roster to Team 1 Initial Responder on a quarterly basis. THEN:

a. The Team 1 Initial Responder should review the roster for accuracy and if needed make changes using Attachment 2, or equivalent form.

b. Attachment 2 should be sent to CAN using the fax number on the attachment.

August 1996 Page 2 EPMP-EPP-06 Rev 03

J 3.2.2 IF no CAN group roster exists, THEN the Team 1 Initial Responder shall maintain and test their method for notifying secondary responders. This can include phone "trees" or pagers.

3.3 ERO Nember Notification Test Actions 3.3.1 Respond to any notification drills by completing Attachment 3 and sending it to EP.

NOTE: Pager tests are not considered notification drills.

3.3.2 Report any pager problems or failures to the NHPC pager coordinator .

3.3.3 Report any changes in home telephone numbers to Emergency Preparedness.

4.0 DEFINITIONS 4.1 Communit Alert Network CAN - A vendor that provides an automated telephone service that activates the NHPC pager system and contacts designated persons with pre-recorded emergency messages.

4.2 Notification Drill An evolution that tests the integrated capability of the ERO notification system, typically consisting of a pager and telephone notification.

5.0 REFERENCES

AND COHHITHENTS None 6.0 RECORD REVIEM AND DISPOSITION The following records generated by this procedure shall be maintained by Nuclear Records Hanagement for the Permanent Plant File in accordance with NIP-RHG-Ol:

None The following records generated by this procedure are not required for retention in the Permanent Plant File:

~ Attachment 2, CAN Database Change Form

~ Attachment 3, Notification Drill Response Form August 1996 Page 3 EPHP-EPP-06 Rev 03

ATTACHNENT 1 CONNUNITY ALERT NETWORK CAN S STEN DESCRIPTION 1.0 CAN is an automated telephone notification system that dials pre-defined telephone numbers when requested by NHPC. The CAN System will dispense a message to each person called, indicating plant status and any requested response.

2.0 The CAN database is divided into four lists, as follows:

List ¹ When called Who is called 1 Unusual event, EP Staff, NRC Resident pager, ERO Initial normal hours Responder pagers 2 Unusual event, EP Staff, NRC Resident pager, ERO Initial off-hours Responder pagers 3 Alert or higher, EP Staff, NRC Resident pager, ERO Initial normal hours Responder pagers 4 Alert or higher, ~ All initial responders (home phone) off-hours ~ ERO initial responder pagers

~ Some secondary responders

~ EP Staff, NRC Resident pager 3.0 EPIP-EPP-20 contains details on the activation of this system.

August 1996 Page 4 EPHP-EPP-06 Rev 03

t E

ATTACHMENT 2 CAN DATABAS CHANGE FON EXANPLE Instructions:

Fill in your group name in the space provided. (Valid group names are listed on the back)

2. Fill in your name and phone number in the "Completed By" space

3. Complete the change table for any, changes needed

4. Fax this page to Can at 518-382-0675 List ¹4 Group Name:

Completed By: Phone: (315)

(A)dd Name Area Telephone (0) elete Code (C)hange August 1996 Page 5 EPHP-EPP-06 Rev 03

V

~ ~ l'

ATTACHMENT 3 NOTIFICATION DRILL RESPONSE FORH Results Summar  :

'Name: Emergency Position:

Team ¹: Date Received:

Pager Activation:

0 Yes (Time Message ): 0 No Telephone Notification:

0 None 0 Drill 0 Unit 1 0 No response required 0 Not a Drill 0 Unit 2 0 Respond-normal location 0 Pager Test 0 Both Units 0 Respond-alternate location 0 Pager Test 0 Pager Test Appropriate number of Secondary Responders indicated they are available to respond:

0 Yes 0 No How long will it take you to get to your emergency response facility (in minutes)7 Comments:

Please return to Emergency Preparedness, NLC August 1996 Page 6 EPHP-EPP-06 Rev 03

I V i

~ ~

I l

NIAGARA MOHAMK POMER CORPORATION NINE MILE POINT NUCLEAR STATION EMERGENCY PLAN MAINTENANCE PROCEDURE PHP-EPP-0101 REVISION 01 UNIT 1 EMERGENCY CLASSIFICATION TECHNICAL BASES TECHNICAL SPECIFICATION REQUIRED Approved by:

N. L. Rademacher

~ ~ Plant Manager Da e Effective Date:

LIST OF EFFECTIVE PAGES

~PN . ~CN II ~PN. ~CN N. ~PN . ~NI N Coversheet . 22 ~ ~ ~ ~ 47 .

23 48 .

ll

~ ~

24 ~ ~ ~ ~ 49 .

25 ~ ~ ~ ~ 50 .

26 . 51 2 ~ ~ 27 ~ 0 I ~ 52 .

3 . 28 ~ ~ ~ ~ 53 4 29 ~ ~ ~ ~ 54 5 . 30 . 55 .

6 . 31 56 .

7 . 32 ~ ~ ~ ~ 57 .

.4; i>

8 . 33.... g 58 .

9 . 59 .

10 . 60 .

ll 36 . 61 12 . 37 . 62 .

13 . 38 . 63 14 . 39 . 64 .

15 . 40 . 65 .

16 . 41 . 66 17 . 42 . 67 18 ~ ~ ~ ~ 43 . 68 .

19 ~ ~ ~ ~ 44 . 69 20 . 45 . 70 .

21 46 . 71 .

November 1996 Page i EPHP-EPP-0101 Rev Ol

fc LIST OF EFFECTIVE PAGES (Cont)

~PN . ~Ch N ~PN. ~CN N. ~PN . ~CN N 72 . 97 .

73 ~ ~ ~ ~ 98 .

74 . 99 .

75 . 100 .

76 . 101 77 . 102 .

78 . 103 .

79 . 104 .

80 ~ ~ ~ ~ 105 .

81 106 .

82 . 107 .

83 108 .

84 ~ ~ ~ ~ 109 .

85 110 .

86 ~ ~ 0 ~ ill 87 . 112 .

88 .

89 ~ ~ ~ ~

90 ~ ~ ~ ~

91 92 .

93 94 .

95 96 .

November 1996 Page ii EPHP-EPP-0101 Rev 01

TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

2.0 PRIMARY RESPONSIBILITY 3.0 PROCEDURE ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3.1 Emergency Preparedness Group . .

4.0 DEFINITIONS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

5.0 REFERENCES

AND COMMITMENTS 2 6.0 RECORD REVIEW AND DISPOSITION . 2 ATTACHMENT 1: UNIT 1 EMERGENCY ACTION LEVEL TECHNICAL BASES 3.

2: FISSION PRODUCT BARRIER LOSS/POTENTIAL LOSS INDICATORS .. 102 ATTACHMENT ATTACHMENT 3:- WORD LIST/DEFINITIONS 105 November 1996 Page iii EPMP-EPP-0101 Rev Ol

1.0 PURPOSE To describe the technical bases for the emergency action levels at Unit 1.

2.0 PRINARY RESPONSIBILITY 2.1 mer enc Pre aredness Grou Monitor/solicit any changes to the technical bases of each emergency action level.

Assess these changes for potential impact on the emergency action level; Maintain the emergency action level technical bases, EPIP-EPP-Ol, and the Emergency Action Level Matrix/Unit l.

3.0 PROCEDURE 3.1 Emer enc Pre aredness Grou 3.1.1 Maintain a matrix of technical bases references for each emergency action level.

3.1.2 Evaluate each technical bases reference change for impact on the affected emergency action level.

3.1.3 Modify EPIP-EPP-Ol, Emergency Action Level (EAL) Matrix/Unit 1 and Attachment 1 of this procedure, as needed.

4.0 DEFINITIONS See Attachment 3.

November 1996 Page 1 EPMP-EPP-0101 Rev Ol

5.0 REFERENCES

AND COMMITMENTS 5.1 Licensee Documentation None 5.2 Standards Re ulations and Codes NUMARC NESP-007, Methodology for Development of Emergency Action Levels.

5.3 Policies Pro rams and Procedures EPIP-EPP-Ol, Classification of Emergency Conditions at Unit.

5.4 Su lemental References Nine Mile Point Unit 1, Plant-Specific EAL Guideline 5.5 Commitments None 6.0 RECORD REVIEW AND DISPOSITION None November 1996 Page 2 EPMP-EPP-0101 Rev 01

ATTACHMENT 1 UNIT 1 EMERGENCY ACTION LEVEL TECHNICAL BASES PURPOSE The purpose of this document is to provide an explanation and rationale for each of the emergency action levels (EALs) included in the EAL Upgrade Program for Nine Mile Point 1 (NHP-1). It is also intended to facilitate the review process of the NMP-1 EALs and provide historical documentation for future reference. This document is also intended to be utilized by those individuals responsible for implementation of EPIP-EPP-01 "Classification of Emergency Conditions Unit 1" as a technical reference and aid in EAL interpretation.

DISCUSSION EALs are the plant-specific indications, conditions or instrument readings which are utilized to classify emergency conditions defined in the NMP-1 Emergency Plan.

The revised EALs were derived from the Initiating Conditions and example EALs given in the NHP-1 Plant-Specific EAL Guideline (PEG). The PEG is the NMP-1 plant interpretation of the NUHARC methodology for developing EALs.

Many of the EALs derived from the NUMARC methodology are fission product barrier based. That is, the conditions which define the EALs are based upon loss or potential loss of one or more of the three fission product barriers.

The primary fission product barriers are:

A. Reactor Fuel Claddin FC : The fuel cladding is comprised of the zirconium tubes which house the ceramic uranium oxide pellets along with the end plugs which are welded into each end of the fuel rods.

B. Reactor Coolant S stem RCS : The RCS is comprised of the reactor vessel shell, vessel head, CRD housings, vessel nozzles and penetrations and all primary systems directly connected to the RPV up to the outermost primary containment isolation valve.

Primar Contai ment PC: The primary containment is comprised of the drywell, suppression chamber (torus), the interconnections between the two, and all isolation valves required to maintain primary containment integrity under accident conditions.

Although the secondary containment (reactor building) serves as an effective fission product barrier by minimizing ground level releases, considered as a fission product barrier for the purpose of emergency it is not classification.

The following criteria serves as the bases for event classification related to fission product barrier loss:

November 1996 Page 3 EPMP-EPP-0101 Rev 01

ATTACHMENT I (Cont)

Unusual vent:

Any loss or potential loss of containment Alert.

Any loss or any potential loss of either fuel clad or RCS Site Area Emer enc  :

Any loss of both fuel clad and RCS or Any potential loss of both fuel clad and RCS or Any potential loss of either fuel clad or RCS with a loss of any additional barrier General Emer enc :

Loss of any two barriers with loss or potential loss of a third Those EALs which reference one or more of the fission product barrier Initiating Condition designators (FC, RCS and PC) in the PEG Reference section of the technical bases are derived from the Fission Product Barrier Analysis.

The analysis entailed an evaluation of every combination of the plant specific barrier loss/potential loss indicators applied to the above criteria.

Where possible, the EALs have been made consistent with and utilize the conditions defined in the NMP-I symptom based Emergency Operating Procedures (EOPs). While the symptoms that drive operator actions specified in the EOPs are not indicative of all possible conditions which warrant emergency classification, they do define the symptoms, independent of initiating events, for which reactor plant safety and/or fission product barrier integrity are threatened. Where these symptoms are clearly representative of one of the PEG Initiating Conditions, they have been utilized as an EAL. This allows for rapid classification of emergency situations based on plant conditions without the need for additional evaluation or event diagnosis. Although some of the EALs presented here are based on conditions defined in the EOPs, classification of emergencies using these EALs is not dependent upon EOP entry or execution. The EALs can be utilized independently or in conjunction with the EOPs.

To the extent possible, the EALs are .symptom based. That is, the action level is defined by values of key plant operating parameters which identify emergency or potential emergency conditions. This approach is appropriate because it allows the full scope of variations in the types of events to be classified as emergencies. But, a purely symptom based approach is not sufficient to address all events for which emergency classification is appropriate. Particular events to which no predetermined symptoms can be ascribed have also been utilized as EALs since they may be indicative of potentially more serious conditions not yet fully realized.

November 1996 Page 4 EPMP-EPP-0101 Rev Ol

~TTACIIII NT (C niSCUSSrON (Cont)

The EALs are grouped into nine categories to simplify their presentation and to promote a rapid understanding by their users. These categories are:

1. Reactor Fuel

2. Reactor Pressure Vessel

3. Primary Containment

4. Secondary Containment

5. Radioactivity Release

6. Electrical Failures

7. Equipment Failures

8. Hazards

9. Other Categories 1 through 5 are primarily symptom based. The symptoms are indicative of actual or potential degradation of either fission product barriers or personnel safety.

Categories 6, 7 and 8 are event based. Electrical Failures are those events associated with losses of either AC or vital DC electrical power. Equipment Failures are abnormal and emergency events associated with vital plant system failures, while Hazards are those non-plant system related events which have affected or may affect plant safety.

Category 9 provides the Emergency Director (Shift Supervisor) the latitude to classify and declare emergencies based on plant symptoms or events which in his judgment warrant classification. This judgment includes evaluation of loss or potential of one or more fission product barriers warranting emergency classification consistent with the NUMARC barrier loss criteria.

Categories are further divided into one or more subcategories depending on the types and number of plant conditions that dictate emergency classifications.

For example, the Reactor Fuel category has five subcategories whose values can be indicative of fuel damage: coolant activity, off-gas activity, containment radiation, other radiation monitors and refueling accidents. An EAL may or may not exist for each sub category at all four classification levels.

Similarly, more than one EAL may exist for a sub category in a given emergency classification when appropriate (i. e., no EAL at the General Emergency level but three EALs at the Unusual Event level).

November 1996 Page 5 EPMP-EPP-0101 Rev Ol

jjjjjjHM M I (j t)

DISCUSSION (Cont)

For each EAL, the following information is provided:

Classification: Unusual Event, Alert, Site Area Emergency, or General Emergency Operating Node Applicability: One or more of the following plant operating conditions are listed: Power Operation, Startup/Hot Standby, Hot Shutdown, Cold Shutdown, Refuel and Defueled EAL: Description of the condition or set of conditions which comprise the EAL Basis: Description of the rationale for the EAL PEG Reference(s): PEG IC(s) and example EAL(s) from which the EAL is derived Basis Reference(s): Source documentation from which the EAL is derived The identified operating modes are defined as follows:

Power 0 erations Reactor is critical and the mode switch is in RUN.

Startu Hot Standb This mode is subsumed in the Power Operations mode.

Hot Shutdown Mode switch is in SHUTDOWN or REFUEL and reactor coolant temperature is >212

'F.

Cold Shutdown Mode switch in SHUTDOMN or REFUEL and reactor coolant temperature is g212 F.

Refuel Node switch in REFUEL and reactor coolant temperature 6212 F.

Defueled RPV contains no irradiated fuel.

November 1996 Page 6 EPHP-EPP-0101 Rev 01

NTTACIIMENT I (C tt 1.0 REACTOR FUEL The reactor fuel cladding serves a's the primary fission product barrier. Over the useful life of a fuel bundle, the integrity of this barrier should remain intact as long as fuel cladding integrity limits are not exceeded.

Should fuel damage occur (breach of the fuel cladding integrity) radioactive fission products are released to the reactor coolant. The magnitude of such a release is dependent upon the extent of the damage as well as the mechanism by which the damage occurred. Once released into the reactor coolant, the highly radioactive fission products can pose significant radiological hazards inplant from reactor coolant process streams. If other fission product barriers were to fail, these radioactive fission products can pose significant offsite radiological consequences.

The following parameters/indicators are indicative of possible fuel failures:

Coolant Activit : During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from either the

'fission of tramp uranium in the fuel cladding or minor perforations in the cladding itself. Any significant increase from these base-line levels is indicative of fuel failures.

Off- as Activit : As with coolant activity, any fuel failures will release fission products to the reactor coolant. Those products which are gaseous or volatile in nature will be carried over with the steam and eventually be detected by the air ejector off-gas radiation monitors.

Containment Radiation Monitors: Although not a direct indication or measurement of fuel damage, exceeding predetermined limits on containment high range radiation monitors under LOCA conditions is indicative of possible fuel failures. In addition, this indicator is utilized as an indicator of RCS loss and potential containment loss.

Other Radiation Monitors: Other process.and area radiation monitoring systems are specifically designed to provide indication of possible fuel damage such as Area Radiation Monitoring Systems.

Refue 'n Accidents: Both area and process radiation monitoring systems designed to detect fission products during refueling conditions as well as visual observation can be utilized to indicate loss or potential loss of spent fuel cladding integrity.

November 1996 Page 7 EPMP-EPP-0101 Rev Ol

sjt H EN 1.0 EAC 0 U Coola t Activit e'.1.1 Unusual Event Wl Coolant activity > 25 pCi/gm I-131 equivalent NUNARC IC:

Fuel clad degradation FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This EAL addresses reactor coolant samples exceeding coolant technical specifications for iodine spiking.

PE6 Reference(s):

SU4.2 Bases Reference(s):

1. Radiological Technical Specifications, Appendix A to Facility Operating License No. DPR-63, Article 3.2.4.a

1. 1. 2 ~lert Coolant activity > 300 pCi/gm 1-131 equivalent NUNARC IC:

N/A FPB Loss/Potential Loss:

Fuel Clad Loss Node Applicability:

Power Operation, Hot Shutdown November 1996 Page 8 EPMP-EPP-0101

~ l he

ATTACHMENT 1 tC t) 1.1.2

~ ~ (Cont)

Basis:

Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This amount of coolant activity is well above that expected for iodine spikes and corresponds to about 2X to 5X fuel clad damage. When reactor coolant activity reaches this level, significant clad heating has occurred and thus the fuel clad barrier is considered lost. Therefore, declaration of an Alert is warranted.

PEG Reference(s):

FC1.1 Basis Reference(s):

1. General Electric NED0-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions 1.2 0 f- as Activit 1.2.1 Unusual Event Valid offgas radiation > hi-hi alarm NUNARC IC:

Fuel clad degradation FPB Loss/Potential Loss:

N/A Node Applicability:

Power operation, Hot shutdown Basis:

Elevated offgas radiation activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This offgas radiation level corresponds to the Technical Specification allowable limit of 500,000 pCi/sec (recombiner discharge gross noble gases beta and/or gamma). The hi-hi alarm setpoint has been conservatively selected because it is operationally significant and is readily recognizable by Control Room operating staff. The system isolates when both RN-12A and 12B alarm.

November 1996 Page 9 EPHP-EPP-0101 Rev 01

~TTA N T 1.2.1 (Cont)

The hi-hi offgas radiation alarm is nominally set in accordance with the Offsite Dose Calculation Manual.

PEG Reference(s):

SU4.1 Basis Reference(s):

1. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Amendment 66, Article 3.6. 15.c

2. Nl-ARP-H1, annunciator Hl-2-7 1.2.2 Alert Valid offgas radiation > 10 x hi-hi alarm NUNARC- IC:

N/A FPB Loss/Potential Loss:

Fuel Clad Loss Node Applicability:

Power Operation, Hot Shutdown Basis:

This EAL is to cover other indications that may indicate loss or otential loss of the fuel clad barrier. Air ejector offgas radiation evels >10 times the nominal hi-hi setpoint is indicative of significant fuel cladding failure and is consistent with the Alert EAL of 300 pCi/gm 1-131'equivalent coolant activity. The hi-hi offgas radiation level corresponds to the Technical Specification allowable limit of 500,000 pCi/sec (recombiner discharge gross noble qases beta and/or gamma). The hi-hi alarm setpoint has been conservatively selected because it is operationally significant and is readily recognized by Control Room operating staff.

The hi-hi offgas radiation alarm is nominally set at 1500 mRem/hr on RN-12A/B. 10 times the hi-hi alarm setpoint is therefore 15,000 mRem/hr.

PEG Reference (s):

FC4.1 Basis Reference (s):

1. Nl-ARP-H1, annunciator Hl-2-7 November 1996 Page 10 EPHP-EPP-0101 Rev Ol

ATTACHMENT 1 (Cont) 1.3 Containment Radiation 1.3.1 Al ert Drywell radiation > 20 R/hr NUNRC:

N/A FPB Loss/Potential Loss:

RCS Loss Node Applicability:

Power Operation, Hot Shutdown Basis:

The drywell radiation reading is a value which indicates the release of reactor coolant to the drywell. The reading is calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operatinq concentrations (i. e., within Technical Specifications) snto the drywell atmosphere. The reading is less than that specified for EAL 1.3.2 because no damage to the fuel clad is assumed. Only leakage from the RCS is assumed in this EAL.

The calculation referenced resulted in an EAL value of 24 R/hr.

However, a Value of 20 R/h was selected as it is observable on existing instrumentation.

It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. Drywell High Range Radiation Monitors have a range of 0 to ES R/hr on recorder RR 201.7-36C pen 1 and 2. They are installed in the following drywell locations:

RAm 201.7-36 Az 340', El 263'"

RAm 201.7-37 Az 310', EL 301'"

PEG Reference(s):

RCS3.1 Basis Reference(s):

1. Nl-RG197-EIL1, Important Design Features of Regulatory Guide 1.97 Instruments

2. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Amendment 72, 76, Table 3.6.11-1

3. Calculation 1H21C003, Rev. 0 November 1996 Page 11 EPMP-EPP-0101

ATTACHMENT 1 (Cont) 1.3.2 S te Area e enc Drywell radiation > 3000 R/hr NUNRC IC:

N/A FPB Loss/Potential Loss:

Fuel Clad Loss, RCS Loss Node Applicability:

Power Operation, Hot Shutdown Basis:

The drywell radiation reading is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell. The reading is calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 pCi/gm dose equivalent I-131 into the drywell atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations allowed within Technical Specifications (including iodine spiking) and are therefore indicative of fuel damage (approximately 2X 5X clad failure depending on core inventory and RCS volume). The reading is higher than that specified for EAL 1.3.1 and, thus, this EAL indicates a loss of both the fuel clad barrier and the RCS barrier.

The calculation referenced resulted in an EAL value of 3090 R/hr.

However, a value of 3000 R/hr was selected existing instrumentation.

as it is observable on It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. Drywell High Range Radiation Monitors have a range of 0 to ES R/hr on recorder RR 201.7-36C pen 1 and 2. They are installed in the following drywell locations:

RAm 201.7-36 Az 340', El 263'"

RAm 201.7-37 Az 310 , EL 301'"

PEG Reference(s):

FC3.1 November 1996 Page 12 EPHP-EPP-0101 Rev 01

~IT N 1.3.2 (Cont)

Basis Reference(s):

1. Nl-RG197-EILl, Important Design Features of Regulatory Guide 1.97 Instruments

2. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Amendment 72, 76, Table 3.6.11-1

3. Calculation 1H21C003, Rev. 0 1.3.3 General Emer enc Drywell radiation > 4.0E6 R/hr NUHARC IC:

N/A FPB Loss/Potential Loss:

Fuel Clad Loss, RCS Loss, Containment Potential Loss Node Applicability:

Power Operation, Hot Shutdown Basis:

The drywell radiation reading is a value which indicates significant fuel damage well in excess of that required for loss of the RCS barrier and the fuel clad bar ier. NUREG-1228 "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents" states that such readings do not exist when the amount of clad damage is less than 20X. A major release of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure into the reactor coolant has occurred. Regardless of whether the primary containment barrier itself is challenged, this amount of activity in containment could have severe consequences if released.

It is, therefore, prudent to treat this as a potential loss of the containment barrier and upgrade the emergency classification to a General Emergency.

The calculation referenced resulted in an EAL value of 3.9E6 R/hr.

However, a value of 4.0E6 R/hr was selected as it is observable on existing instrumentation.

November 1996 Page 13 EPHP-EPP-0101 Rev 01

1.3.3 (Cont)

It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. Drywell High Range Radiation Nonitors have a range of 0 to E8 R/hr on recorder RR 201.7-36C pen 1 and 2. They are installed in the following drywell locations:

RAm 201.7-36 Az 340 , El 263'"

RAm 201.7-37 Az 310, EL 301'"

PEG Reference(s):

PC3.1 Basis Reference(s):

1. Nl-RG197-EIL1, Important Design Features of Regulatory Guide 1.97 Instruments

2. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Amendment 72, 76, Table 3.6. 11-1

3. Calculation 1H21C003, Rev. 0 1.4 Other Rad at o Nonitors 1.4.1 Unusual Event Any sust'ained ARN reading > 100 x alarm (OP-50A) or offscale hi resulting from an uncontrolled process NUNARC IC:

Unexpected increase in plant radiation or airborne concentration.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

Valid elevated area radiation levels usually have long lead times relative to the potential for radiological release beyond the site boundary, thus impact to public health and safety is very low.

November 1996 Page 14 EPMP-EPP-0101 Rev Ol

S 1.4.1 (Cont)

This EAL addresses unplanned increases in radiation levels inside the plant. These radiation levels represent a degradation in the control of radioactive material and a potential degradation in the level of safety of the plant. Area radiation levels above 100 times the alarm setpoint have been selected because they are readily .identifiable on ARH instrumentation. The ARM alarm setpoint is considered to be a bounding value above the maximum normal radiation level in an area.

Since ARH setpoints are nominally set one decade over normal levels, 100 times the alarm setpoint provides an appropriate threshold for emergency classification. For those ARMS whose upper range limits are less than 100 times the alarm setpoint, a value of offscale high is used. This EAL escalates to an Alert, if the increases impair the level of safe plant operation.

PEG Reference(s):

AU2.4 Basis Reference(s):

1. Nl-EOP-5/6, Secondary Containment Control / Radioactivity Release Control

2. OP-50A, Area Radiation Monitoring System, Attachments 2 and 3 1.4.2 Alert Sustained RB Vent Monitor RN07A5 or B5 > 5 mR/hr OR Any sustained refuel floor rad monitor > S.O R/hr or offscale hi, Table l.

Table 1 Refuel Floor Rad Monitors West End of Shield Wall, RB 340 (¹18)

Rx Bldg. - East Wall El 340'¹25)

Refuel Bridge (high range) (Process Mon.)

Refuel Bridge (low range) (¹29)

NUNARC IC:

Major damage to irradiated fuel or loss of water level that has or will result in the uncovering of irradiated fuel outside the reactor vessel.

FPB Loss/Potential Loss:

N/A Node Applicability:

All November 1996 Page 15 EPHP-EPP-0101 Rev 01

TTACH (Cont) 1.4.2 (Cont)

Basis:

This EAL is defined 'by the specific areas where irradiated fuel is located such as reactor cavity, reactor vessel, or spent fuel pool.

Sufficient time exists to take corrective actions for these conditions and there is little potential for substantial fuel damage. NUREG/CR-4982 "Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82" indicates that even if corrective actions are not taken, no prompt fatalities are predicted and the risk of injury is low. In addition, NRC Information Notice No. 90-08, "KR-85 Hazards from Decayed Fuel" presents the following in its discussion:

"In the event of a serious accident involving decayed spent fuel, protective actions would be needed for personnel on site, while offsite doses (assuming an exclusion area radius of one mile from the lant site) would be well below the Environmental Protection Agency's rotective Action Guides. Accordingly, it is important to be able to properly survey and monitor for Kr-85 sn the event of an accident spent fuel." with'ecayed Thus an Alert Classification for this event is appropriate.

Esca/ation, judgment in if appropriate, would occur via Emergency Director EAL Category 9.0.

The basis for the reactor building ventilation monitor setpoint (5 mR/hr) is a spent fuel handling accident and is, therefore, appropriate for this EAL.

Area radiation levels on the refuel floor at or above the Maximum Safe Operating value (8.0 R/hr) are indicative of radiation fields which may limit personnel access. Access to the refuel floor is required in order to visually observe water level in the spent fuel pool. Without access to the refuel floor, it would not be possible to determine the applicability of EAL 1.5.2. For those radiation monitors whose upper range limits are less than 8.0 R/hr, a value of offscale high is used.

PEG Reference(s):

AA2.1 Bases Reference(s):

1. NUREG-0818, Emergency Action Levels for Light Water Reactors

2. NUREG/CR-4982; Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82, July 1987

3. NRC Information Notice No. 90-08, KR-85 Hazards from Decayed Fuel

4. Nl-ARP-Ll, annunciator Ll-4-3

5. Niagara Mohawk Power Corporation Memo File Code NMP31027, Exposure Guidelines for Unusual/Accident Conditions November 1996 Page.16 EPMP-EPP-0101 Rev Ol

STT HIIE 1.4.3 jg eat Sustained area radiation levels > 15 mR/hr in either:

Control Room OR Central Alarm Station (CAS) and Secondary Alarm Station (SAS)

NUSLRC IC:

Release of radioactive material or increases in radiation levels within the facility that impedes operation of systems required to maintain safe operations or to establish or maintain cold shutdown.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This EAL addresses increased radiation levels that impede necessary access to operating stations requiring continuous occupancy to maintain safe plant operation or perform a safe plant shutdown. Areas requiring continuous occupancy include the Control Room, the central alarm station (CAS) and the secondary security alarm station (SAS).

The security alarm stations are included in this EAL because of their importance to permitting access to areas required to assure safe plant operations.

The value of 15 mR/hr is derived from the GDC 19 value of 5 rem in 30 days with adjustment for expected occupancy times. Although Section III.D.3 of NUREG-0737, Clarification of TMI Action Plan Requirements", provides that the 15 mR/hr value can be averaged over the 30 days, the value is used here without averaging. A 30 day duration implies an event potentially more significant than an Alert.

It is the impaired ability to operate the plant that results in the actual or potential degradation of the level of safety of the plant.

The cause or magnitude of the increase in radiation levels is not a concern of this EAL. The Emergency Director must consider the source or cause of the increased radiation levels and determine EALs may be involved.

if any other For example, a dose rate of 15 mR/hr in the Control Room may be a problem in itself. However, the increase may also be indicative of high dose rates in the containment due to a LOCA. In this latter case, a Site Area Emergency or a General Emergency may be indicated by other EAL categories.

This EAL could result in declaration of an Alert at NHP-1 due to a radioactivity release or radiation shine resulting from a major accident at the NNP-2 or JAFNPP. Such a declaration would be appropriate if the increase impairs safe plant operation.

November 1996 Page 17 EPMP-EPP-0101 Rev 01

1.4.3 (Cont)

This EAL is not intended to apply to anticipated temporary radiation increases due to planned events (e. g., radwaste container movement, depleted resin transfers, etc.).

PEG Reference(s):

AA3.1 Basis Reference(s):

1. GDC 19

2. NUREG-0737, "Clarification of TMI Action Plan Requirements",

Section III.D.3 1.4.4 Alert Sustained area radiation levels > 8 R/hr in any areas, Table 2 AND Access is required for safe operation or shutdown Table'2 Plant Safet Function Areas Reactor Building Turbine Building Screen and Pump House Off Gas Building NUMARC IC:

Release of. radioactive material or increases in radiation levels within the facility that impedes operation of systems required to maintain safe operations or to establish or maintain cold shutdown.

FPB Loss/Potential Loss:

N/A Node Applicability:

All November 1996 Page 18 EPMP-EPP-0101 Rev 01

1.4.4 (Cont)

Basis:

This EAL addresses increased radiation levels in areas requiring infrequent access in order to maintain safe plant operation or perform a safe plant shutdown. Area radiation levels at or above 8 R/hr are indicative of radiation fields which may limit personnel access. This bases of the value is described in NHPC memo File Code NHP31027 "Exposure Guidelines For Unusual/Accident Conditions". The areas selected are consistent with those listed in other EALs and represent those structures which house systems and equipment necessary for the safe operation and shutdown of the plant.

It is the impaired ability to operate the plant that results in the actual or potential degradation of the level of safety of the plant.

The cause or magnitude of the increase in radiation levels is not a concern of this EAL. The Emergency Director must consider the source or cause of the increased radiation levels and determine EAL may be involved.

if any other For example, a dose rate of 8 R/hr may be a problem in itself. However, the increase may also be indicative of high dose rates in the containment due to a LOCA. In this latter case, a Site Area Emergency or a General Emergency may be indicated by other EAL categories.

This EAL could result in declaration of an Alert at NHP-I due to a radioactivity release or radiation shine resulting from a major accident at the NHP-2 or JAFNPP. Such a declaration would be appropriate if the increase impairs safe plant operation.

This EAL"is not meant to apply to increases in the containment radiation monitors as these are events which are addressed in other EALs. Nor is it intended to apply to anticipated temporary radiation increases due to planned events (e. g., radwaste container movement, deplete resin transfers, etc.).

PEG Reference(s):

AA3.2 Basis Reference(s):

Niagara Mohawk Power Corporation Memo File Code NHP 31027, Exposure Guidelines for Unusual/Accident Conditions November 1996 Page 19 EPHP-EPP-0101 Rev 01

~TIA II T 1.5 Refue cc de ts 1.5.1 Unusual Event Spent fuel pool/ reactor cavity water level cannot be restored and maintained above the spent fuel pool low water level alarm.

KUMARC IC:

Unexpected increase in plant radiation or airborne concentration.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

The above event has a long lead time relative to the potential for radiological release outside the site boundary, thus impact to public health and safety is very low. However, in light of recent industry events, classification as an Unusual Event is warranted as a precursor to a more serious event.

The spent fuel pool low water level alarm setpoint is actuated by LS-26C which alarms at El 338'". The definition of "... cannot be restored and maintained above ..." allows the operator to visually observe the low water level condition, if possible, and to attempt water level restoration instructions as long as water level .remains above the top of irradiated fuel. Water level restoration instructions are performed in accordance with procedure Nl-SOP-20, Loss of SFP/Rx Cavity Level/Decay Heat Removal.

When the fuel transfer canal is directly connected to the spent fuel pool and reactor cavity, there could exist the possibility of uncovering irradiated fuel in the fuel transfer canal. Therefore, this EAL is applicable for conditions in which irradiated fuel is being transferred to and from the RPV and spent fuel pool.

PEG Reference(s):

AU2.1 Basis Reference(s):

None November 1996 Page 20 EPMP-EPP-0101 Rev Ol

alert Imminent report of actual visual observation of irradiated fuel uncovered NUNRC IC:

Major damage to irradiated fuel or loss of water level that has or will result in the uncovering of irradiated fuel outside the reactor vessel.

FPB Loss/Potential Loss:

N/A Node Appli'cability:

All Basis:

This EAL is defined by the specific areas where irradiated fuel is located such as reactor cavity, reactor vessel, or spent fuel pool.

Sufficient time exists to take corrective actions for these'conditions and there is little potential for substantial fuel damage. NUREG/CR-4982 "Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82" indicates that even if corrective actions are not taken, no prompt fatalities are predicted and the risk of injury is low. In addition, NRC Information Notice No. 90-08, KR-85 Hazards from Decayed Fuel" presents the following it its discussion:

"In the event of a serious accident involving decayed spent fuel, protective actions would be needed for personnel on site, while offsite doses (assuming an exclusion area radius of one mile from the plant site) would be well below the Environmental Protection Agency's Protective Action Guides. Accordingly, it is important to be able to properly survey and monitor for Kr-85 in the event of an accident with decayed spent fuel."

Thus, an Alert Classification for this event is appropriate.

Escalation, if appropriate, would occur by Emergency Director judgment in EAL Category 9.0.

There is no indication that water level in the spent fuel pool has dropped to the level of the fuel other than by visual observation by personnel on the refueling floor. When the fuel transfer canal is directly connected to the spent fuel pool and reactor cavity, there could exist the possibility of uncovering irradiated fuel in the fuel transfer canal. Therefore, this EAL is applicable for conditions in which irradiated fuel is being transferred to and from the RPV and spent fuel pool. NI-SOP-20, Loss of SFP/Rx Cavity Level'/Decay Heat Removal, provides appropriate instructions to report a visual observation of irradiated fuel uncovery.

November 1996 Page 21 EPMP-EPP-0101 Rev 01

1.5.2 (Cont)

This EAL applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage.

PEG Reference(s):

AA2.2 Basis Reference(s):

1. NUREG-0818, Emergency Action Levels for Light Water Reactors

2. NUREG/CR-4982, Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82, July 1987

3. NRC Information Notice No. 90-08, KR-85 Hazards 'from Decayed Fuel

4. Nl-SOP-20, Loss of SFP/Rx Cavity Level/Decay Heat Removal 2.0 REACTO RESSU E VESSEL PV The reactor pressure vessel provides a volume for the coolant which covers the reactor core. The RPV and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel cladding integrity fail.

There are two RPV parameters which are indicative of conditions which may pose a threat to RPV or fuel cladding integrity:

~ .RPV Water Level: RPV water level is directly related. to the status of adequate core cooling, and therefore fuel cladding integrity. Excessive ( > Tech. Spec.) reactor coolant to drywell leakage indications are utilized to indicate potential pipe cracks which may propagate to an extent threatening fuel clad, RPV and primary containment integrity. Conditions under which all attempts at establishing adequate core cooling have failed require primary containment flooding.

~ Reactor Power Reactivit Control: The inability to control reactor power below certain levels can pose a direct threat to reactor fuel, RPV and primary containment integrity.

2.1 PV Water eve E.l.l ~U1 I Unidentified drywell leakage ~ 10 gpm OR Reactor coolant to drywell identified leakage > 25 gpm November 1996 Page 22 EPHP-EPP-0101 Rev 01

2.1.1

~ ~ (Cont)

NUNARC IC:

RCS leakage FPB Loss/Potential Loss:

N/A Node Applicability:

Power Operation, Hot Shutdown Basis:

The conditions of this EAL may be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for unidentified drywell leakage was selected because it with normal Control Room indications and is consistent with the is observable 'he Technical Specification threshold for leaks beyond which increased risk of crack propagation exists. The 25 gpm value for identified reactor coolant to drywell leakage is set at a higher. value because of the significance of identified leakage in comparison to unidentified or pressure boundary leakage.

Only operating modes in which there is fuel in the reactor coolant system and the system is pressurized are specified.

PEG Reference(s):

SU5.1 Basis Reference(s):

None 2.1.2 Site Area Emer enc RPV water level cannot be restored and maintained > -84 in. (TAF)

NUMARC IC:

Loss of reactor vessel water level has or will uncover fuel in the reactor vessel.

FPB Loss/Potential Loss:

Fuel Clad Potential Loss, RCS Loss November 1996 Page 23 EPNP-EPP-0101 Rev Ol

TTAC EN (Cont) 2.1.2 (Cont)

Node Applicability:

Power Operation, Hot Shutdown, Cold Shutdown, Refuel Basis:

The RPV water level used in this EAL is the top of active fuel (TAF).

This value corresponds to the level which is used in EOPs to indicate challenge to core cooling and loss of the fuel clad barrier. This is the minimum water level to assure core cooling without further degradation of the clad. Severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured water level is not maintained above TAF.

if RPV Uncovery of the fuel irrespective of the event that causes fuel uncovery is justification alone for declaring a Site Area Emergency.

This includes events that could lead to fuel uncovery in any plant operating mode including cold shutdown and refuel. Escalation to a General Emergency occurs through radiological effluence addressed in EAL 1.3.3 for drywell radiation and in the EALs defined for Category 5.0, Radioactivity Release.

The terminology of "cannot be restored and maintained" is intended to be consistent with the interpretation that:

"The value of the identified parameter(s) is/is not able to be returned to above/below specified limits. This determination includes making an evaluation that considers both current and future system performance in relation to the current value and trend of the parameter(s). Neither implies that the parameter must actually exceed the limit before the classification is made nor that the classification must be made before the limit is reached. Does not imply any specific time interval but does not permit prolonged operation beyond a limit without making the specified classification."

This definition would require the emergency classification be made prior to water level dropping below TAF if, based on an evaluation of the current trend of RPV water level and in consideration of current and future injection system performance, that RPV water level will not likely be restored and maintained above TAF. This definition, however, also provides the latitude, based on that same evaluation, not to declare the SAE for those situations in which the RPV water level transiently drops below TAF in the process of RPV water level restoration.

November 1996 Page 24 EPHP-EPP-0101 Rev Ol

ATTACHMENT 1 (Cont) 2.1.2 (Cont)

PEG Reference(s):

SS5.1 FC2.1 RCS4.1 Bases Reference(s):

1. NI-ODP-PR0-0302, EOP Technical Bases 2.1.3 General Emer e c Drywell Flooding required NUMARC IC:

N/A FPB Loss/Potential Loss:

Fuel Clad Loss, RCS Loss, Containment Potential Loss Node Applicability:

Power Operation, Hot Shutdown Basis:

The condition in this EAL represents a potential for imminent melt sequences which, if not corrected, could lead to RPV failure and increased potential for primary containment failure. If the EOPs have been ineffective in restoring RPV water level above the top .of active fuel, loss of the fuel clad barrier may be imminent. Therefore, declaration of a General Emergency is appropriate when entry to the Drywell Flooding EOP is required.

PEG Reference(s):

PC4.1 Basis Reference(s):

1. Nl-ODP-PR0-0302, EOP Technical Bases November 1996 Page 25 EPHP-EPP-0101 Rev 01

AC (Cont) 2.2 Reactor Power Reactivit Control 2.2.1 /~et

~n RPS scram setpoint has been exceeded AND Automatic scram fails to result in a control rod pattern which assures reactor shutdown under all conditions without boron..

NUMARC IC:

Failure of Reactor Protection system instrumentation to complete or initiate an automatic reactor trip once a Reactor Protection system setpoint has been exceeded and manual trip was successful while in power operations or hot standby.

FPB Loss/Potential Loss:

N/A Node Applicability:

Power Operation Basis:

This condition indicates a failure of the Reactor Protection System to scram the reactor automatically, and maintain it in a shutdown under all conditions without boron. This is consistent with the entry conditions into Nl-EOP-03, "Failure to Scram".

If a manual scram does not result in reactor power being reduced below the APRM downscale setpoint (6X) or torus temperature exceeds the Boron Injection Initiation Temperature (110'F) escalation to a Site Area Emergency is required. A manual scram is any set of actions by the reactor operators at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor subcritical including manual scram push buttons, ARI and mode switch.

November 1996 Page 26 EPHP-EPP-0101 Rev Ol

(Cont) 2.2.1 (Cont)

PEG Reference(s):

SA2.1 Basis Reference(s):

1. Nl-ODP-PR0-0302, EOP Technical Bases

2. "Hethodology for Development of Emergency Action Levels" NUHARC/NESP-007 Rev 2-guestions and Answers, June 1993 2.2.2 Site Area Emer enc

~An RPS scram setpoint has been exceeded AND Automatic and manual scrams fail to result in a control rod pattern which assures reactor shutdown under all conditions without boron.

AND Either:

Reactor power >6X OR Torus temperature >110 F NUMARC IC:

Failure of Reactor Protection system instrumentation to complete or initiate an automatic reactor trip once a Reactor Protection system setpoint has been exceeded and manual scram trip was not successful.

FPB Loss/Potential Loss:

N/A Mode Applicability:

Power Operation Basis:

This condition indicates failure of the Reactor Protection System to shut down the reactor (automatically or manually) and maintain shutdown under all conditions without boron. Under these conditions it the reactor is producing more heat than can be removed using available safety systems. A Site Area Emergency is indicated because conditions exist leading to imminent or potential loss of both the fuel clad and the primary containment.

The failure of. automatic initiation of a reactor scram followed by an unsuccessful manual initiating actions which can be rapidly taken at the reactor control console does not, by itself, lead to imminent loss of either fuel clad or primary containment barriers. It is the continued criticality under conditions requiring a rector scram alone with the continued addition of heat to the containment which poses the imminent threat to primary containment or fuel clad barriers. In accordance with the EOPs, Liquid Poison System is initiated based on heat addition to containment in excess of safety system capability under failure to scram conditions.

November 1996 Page 27 EPHP-EPP-0101 Rev Ol

2.2.2 (Cont)

An immediate manual scram is any set of actions by the reactor operator at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor subcritical, including manual scram pushbuttons, ARI and mode switch.

PEG Reference(s):

SS2.1 Basis Reference(s):

1. Nl-ODP-PR0-0302, EOP Technical Bases

2. "Methodology for Development of Emergency Action Level" NUNRC/NESP-007 Revision 2 guestions and Answers, June 1993 2.2. S~B

~ RPS AND scram setpoint has been exceeded Automatic and manual scrams fail to result in a control rod pattern which assures reactor shutdown under all conditions without boron AND Either:

RPV water level cannot be restored and maintained > -108 in.

OR Torus temperature and RPV pressure cannot be maintained < HCTL.

NUNRC IC:

Failure of the Reactor Protection System to complete an automatic trip and manual trip was not successful and there is indication of an extreme challenge to the ability to cool the core.

FPB Loss/Potential Loss:

N/A Node Applicability:

Power Operation Basis:

Under the conditions of this EAL, the efforts to bring the reactor subcritical have been unsuccessful and, as a result, the reactor is producing more heat than the maximum decay heat load for which the safety systems were designed.

November 1996 Page 28 EPHP-EPP-0101 Rev 01

~TT Hll N (C 2.2.3 (Cont)

An extreme challenge to the ability to cool the core .is indicated when RPV water level cannot be restored and maintained above the Minimum Steam Cooling RPV Water Level (-108 in.). This RPV water level is used in the EOPs to define the lowest RPV water level in a failure-to-scram event above which adequate core cooling can be maintained without sufficient steam cooling flow. This situation could be precursor for a core melt sequence.

An extreme challenge to the primary containment is indicated when the inability to remove heat during the early stages of this sequence results in heatup of the containment. The Heat Capacity Temperature Limit (HCTL) is a measure of the maximum heat load which the primary containment can withstand; This situation could be a precursor for containment failure.

In this situation, core degradation can occur rapidly For this reason, the General Emergency declaration is intended to be anticipatory of the loss of two fission product barriers and a potential loss of a third thus permitting the maximum offsite intervention time.

An immediate manual scram is any set of actions by the reactor operator at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor subcritical including manual scram push buttons, ARI and mode switch.

PEG Reference(s):

SG2.1 Basis Reference(s):

1. NI-ODP-PR0-0302, EOP Technical Bases 3.0 P Y CONTAINMENT PC The primary containment structure is a pressure suppression system.

It forms a fission product barrier designed to limit the release of radioactive fission products generated from any postulated accident so as to preclude exceeding offsite exposure limits.

The primary containment structure is a low leakage pressure suppression system housing the reactor pressure vessel (RPV), the reactor coolant recirculation piping and other branch connections of the reactor primary system. The primary containment 'is equipped with isolation valves for most systems which penetrate the containment boundary. These valves automatically actuate to isolate systems under emergency conditions.

November 1996 Page 29 EPMP-EPP-0101 Rev 01

TTACHM N (Cont) 3.0 (Cont)

There are four primary containment parameters which are indicative of conditions which may pose a threat to primary containment integrity or indicate degradation of RPV or reactor fuel integrity.

~ Co tai ment Pressure: Excessive primary containment pressure is also indicative of either primary system leaks into containment or loss of containment cooling function. Primary containment pressures at or above specified limits pose a direct threat to primary containment integrity and the pressure suppression function.

~ or s em e ature: Excessive torus water temperatures can result in a loss of the pressure suppression capability of containment and thus be indicative of severely degraded RPV and containment conditions.

~ Co b st ble Gas Co centrations: The existence of combustible gas concentrations in containment pose a severe threat to containment integrity and are indicative of severely degraded reactor core and/or RPV conditions.

~ Conta nment Isolation Status: The existence of an unisolable steam line break outside .containment constitutes a loss of containment integrity as well as a loss of RCS boundary. Should a loss of fuel cladding integrity occur, the potential for release of large amounts of radioactive materials to the environment exists.

3.1 Containme t ressure 3.1.1 alert Drywell pressure cannot be maintained < 3e5 psig due to coolant leakage NUNRC IC:

N/A FPB Loss/Potential Loss:

RCS Loss Node Applicability:

Power Operation, Hot Shutdown November 1996 . Page 30 EPMP-EPP-0101 I

Rev Ol

STTACIIIIENT I tC 3.1.1

~ ~ (Cont)

Basis:

The primary containment pressure value is the drywell high pressure scram setpoint and is indicative of a LOCA event. The term "cannot be maintained below" is intended to be consistent with the conditions specified in the Primary Containment Control EOP indicative of a high energy release into containment for which normal containment cooling systems are insufficient.

PEG Reference(s):

RCS2.1 Basis Reference(s):

1. Nl-ARP-F1, annunciator 1-5

2. Nl-ARP-F4, annunciator 1-4

3. Nl-EOP-4, Primary Containment Control 3.1.2 S te Area Emer enc Drywell pressure cannot be maintained < 3.5 psig AND Coolant activity ) 300 pCi/gm I- 131 equivalent NUMARC IC:

N/A FPB Loss/Potential Loss:

Fuel Clad Loss, RCS Loss Node Applicability:

Power Operation, Hot Shutdown Basis:

The primary containment pressure value is the drywell high pressure scram setpoint and is indicative of a LOCA event. The term "cannot be maintained below" is intended to be consistent with the conditions specified in the Primary Containment Control EOP indicative of a high energy release into containment for which normal containment cooling systems are insufficient.

November 1996 Page 31 EPNP-EPP-0101 Rev Ol

ATTACHHENT 1 (Cont) 3.1.2 (Cont)

Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This amount of coolant activity is well above that expected for iodine spikes and corresponds to about 2X to 5X fuel clad damage. Qhen reactor coolant activity reaches this level, significant clad heating has occurred and thus the fuel clad barrier is considered lost.

The combination of these conditions represents a loss of two fission product barriers and, therefore, declaration of a Site Area Emergency is warranted.

PEG Reference(s):

FC1.1 RCS2.1 Bases Reference(s):

1. Nl-ARP-Fl, annunciator 1-5

2. Nl-ARP-F4, annunciator 1-4

3. General Electric NED0-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions

4. Nl-EOP-4, Primary Containment Control 3.1.3 General Emer enc Primary containment venting is required due to PCPL NUMARC IC:

N/A FPB Loss/Potential Loss:

Fuel Clad Loss, RCS Loss, Containment Loss Mode Applicability:

Power Operation, Hot Shutdown Basks:

Loss of primary containment is indicated when proximity to the Primary Containment Pressure Limit (PCPL) requires venting irrespective of the offsite radioactivity release rate. To reach the PCPL, primary containment pressure must exceed that predicted in any plant design bases accident analysis. A loss of the RCS barrier must have occurred with a potential loss of the fuel clad barrier.

November 1996 Page 32 EPHP-EPP-0101 Rev Ol

~TIACIIII NT I (C t) 3.1.3

~ ~ (Cont)

PEG Reference(s):

PC1.3 PC2.2 Bases Reference(s):

1. NI-ODP-PR0-0302, EOP Technical Bases 3.2 Torus Tem erature 3.2.1 Site Area Emer enc Torus temperature and RPV pressure cannot be maintained < HCTL (non-ATWS)

NUNARC IC:

Complete loss of function needed to achieve or maintain hot shutdown with reactor coolant > 212 F.

FPB Loss/Potential Loss:

N/A Node Applicability:

Power Operation, Hot Shutdown Basis:

This EAL addresses complete loss of functions, including ultimate heat sink, required for hot shutdown with the reactor at pressure and temperature. Under these conditions, there is an actual major failure of a system intended for protection of the public. Thus, declaration of a Site Area Emergency is warranted.

Functions required for hot shutdown consist of the ability to achieve reactor shutdown and to discharge decay heat energy from the reactor to the ultimate heat sink. Inability to remove decay heat energy is reflected .in an increase in torus temperature. Elevated torus temperature is addressed by the Heat Capacity Temperature Limit (HCTL). The HCTL is a function of RPV pressure and torus water temperature. If RPV pressure and torus temperature cannot be maintained below the HCTL, primary containment integrity is challenged and declaration of a Site Area Emergency is warranted.

"non-ATWS" has been added parenthetically to discriminate from General Emergency EAL 2.2.4.

November 1996 Page 33 EPNP-EPP-0101 Rev Ol

~ITA HM NT tC t) 3.2.1 (Cont)

PEG Reference(s):

SS4.1 Basis Reference(s):

1. Nine Mile Point Nuclear Station Unit 1 Appendix 'R'eview Safe Shutdown Analysis, Figure V-I Addresses: "Hot Shutdown Systems" "Functional Perf. Criteria Req. for Station Shutdown" 3.3 Combustible Gas Concentratio 3.3.1 Site Area Emer enc Z 4A Hz exists in DW or torus NUNRC IC:

N/A FPB Loss/Potential Loss:

Fuel Clad Loss, RCS Loss Node Applicability:

Power Operation, Hot Shutdown Basis:

4X hydrogen concentration is the lowest hydrogen concentration which, in the presence of sufficient oxygen, can support upward flame propagation. This hydrogen concentration is generally considered the lower boundary of the range in which localized deflagrations may occur. To generate such a concentration of combustible gas, loss of both the fuel clad and RCS barriers must have occurred. Therefore, declaration of a Site Area Emergency is warranted.

If hydrogen concentrations increase in conjunction with the presence of oxygen to global deflagration levels (i.e. ~ 6X hydrogen and > 5X oxygen), venting of the containment irrespective of the offsite radioactive release rate would be required by EOPs and declaration of a General Emergency required.

November 1996 Page 34 EPHP-EPP-0101 Rev 01

BTTACIIIIENT I lC 3.3.1 (Cont)

PEG Reference(s):

SS5.2 Basis Reference(s):

1. Nl-ODP-PR0-0302, EOP Technical Bases 3.3.2 Gene al Emer enc Primary containment venting is required due to combustible gas concentrations NUNARC IC:

N/A FPB Loss/Potential Loss:

Fuel Clad Loss, RCS Loss, Containment Loss Node Applicability:

All Basis:

6X hydrogen concentration in the presence of 5X oxygen concentration is the lowest concentration at which a deflagration inside of the primary containment could occur. When hydrogen and oxygen concentrations reach or exceed combustible limits, imminent loss of the containment barrier exists. To generate such levels of combustible gas, loss of the fuel clad and RCS barriers must have occurred. Venting of the containment irrespective of the offsite radioactive release rate is required by EOPs for this condition.

PEG Reference(s):

PC1.4 PC2.2 Basis Reference(s):

1. Nl-ODP-PR0-0302, EOP Technical Bases November 1996 Page 35 EPHP-EPP-0101 Rev 01

SIIA IIMENT I lC t) 3.4 Containment Isolation Status 3.4.1 Site rea r enc HSL, EC steam line or Reactor Water Clean-up Isolation failure AND A release pathway, outside normal process system flowpaths from the unisolable system, exists outside primary containment.

NUNRC IC:

N/A FPB Loss/Potential Loss:

RCS Loss, Containment Loss Node Applicability:

Power Operation, Hot Shutdown Basis:

This EAL covers containment isolation failures allowing a direct flow path to the environment. A release pathway outside primary containment exists when steam flow is not prevented by downstream isolations. In the case of a failure of both isolation valves to close but in which no downstream flowpath exists, declaration under this EAL would not be required. The conditions of this EAL represent the loss of both the RCS barrier and the primary containment barrier and thus justifies declaration of a Site Area Emergency.

PEG Reference(s):

PC2.1 Basis Reference(s):

None November 1996 Page 36 E PUP-EPP-0101 Rev 01

(Cont) 3.1.2 e c HSL, EC steam line isolation failure or Reactor Mater Clean-up isolation failure AND A release pathway, outside normal process system flowpaths from the unisolable system, exists outside primary containment AND any:

~ Coolant activity > 300 pCi/gm I-131 equivalent

~ RPV water level < -84 in. (TAF)

~ DM radiation > 3000 R/hr NUNARC IC:

N/A FPB Loss/Potential Loss:

Fuel Clad Loss/Potential Loss, RCS Loss, Containment Loss Node Applicability:

Power Operation, Hot Shutdown Basis:

The conditions of this EAL include the containment isolation failures allowing a direct flow path to the environment. A release pathway outside primary containment exists when steam flow is not prevented by downstream.i'solations. In the case of a failure of both isolation valves to close but in which no downstream flowpath exists, declaration under this EAL would not be required. Containment isolation failures which result in a release pathway outside primary containment are the bases for declaration of Site Area Emergency in EAL 3.4.1.

Mhen isolation failures are accompanied by elevated coolant activity, RPV water level below TAF, or high drywell radiation, declaration of a General Emergency is appropriate due to loss of the primary containment barrier, RCS barrier, and loss or potential loss of the fuel clad barrier.

Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This amount of'coolant activity is well above that expected for iodine spikes and corresponds to about 2X to 5X fuel clad damage. When reactor coolant activity reaches this level, significant clad heating has occurred and thus the fuel clad barrier is considered lost.

November 1996 Page 37 EPHP-EPP-0101 Rev 01

3.4.2 (Cont)

The RPV water level used in this EAL is the top of active fuel (TAF).

This value corresponds to the level which is used in EOPs to indicate challenge to core cooling and loss of the fuel clad barrier. This is the minimum water level to assure core cooling without further degradation of the clad. Severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured water level is not maintained above TAF.

if RPV The drywell radiation reading is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell. The reading is calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 pCi/gm dose equivalent I-131 into the drywell atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations allowed within Technical Specifications (including iodine spiking) and are therefore indicative of fuel damage (approximately 2X - 5X clad failure depending on core inventory and RCS volume).

It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. Drywell High Range Radiation Monitors have a range of 0 to E8 R/hr on recorder RR 201.7-36C pen 1 and 2. They are installed in the following drywell locations:

RAm 201.7-36 Az 340', El 263'"

RAm 201.7-37 Az 310', EL 301'"

PEG Reference(s):

PC2.1 and FC1.1 PC2.1 and FC2.1 PC2.1 and FC3.1 November 1996 Page 38 EPMP-EPP-0101 Rev 01

STT CIIIIENT I lC t) 3.4.2 (Cont)

Basis Reference(s):

1. General Electric NED0-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions

2. Nl-ODP-PR0-0302, EOP Technical Bases

3. Nl-RG197-EIL1, Important Design Features of Regulatory Guide 1.97 Instruments

4. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Amendment 72, 76, Table 3.6.11-1

5. Calculation 1H21C003, Rev 0 4.0 SECONDARY CONT INNENT SC The secondary containment is comprised of the reactor building and associated ventilation, isolation and effluent systems. The secondary containment serves as an effective fission product barrier and is designed to minimize any ground level release of radioactive materials which might result from a serious accident.

The reactor building provides secondary containment during reactor operation and serves as primary containment when the reactor is shutdown'nd the drywell is open, as during refueling. Because the secondary containment is an integral part of the complete containment system, conditions which pose a threat to vital equipment located in the secondary containment are classifiable as emergencies.

There are two secondary containment parameters which are indicative of a direct release into secondary containment:

Secondar Containment Tem eratures: Abnormally high secondary containment area temperatures can also pose a threat to the operability of vital equipment located inside secondary containment including RPV water level instrumentation. High area temperatures may limit personnel accessibility to vital areas.

High area temperatures may also be indicative of either primary system discharges into secondary containment or fires.

Secondar Containment Area Radiation Levels: Abnormally high area radiation levels in secondary containment, although not necessarily posing a threat to equipment operability, may pose a threat to personnel safety and the ability to operate vital equipment due to a lack of accessibility. Abnormally high area radiation levels may also be the result of a primary system discharging into the secondary containment and be indicative of precursors to significant radioactivity release to the environment.

November 1996 Page 39 EPNP-EPP-0101 Rev 01

4.1 eactor Bu d Tem erature 4.1.1 S te ea e c Primary system is discharging outside PC AND RB general area temperatures are > 135 F in two or more areas, N1-EOP-5 NUMARC IC:

N/A FPB Loss/Potential Loss:

RCS Loss, Containment Loss Node Applicability:

Power Operation, Hot Shutdown Basis:

The presence of elevated area temperatures in the secondary containment may be indicative of an unisolable primary system leakage outside the primary containment. These conditions represent a loss of the containment barrier and a potential loss of the RCS barrier.

PEG Reference(s):

PC2.3 RCS1.3 Basis Reference(s):

1. Nl-ODP-PR0-0302, EOP Technical Bases

2. Nl-EOP-5 4.1.2 General Emer enc Primary system is discharging outside PC AND RB general area temperatures are >135 F in two or more areas, Nl-EOP-5 AND any:

Coolant activity > 300 pCi/gm 1-131 equivalent

~ RPV water level < -84.in. (TAF)

~ DM radiation > 3000 R/hr November 1996 Page 40 EPHP-EPP-0101 Rev 01

~TIA HN I ( t) 4.1.2 (Cont)

NUNRC IC:

N/A FPB Loss/Potential Loss:

Fuel Clad Loss/Potential Loss, RCS Loss, Containment Loss Node Applicability:

Power Operation, Hot Shutdown Basis:

The presence of elevated area temperatures in the secondary containment may be indicative of an unisolable primary system leakage outside the primary containment. These conditions represent a loss of containment barrier and a potential loss of the RCS barrier. 'he When secondary containment area temperatures are accompanied by elevated coolant activity, RPV water level below TAF, or high drywell radiation, declaration of a General Emergency is appropriate due to loss of the primary containment barrier, RCS barrier, and loss or potential loss of the fuel clad barrier.

Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This amount of coolant activity is well above that expected for iodine spikes and corresponds to about 2X to 5X fuel clad damage. When reactor coolant activity reaches this level, significant clad heating has occurred and thus the fuel clad barrier is considered lost.

The RPV water level used in this EAL is the top of active fuel (TAF).

This value corresponds to the level which is used in EOPs to indicate challenge to core cooling and loss of the fuel clad barrier. This is the minimum water level to assure core cooling without further degradation of the clad. Severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured water level is not maintained above TAF.

if RPV The drywell radiation reading is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell. The reading is calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 pCi/gm dose equivalent I-131 into the drywell atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations allowed within Technical Specifications (including iodine spiking) and are therefore indicative of fuel damage (approximately 2X - 5X clad failure depending on core inventory and RCS volume).

November 1996 Page 41 EPHP-EPP-0101 Rev 01

~HN 1 4.1.2 (Cont)

It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. Drywell High Range Radiation Nonitors have a range of 0 to E8 R/hr on recorder RR 201.7-36C pen 1 and 2. They are installed in the following drywell locations:

RAm 201.7-36 Az 340, El 263'"

RAm 201.7-37 Az 310, EL 301'"

PEG Reference(s):

PC2.3 and FC1.1 PC2.3 and FC2. 1 PC2.3 and FC3.1 Basis Reference(s):

1. Nl-ODP-PR0-0302, EOP Technical Bases

2. General Electric NED0-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions

3. Nl-RG197-EILl, Important Design Features of Regulatory Guide 1.97 Instruments

4. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Amendment 72, 76, Table 3.6.11-1

5. Calculation 1H21C003, Rev 0 6 .. N1-EOP-5 4.2 eactor Build adiatio Level 4.2.1 Site Area er e c Primary system is discharging outside PC AND RB area radiation levels are > 8.0 R/hr in two or more areas, Nl-EOP-5 NVNARC IC:

N/A FPB Loss/Potential Loss:

RCS Loss, Containment Loss Node Applicability:

Power Operation, Hot Shutdown November 1996 Page 42 EPHP-EPP-0101 Rev 01

ST A T (C 4.2.1 (Cont)

Basis:

The presence of elevated area radiation levels in the secondary containment may be indicative of an unisolable primary system leakage outside the primary containment. These conditions represent a loss of the containment barrier and a potential loss of the RCS barrier.

PEG Reference(s):

PC2.3 RCS1.3 Basis Reference(s):

1. Nl-ODP-PR0-0302, EOP Technical Bases

2. Nl-EOP-5 4.2.2 General Emer enc Primary system is discharging outside PC AND RB area radiation levels are > 8.0 R/hr in two or more areas, Nl-EOP-5 AND any:

~ Coolant activity > 300 pCi/gm 1-131 equivalent

~ RPV water level < -84 in. (TAF)

~ DW radiation > 3000 R/hr NUNARC IC:

N/A FPB Loss/Potential Loss:

Fuel Clad Loss/Potential Loss, RCS Loss, Containment Loss Node Applicability:

Power Operation, Hot Shutdown Basis:

The presence of elevated area radiation levels in the secondary containment may be indicative of an unisolable primary system leakage outside the primary containment. These conditions represent a loss of the containment barrier and a potential loss of the RCS barrier.

When secondary containment radiation levels are accompanied by elevated coolant activity, RPV water level below TAF, or high drywell radiation, declaration of a General Emergency is appropriate due to loss of the primary containment barrier, RCS barrier, and loss or potential loss of the fuel clad barrier.

November 1996 Page 43 EPHP-EPP-0101 Rev Ol

(Cont) 4.2.2 (Cont)

Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This amount of coolant activity is well above that expected for iodine spikes and corresponds to about 2X to 5X fuel clad damage. Mhen reactor coolant activity reaches this 1'evel, significant clad heating has occurred and thus the fuel clad barrier is considered lost.

The RPV water level used in this EAL is the top of active fuel (TAF).

This value corresponds to the level which is used in EOPs to indicate challenge to core cooling and loss of the fuel clad barrier. This is the minimum water level to assure core cooling without further degradation of the clad. Severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured if RPV water level is not maintained above TAF.

The drywell radiation reading is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell. The reading is calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 pCi/gm dose equivalent I-131 into the drywell atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations allowed within Technical Specifications (including iodine spiking) and are therefore indicative of fuel damage (approximately 2X 5X clad failure depending on core inventory and RCS volume).

It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. Drywell High Range Radiation Monitors have a range of 0 to E8 R/hr on recorder RR 201.7-36C pen 1 and 2. They are installed in the following drywell locations:

RAm 201.7-36 Az 340', El 263'"

RAm 201.7-37 Az 310, EL 301'"

PEG Reference(s):

PC2.3 and FCl.l PC2.3 and FC2.1 PC2.3 and FC3.1 November 1996 Page 44 EPHP-EPP-0101 Rev Ol

BITA HMENT I 4.2.2 (Cont)

Basis Reference(s):

1. Nl-ODP-PR0-0302, EOP Technical Bases

2. General Electric NED0-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions

3. Nl-RG197-EILI, Important Design Features of Regulatory Guide 1.97 Instruments

4. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Amendment 72, 76, Table 3.6.11-1

5. Calculation 1H21C003, Rev 0

6. N1-EOP-5 5.0 RADIOACTIVITY RELEASE Many EALs are based on actual or potential degradation of fission product bar riers because of the increased potential for offsite radioactivity release. Degradation of fission product barriers though, is not always apparent via non-radiological symptoms.

Therefore, direct indication of increased radiological effluents are appropriate symptoms for emergency classification.

At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions.

There are two basic indications of radioactivity release rates which warrant emergency classifications.

Effluent Monitors: Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits.

Dose Pro 'ection and or nvironmental Measurements: Projected offsite doses (based on effluent monitor readings) or actual offsite field measurements indicating doses or dose rates above classifiable limits.

November 1996 Page 45 EPMP-EPP-0101 Rev 01

~IT II T I (C 5.1 ff uent No itors 5.1.1 Unusua vent A valid reading from an unplanned release on any monitors from Table 3 "UE" column for > 60 min. unless sample analysis can confirm release rates < 2 x technical specifications within this time period.

Table 3 ffluent Nonitor Class icat on Thresholds Monitor UE Al ert SAE GE Stack (RN10A/B) >300 cps ~3.0E4 cps >5.0 E6 cps N/A EC Vent >10 mR/hr ~30 mR/hr h310 mR/hr N/A SW Effluent >900 cpm ~90,000 cpm N/A N/A RW Discharge ~2 x batch 2200 x batch N/A N/A NUNARC IC:

Any unplanned release of gaseous or liquid radioactivity to the environment that exceeds two times the radiological Technical Specifications for 60 minutes or longer.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

Valid means that a radiation monitor reading has been confirmed by the operators to be correct. Unplanned releases in excess of two times the site technical specifications that continue for 60 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of safety. The final integrated dose (which is very low in the Unusual Event emergency class) is not the primary concern; it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes. Therefore, it is not intended that the release be averaged over 60 minutes. For example, a release of 4 times T/S for 30 minutes does not exceed this initiating condition. Further, the Emergency Director should not wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 60 minutes.

November 1996 Page 46 EPMP-EPP-0101 Rev 01

ETE MENT EE 5.1.1 (Cont)

Two times the monitors alarm setpoints have been selected for use in this EAL. The alarm setpoints for the listed monitors are conservatively set to ensure Technical Specification radioactivity release limits are not exceeded. The value shown for the UE level is two times the high alarm setpoint for the Emergency Condenser vent monitor and the Service Water effluent monitor, and two times the high-high alarm setpoint for the main stack (OGESM) monitor.

The following radiation monitors are not included in this EAL:

Reactor Building Vent Monitors: Reactor building ventilation discharges to the main stack. Radioactivity release from the reactor building would, therefore, be assessed by the main stack monitor.

Containment Spray Raw Water Monitors: These monitors detect radiation in the discharge from their respective processes. The monitors are located upstream of the Service Water monitor. Therefore, the Service Water radiation monitor adequately detects offsite radioactivity releases from these systems.

PEG Reference(s):

AU1.1 Basis Reference(s):

1. Nl-OP-50B Process Radiation Monitoring System

2. Nl-ARP-Hl Annunciator Hl-1-8

3. Nl-CSP-(308, Attachment 2

4. Nl-CSP-9215, Service Water Alarm Setpoint Determination, Attachment 2

5. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications

6. Calculation 1H21C003, Rev 0 November 1996 Page 47 EPMP-EPP-0101 Rev 01

CH N (Cont) 5.1. 2 ~lert A valid reading from an unplanned release on any monitors from Table 3 "Alert" column for > 15 min. unless dose assessment can confirm releases are below Table 4 column "Alert" within this time period.

Table 3 fflue t Nonito Class f cation T resholds Monitor UE Alert SAE GE Stack (RN10A/B) 2300 cps ~3.0E4 cps >5.0 E6 cps N/A EC Vent hlO mR/hr Z30 mR/hr 2310 mR/hr N/A SW Effluent >900 cpm ~90,000 cpm N/A N/A RW Discharge h2 x batch Z200 x batch N/A N/A Table 4 Dose Pro ection Env. Neasurement Classification Thresholds alert SAE GE TEDE 10 mRem 100 mRem 1000 mRem CDE Thyroid N/A 500 mRem 5000 mRem External exposure rate 10 mRem 100 mRem/hr 1000 mRem/hr Thyroid exposure rate N/A 500 mRem/hr 5000 mRem/hr (for 1 hr. of inhalation)

NUNRAC IC:

Any unplanned release of gaseous or liquid radioactivity to the environment that exceeds 200 times radiological Technical Specifications for 15 minutes or longer.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

Valid means that a radiation monitor reading has been confirmed by the operators to be correct. This event escalates from the Unusual Event by increasing the magnitude of the release by a factor of 100 over the Unusual Event level (i. e., 200 times Technical Specifications).

Prorating the 500 mR/yr bases of the 10CFR20 non-occupational DAC limits for both time (8766 hr/yr) and the 200 multiplier, the associated site boundary dose rate would be 10 mR/hr. The required release duration was reduced to 15 minutes in recognition of the inct eased severity.

The following radiation monitors are not included in this EAL:

November 1996 Page 48 EPHP-EPP-0101 Rev 01

~TT 7 5.1e2 (Cont)

Reactor Building Vent Honitors: Reactor building ventilation discharges to the main stack. Radioactivity release from the reactor building would, therefore, be assessed by the main stack monitor.

Containment Spray Raw Mater Honitors: These monitors detect radiation in the discharge from their respective processes. The monitors are located upstream of the Service Mater monitor. Therefore, the Service Water radiation monitor adequately detects offsite radioactivity releases from these systems.

PEG Reference(s):

AA1.1 Basis Reference(s):

1. Nl-OP-50B, Process Radiation Honitoring System

2. Nl-ARP-H1, Annunciator Hl-1-8

3. Nl-CSP-f308, Attachment 2

4. Nl-CSP-(215, Service'Water Alarm Setpoint Determination, Attachment 2

5. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications

6. Calculation 1H21C003, Rev 0 5.1.3 Site Area Emer enc A valid reading from an unplanned release on any monitors from Table 3 "SAE column for > 15 min. unless dose assessment can confirm releases are below Table 4 column "SAE" within this time period.

Table 3 flue t Honitor Classi cation Thresholds Honitor UE Al ert SAE GE Stack (RNlOA/B) ~300 cps ~3.0E4 cps >5.0 E6 cps N/A EC Vent ~10 mR/hr ~30 mR/hr h310 mR/hr N/A SM Effluent ~900 cpm ~90,000 cpm N/A N/A RW Discharge ~2 'x batch ~200 x batch N/A N/A Table 4 Dose Pro ectio Env. Neasurement Classification Thresholds Alert SAE GE TEDE 10 mRem 100 mRem 1000 mRem CDE Thyroid N/A 500 mRem 5000 mRem External exposure rate 10 mRem/ 100 mRem/hr 1000 mRem/hr Thyroid exposure rate N/A 500 mRem/hr 5000 mRem/hr (for 1 hr. of inhalation)

E November 1996 Page 49 EPHP-E PP-0101

STIA IIIIENT 1 (C t) 5.1.3 (Cont)

NUNRAC IC:

Boundary dose resulting from an actual or imminent release of gaseous radioactivity exceeds 100 mRem TEDE or 500 mRem CDE Thyroid for the actual or projected duration of the release.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

Valid means that a radiation monitor reading has been confirmed by the to be correct. The SAE values of Table 5.1 are based on the 'perators boundary dose resulting from an actual or imminent release of gaseous radioactivity that exceeds 100 mR whole body or 500 mR child thyroid for the actual or projected duration of the release. The 100 mR integrated dose is based on the proposed 10CFR20 annual average population exposure. The 500 mR integrated child thyroid dose was established in consideration of the 1:5 ratio of the EPA Protective Action Guidelines for whole body thyroid.

These values provide a desirable gradient (one order of magnitude) between the Alert, Site Area Emergency, and General Emergency classifications. It is deemed that exposures less than this limit are not consistent with the Site Area Emergency class description.

Integrated doses are generally not monitored in real-time.. In establishing this emergency action level, a duration of one hour is assumed based on site boundary doses for either whole body or child thyroid, whichever is more limiting (depends on source term assumptions).

The FSAR source terms applicable to each monitored pathway are used in determining indications for the monitors on that pathway.

The values are derived from Calculation 1H21C003, Rev. 0.

PEG Reference(s):

AS1.1 November 1996 Page 50 EPHP-EPP-0101 Rev 01

~TTA II NT lC 5.1.3 (Cont)

Basis Reference(s):

1. Nl-OP-50B, Process Radiation Nonitoring System

2. NI-ARP-H1, Annunciator Hl-1-8

3. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications

4. Calculation 1H21C003, Rev. 0 5.2 Dose Pro ections Environmental Neasurements 5.2.1 Unusual Event Confirmed sample analyses for gaseous or liquid release rates > 2 x technical specifications limits for > 60 min.

NUMARC IC:

Any unplanned release of gaseous or liquid radioactivity to the environment that exceeds two times the radiological Technical Specifications for 60 minutes or longer.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

Confirmed sample analyses in excess of two times the site technical specifications that continue for 60 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of safety. The final integrated dose (which is very low in the Unusual Event emergency class) is not the primary concern; it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes. Therefore, it is not intended that the release be aver aged over 60 minutes. For example, a release of 4 times T/S for 30 minutes does not exceed this initiating condition.

Further, the Emergency Director should not wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 60 minutes.

November 1996 Page 51 EPNP-EPP-0101 Rev Ol

PEG Reference(s):

AU1.2 Basis Reference(s):

1. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Article 3.6.15.a(l) and Article 3.6.15.b(l) (a) and (b) 5.2.2 Alert Confirmed sample analyses for gaseous or liquid release rates > 200 x technical specifications limits for > 15 min.

NUNRC IC:

Any unplanned release of gaseous or liquid radioactivity to the environment that exceeds 200 times radiological Technical Specifications for 15 minutes or longer.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

Confirmed sample analyses in excess of two hundred times the site technical specifications that continue for 15 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of .safety. This event escalates from the Unusual Event by increasing the magnitude of the release by a factor of 100 over the Unusual Event level (i. e., 200 times Technical Specifications).

Prorating the 500 mR/yr bases of the 10CFR20 non-occupational DAC limits for both time (8766 hr/yr) and the 200 multiplier, the associated site boundary dose rate would be 10 mR/hr. The required release duration was reduced to 15 minutes in recognition of the increased severity.

PEG Reference(s):

AA1. 2 Basis Reference(s):

1. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Article 3.6. 15.a(1) and Article 3.6.15.b(1)(a) and (b)

November 1996 Page 52 EPMP-EPP-0101 Rev Ol

SIIAANN I 1 c 5.2.3 g e~t Dose projections or field surveys resulting from actual or imminent release which indicate doses / dose rates > Table 4 column "Alert" at the site boundary or beyond Table 4 ose Pro 'ection Env. easur ement Classification Thresholds Alert SAE GE TEDE 10 mRem 100 mRem 1000 mRem CDE Thyroid N/A 500 mRem 5000 mRem External exposure 10 mRem/hr 100 mRem/hr 1000 mRem/hr rate Thyroid exposure N/A 500 mRem/hr 5000 mRem/hr rate (for 1 hr. of inhalation)

NUNARC IC:

Any unplanned release of gaseous or liquid radioactivity to the environment that exceeds 200 times radiological Technical Specifications for 15 minutes or longer.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

Offsite integrated doses in excess of 10 mR TEDE or dose rates in excess of 10 mR/hr TEDE represent an uncontrolled situation and hence, a potential degradation in the level of safety. This event escalates from the Unusual Event by increasing the magnitude of the release by a factor of 100 over the Unusual Event level (i. e., 200 times Technical Specifications). Prorating the 500 mR/yr bases of 10CFR20 for both time (8766 hr/yr) and the 200 multiplier, the associated site boundary dose rate would be 10 mR/hr.

November 1996 Page 53 EPHP-EPP-0101 Rev 01

BTTTA H N 1 (C 5.2.3 (Cont)

Basis (Cont)

As previously stated, the 10 mR/hr value is based on a proration of 200 times the 500 mR/yr bases of 10CFR20, rounded down to 10 mR/hr.

Imminent is intended to mean that a release will occur.

PEG Reference(s):

AAl.2 Basis Reference(s):

1. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Article 3.6.15.a(1) and Article 3.6.15.b(l)(a) and (b) 5.2.4 Site rea Emer enc Dose, projections or field surveys resulting from actual or imminent release which indicate doses / dose rates > Table 4 column "SAE" at the site boundary or beyond Table 4 Dose Pro 'ection Env. Measurement Classification Thresholds Alert SAE GE TEDE 10 mRem 100 mRem 1000 mRem CDE Thyroid N/A 500 mRem 5000 mRem External exposure 10 mRem/hr 100 mRem/hr 1000 mRem/hr rate Thyroid exposure N/A 500 mRem/hr 5000 mRem/hr rate (for 1 hr. of inhalation)

NUNRC IC:

Boundary dose resulting from an actual or imminent release of gaseous radioactivity exceeds 100 mRem TEDE or 500 mRem CDE Thyroid for the actual or projected duration .of the release.

FPB Loss/Potential Loss:

N/A November 1996 Page 54 EPMP-EPP-0101 Rev Ol

ATIA HIIEET I EC 5.2.4 (Cont)

Node Applicability:

All Basis:

The 100 mR integrated TEDE dose in this EAL is based on the proposed 10CFR20 annual average population exposure. This value also provides a desirable gradient (one order of magnitude) between the Alert, Site Area Emergency, and General Emergency classes. It is deemed that exposures less than this limit are not consistent with the Site Area Emergency class description. The 500 mR integrated CDE thyroid dose was established in consideration of the 1:5 ratio of the EPA Protective Action Guidelines for whole body thyroid. In establishing the dose rate emergency action levels, a duration of one hour is assumed. Therefore, the dose rate EALs are based on a site boundary dose rate of 100 mR/hr TEDE or 500 mR/hr CDE thyroid, whichever is more limiting.

Imminent is intended to mean that a release will occur.

PEG Reference(s):

AS1.3 AS1.4 Basis Reference(s):

1. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications 5.5.5 Dose projections or field surveys resulting from actual or imminent release which indicate doses / dose rates > Table 4 column "GE" at the site boundary or beyond Table 4 Dose Pro 'ection Env. Measurement Classification Thresholds Alert SAE GE TEDE 10 mRem 100 mRem 1000 mRem CDE Thyroid N/A 500 mRem 5000 mRem External exposure 10 mRem/hr 100 mRem/hr 1000 mRem/hr rate Thyroid exposure N/A 500 mRem/hr 5000 mRem/hr rate (for 1 hr. of inhalation)

November 1996 Page 55 EPHP-EPP-0101 Rev 01

AC N (Cont) 5.2.5 (Cont)

NUNRC IC:

Boundary dose resulting from an actual or imminent release of gaseous radioactivity exceeds 1000 mRem TEDE or 5000 mRem COE Thyroid for the actual or projected duration of the release using actual meteorology.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

The General Emergency values of Table 5.2 are based on the boundary dose resulting from an actual or imminent release of gaseous radioactivity that exceeds 1000 mR TEDE or 5000 mR CDE thyroid for the actual or projected duration of the release. The 1000 mR TEOE and the 5000 mR CDE thyroid integrated dose are based on the EPA protective action guidance which indicates that public protective actions are indicated if the dose exceeds I rem TEDE or 5 rem CDE thyroid. This is consistent with the emergency class description for a General Emergency. This level constitutes the upper level of the desirable gradient for the Site Area Emergency. Actual meteorology is specifically identified since it gives the most accurate dose assessment. Actual meteorology (including forecasts) should be used whenever'possible. In establishing the dose rate emergency action levels, a duration of one hour is assumed. Therefore, the dose rate EALs are based, on a site boundary dose rate of 1000 mR/hr TEDE or 5000 mR/hr CDE thyroid, whichever is more limiting.

Imminent is intended to mean that a release will occur.

PEG Reference(s):

AG1.3 AG1.4 Basis Reference(s):

1. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications November 1996 Page 56 EPMP-EPP-0101 Rev 01

AT(A I(ME)IT I (t I) 6.0 C IC L AILVRES Loss of vital plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.

The events of this category have been grouped into the following two loss of electrical power types:

~ Loss of AC Power Sources: This category includes losses of onsite and/or offsite AC power sources including station blackout events.

oss of DC Power Sources: This category involves total losses of vital plant 125 vdc power sources.

6.1 Loss of AC Power Sources 6.1.1 Unusual Event Loss of power for > 15 min. to all:

~ T-101N

~ T-101S

~ T-10 backfed from offsite through T-1 or T-2 NUNARC IC:

Loss of all offsite power to establish busses for greater than 15 minutes.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

Prolonged loss of all offsite AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete loss of AC power (station blackout). Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

November 1996 Page 57 EPMP-EPP-0101 Rev Ol

~AH N (C 6.1.1 (Cont>

Backfeeding of the Station Transformer T10 has been included to allow for those conditions in which maintenance is being performed on the Station Reserve Transformers or 115 kv system. It is recognized that this is not a readily available source of offsite emergency power under emergency conditions and should only be taken credit for those conditions under which backfeeding has already been established.

PEG Reference(s):

SU1.1 Basis Reference(s):

1. Nl-OP-45, Emergency Diesel Generators

2. Nl-OP-30, 4.16 Kv, 600V, and 480V House Service 6.1. 2 gert Loss of all emergency bus AC power for >15 min.

NVNARC IC:

Loss of all offsite power and loss of all onsite AC power to essential busses during cold shutdown, refueling o} defueled mode.

FPB Loss/Potential Loss:

N/A Node Applicability:

Cold Shutdown, Refuel, Defuel Basis:

Loss of all AC power compromises all plant safety systems requiring electric power. This EAL is indicated by:

Loss of power to all:

T-101N T-101 S T-10 backfed through T-1 or T-2 AND failure of both DGs to power emergency buses AND failure to restore power to PB102 or PB103 in g 15 min.

AND Failure of both DGS to power emergency buses AND Failure to restore power to PB102 or PB103 in g 15 min.

November 1996 Page 58 EPMP-EPP-0101 Rev Ol

AC (Cont) 6.1.2 (Cont)

Mhen in cold shutdown, refueling, or defueled mode this event is classified as an Alert. This is because of the significantly reduced decay heat, lower temperature and pressure, thus increasing the time to restore one of the emergency busses, relative to that specified for the Site Area Emergency EAL. Escalating to the Site Area Emergency, if appropriate, is by Abnormal Rad Levels/Radiological Effluent, or Emergency Director Judgment ICs. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Backfeeding of the Normal Station Transformer has been included to allow for those conditions in which maintenance is being performed on the Station Reserve Transformers or 115 kv system. It is recognized that this .is not a readily available source of emergency power under emergency conditions and should only be taken credit for those conditions under which backfeeding has already been established.

PE6 Reference(s):

SA1.1 Basis Reference(s):

1. NI-OP-30, 4.16 Kv, 600V, and 480V House Service

2. Nl-OP-45, Emergency Diesel Generators 6.1.3

~ ~ alert Available emergency bus AC power reduced to only one of the following sources for >15 min.:

DGI02 (PB102)

DG103 (PB103)

T-101N T-101S KUNARC IC:

AC power capability to essential busses reduced to a single power source for greater than 15 minutes such that any additional single failure would result in station blackout with reactor coolant >212 'F.

FPB Loss/Potential Loss:

N/A Node Applicability:

Power Operation, Hot Shutdown November 1996 Page 59 EPNP-EPP-0101 Rev Ol

~IIN I (C

6. 1.3 (Cont)

Basis:

The condition indicated by this EAL is the degradation of the offsite power with a concurrent failure of one emergency generator to supply power to its emergency busses. The subsequent loss of this single power source would escalate the event to a Site Area Emergency.

PEG Reference(s):

SA5.1 Basis Reference(s):

1. Nl-OP-45, Emergency Diesel Generators

2. Nl-OP-30, 4.16 Kv, 600V, and 480V House Service 6.1.4 Site A ea er e c Loss of all emergency bus AC power for >15 min.

NUNRC IC:

Loss of all offsite power and loss of all onsite AC power to essential busses with reactor coolant >212 F.

FPB Loss/Potential Loss:

N/A Node Applicability:

Power Operation, Hot Shutdown Basis:

Loss of all AC power compromises all plant safety systems requiring electric power. This EAL is indicated by:

Loss of power to T-101N and T-101S, and T-10 backfed through T-1 or T-2 AND failure of both DGs to power any emergency buses AND failure to restore power to PB102 or PB103 in g 15 min.

Prolonged loss of all AC power will cause core uncovery and loss of containment integrity, thus this event can escalate to a General Emergency. The time duration selected, 15 minutes, excludes transient or momentary power losses.

November 1996 Page 60 EPMP-EPP-0101 Rev 01

~TT II I I TC 6.1.4 (Cont)

PEG Reference(s):

SS1.1 Basis Reference(s):

1. Nl-OP-45, Emergency Diesel Generators

2. Nl-OP-30 4.16 Kv, 600V, and 480V House Service

3. Nl-SOP-18, Station Blackout

5. I. I Loss of all emergency bus AC power AND either:

Power restoration to any emergency bus is not likely in < 4 hrs OR RPV water level cannot be restored and maintained > -84 in. (TAf)

NUNARC IC:

Prolonged loss of all offsite power and prolonged loss of all onsite AC power with reactor coolant >212 'F.

FPB Loss/Potential Loss:

N/A Node Applicability:

Power Operation, Hot Shutdown Basis:

Loss of all AC power compromises all plant safety systems requiring electric power. Prolonged loss of all AC power will lead to loss of fuel clad, RCS, and containment. Although this EAL may be viewed as redundant to the RPV Water Level EALs, its inclusion is necessary to better assure timely recognition and emergency response.

This EAL is specified to assure that in the unlikely event of prolonged station blackout, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency occurs as early as is appropriate, based on a reasonable assessment of the event trajectory.

The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of the situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing public protective actions.

'I November 1996 Page 61 EPHP-EPP-0101 Rev Ol

~Tl H ENT (C 6.1.5 (Cont)

In addition, under these conditions, fission product barrier monitoring capability may be degraded. Although to predict when power can be restored, the Emergency Director should it may be difficult declare a General Emergency based on two major considerations:

l. Are there any present indications that core cooling is already degraded to the point that Loss or Potential Loss of fission product barriers is imminent?

2. If there are no present indications of such core cooling degradation, how likely is it that power can be restored in time to assure that a loss of two barriers with a potential loss of the third barrier can be prevented'hus, indication of continuing core cooling degradation must be based on fission product barrier monitoring with particular emphasis on Emergency Director judgment as it relates to imminent loss or potential loss of fission product bar riers and degraded ability to monitor fission product barriers.

The time to restore AC power is based on site blackout coping analysis performed in conformance. with 10CFR50.63 and Regulatory Guide 1.155, "Stat'ion Blackout", with appropriate allowance for offsite emergency response.

The terminology of "cannot be restored and maintained" is intended to be consistent with the interpretation that:

"The value of the identified parameter(s) is/is not able to be returned to above/below specified limits. This determination includes making an evaluation that considers both current and future system performance in relation to the current value and trend of the parameter(s). Neither implies that the parameter must actually exceed the limit before the classification is made nor that the classification must be made before the limit is reached. Does not imply any specific time interval but does not permit prolonged operation beyond a limit without making the specified classification."

This definition would require the emergency classification be made prior to water level dropping below TAF if, based on an evaluation of the current trend of RPV water level and in consideration of current and future injection system performance, that RPV water level will not likely be restored and maintained above TAF. This definition, however, also provides the latitude, based on that same evaluation, not to declare the SAE for those situations in which the RPV water level transiently drops below TAF in the process of RPV water level restoration.

November 1996 Page 62 EPHP-EPP-0101 Rev 01

~IIA I III I (I 6.1.5 (Cont)

PEG Reference(s):

SG1.1 Basis Reference(s):

1. Nl-OP-45, Emergency Diesel Generators

2. Nl-OP-30 4.16 Kv, 600V, and 480V House Service

3. Nl-SOP-18, Station Blackout, pg. 1

4. Nl-ODP-PR0-0302, EOP Technical Bases 6.2 Loss of DC Power Sources 6.2.2 ~EI I

< 106 vdc on battery board 11 and 12 for >15 min.

NUMARC IC:

Unplanned loss of required DC power during cold shutdown or refueling mode for greater than 15 minutes.

FPB Loss/Potential Loss:

N/A Mode Applicability:

Cold Shutdown, Refuel Basis:

The purpose of this EAL is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations. This EAL is intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss.

The bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value incorporates a margin of at least 15 minutes of operation before the onset of inability to operate loads.

November 1996 Page 63 EPHP-EPP-0101 Rev 01

S IIII N 6.2.1 (Cont)

PEG Reference(s):

SU7.1 Basis Reference(s):

1. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Basis for articles 3.6.3 and 4.6.3

2. Nl-OP-47A, 125 vdc Power System 6.2.2 Site rea Emer enc

< 106 vdc on battery board 11 and 12 for > 15 min.

NUMARC IC:

Loss of all vital DC power with reactor coolant > 212'F.

FPB Loss/Potential Loss:

N/A Mode Applicability:

Power Oper ation, Hot Shutdown Basis:

Loss of all DC power compromises ability to monitot and control plant safety functions. Prolonged loss of all DC power will cause core uncovering and loss of containment integrity when there is significant decay heat and sensible heat in the reactor system. Escalation to a General Emergency would occur by other EAL categories. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

The bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value incorporates a margin of at least 15 minutes of operation before the onset of inability to operate loads.

PEG Reference(s):

SS3.1 Basis Reference(s):

1. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Basis for articles 3.6.3 and 4.6.3

2. Nl-OP-47A, 125 vdc Power System November 1996 Page 64 EPHP-EPP-0101 Rev Ol

E IIEET 7.0 U PNE F LURES Numerous plant system related equipment failure events which warrant emergency classification, based upon their potential to pose actual or potential threats to plant safety, have been identified in this category.

The events of this category have been grouped into the following event types:

Technical S ecifications: Only one EAL falls under this event type related to the failure of the plant to be brought to the required plant operating condition required by technical specifications.

~ S stem Failures or Control Room Evacuation: This category includes events which are indicative of losses of operability of safety systems such as ECCS, isolation functions, Control Room habitability or cold and hot shutdown capabilities.

oss of Indication Alarm or Communication Ca abilit : Certain events which degrade the plant operators ability to effectively assess plant conditions or communicate with essential personnel within or external to the plant warrant emergency classification.

Under this event type are losses of annunciators and/or communication equipment.

Technical S ecifications 7.1.1 Unusual Event Plant is not brought to required operating mode within Technical Specifications LCO Action Statement Time NUNRC IC:

Inability to reach required shutdown within Technical Specification Limits.

FPB Loss/Potential Loss:

N/A Node Applicability:

Power Operation, Hot Shutdown November 1996 Page 65 EPHP-EPP-0101 Rev 01

ATTACHMENT I (Cont) 7.1. 1 (Cont)

Basis:

Limiting Conditions of Operation (LCOs) require the plant to be brought to a required shutdown mode when the Technical Specification required configuration cannot be restored. Depending on the circumstances, this may or may not be an emergency or precursor to a more severe condition. In any case, the initiation of plant shutdown required by the site Technical Specification requires a one hour report under 10CFR50.72 (b) non-emergency events. The plant is within its safety envelope when being shut down within the allowable action statement time in the Technical Specifications. An immediate Notification of an Unusual Event is required when the plant is not brought to the required operating mode within the allowable action statement time in the Technical Specifications. Declaration of an Unusual Event is based on the time at which the LCO-specified action statement time period elapses under the site Technical Specifications and is.not related to how long a condition may have existed. Other required Technical Specification shutdowns that involve precur sors to more serious events are addressed by other EALs.

PEG Reference(s):

SU2.1 Basis Reference(s):

1. Radiological Technical Specifications, Appendix A to Facility Operating License No. DPR-63, article 3.0.1 7.2 S stem Fai ures or Contro Room Evacuation 7.2.l ~u Report of main turbine failure resulting in casing penetration or damage to turbine seals or generator seals NUNRC IC:

Natural and destructive phenomena affecting the protected area.

FPB Loss/Potential Loss:

N/A Node Applicability:

Power Operation, Hot Shutdown November 1996 Page 66 EPMP-EPP-0101 Rev Ol

ETTA IIIIENT (C 7.2.1 (Cont)

Basis:

This EAL is intended to address main turbine rotating component failures of sufficient magnitude to cause observable damage to the turbine casing or to the seals of the turbine generator. Of major concern is the potential for leakage of combustible fluids lubricating oils) and gases (hydrogen

~

g cooling) to the plant environs.

ctual 'fires and flammable gas build up are appropriately classified through other EALs. This EAL is consistent with the definition of an Unusual Event while maintaining the anticipatory nature desired and recognizing the risk to non-safety related equipment.

PEG Reference(s):

HU1.6 Basis Reference(s):

None 7.2.2 ~1ert Entry into Nl-SOP-9.1, "Control Room evacuation" NUNARC IC:

Control room evacuation has been initiated.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

With the Control Room evacuated, additional support, monitoring and direction through the Technical Support Center and/or other Emergency Operations Facility is necessary. 1nability to establish plant control from outside the Control Room will escalate this event to a Site Area Emergency.

PEG Reference(s):

HA5.1 Basis Reference(s):

1. Nl-SOP-9.1, Control Room Evacuation November 1996 Page 67 EPNP-EPP-0101 Rev Ol

~TI H T I 7.2.3 alert Reactor coolant temperature cannot be maintained < 212 F NUNRC IC:

Inability to maintain plant in cold shutdown.

FPB Loss/Potential Loss:

N/A Node Applicability:

Cold Shutdown, Refuel Basis:

This EAL addresses complete loss of functions required for core cooling during refueling and cold shutdown modes. Escalation to Site Area Emergency or General Emergency would be through other EALs.

A reactor coolant temperature increase that approaches or exceeds the cold shutdown technical specification limit warrants declaration of an Alert irrespective of the availability of technical specification required functions to maintain cold shutdown. The concern of this EAL is the loss of ability to maintain the plant in cold shutdown which is defined by reactor coolant temperature and not the operability of equipment which supports removal of heat from the reactor.

PEG Reference(s):

SA3. 1 Basis Reference(s):

1. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Amendment 99, Article l.l.a 7.2.4 Site Area Emer enc Entry into Nl-SOP-9.1, "Control Room Evacuation".

AND Plant control cannot be established per Nl-SOP-9.1, "Control Room Evacuation" in < 15 min.

November 1996 Page 68 EPMP-EPP-0101 Rev Ol

~TTA IIII NT 7.2.4 (Cont)

NUNRC IC:

Control room evacuation has been initiated and plant control cannot be established.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This EAL indicates that expeditious transfer of safety systems has not occurred but fission product barrier damage may not yet be indicated.

The time interval for transfer is based on analysis or assessments as to how quickly control must be reestablished without core uncovering and/or core damage. In cold shutdown and refueling modes, operator concern is directed toward maintaining core cooling such as is discussed in Generic Letter 88-17, "Loss of Oecay Heat Removal." In power operation, hot standby, and hot shutdown modes, operator concern is primarily directed toward monitoring and controlling plant parameters dictated by the EOPs and thereby assuring fission product barrier integrity.

PEG Reference(s):

HS2.1 Basis Reference(s):

1. Generic Letter 88-17, Loss of Oecay Heat Removal"

2. Nl-SOP-18, Station Blackout

3. Nl-SOP-9.1, Control Room Evacuation 7.3 Loss of Indications Alarm Communication Ca abilit 7.3.1 Unusual Event Unplanned loss of all annunciators or indicators on all panels L, K, H, F, G for > 15 min.

ANO Increased surveillance is required for safe plant oper ation NVNARC IC:

Unplanned loss of most or all safety system annunciation or indication in the control room for greater than 15 minutes.

November 1996 Page 69 EPMP-EPP-0101 Rev 01

7.3.1 (Cont)

FPB Loss/Potential Loss:

N/A Mode Applicability:

Power Operation, Hot Shutdown Basis:

This EAL recognizes the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment. Recognition of the availability of computer based indication equipment is considered (SPDS, plant computer, etc.).

Unplanned" loss of annunciators or indicators excludes scheduled maintenance and testing activities.

It is not intended that plant personnel perform a detailed count of instrumentation lost but the use of judgment by the Shift Supervisor as the threshold for determining the severity of the plant conditions. This judgment is supported by the specific opinion of the Shift Supervisor that additional operating personnel will be required to provide increased monitoring of system operation to safely operate the plant.

It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptable power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by thei}

specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10CFR50.72. If the shutdown is not in compliance with the Technical Specification'ction, the Unusual Event is based on EAL 7.1.1, Inability to Reach Required Shutdown Within Technical Specification Limits.

Annunciators or indicators for this EAL must include those identified in the Abnormal Operating procedures, in the Emergency Operating Procedures, and in other EALs (e. g., area, process, and/or effluent rad monitors, etc.).

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Due to the limited number of safety systems in operation during cold shutdown, refueling, and defueled modes, this EAL is not applicable during these modes of operation.

November 1996 Page 70 EPHP-EPP-0101 Rev 01

HIILHHNT t lC 7.3. 1 (Cont)

This Unusual Event will be escalated to an Alert if a transient is in progress during the loss of annunciation or indication.

PEG Reference(s):

SU3.1 Basis Reference(s):

1. Nl-OP-42, Process Computer/SPDS 7.3.2 Unusual Event Loss of all communications capability affecting the ability to either:

Perform routine onsite operations OR Notify offsite agencies or personnel NUNRC IC:

Unplanned loss of all onsite or offsite communications capabilities.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

The purpose of this EAL is to recognize a loss of communications capability that either defeats the plant operations staff ability to perform routine tasks necessary for plant operations or the ability to communicate problems with offsite authorities. The loss of offsite communications ability is expected to be significantly more comprehensive than the condition addressed by 10CFR50.72.

November 1996 Page 71 EPHP-EPP-0101 Rev Ol

STTII ENT (C 7.3.2 (Cont)

The onsite communications loss must encompass the loss of all means of routine communications, Table 7.1.

Table 7.1 Commu icat'o s S stems

~Sstem Oesite Offsite PBX Gaitronics Portable headsets Station radios ENS REGS UHF radios The offsite communications loss must encompass the loss of all means of communications with offsite authorities, Table 7.1. This EAL is intended to be used only when extraordinary means are being utilized to make communications possible (relaying of information from radio transmissions, individuals being sent to offsite locations, etc.).

PEG Reference(s):

SU6.1 Basis Reference(s):

1. Nl-OP-51, Communications System 7.3.3 A1ert Unplanned loss of all annunciators ot indicators on all panels L, K, H, F, G for > 15 min.

AND Increased surveillance is required for safe plant operation AND either:

Plant transient in progress OR plant computer and SPDS are unavailable NVSLRC IC:

Unplanned loss of most or all safety system annunciation or indication in control room with either (1) a significant transient in progress, or (2) compensatory non-alarming indicators are unavailable.

FPB Loss/Potential Loss:

N/A November 1996 Page 72 EPHP-EPP-0101 Rev 01

SIIA HNEN 1 7.3.3 (Cont)

Node Applicability:

Power Operation, Hot Shutdown Basis:

This EAL recognizes the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment during a transient. Recognition of the availability of computer based indication equipment is considered (SPDS, plant computer, etc.).

Unplanned" loss of annunciators or indicators does not include scheduled maintenance and testing activities.

It is not intended that plant personnel perform a detailed count of the instrumentation lost but the use of the value as a judgment by the shift supervisor as the threshold for determining the severity of the plant conditions. This judgment is supported by the specific opinion of the Shift Supervisor that additional operating personnel will be required to provide increased monitoring of system operation to safely operate the plant.

It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptable power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant 'conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10CFR50.72.

Annunciators or indicators for this EAL includes those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures, and in other EALs (e. g., area, process, and/or effluent rad monitors, etc.).

"Transient" includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25X thermal power change, ECCS injections, or thermal power oscillations of lOX or greater.

If both a major portion of the annunciation system and all computer monitoring are unavailable to the extent that the additional operating personnel are required to monitor indications, the Alert is required.

November 1996 Page 73 EPMP-EPP-0101 Rev 01

II ENT 7.3.3 (Cont)

Due to the limited number of safety systems in operation during cold shutdown, refueling and defueled modes, no EAl is indicated during these modes of operation.

This Alert will be escalated to a Site Area Emergency if the operating crew cannot monitor the transient in progress.

PEG Reference(s):

SA4.1 Basis Reference(s):

1. NI-OP-42, Process Computer/SPDS 7.3.4 Site Area Emer enc Loss of all annunciators or indicators on all panels L, K, H, F, G AND Plant computer and SPDS are unavailable AND Indications to monitor all RPV and primary containment EOP parameters are lost AND Plant transient is in progress NUNRC IC:

Inability to monitor a significant transient in progress.

FPB Loss/Potential Loss:

N/A Node Applicability:

Power Operation, Hot Shutdown Basis:

This EAL recognizes the inability of the Control Room staff to monitor the plant response to a transient. A Site Area Emergency is considered to exist if the Control Room staff cannot monitor safety functions needed for protection of the public.

Annunciators for this EAL should be limited to include those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures, and in other EALs (e. g., rad monitors, etc.).

November 1996 Page 74 EPHP-EPP-0101 Rev,01

~TT II NT Tg 7.3.4 (Cont)

Transient" includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25K thermal power change, ECCS injections, or thermal power oscillations of 10X or greater.

Indications needed to monitor safety functions necessary for protection of the public must include Control Room indications, computer generated indications and dedicated annunciation capability.

The specific indications should be those used to determine such functions as the ability to shut down the reactor, maintain the core cooled and in a eoolable geometry, to remove heat from the core, to maintain the reactor coolant system intact, and to maintain containment intact.

Planned" actions are excluded from this EAL since the loss of instrumentation of this magnitude is of such significance during a transient that the cause of the loss is not an ameliorating factor.

PEG Reference(s):

SS6.1 Basis Reference(s):

1. Nl-OP-42, Process Computer/SPDS

2. Nl-ODP-PR0-0302, EOP Technical Bases, 8.0 Q~ZDS Hazards are those non-plant system related events which can directly or indirectly impact plant operation or reactor plant and personnel safety.

The events of this category have been grouped into the following types:

~ Securit Threats: This category includes unauthorized entry attempts into the Protected Area as well as bomb threats and sabotage attempts. Also addressed are actual security compromises threatening loss of physical control of the plant.

~NT I:

personnel and Tt reactor safety.

d INITI tt d t Appropriate for classification are fires within the site Protected Area or which may affect operability of vital equipment.

November 1996 Page 75 EPHP-EPP-0101 Rev 01

(Cont) 8.0 (Cont) occurring t: h -

events which d t th - t lit can cause damage to plant facilities such E t h h t, as aircraft crashes, missile impacts, toxic or flammable gas leaks or explosions from whatever source.

tht tornadoes which have potential to cause damage to plant structures or equipment significant or plant safety.

enough k

to threaten personnel 8.1 Securit reats 8.1.1 Unusual Event Bomb device or other indication of attempted sabotage discovered within plant Protected Area OR Any security event which represents a potential degradation in the level of safety of the plant.

NUNRC IC:

Confirmed security event which indicates a potential degradation in the level of safety of the plant.

FPB Loss/Po'tential Loss'.

N/A Node Applicability:

All Basis:

This EAL is based on the Nine Nile Point Nuclear Station Physical Security and Safeguards Contingency Plans. Security events which do not represent at least a potential degradation in the level of safety of the plant, are reported under'0CFR73.71 or in some cases under 10CFR50.72.

The plant Protected Area boundary is within the security isolation zone and is defined in the security plan. Bomb devices discovered within the plant vital area would result in EAL escalation.

PEG Reference(s):

HU4. 1 HU4.2 Basis Reference(s):

1. Nine Hile Point Nuclear Station Physical Security and Safeguards Contingency Plans November 1996 Page 76 EPMP-EPP-0101

~TT tC t) 8.1.2 alert Intrusion into plant Protected Area by an adversary OR Any security event which represents an actual substantial degradation of the level of safety of the plant.

NUNARC IC:

Security event in a plant protected area.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This class of security events represents an escalated threat to plant safety above that contained in the Unusual Event. For the purposes of this EAL, the intrusion by unauthorized personnel inside the Protected Area boundary can be considered a significant security threat.

Intrusion into a vital area by unauthorized personnel will escalate this event to a Site Area Emergency.

NMP-1 and NMP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also see SN Drawing No. 12187-SK-032483-25, Issue No.

1, Site Facilities Layout Status as of 8/1/89.

PEG Reference(s):

HA4.1 HA4.2 Basis Reference(s):

1. Nine Mile Point Nuclear Station Physical Security and Safeguards Contingency Plans

2. SEW Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89 November 1996 Page 77 EPMP-EPP-0101 Rev Ol

STIA IIIIINT (C 8.1.3 Site rea r enc Intrusion into a plant security vital area by an adversary OR Any security event which represents actual or likely failures of plant systems needed to protect the public.

NUNRC IC:

Security event in a plant vital area.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This class of security events represents an escalated threat to plant safety above that contained in the Alert in that unauthorized personnel have progressed from the Protected Area to the vital area.

PEG Reference(s):

HS1.1 HS1.2 Basis Reference(s):

I. Nine Mile Point Nuclear Station Physical Security and Safeguards Contingency Plans 8.1.4 General Emer enc Security event which results in either:

Loss of plant control from the Control Room OR Loss of remote shutdown capability NUNRC IC:

Security event resulting in loss of ability to reach and maintain cold shutdown.

FPB Loss/Potential Loss:

N/A November 1996 Page 78 EPMP-EPP-0101 Rev Ol

~TT H ENT (C 8.1.4 (Cont)

Node Applicability:

All Basis:

This EAL encompasses conditions under which unauthorized personnel have taken physical control of vital areas required to reach and maintain safe shutdown.

PEG Reference(s):

HGl. 1 HG1.2 Basis Reference(s):

None 8.2 Fire or Ex losio 8.2.1 Unusual Event Confirmed fire in or contiguous to any plant area, Table 5 or Table 6, not extinguished in < 15 min. of Control Room notification Table 5 Plant Areas

~ RadWaste Solidification and Storage Bldg.

~ Security West Bldg.

Table 6 Plant Vita A eas Reactor Building ControT Room Diesel Generator Engine and Board Rooms Battery Rooms Battery Board Rooms Cable Spreading Room Central Alarm Station Secondary Alarm Station Security Uninterruptible Power System Room Telephone Rooms Hain Steam Isolation Valve Room NUMARC IC:

Fire within protected area boundary not extinguished within 15 minutes of detection.

November 1996 Page 79 EPHP-EPP-0101 Rev 01

ACH N (Cont) 8.2.1 (Cont)

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

The purpose of this EAL is to address the magnitude and extent of fires that may be potentially significant precursors to damage to safety systems. This excludes such items as fires within administration buildings, waste-basket fires, and other small fires of no safety consequence.

PEG Reference(s):

HU2.1 Basis Reference(s):

~,

1. Nine Mile Point Nuclear Station Physical Security and Safeguards Contingency Plans

2. NUREG 0737, Section II.B.2-2 8.2.2 gert Fire or explosion in any plant area, which results in damage to plant equipment or structures needed for safe plant operation, Table 5 or Table 6.

Table 5 Plant Areas RadWaste Solidification and Storage Bldg.

Security West Bldg.

Table 6 Plant Vital Areas Reactor Building Control Room Diesel Generator Engine and Board Rooms Battery Rooms Battery Board Rooms Cable Spreading Room Central Alarm Station Secondary Alarm Station Security Uninterruptible Power System Room Telephone Rooms Main Steam Isolation Valve Room November 1996 Page 80 EPMP-EPP-0101 Rev 01

8.2.2 (Cont)

NUNRC IC:

Fire or explosion affecting the operability of plant safety systems required to establish or maintain safe shutdown.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

The listed areas contain functions and systems required for the safe shutdown of the plant. The NHP-1 safe shutdown analysis was consulted for equipment and plant areas required for the applicable mode.

With regard to explosions, only those explosions of sufficient force to damage permanent structures or equipment required for safe operation within the identified plant areas should be considered. As used here, an explosion is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts significant energy to nearby structures and materials. No -

attempt is made in this EAL to assess the actual magnitude of the damage. The declaration of an Alert and the activation of the TSC will provide the Emergency Director with the resources needed to perform damage assessments. The Emergency Director also needs to consider any security aspects of the explosions.

PEG Reference(s):

HA2.1 Basis Reference(s):

1. Nl-SOP-9, Fire In Plant

2. Nine Mile Point Nuclear Station FSAR, Section 10

3. NUREG 0737, Section II.B.2-2 November 1996 Page 81 EPHP-EPP-0101 Rev 01

A~IT IIIIENT (C 8.3 8.3.l U~t a ade ve ts Vehicle crash into or projectile which impacts plant structures or systems within Protected Area boundary NUNARC IC:

Natural and destructive phenomena affecting the protected area.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

The Protected Area boundary is within the security isolation zone and is defined in the site security plan. NMP-1 and NMP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also, refer to SEW Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89.

This EAL addresses such items as plane, helicopter, train, car, truck, or barge crash, or impact of other projectiles that may potentially damage plant structures containing functions and systems required for safe shutdown of the plant. If the crash is confirmed to affect a plant vital area, the event may be escalated to Alert.

PE6 Reference(s):

HU1.4 Basis Reference(s):

1. USAR Figure 1.2-1

2. SEW Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89 November 1996 Page 82 EPMP-EPP-0101 Rev 01

ATTACHMENT 1 (Cont) 8.3.2 Unusual ve t Report by plant personnel of an explosion within Protected Area boundary resulting in visible damage to permanent structures or equipment NVNARC IC:

Natural'nd destructive phenomena affecting the protected area.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

The Protected Area boundary is within the security isolation zone and is defined in the site security plan. NMP-1 and NMP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also, refer to S&W Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89.

For this EAL, only those explosions of sufficient force to damage permanent structures or equipment within the Protected Area should be considered. As used here, an explosion is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts significant energy to near by structures and materials. No. attempt is made in this EAL to assess the actual magnitude of the damage. The occurrence of the explosion with reports of evidence of damage (e. g., deformation, scorching) is sufficient for declaration. The Emergency Director also needs to consider any security aspects of the explosion.

PEG Reference(s):

HU1.5 Basis Reference(s):

l. USAR Figure 1.2-1

2. SSW Drawing No; 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89 November 1996 Page 83 EPMP-EPP-0101 Rev Ol

tt 3.3.3 ~33 ~TTHHH 3 3 Report or detection of a release of toxic or flammable gases that could enter or have entered within the Protected Area boundary in amounts that could affect the health of plant personnel or safe plant operation OR Report by local, county or state officials for potential evacuation of site personnel based on offsite event NUNRC IC:

Release of toxic or flammable gases deemed detrimental to safe operation of the plant.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This EAL is based on releases in concentrations within the site boundary that will affect the health of plant personnel or affecting the safe operation of the plant with the plant being within the evacuation area of an offsite event (i. e., tanker truck accident releasing toxic gases, etc.). The evacuation area is as determined from the DOT Evacuation Tables for Selected Hazardous Materials, in the DOT Emergency Response Guide for Hazardous Materials.

NMP-1 and NMP-2 share no common safety systems, but their - respective Protected Area boundaries share common borders in some places.

Therefore it is possible that a toxic or flammable gas incident happening on one site could affect the other site.

Should an explosion occur within a specified plant area,. an Alert would be declared based on EAL 8.2.2 PEG Reference(s):

HU3.1 HU3.2 Basis Reference(s):

None November 1996 Page 84 EPMP-EPP-0101 Rev Ol

~ITACH IIT I 8.3.4 A1eet Vehicle crash or projectile impact which precludes personnel access to or damages equipment in plant vital areas, Table 6 Table 6 lant Vital Areas Reactor Building Control Room Diesel Generator Engine and Board Rooms Battery Rooms Battery Board Rooms Cable Spreading Room Central Alarm Station Secondary Alarm Station Security Uninterruptible Power System Room Telephone Rooms Main Steam Isolation Valve Room NUMARC IC:

Natural and destructive phenomena affecting the'plant vital area.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This EAL addresses events that may have resulted in a plant vital area being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage.

NMP-1 and NMP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also see SN Drawing No. 12187-SK-032483-25, Issue No.

1, Site Facilities Layout Status as of 8/1/89.

This EAL addresses such items as plane, helicopter, train, car, or truck crash, or impact of other projectiles into a plant vital area.

November 1996 Page 85 EPMP-EPP-0101 Rev 01

~IIBT 8.3.4 (Cont)

PEG Reference(s):

HA1.5 Basis Refer ence(s):

l. USAR Figure 1.2-1

2. SN Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89

3. NUREG 0737, Section II.B.2-2 8.3.5 Alert Confirmed report or detection of toxic or flammable gases within a plant vital area, Table 6, in concentrations that will be life threatening to plant personnel or preclude access to equipment needed for safe plant operation Table 6 Pl nt 'tal Areas Reactor Building Control Room Diesel Generator Engine and Board Rooms Battery Rooms Battery Board Rooms Cable Spreading Room Central Alarm Station Secondary Alarm Station Security Uninterruptible Power System Room Telephone Rooms Hain Steam Isolation Valve Room NUMARC IC:

Release of toxic or flammable gases within a facility structure which jeopardizes operation of systems required to maintain safe operations or to establish or maintain cold shutdown.

FPB Loss/Potential Loss:

N/A Node Applicability:

All November 1996 Page 86 EPMP-EPP-0101 Rev 01

SITA HNEIIT I (C 8.3.5 (Cont) 0 Basis:

This EAL is based on gases that have entered a plant structure precluding access to equipment necessary for the safe operation of the plant. This EAL applies to buildings and areas contiguous to plant vital areas or other significant buildings or areas. The intent of this EAL is not to include buildings (i. e., warehouses) or other areas that are not contiguous or immediately adjacent to plant vital areas. It is appropriate that increased monitoring be done to ascertain whether consequential damage has occurred.

PEG Reference(s):

HA3. 1 HA3.2 Basis Reference(s):

l. USAR Figure III-6, Station Floor Plan Elevation 281'-0" and 291'-0" 8.4 Natural Events S.Ll U~1E Earthquake felt inplant based upon a consensus of Control Room Operators on duty.

AND either:

NHP-'I seismic instrumentation actuated OR Confirmation of earthquake received on NNP-2 or JAFNPP seismic instrumentation NUNRC IC:

Natural and destructive phenomena affecting the protected area.

FPB Loss/Potential Loss:

N/A Node Applicability:

E All November 1996 Page 87 EPHP-EPP-0101 Rev 01

~IT HM NT 1 8.4. 1 (Cont)

Basis:

NMP-1 seismic instrumentation actuates at 0.01 g.

Damage to some portions of the site may occur but it should not affect ability of safety functions to operate. Methods of detection can be based on instrumentation validated by a reliable source, operator assessment, or indication received from NMP-2 or JAFNPP instrumentation. As defined in the EPRI-sponsored "Guidelines for Nuclear Plant Response to an Earthquake", dated October 1989, a "felt earthquake" is:

An earthquake of sufficient intensity such that: (a) the inventory ground motion is felt at the nuclear plant site and recognized as an earthquake based on a consensus of Control Room operators on duty at the time, and (b) for plants with operable seismic instrumentation, the seismic switches of the plant are activated. For most plants with seismic instrumentation , the seismic switches are set at an acceleration of about 0.01 g" PEG Reference(s):

HU1.1 Basis Reference(s):

1. Nl-ARP-H2 annunciator H2-1-6

2. Nl-SOP-11, Earthquake

3. EPRI document, "Guidelines for Nuclear Plant Response to an Earthquake" 8.4.2 U~E Report by plant personnel of tornado striking within plant Protected Area boundary NUNARC IC:

Natural and destructive phenomena affecting the protected area.

FPB Loss/Potential Loss:

N/A Node Applicability:

All November 1996 Page 88 EPMP-EPP-0101 Rev Ol

SIT CIIMENT I tC t) 8.4.2 (Cont)

Basis:

This EAL is based on the assumption that a tornado striking (touching down) within the protected boundary may have potentially damaged plant structures containing functions or systems required for safe shutdown of the plant. If such damage is confirmed visually or by other in-plant indications, the event may be escalated to Alert.

NAP-I and NHP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also see S&W Drawing No. 12187-SK-032483-25, Issue No.

1, Site Facilities Layout Status as of 8/1/89.

PEG Reference(s):

HU1.2 Basis Reference(s):

1. USAR Figure 1.2-1

2. SstW Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89 8.4.3 Unusua ve t Lake water level > 248 'ft OR forebay water level < 238.8 ft NUNARC IC:

Natural and destructive phenomena affecting the protected area.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This covers high and low lake water level conditions that could be precursors of more serious events. The high lake level is based upon the maximum attainable uncontrolled lake water level. The low level is based on intake forebay level and corresponds to the minimum intake water level for operability of Emergency Service Water, Emergency Diesel Generator cooling water, Containment Spray Raw Water and Diesel and Electric Fire Pump.

November 1996 Page 89 EPHP-EPP-0101 Rev Ol

HTl Hll Tl CC" 8.4.3 (Cont)

PEG Reference(s):

HUl. 7 Basis Reference(s):

1. Nl-ARP-H2, Annunciator H2-1-3

2. Nl-SOP-7, Service Water Failure/Low Intake Level

3. DER 1-92-0-0489 8.4.4 Aler t Earthquake felt in plant based upon a consensus of Control Room Operators on duty AND NAP-I seismic instrumentation indicates > O.ll g NUMARC IC:

Natural and destructive phenomena affecting the plant vital area.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This EAL addresses events that may have resulted in a plant vital area being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage.

This EAL is based on the FSAR design operating bases earthquake of O.ll g. Seismic events of this magnitude can cause damage to plant safety functions.

PEG Reference(s):

HA1.1 Basis Reference(s):

1. Nl-ARP-H2, annunciator H2-1-6

. 2. Nl-SOP-ll, Earthquake November 1996 Page 90 EPMP-EPP-0101 Rev Ol

~AITA 1 111 1 (1 t) 5.4.5 ~1ert Sustained winds > 125 mph OR Tornado strikes a plant vital area, Table 6 Table 6 tt Vtt1 A Reactor Building Control Room Diesel Generator Engine and Board Rooms Battery Rooms Battery Board Rooms Cable Spreading Room Central Alarm Station Secondary Alarm Station Security Uninterruptible Power System Room Telephone Rooms Hain Steam Isolation Valve Room NUMARC IC:

Natural and destructive phenomena affecting the plant vital area.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This EAL addresses events that may have resulted in a plant vital area being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage.

This EAL is based on the FSAR design bases of 125 mph. Wind loads of this magnitude can cause damage to safety functions.

NHP-1 and NHP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also see S8W Drawing No. 12187-SK-032483-25, Issue No.

1, Site Facilities Layout Status as of 8/1/89.

PEG Reference(s):

HA1.2 November 1996 Page 91 EPHP-EPP-0101 Rev Ol

CHM N (Cont) 8.4.5 (Cont)

Basis Reference(s):

1. FSAR Section VI.C.l.l, Wind and Snow Loadings, 6/91

2. Nl-SOP-10, High Winds

3. USAR Figure 1.2-1

4. SEW Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89

5. NUREG 0737, Section II.B.2-2 8.4.6 alert Any natural event which results in a report of visible structural damage or assessment by Control Room personnel of actual damage to equipment needed for safe plant operation, Table 6.

Table 6 Plant V'ta A} eas Reactor Building Control Room Diesel Generator Engine and Board Rooms Battery Rooms Battery Board Rooms Cable Spreading Room Central Alarm Station Secondary Alarm Station Security Uninterruptible Power System Room Telephone Rooms Hain Steam Isolation Valve Room NUNARC IC:

Natural and destructive phenomena affecting the plant vital area.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This EAL addresses events that may have resulted in a plant vital area being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage.

November 1996 Page 92 EPHP-EPP-0101 Rev Ol

8.4.6 (Cont)

This EAL specifies areas in which structures containing systems and functions required for safe shutdown of the plant are located.

PEG Reference(s):

HA1.3 Basis Reference(s):

1. USAR Figure III-6, Station Floor Plan Elevation 281'-0" and 291'-0"

2. NUREG 0737, Section II.B.2-2 8.4.7 Alert Lake water level > 254 ft OR forebay water level < 236 ft NUNARC IC:

Natural and destructive phenomena affecting the plant vital area.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This EAL addresses events that may have resulted in a plant vital area being subjected to levels beyond design limits, and thus damage may be assumed to have occurred to plant safety systems. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage.

This EAL covers high and low lake water level conditions that exceed levels which threaten vital equipment. The high lake level is based upon the maximum probable flood. level. The low forebay water level corresponds to the minimum level before damage may occur to the service water pumps.

November 1996 Page 93 EPMP-EPP-0101 Rev,01

~TT TNT 8.4.7 (Cont)

PEG Reference(s):

HA1.7 Basis Reference(s):

1. NI-SOP-7, Service Water Failure/Low Intake Level

2. DER 1-92-0-0489 9.0 ~OT <~E The EALs defined in categories 1.0 through 8.0 specify the predetermined symptoms or events which are indicative of emergency or potential emergency conditions, and which warrant classification.

While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Shift Supervisor or Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria, based upon their judgment.

N.l.l ~V Any event, as determined by the Shift Supervisor or Emergency Director that could lead to or has led to a potential degradation of the level of safety of the plant.

NUNARC IC:

Emergency Director Judgement FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the Unusual Event emergency class.

November 1996 Page 94 EPMP-EPP-0101 Rev 01

~TTA All T l 9.1.1 (Cont)

From a broad perspective, one area that may warrant Emergency Director judgment is related to likely or actual breakdown of site specific event mitigating actions. Examples to consider include inadequate emergency response procedures, transient response either unexpected or not understood, failure or unavailability of emergency systems during an accident in excess of that assumed in accident analysis, or insufficient availability of equipment and/or support personnel.

Another example to consider would be exceeding a plant safety limit as defined in Technical Specifications.

PEG Reference(s):

HU5.1 Basis Reference(s):

None 9.1.2 Unusual Event Any event, as determined by the Shift Supervisor or Emergency Director, that could lead to or has led to a loss or potential loss of containment. (Attachment 2)

Loss of containment indicators may include a rapid unexplained decrease following initial increase in containment pressure.

NUNARC IC:

N/A FPB Loss/Potential Loss:

Containment Loss/Potential Loss Node Applicability:

Power Operations, Hot Shutdown Basis:

This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the containment barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this EAL as a factor in Emergency Director judgment that the barrier may be considered lost or potentially lost.

November 1996 Page 95 EPHP-EPP-0101 Rev Ol

SIMHHII IIT ( t) 9.1.2 (Cont)

PEG Refers ence(s):

PC6.1 Basis Reference(s):

None 9.1.3 gert Any event, as determined by the Shift Supervisor or Emergency Director, that could cause or has caused actual substantial degradation of the level of safety of the plant.

NUSLRC IC:

Emergency Director Judgement FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the Alert emergency class.

PEG Reference(s):

HA6.1 Basis Reference(s):

None November 1996 Page 96 EPHP-EPP-0101 Rev 01

9.1.4 alert Any event, as determined by the Shift Supervisor or Emergency Director, that could lead or has led to a loss or potential loss of either fuel clad or RCS barrier. (Attachment 2)

NUNRC IC:

N/A FPB Loss/Potential Loss:

Loss or Potential Loss of Either Fuel Clad or RCS Barrier Node Applicability:

Power Operations, Hot Shutdown Basis:

This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the fuel clad or RCS barriers are lost or potentially lost. In addition, the inability to monitor the barriers should also be considered in this EAL as a factor in Emergency Director judgment that the barriers may be considered lost or potentially lost.

PEG Reference(s):

FC5.1 RCS6.1 Basis Reference(s):

None 9.1.5 Site Area Emer enc As determined by the Shift Supervisor or Emergency Director, events are in progress which indicate actual or likely failures of plant systems needed to protect the public. Any releases are not expected to result in exposures which exceed EPA PAGs.

NUNRC IC:

Emergency Director Judgement FPB Loss/Potential Loss:

N/A Node Applicability:

All November 1996 Page 97 EPMP-EPP-0101 Rev 01

9.1.5 (Cont)

Basis:

This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency class description for Site Area Emergency.

PEG Reference(s):

HS3.1 Basis Reference(s):

None 9.1.6 Site ea er enc Any event, as determined by the Shift Supervisor or Emergency Director, that could lead or has led to either:

Loss or potential loss of both fuel clad and RCS barrier, Attachment 2 OR Loss or potential loss of either fuel clad or RCS barrier in conjunction with a loss of containment, Attachment 2 Loss of cohtainment indicators may include a rapid unexplained decrease following initial increase in containment pressure NUNARC IC:

N/A FPB Loss/Potential Loss:

Loss or potential loss of both fuel clad and RCS barrier OR Loss or potential loss of either fuel clad or RCS barrier in conjunctions with a loss of containment Node Applicability:

Power Operations, Hot Shutdown November 1996 Page 98 EPMP-EPP-0101 Rev 01

~(( Hll (( ( ((

9.1.6 (Cont)

Basis:

This EAL addresses unanticipated conditions affecting fission product barriers which are not addressed explicitly elsewhere. Declaration of an emergency is warranted because conditions exist which are believed by .the Emergency Director. to fall under the emergency class description for Site Area Emergency.

Rapid unexplained loss of pressure (i. e., not attributable to drywell spray or condensation effects) following an initial pressure increase may indicate a loss of containment integrity.

PEG Reference(s):

FC5.1 RCS6.1 PC6.1 PC1.1 PC1.2 Basis Reference(s):

None 9.1.7

~ ~ General Emer enc As determined by the Shift Supervisor or Emergency Director, events are in progress which indicate actual or imminent core damage and the potential for a large release of radioactive material in excess of EPA PAGs outside the site boundary.

NUNRC IC: .

Director Judgement

'mergency FPB Loss/Potential Loss:

N/A Node Applicability:

All November 1996 Page 99 EPHP-EPP-0101 Rev Ol

~H" T 9.1.7 (Cont)

Basis:

This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to be consistent with the General Emergency classification description.

Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the site boundary.

PEG Reference(s):

HG2.1 Basis Reference(s):

9.l.S ~t None Any event, as determined by the Shift Supervisor or Emergency Director, that could lead or has led to a loss of any two fission product barriers and loss or potential loss of the third (Attachment 2).

Loss of containment indicators may include a rapid unexplained decrease following initial increase in containment pressure:

NUNRC IC:

N/A FPB Loss/Potential Loss:

Loss of any two fission product barriers and loss or potential loss of the third Node Applicability:

Power Operations, Hot Shutdown Basis:

This EAL addresses unanticipated conditions affecting fission. product barriers which are not addressed explicitly elsewhere. Declaration of an emergency is warranted because conditions exist which are believed by the Emergency Director to fall under the emergency class description for the General Emergency class.

November 1996 Page 100 EPMP-EPP-0101 Rev 01

ETTA HIIEN 1 ( t) 9.1.8 (Cont)

Rapid unexplained loss of pressure (i. e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of containment integrity.

PE6 Reference(s):

FC5.1 RCS6.1 PC6.1 PC1.1 PC1.2 Basis Reference(s):

None November 1996 Page 101 EPMP-EPP-0101 Rev 01

ATTACHMENT 2 FISSION PRODUCT BARRIER LOSS 8c POTENTIAL LOSS INDICATORS November 1996 Page 102 EPMP-EPP-0101 Rev, 01

Fission Product Bar rier Loss/Potential Loss Matrix (Those thresholds for which loss or potential is determined to be imminent, classify as'though the threshold(s) has been exceeded)

Fuel Cladding

~P "ti

~ RPV water level cannot be restored and maintained > -84 in. (TAF)

~ Emergency Director Judgment

~oss

~ RPV water level cannot be restored and maintained > -84 in. (TAF)

~ Coolant activity > 300 pCi/gm 1-131 equivalent

~ Valid offgas radiation > 10 x hi-hi alarm

~ Drywell radiation > 3000 R/hr

~ Emergency Director Judgment RCS Potential Loss

~ RCS leakage greater than 50 gpm inside the drywell

~ Primary system is discharging outside PC AND RB area radiation levels are > 8.0 R/hr in two or more areas, Nl-EOP-5

~ Primary system is discharging outside PC AND RB general area temperatures are > 135 F in two or more areas, Nl-EOP-5

~ Emergency Director Judgment

~oss

~ RPV water level cannot be restored and maintained > -84 in. (TAF)

~ Primary containment pressure cannot be maintained < 3.5 psig due to coolant leakage

~ Drywell radiation > 20 R/hr

~ Emergency Director Judgment November 1996 Page 103 EPMP-EPP-0101 Rev 01

Fission Product Barrier Loss/Potential Loss Hatrix (Those thresholds for which loss or potential is determined to be imminent, classify as though the threshold(s) has been exceeded)

Containment Potential Loss

~ Drywell radiation > 4.0E6 R/hr

~ Emergency Director Judgment

~oss

~ Primary containment venting is required due to PCPL

~ Primary containment venting is required due to combustible gas concentrations

~ HSL, EC steam line or RWCU isolation failure resulting in a release pathway outside primary containment

~ Primary system is discharging outside PC AND RB area radiation levels are > 8.0 R/hr in two or more areas, Nl-EOP-5

~ Primary system is discharging outside PC AND

~ RB general area temperatures are > 135'F in two or more areas, Nl-EOP-5

~ Emergency Director Judgment of containment indication may include rapid unexplained decrease

'oss following initial increase in containment pressure November 1996 Page 104 EPHP-EPP-0101 Rev Ol

~ITACII E WORD IST DE INITIONS ctuate To put into operation; to move to action; commonly used to refer to automated, multi-faceted operations. "Actuate ECCS".

~dversar As applied to security EALs, an individual whose intent is to commit sabotage, disrupt Station operations or otherwise commit a crime on station property.

'I de uate Core'ool n Heat removal from the reactor sufficient to prevent rupturing the fuel clad.

lert Events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

va lable The state or condition of being ready and able to be used (placed into operation) to accomplish the stated (or implied) action or function. As applied to a system, this requires the operability of necessary support systems (electrical power supplies, cooling water, lubrication, etc.).

Can Cannot be dete i ed The current value or status of an identified parameter relative to that specified can/cannot be ascertained using all available indications (direct and indirect, singly or in combination).

Can Cannot be mai ta ed above below The value of the identified parameter(s) is/is not able to be kept above

/below specified limits. This determination includes making an evaluation that considers both current and future system performance in relation to the current value and trend of the parameter(s). Neither implies that the parameter must actually exceed the limit before the action is taken nor that the action must be taken before the limit is reached.

November 1996 Page 105 EPHP-EPP-0101 Rev 01

~HENT a a not be estored and maintained above below The value of the identified parameter(s) is/is not able to be returned to above/below specified limits. This determination includes making an evaluation that considers both current and future systems performances in relation to the current value and trend of the parameter(s). Neither implies that the parameter must actually exceed the limit before the classification is made nor that the classification must be made before the limit is reached.

This does not imply any specific time interval but does not permit prolonged operation beyond a limit without taking the specified classification.

As applied to loss of electrical power sources (ex.: Power cannot be restored to any vital bus in < 4 hrs) the specified power source cannot be returned to service within the specified time. This determination includes making an evaluation that considers both current and future restoration capabilities.

This implies that the declaration should be made as soon as the determination is made that the power source cannot be restored within the specified time.

C1cse To position a valve or damper so as to prevent flow of the process fluid.

To make an electrical connection to supply power.

Co at o To validate, through visual observation or physical inspection, that an assumed condition is as expected or required, without taking action to alter the as found" configuration.

Conti uous Being in actual contact; touching along a boundary or at a point g~otyol Take action, as necessary, to maintain the value of a specified parameter within applicable limits; to fix or adjust the time, amount, or rate of; to regulate or restrict.

~Dec ease To become progressively less in size, amount, number, or intensity.

~Dschae e Removal of a fluid/gas from a volume or system.

November 1996 Page 106 EPMP-EPP-0101 Rev Ol

~TTA HN N s t)

~rv~e That component of the BWR primary containment which houses the RPV and associated piping.

/pter To go into.

~stab 1s To perform actions necessary to meet a stated condition. "Establish communication with the Control Room."

Evacuate To remove the contents of; to remove personnel from an area.

Exceeds To go or be beyond a stated or implied limit, measure, or degree.

To have being with respect to understood limitations or conditions.

Fai1ure A state of inability to perform a normal function.

Genera Emer e c Events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.

Logic term which indicates that taking the action prescribed is contingent upon the current existence of the stated condition(s). If conditions do not exist, the prescribed action is not to be taken and the identified execution of 'operator actions must proceed promptly in accordance with subsequent instructions.

November 1996 Page 107 EPHP-EPP-0101 Rev Ol

~cease To become progressively greater in size, amount, number or intensity.

~d~cte To point out or point to; to display the value of a process variable; to be a sign or symbol.

Initiate The act of placing equipment or a system into service, either manually or automatically. Activation of an function or protective feature (i.e. initiate a manual scram).

I ectio The act of forcing a fluid into a volume or vessel.

Ino erable Not able to perform it's intended function

~Itruai o The act of entering without authorization oss Failure of operability or lack of access to.

~aintain Take action, as necessary, to keep the value of the specified parameter within the applicable limits.

aximum Sa e 0 e at arameter The highest value of the identified operating parameter beyond which, required personnel access or continued operation of equipment important to safety cannot be assured.

November 1996 Page 108 EPHP-EPP-0101 Rev Ol

NT( CIIMINT (C I)

JfoOjLtto Observe and evaluate at a frequency sufficient to remain apprised of the value, trend, and rate of change of the specified parameter.

To give notice of or report the occurrence of; to make known to; to inform specified personnel; to advise; to communicate; to contact; to relay.

~0en To position a valve or damper so as to allow flow of the process fluid.

To break an electrical connection which removes a power supply from an electrical device.

To make available for entry or passage by turning back, removing, or clearing away.

~0e ab1e Able to perform it's intended function Perform To carry out an action; to accomplish; to affect; to reach an objective.

r mar Conta nment The airtight volume immediately adjacent to and surrounding the RPV, consisting of the drywell and wetwell in a BWR plant.

Primar S stem The pipes, valves, and other equipment which connect directly to the RPV or reactor coolant system such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

memove To change the location or position of.

November 1996 Page 109 EPHP-EPP-0101 Rev Ol

[Lee o~t To describe as being in a specific state.

jLeeLu~e To demand as necessary or essential.

geateee Take the appropriate action requires to return the value of an identified parameter to within applicable limits.

Rise Describes an increase in a parameter as the result of an operator or automatic'ction.

S~am le To perform an analysis on a specified media to determine its properties.

~Sc an~

To take action to cause shutdown of the reactor by rapidly inserting a control rod or control rods (BWR).

Secondar Co ta nment The airtight volume immediately adjacent to or surrounding the primary containment in a BWR plant.

~Sut dowg To perform operations necessary to cause equipment to cease or suspend operation; to stop. Shut down unnecessary equipment."

~Sut~do As applied to the BWR reactor, subcritical with reactor power below the heating range.

November 1996 Page 110 EPHP-EPP-0101 Rev 01

~IT HIIENT Site ea r enc Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels except near the site boundary.

Sustained Prolonged. Not intermittent or of transitory nature Torus The volume of water in a BWR plant intended to condense steam discharged from a primary system break inside the drywell.

g~asi eret Events of off-normal nature such as; scrams, runbacks involving >25X thermal power changes, ECCS injections or thermal power oscillations of >10X.

Xeiu To de-energize a pump or fan motor; to position a breaker so as to interrupt or prevent the flow of current in the associated circuit; to manually activate a semi-automatic feature.

~Ud An evolution lacking control but is not the result of operator action.

U armed Not as an expected result of deliberate action.

Indicates that the associated prescribed action is to proceed only so long as the identified condition does not exist.

Unusual Eve t Events are in process or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive'material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

November 1996 Page 111 EPMP-EPP-0101 Rev 01

Va d Supported or corroborated on a sound bases.

To open an effluent (exhaust) flowpath from an enclosed volume; to reduce pressure in an enclosed volume.

Ver~i To confirm a condition and take action to establish that condition required. Verify reactor trip."

if ita rea Any plant area which contains vital equipment.

November 1996 Page ll2 EPNP-EPP-0101 Rev Ol

NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT NUCLEAR STATION EMERGENCY PLAN MAINTENANCE PROCEDURE EPMP-EPP-0102 REVISION 02 UNIT 2 EMERGENCY CLASSIFICATION TECHNICAL BASIS TECHNICAL SPECIFICATION REQUIRED Approved by:

K. A. Dahlberg Plant Manager - Unit 2 Date 06/19/98 Effective Date:

Fe LIST OF EFFECTIVE PAGES

~PN . ~h Pacae No. Cha~che No, Pacae No. Chan e No.

Coversheet . 22 47 1 ~ ~ ~ ~ 23 48 24 49 25 50 .

26 . 51 2 ~ ~ 27 52 3 ~ ~ 28 . 53 4 . 29 54 5 30 55 6 . 31 56 7 ~ ~ 32 57 .

8 . 33 58 .

9 34 . 59 10 . 35 60 ll 36 61 12 . 37 62 13 38 . 63 14 . 39 64 .

15 . 40 65 16 . 41 66 17 . 42 67 .

18 . 43 68 19 44 . 69 20 . 45 70 21 46 71 Hay 1998 Page i EPHP-EPP-0102 Rev 02

LIST OF EFFECTIVE PAGES (Cont)

~PII . ~h Pacae No. Chan<he No.

72 . 97 73 98 0 ~ 0 74 . 99 ~ ~ ~

75 . 100 76 ~ ~ ~ ~ 101 77 . 102 .

78 . 103 79 . 104 80 . 105 .

81 106 82 107 83 108 .

84 . 109 85 . 110 86 111 87 . 112 88 . 113 89 114 .

90 . 115 91 116 .

92 . 117 93 . "-118 94 .

95 .

96 .

May,1998 Page ii EPMP-EPP-0102 Rev 02

P TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

2.0 PRIMARY RESPONSIBILITY 3.0 PROCEDURE ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3. 1 Emergency Preparedness Group . . . . . . . . . . .

4.0 DEFINITIONS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

5.0 REFERENCES

AND COMMITMENTS 6.0 RECORD REVIEW AND DISPOSITION . 2 PURPOSE . ~ ~ 3 DISCUSSION 3 ATTACHMENT 1: UNIT 2 EMERGENCY ACTION LEVEL TECHNICAL BASIS 3 1.0 REACTOR FUEL 8 2.0 REACTOR PRESSURE VESSEL (RPV) 24 3.0 PRIMARY CONTAINMENT (PC) 31 4.0 SECONDARY CONTAINMENT (SC) 40

5. 0 RADIOACTIVITY RELEASE . 47 6.0 ELECTRICAL FAILURES . 60

7. 0 E(UI PMENT FAILURES 69 8.0 HAZARDS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ 80 9.0 OTHER ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 100 ATTACHMENT 2: FISSION PRODUCT BARRIER LOSS/POTENTIAL LOSS INDICATOR .. 107 ATTACHMENT 3: WORD LIST/DEFINITIONS 110 May 1998 Page iii EPMP-EPP-0102 Rev 02

1.0 PURPOSE To describe the Technical Basis for the Emergency Action Levels at Unit 2 ~

2.0 PRIMARY RESPONSIBILITY 2.1 Emer enc Pre aredness Grou Monitor/solicit any changes to the Technical Basis of each Emergency Action Level Assess these changes for potential impact on the Emergency Action Level Maintain the Emergency Action Level (EAL) Technical Basis, EPIP-EPP-02, and the Emergency Action Level Matrix/Unit 2.

3.0 PROCEDURE 3.1 Emer enc Pre aredness Grou 3.1.1 Maintain a matrix of Technical Basis references for each Emergency Action Level.

3.1.2 Evaluate each Technical Basis Reference Change for impact on the Affected Emergency Action Level.

3.1,3 Modify EP IP-EPP-02, Emergency Acti on Level Hatri x/Uni t, and Attachment 1 of this procedure, as needed.

4.0 DEFINITIONS See Attachment 3.

5.0 REFERENCES

AND COMMITMENTS 5.1 Licensee Documentation None 5.2 Standards Re ulations and Codes NUMARC NESP-007, Methodology for Development of Emergency Action Levels 5.3 Policies Pro rams and Procedures EPIP-EPP-02, Classification of Emergency Condition at Unit 2.

Hay 1998 Page 1 EPHP-EPP-0102 Rev 02

5.4 Su lemental References Nine Hile Point Unit 2 Plant-Specific EAL Guideline 5.5 Commitments None 6.0 RECORD REVIEM AND DISPOSITION None Hay 1998 Page 2 EPHP-EPP-0102 Rev 02

ATTACHMENT 1 UNIT 2 EMERGENCY ACTION LEVEL TECHNICAL BASIS PURPOSE The purpose of this document is to provide an explanation and rationale for each of the emergency action levels (EALs) included in the EAL Upgrade Program for Nine Mile Point 2 (NHP-2). It is also intended to facilitate the review process of the NMP-2 EALs and provide historical documentation for future reference. This document is also intended to be utilized by those individuals responsible for implementation of EPIP-EPP-02 "Classification of Emergency Conditions Unit 2" as a technical reference and aid in EAL interpretation.

DISCUSSION EALs are the plant-specific indications, conditions or instrument readings which are utilized to classify emergency conditions defined in the NHP-2 Emergency Plan.

While the upgraded EALs are site specific, an objective of the upgrade project was to ensure conformity and consistency between the sites to the extent possible.

The revised EALs were derived from the Initiating Conditions and example EALs given in the NMP-2 Plant-Specific EAL Guideline (PEG). The PEG is the NMP-2 plant interpretation of the NUHARC methodology for developing EALs.

Many of the EALs derived from the NUHARC methodology are fission product barrier based. That is, the conditions which define the EALs are based upon loss or potential loss of one or more of the three fission product barriers.

The primary fission product barriers are:

A. Reactor fuel Claddin FC ; The fuel cladding is comprised of the zirconium tubes which house the ceramic uranium oxide pellets along with the end plugs which are welded into each end of the fuel rods.

Reactor Coolant S stem RCS : The RCS is comprised of the reactor vessel shell, vessel head, CRD housings, vessel,nozzles and penetrations and all primary systems directly connected to the RPV up to the outermost primary containment isolation valve.

C. Primar Containment PC  : The primary containment is comprised of the drywell, suppressi.on chamber, the interconnections between the two, and all isolation valves required to maintain primary containment integrity under accident conditions.

Although the secondary containment (reactor building) serves as an effective fission product barrier by minimizing ground level releases, it is not considered as a fission product barrier for the purpose of emergency classification.

May 1998 Page 3 EPHP-EPP-0102 Rev 02

DISCUSSION (Cont)

The following criteria serves as the basis for event classification related to fission product barrier loss:

Unusual Event:

Any loss or potential loss of containment Alert:

Any loss or any potential loss of either fuel clad or RCS Site Area Emer enc  :

Any loss of both fuel clad and RCS or Any potential loss of both fuel clad and RCS or Any potential loss of either fuel clad or RCS with a loss of any additional barrier General Emer enc  :

Loss of any two barriers with loss or potential loss of a third Those EALs which reference one or more of the fission product barrier Initiating Condition (IC) designators (FC, RCS and PC) in the PEG Reference section of the technical basis are derived from the Fission Product Barrier Analysis. The analysis entailed an evaluation of every combination of the plant specific barrier loss/potential loss indicators applied to the above criteria.

Where possible, the EALs have been made consistent with and utilize the conditions defined in the NMP-2 symptom based Emergency Operating Procedures (EOPs). While the symptoms that drive operator actions specified in the EOPs are not indicative of all 'possible conditions which warrant emergency classification, they do define the symptoms, independent of initiating events, for which reactor plant safety and/or fission product barrier integrity are threatened. Where these symptoms are clearly representative of one of the PEG Initiating Conditions,. they have been utilized as an EAL. This allows for rapid classification of emergency situations based on plant conditions without the need for additional evaluation or event diagnosis. Although some of the EALs presented here are based on conditions defined in the EOPs, classification of emergencies using these EALs is not dependent upon EOP entry or execution. The EALs can be utilized independently or in conjunction with the EOPs.

May 1998 Page 4 EPMP-EPP-0102 Rev 02

DISCUSSION (Cont)

To the extent possible, the EALs are symptom based. That is, the action level is defined by values of key plant operating parameters which identify emergency or potential emergency conditions. This approach is appropriate because it allows the full scope of variations in the types of events to be classified as emergencies. But, a purely symptom based approach is not sufficient to address all events for which emergency classification is appropriate. Particular events to which no predetermined symptoms can be ascribed have also been utilized as EALs since they may be indicative of potentially more serious conditions not yet fully realized.

The EALs are grouped into nine categories to simplify their presentation and to promote a rapid understanding by their users. These categories are:

1. Reactor Fuel

2. Reactor Pressure Vessel

3. Primary Containment

4. Secondary Containment

5. Radioactivity Release

6. Electrical Failures

7. Equipment Failures

8. Hazards

9. Other Categories 1 through 5 are primarily symptom based. The symptoms are indicative of actual or potential degradation of either fission product barriers or personnel safety.

Categories 6, 7 and 8 are event based. Electrical Failures are those events associated with losses of either AC or vital DC electrical power. Equipment Failures are abnormal and emergency events associated with vital plant system failures, while Hazards are those non-plant system related events which have affected or may affect plant safety.

Category 9 provides the Emergency Director the latitude to classify and declare emergencies based on plant symptoms or events which in his judgment warrant classification. This judgment includes evaluation of loss or potential of one or more fission product barriers warranting emergency classification consistent with the NUHARC barrier loss criteria.

Hay 1998 Page 5 EPHP-EPP-0102 Rev 02

DISCUSSION (Cont)

Categories are further divided into one or more subcategories depending on the types and number of plant conditions that dictate emergency classifications.

For example, the Reactor Fuel category has five subcategories whose values can be indicative of fuel damage: coolant activity, off-gas activity, containment radiation, other radiation monitors and refueling accidents. An EAL may or may not exist for each sub category at all four classification levels.

Similarly, more than one EAL may exist for a sub category in a given emergency classification when appropriate (i. e., no EAL at the General Emergency level but three EALs at the Unusual Event level).

For each EAL, the following information is provided:

Classification: Unusual Event, Alert, Site Area Emergency, or General Emergency Operating Mode Applicability; One or more of the following plant operating conditions are listed: Power Operation, Startup/Hot Standby, Hot Shutdown, Cold Shutdown, Refuel and Defueled EAL: Description of the condition or set of conditions which comprise the EAL Basis: Description of the rationale for the EAL PEG Reference(s): PEG IC(s) and example EAL(s) from which the EAL is derived

~ Basis Reference(s): Source documentation from which the EAL is derived The identified operating modes are defined as follows:

Power 0 erations Reactor is critical and the mode switch is in RUN.

Startu Hot Standb Reactor is critical and the mode switch is in STARTUP/HOT STANDBY.

Hot Shutdown Mode switch is usually in SHUTDOWN and reactor coolant temperature is >200 'F.

May. 1998 Page 6 EPMP-EPP-0102 e

Rev 02

DISCUSSION (Cont)

Cold Shutdown Mode switch usually in SHUTDOWN and reactor coolant temperature is ~200 'F.

Refuel Mode switch in REFUEL (with vessel head closure bolts less than fully tensioned or with head removed)

OR Mode switch in SHUTDOWN and reactor coolant temperature is <140 'F.

Defueled RPV contains no irradiated fuel.

May 1998 Page 7 EPMP-EPP-0102 Rev 02

1.0 REACTOR FUEL The reactor fuel cladding serves as the primary fission product barrier. Over the useful life of a fuel bundle, the integrity of this barrier should remain intact as long as fuel cladding integrity limits are not exceeded.

Should fuel damage occur (breach of the fuel cladding integrity) radioactive fission products are released to the reactor coolant. The magnitude of such a release is dependent upon the extent of the damage as well as the mechanism by which the damage occurred. Once released into the reactor coolant, the highly radioactive fission products can pose significant radiological hazards inplant from reactor coolant process streams. If other fission product barriers were offsite to fail, these radioactive fission products can pose significant radiological consequences.

The following parameters/indicators are indicative of possible fuel failures:

Coolant Activit  : During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from either the fission of tramp uranium in the fuel cladding or minor perforations in the cladding itself. Any significant increase from these base-line levels is indicative of fuel failures.

Off- as Activit : As with coolant activity, any fuel failures will release fission products to the reactor coolant. Those products which are gaseous or volatile in nature will be carried over with the steam and eventually be detected by the air ejector off-gas radiation monitors.

Containment Radiation Monitors: Although not a direct indication or measurement of fuel damage, exceeding predetermined limits on containment high range radiation monitors under LOCA conditions is indicative possible fuel failures. In addition, this indicator is utilized as an indicator of RCS loss and potential containment loss.

Other Radiation Monitors: Other process and area radiation monitoring systems are specifically designed to provide indication of possible fuel damage such as Area Radiation Monitoring Systems.

Refuelin Accidents: Both area and process radiation monitoring systems designed to detect fission products during refueling conditions as well as visual observation can be utilized to indicate loss or potential loss of spent fuel cladding integrity.

May 1998 Page 8 EPMP-,EPP-0102 Rev 02

Cool ant Acti vi t 1.1.1 Unusual Event Coolant activity > 0.2 pCi/gm 1-131 equivalent or >100/Ebar pCi/gm NUMARC IC:

Fuel clad degradation FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential- precursor of more serious problems. This EAL addresses reactor coolant samples exceeding coolant technical specifications for iodine spiking.

PEG Reference(s):

SU4.2 Basis Reference(s):

1. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, Article 3.4.5.a and b May 1998 Page 9 EPMP-EPP-0102 Rev 02

Coolant activity > 300 pCi/gm 1-131 equivalent NUMARC IC:

N/A FPB Loss/Potential Loss:

Fuel clad loss l

Mode Appl i cabi 1 i ty:

Power oper ation, startup/hot standby, hot shutdown Basis:

Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This amount of coolant activity is well above that expected for iodine spikes and corresponds to about 2N to 5N fuel clad damage. When reactor coolant activity reaches this level, significant clad heating has occurred and thus the fuel clad barrier is considered lost. Therefore, declaration of an Alert is warranted.

PEG Reference(s): . ~

FC1.1 Basis Reference(s):

1. General Electric NED0-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions May 1998 Page 10 EPMP-EPP-0102 Rev 02

1. 2 Off- as Acti vi t 1.2.1 Unusual Event Valid offgas radiation high alarm (at >DRMS red) for >15 min.

NUMARC IC:

Fuel clad degradation Mode Applicability:

Power Operation, Startup/hot standby, hot shutdown FPB Loss/Potential Loss:

N/A Basis:

Elevated offgas radiation activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. The Technical Specification allowable limit is an offgas level not to exceed 350,000 pCi/sec. The DRHS alarm setpoint has been conservatively selected because it is operationally significant and is readily recognizable by Control Room operating staff. 15 minutes is allotted for operator action to reduced the offgas radiation levels and exclude transient conditions.

The hi offgas radiation alarm is set using methodology outlined in the ODCH.

PEG Reference(s):

SU4.1 Basis Reference(s):

1. NUREG-1253 Technical Specifications Nine Hile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Article 3. 11.2.7

2. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7. 10-1

3. NUREG-1253 Technical Specifications Nine Hile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Article 3.4.5.a and b Hay 1998 Page 11 EPHP-EPP-0102 Rev 02

1.2.1 (Cont)

4. NUREG-1253 Technical Specifications Nine Nile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Article 3.4.5c.2 and 3

5. N2-0P-42, annunciator 851253, pg. 115 Containment Radiation 1.3.1 Alert Drywell area radiation > 41 R/hr NUMARC IC:

N/A FPB Loss/Potential Loss:

RCS Loss FPB Loss/Potential Loss:

RCS loss Node Applicability:

Power operation, startup/hot standby, hot shutdown Basis:

The drywell radiation reading is a value which indicates the release of reactor coolant to the drywell. The reading is calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i. e., within Technical Specifications) into the drywell atmosphere. The reading is less than that specified for EAL 1.3.2 because no damage to the fuel clad is assumed. Only leakage from the RCS is assumed in this EAL.

Hay 1998 Page 12 EPHP-EPP-0102 Rev 02

1.3.'2 (Cont)

It is important to recognize that the radiation monitor may be sensi.tive to shine from the RPV or RCS piping. Drywell High Range Radiation .Monitors are installed in the following locations:

2CEC*Pn18800: ORMS 2RMS*RE1B/0 RMS*RUZ1B RMS*RUZ1D 2CEC*Pnl880B'BMS 2RHS*RElA/C RMS*RUZ1A RHS*RUZlC PEG Reference(s):

RCS3.1 Basis Re'ference(s):

1. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7.5-1

2. Calculation PR-C-24-0 1.3.2 Site Area Emer enc Drywell area radiation > 3100 R/hr NUHARC IC:

N/A FPB Loss/Potential Loss:

Fuel clad loss, RCS loss Node Applicability:

Power operation, star tup/hot standby, hot shutdown Basis:

The drywell radiation reading is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell. The reading is calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 pCi/gm dose eq'uivalent 1-131 into the drywell atmosphere. Reactor coolant Hay 1998 Page 13 EPHP-EPP-0102 Rev 02

1.3.2 (Cont) concentrations of this magnitude are several times larger than the maximum concentrations allowed within Technical Specifications (including iodine spiking) and are therefore indicative of fuel damage (approximately 2% 5N clad failure depending on core inventory and RCS volume). The reading is higher than that specified for EAL 1.3. 1 and, thus, this EAL indicates a loss of both the fuel clad barrier and the RCS barrier.

It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. Drywell High Range Radiation Monitors are installed in the following locations:

2CEC*Pnl880D: ORMS 2RHS*RE1B/D RMS*RUZ1B RHS*RUZ1D 2CEC*Pnl880B: DRHS 2RHS*RElA/C RMS*RUZ1A RHS*RUZ 1 C PEG Reference(s):

FC3.1 Basis Reference(s):

1. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7.5-1

2. Calculation PR-C-24-0 1.3.3 General Emer enc Drywell area radiation > 5.2E6 R/hr NUHARC IC:

N/A FPB Loss/Potential Loss:

Fuel clad loss, RCS loss, Containment potential loss Node Applicability:

Power operation, startup/hot standby, hot shutdown May 1998 Page 14 EPHP-EPP-0102 Rev 02

1.3.2

~ ~ (Cont)

Basis:

The drywell radiation reading is a value which indicates significant fuel damage well in excess of that required for loss of the RCS barrier and the fuel clad barrier. NUREG-1228 "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents" states that such readings do not exist when the amount of clad damage is less than 20%. A major release of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure into the reactor coolant has occurred. Regardless of whether the primary containment barrier itself is challenged, this amount of activity in containment could have severe consequences if released.

It is, therefore, prudent to treat this as a potential loss of the containment barrier and upgrade the emergency classification to a General Emergency.

It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. Drywell High Range Radiation Monitors are installed in the following locations:

2CEC*Pnl880D: DRHS 2RMS"RE18/D RMS*RUZ1B RHS*RUZ1D 2CEC*Pnl880B: ORMS 2RHS*RE1A/C RMS*RUZ1A RMS*RUZIC PEG Reference(s):

PC3.1 Basis Reference(s):

1. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7.5-1

2. Calculation PR-C-24-0, Rev. 4 Hay 1998 Page 15 EPHP-EPP-0102 Rev 02

1.4 Other Radiation Monitors 1.4.1 Unusual Event Any sustained ARH reading > 100 x ORMS high radiation alarm (red) or offscale high (DETECTOR SATURATION) resulting from an uncontrolled process NUHARC IC:

Unexpected increase in plant radiation or airborne concentration.

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

Valid elevated area radiation levels usually have long lead times relative to the potential for radiological release beyond the site boundary, thus impact to public health and safety is very low.

This EAL addresses unplanned increases in radiation levels inside the plant. These radiation levels represent a degradation in the control of radioactive material and a potential degradation in the level of safety of the plant. Area radiation levels above 100 times the high radiation alarm setpoint have been selected because they are readily identifiable on ARH instrumentation. The ARH alarm setpoint is considered to be a bounding value above the maximum normal radiation level in an area. Since ARH setpoints are nominally set one decade over normal levels, 100 times the alarm setpoint provides an appropriate threshold for emergency classification. For those ARMS whose upper range limits are less than 100 times the high radiation alarm setpoint, a value of offscale high is used. This EAL escalates to an Alert, if the increases impair the level of safe plant operation.

PEG Reference(s):

AU2.4 Basis Reference(s):.

1. N2-0P-79, Radiation Monitoring System

2. Calculation PR-C-25-1 Hay 1998 Page 16 EPHP-EPP-0102 Rev 02

1.4.2

~ ~ Alert Valid Rx Bldg. above Refueling Floor Radiation Monitor 2HVR*RE14A or B, Gaseous Radiation Monitors (channel 1) isolation OR Any sustained refuel floor rad monitor > 8.0 R/hr Table 1 Table 1 Refuel Floor Rad Monitors RMS111, RB 354'est of Spent Fuel Pool RMS112, RB 354'ast of Spent Fuel Pool NUMARC IC:

Major damage to irradiated fuel or loss of water level that has or will result in the uncovering of irradiated fuel outside the reactor vessel.

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

This EAL is defined by the specific areas where irradiated fuel is located such as reactor cavity, reactor vessel, or spent fuel pool.

Sufficient time exists to take corrective actions for these conditions and there is little potential for substantial fuel damage. NUREG/CR-4982 "Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82" indicates that even if corrective actions are not taken, no prompt fatalities are predicted and the risk of injury is low. In addition, NRC Information Notice No. 90-08, "KR-85 Hazards from Decayed Fuel" presents the following in its discussion:

"In the event of a serious accident involving decayed spent fuel, protective actions would be needed for personnel on site, while offsite doses (assuming an exclusion area radius of one mile from the plant site) would be well below the Environmental Protection Agency's Protective Action Guides. Accordingly, it is important to be able to properly survey and"monitor for Kr-85 in the event of an accident with decayed spent fuel."

Thus, an Alert Classification for this event is appropriate.

Escalation, if appropriate,9.0.would occur via Emergency Director judgment in EAL Category May 1998 Page 17 EPMP-EPP-0102 Rev 02

1.4.2 (Cont)

The basis for the reactor building ventilation monitor setpoint is a spent fuel handling accident (isolation setpoint) and is, therefore, appropriate for this EAL. Technical Specification requires isolation at < 2.36 E-3 pCi/cc).

Area radiation levels on the refuel floor at or above the Maximum Safe Operating value (8.0 R/hr) are indicative of radiation fields which may limit personnel access. Access to the refuel floor is required in order to visually observe water level in the spent fuel pool. Without access to the refuel floor, it would not be possible to determine the applicability of EAL 1.5.2. Area radiation levels on the refuel floor at or above the Maximum Safe Operating value could also adversely affect equipment whose operation may be needed to assure adequate core cooling or shutdown the reactor.

PEG Reference(s):

AA2.1 Basis Reference(s):

1. NUREG-0818, Emergency Action Levels for Light Water Reactors

2. NUREG/CR-4982, Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82, July 1987

3. NRC Information Notice No. 90-08, KR-85 Hazards from Decayed Fuel

4. N2-0P-79, Radiation Monitoring System

5. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.2-2

6. N2-0P-61B, Standby Gas Treatment 1.4.3 Alert Sustained area radiation levels > 15 mR/hr in either:

Control Room OR Central Alarm Station (CAS) and Secondary Alarm Station (SAS)

NUMARC IC:

Release of radioactive material or increases in radiation levels within the facility that impedes operation of systems required to maintain safe operations or to establish or maintain cold shutdown.

May 1998 Page 18 EPMP-EPP-0102 Rev 02

1.4.3 (Cont)

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This EAL addresses increased radiation levels that impede necessary access to operating stations requiring continuous occupancy to maintain safe plant operation or perform a safe plant shutdown. Areas requiring continuous occupancy include the Control Room, the central alarm station (CAS) and the secondary security alarm station (SAS).

The security alarm stations are included in this EAL because of their importance to permitting access to areas required to assure safe plant operations.

The value of 15 mR/hr is derived from the GOC 19 value of 5 rem in 30 days with adjustment for expected occupancy times. Although Section III.D.3 of NUREG-0737, "Clarification of THI Action Plan Requirements", provides that the 15 mR/hr value can be averaged over the 30 days, the value is used here without averaging. A 30 day duration implies an event potentially more significant than an Alert.

It is the impaired ability to operate the plant that results in the actual or potential degradation of the level of safety of the plant.

The cause or magnitude of the increase in radiation levels is not a concern of this EAL. The Emergency Director must consider the source or cause of the increased radiation levels and determine if any other EALs may be involved. For example, a dose rate of 15 mR/hr in the Control Room may be a problem in itself. However, the increase may also be indicative of high dose rates in the containment due to a LOCA. In this latter case, a Site Area Emergency or a General Emergency may be indicated by other EAL categories.

This'AL could result in declaration of an Alert at NHP-2 due to a radioactivity release or radiation shine resulting from a major accident at the NHP-1 or JAFNPP. Such a declaration would be appropriate if the increase impairs safe plant operation.

This EAL is not intended to apply to anticipated temporary radiation increases due to pl,awned events (e. g., radwaste container movement, depleted resin transfers, etc.).

PEG Reference(s):

AA3.1 May 1998 Page 19 EPHP-EPP-0102 Rev 02

1.4.3 (Cont)

Basis Reference(s):

1. GDC 19

2. NUREG-0737, "Clarification of THI Action Plan Requirements",

Section III.D.3 1.4.4 Alert Sustained area radiation levels > 8.0 R/hr in any areas, Table 2 AND Access is required for safe operation or shutdown Table 2 Plant Safet Function Areas Control Building Normal Switchgear Building South Aux. Bay North Aux. Bay RadWaste Building Reactor Building Turbine Building Diesel Generator Building Screenwell Building/ Service Water Pump Bays Condensate Storage Tanks Building Standby Gas Treatment Building NUMARC IC:

Release of radioactive material or increases in radiation levels within the facility that impedes operation of systems required to maintain safe operations or to establish or maintain cold shutdown.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This EAL addresses i.ncreased radiation. levels in areas requiring infrequent access in order to maintain safe plant operation'or perform a safe plant shutdown. Area radiation levels at or above 8 R/hr are indicative of radiation fields which may limit personnel access or adversely affect equipment whose operation may be needed to assure adequate core cooling or shutdown the reactor. This basis of the value is described in NHPC memo File Code NHP31027 "Exposure Guidelines For Unusual/Accident Conditions". The areas selected are Hay 1998 Page 20 EPHP-EPP-0102 Rev 02

1.4.4 (Cont) consistent with th'ose listed in other EALs and represent those structures which house systems and equipment necessary for the safe operation and shutdown of the plant. Guidelines For Unusual/Accident Conditions". The areas selected are consistent with those listed in other EALs and represent those structures which house systems and equipment necessary for the safe operation and shutdown of the plant.

It is the impaired ability to operate the plant that results in the actual or potential degradation of the level of safety of the plant.

The cause or magnitude of the increase in radiation levels is not a concern of this EAL. The Emergency Director must consider the source or cause of the increased radiation levels and determine if any other EAL may be involved. For example, a dose rate of 8 R/hr may be a problem in itself. However, the increase may also be indicative of high dose rates in the containment due to a LOCA. In this latter case, a Site Area Emergency or a General Emergency may be indicated by other EAL categories.

This EAL could result in declaration of an Alert at NMP-2 due to a radioactivity release or'adiation shine resulting from a major accident at the NMP-1 or JAFNPP. Such a declaration would be

- appropriate if the increase impairs safe plant operation.

This EAL is not meant to apply to increases in the containment radiation monitors as these are events which are addressed in other EALs. Nor is it intended to apply to anticipated temporary radiation increases due to planned events (e. g., radwaste container movement, deplete resin transfers, etc.).

PEG Reference(s):

AA3.2 Basis Reference(s):

1. Niagara Mohawk Power Corporation memo File Code NMP31027 "Exposure Guidelines For Unusual/Accident Conditions", Revision 1, 3/18/93 1.5 Refuelin Accidents 1.5.1 Unusual Event Spent fuel pool/reactor cavity water level cannot be restored and maintained above the spent fuel pool low water level alarm NUMARC IC:

Unexpected increase in plant radiation or airborne concentration.

May 1998 Page 21 EPMP-EPP-0102 Rev 02

1.5.1 (Cont)

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

, The above event has a long lead time relative to the potential for radiological release outside the site boundary, thus impact to public health and safety is very low. However, in light of recent industry events, classification as an Unusual Event is warranted as a precursor to a more serious event.

The spent fuel pool low water level is indicated by annunciators 873317 and 875117 which alarm at El 352'". The definition of "...

cannot be restored and maintained above ..." allows the operator to visually observe the low water level condition, if possible, and to attempt water level restoration instructions as long as water level remains above the top of irradiated fuel. Water level restoration instructions are performed in accordance with N2-0P-38.

When the fuel transfer canal is directly connected to the spent fuel pool and reactor cavity, there could exist the possibility of uncovering irradiated fuel in the fuel transfer canal. Therefore, this EAL is applicable for conditions in which irradiated fuel is being transferred, to and from the RPV and spent fuel pool.

PEG Reference(s):

AU2.1 Basis Reference(s):

1. N2-0P-38, Spent Fuel Pool Cooling and Cleanup System 1.5.2 Alert Imminent or report of actual observation of the uncovering of irradiated fuel.

NUMARC IC:

Hajor damage to irradiated fuel or loss of water level that has or will result in the uncovering of irradiated fuel outside the reactor vessel.

Hay 1998 ,Page 22 EPHP-EPP-0102 Rev 02

1.5.2 (Cont)

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This EAL is defined by the specific areas where irradiated fuel is located such as reactor cavity, reactor vessel, or spent fuel pool.

Sufficient time exists to take corrective actions for these conditions and there is little potential for substantial fuel damage. NUREG/CR-4982 "Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82" indicates that even if corrective actions are not taken, no prompt fatalities are predicted and the risk of injury is low. In addition, NRC Information Notice No. 90-08, "KR-85 Hazards from Decayed Fuel" presents the following in its discussion:

"In the event of a serious accident involving decayed spent fuel, protective actions would be needed for personnel on site, while offsite doses (assuming an exclusion area radius of one mile from the plant site) would be well below the Environmental Protection Agency's Protective Action Guides. Accordingly, it is important to be able to properly survey and monitor for Kr-85 in the event of an accident with decayed spent fuel."

Thus, an Alert Classification for this event is appropriate.

Escalation, if appropriate, would occur by Emergency Director judgment in EAL Category 9.0.

There is no indication that water level in the spent fuel pool has dropped to the level of the fuel other than by visual observation by personnel on the refueling floor. When the fuel transfer canal is directly connected to the spent fuel pool and reactor cavity, there could exist the possibility of uncovering irradiated fuel in the fuel transfer canal. Therefore, this EAL is applicable for conditions in which irradiated fuel is being transferred to and from the RPY and spent fuel pool, This EAL applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage.

PEG Reference(s):

AA2.2 Hay 1998 Page 23 EPHP-EPP-0102 Rev 02

1.5.2 (Cont)

Basis Reference(s):

1. NUREG-0818, Emergency Action Levels for Light Water Reactors

2. NUREG/CR-4982, Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82, July 1987

3. NRC Information Notice No. 90-08, KR-85 Hazards from Decayed Fuel 2.0 REACTOR PRESSURE VESSEL RPV The reactor pressure vessel provides a volume for the coolant which covers the reactor core. The RPV and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel cladding integrity fail.

There are two RPV parameters which are indicative of conditions which may pose a threat to RPV or fuel cladding integrity:

~ RPV Water Level: RPV water level is directly related to the status of adequate core cooling, and therefore fuel cladding integrity. Excessive ( > Tech. Spec.) reactor coolant to drywell leakage indications are utilized to indicate potential pipe cracks which may propagate to an extent threatening fuel clad, RPV and primary containment integrity. Conditions under which all attempts at establishing adequate core cooling have failed require primary containment flooding.

~ Reactor Power Reactivit Control: The inability to control reactor power below certain levels can pose a direct threat to reactor fuel, RPV and primary containment integrity.

2.1 RPV Water Level 2.1.1 Unusual Event Unidentified drywell leakage > 10 gpm OR Reactor coolant to drywell identified leakage > 25 gpm NUMARC IC:

RCS leakage FPB Loss/Potential 'Loss:

N/A Mode Applicability:

Power operation, startup/hot standby, hot shutdown Hay 1998 Page 24 EPMP-EPP-0102 Rev 02

2.1.

~ ~ 1 (Cont)

~

Basis:

The conditions of this EAL may be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for the unidentified drywell leakage was selected because it is observable

'ith normal Control Room indications and is consistent with the Technical Specification threshold for leaks beyond which increased risk of crack propagation exists. The 25 gpm value for identified reactor coolant to drywell leakage is set at a higher value because of the significance of identified leakage in comparison to unidentified or pressure boundary leakage.

Only operating modes in which there is fuel in the reactor coolant system and the system is pressurized are specified.

PEG Reference(s):

SU5.1 Basis Reference(s):

None 2.1.2 Site Area Emer enc RPV water level cannot be restored and maintained > top of active fuel.

NUMARC IC:

Loss of reactor vessel water level has or will uncover fuel in the reactor vessel.

FPB Loss/Potential Loss:

Fuel clad potential loss, RCS loss Mode Applicability:

Power Operation, Startup/Hot Standby, Hot Shutdown, Cold Shutdown, Refuel May 1998 Page 25 EPMP-EPP-0102 Rev 02

2. 1. 2 (Cont)

Basis:

The RPV water level used in this EAL is the top of active fuel (TAF).

This value corresponds to the level which is used in EOPs to indicate challenge to core cooling and loss of the fuel clad barrier. This is the minimum water level to assure core cooling without further degradation of the clad. Severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured if RPV water level is not maintained > TAF.

Sustained uncovery of the fuel irrespective of the event that causes fuel uncovery is justification alone for declaring a Site Area Emergency. This includes events that could lead to fuel uncovery in any plant operating mode including cold shutdown and refuel.

Escalation to a General Emergency occurs through radiological effluence addressed in EAL 1.3.3 for drywell radiation and in the EALs defined for Category 5.0, Radioactivity Release.

The terminology of "cannot be restored and maintained" is intended to be consistent with the interpretation that:

"The value of the identified parameter(s) is/is not able to be returned to above/below specified limits. This determination includes making an evaluation that considers both current and future systems performance in relation to the current value and trend of the parameter(s). Neither implies that the parameter must actually exceed the limit before the classification is made nor that the classification must be made before the limit is reached. Does not imply any specific time interval but does not permit prolonged operation beyond a limit without making the specified classification."

This definition would require the emergency classification be made prior to water level dropping below TAF if, based on an evaluation of the current trend of RPV water level and in consideration of current and future injection system performance, that RPV water level will not likely be restored and maintained above TAF. This definition however, also provides the latitude, based on that same evaluation, not to declare the SAE for those situations in which the RPV water level transiently drops below TAF in the process of RPV water level restoration.

PEG Reference(s):

SS5.1 FC2. 1 RCS4.1 Basis Reference(s):

1. N2-EOP-RPV, RPV Control Hay 1998 Page 26 EPHP-EPP-0102 Rev 02

2.1.3 General Emer enc Primary Containment Flooding required NUNARC IC:

N/A FPB Loss/Potential Loss:

Fuel clad loss, RCS loss, Containment potential loss Node Applicability:

Power operation, startup/hot standby, hot shutdown Basis:

The condition in this EAL represents imminent melt sequences which, if not corrected, could lead to RPV failure and increased potential for primary containment failure. If the EOPs are ineffective in restoring RPV water level above the top of active fuel, loss of the fuel clad barrier is imminent., Therefore, declaration of a General Emergency is appropriate when entry to the Primary Containment Flooding EOP is required.

PEG Reference(s):

PC4.1 Basis Reference(s):

1. N2-EOP-RPV, RPV Control 2.2 Reactor Power Reactivit Control 2.2.1 Alert

~An RPS scram setpoint has been exceeded AND Automatic scram fails to result in a control rod pattern which assures reactor shutdown under all conditions without boron.

NUNARC IC:

Failure of Reactor Pt otection system instrumentation to complete or initiate an automatic reactor trip once a Reactor Protection system setpoint has been exceeded and manual trip was successful while in power operations or hot standby.

Hay 1998 Page 27 EPHP-EPP-0102 Rev 02

2.2. 1 (Cont)

FPB Loss/Potential Loss:

N/A Mode Applicability:

Power operation, startup/hot standby Basis:

This condition indicates a failure of the Reactor Protection System to scram the reactor automatically, and maintain it in a shutdown under all conditions without boron. This is consistent with the entry requirements of N2-EOP-C5, "Level/Power Control".

If a manual scram does not result in reactor power being reduced below the APRM downscale setpoint (4%) or suppression pool temperature exceeds the Boron Injection Initiation Temperature (110 'F) escalation to a Site Area Emergency is required. A manual scram is any set of action by the reactor operators at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor subcritical including manual scram push buttons, ARI and mode switch.

PEG Reference(s):

SA2.1 Basis Reference(s):

1'. N2-EOP-RPV, RPV Control, Section RL

2. "Methodology for Development of Emergency Action Levels" NUMARC/NESP-007 Revision 2 - guestions and Answers, June 1993 2.2.2 Site Area Emer enc

~An RPS scram setpoint has been exceeded ANO Automatic and manual scrams fail to result in a control rod pattern which assures reactor shutdown under all conditions without boron AND Either:

Reactor power >4%

OR Suppression pool temperature >110'F NUMARC IC:

Failure of Reactor Protection system instrumentation to complete or initiate an automatic reactor trip once a Reactor Protection system setpoint has been exceeded and manual scram trip was not successful.

May 1998 Page 28 EPMP-EPP-0102 Rev 02

2.2.2 (Cont)

FPB Loss/Potential Loss:

N/A Mode Applicability:

Power operation, startup/hot standby Basis:

This condition indicates failure of the Reactor Protection System to shutdown the reactor (automatically or manually) and maintain it shutdown under all conditions without boron. Under these conditions, the reactor is producing more heat than can be removed using available safety systems. A Site Area Emergency is indicated because conditions exist leading to imminent or potential loss of both the fuel clad and the Primary Containment.

The failure of automatic initiation of a reactor scram followed by unsuccessful manual initiation actions which can be rapidly taken at the reactor control console does not, by itself, lead to imminent loss of either fuel clad or primary containment barriers. It is the continued criticality under conditions requiring a reactor scram along with the continued addition of heat to containment which poses the imminent threat to primary containment or fuel clad barriers. In accordance with the EOPs, SLC is initiated based on heat addition to containment in excess of safety system capability under failure to scram conditions.

An immediate manual scram is any set of actions by the reactor operator at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor subcritical including manual scram push buttons, ARI and mode switch.

PEG Reference(s):

SS2.1 Basis Reference(s):

1. N2-EOP-RPV, RPV Control, Section RL

2. "Methodology for Development of Emergency Action Levels" NUMARC/NESP-007 Revision 2 - guestions and Answers, June 1993 May 1998 Page 29 EPMP-EPP-0102 Rev 02

2.2.3 General Emer enc

~An RPS scram setpoint has been exceeded AND Automatic and manual scrams fail to result in a control rod pattern which assures reactor shutdown under all conditions without boron AND Either:

RPV water level cannot be restored and maintained > Minimum Steam Cooling RPV Water Level OR Suppression pool temperature and RPV pressure cannot be maintained <HCTL.

NUMARC IC:

Failure of the Reactor Protection System to complete an automatic trip and manual trip was not successful and there is indication of an extreme challenge to the ability to cool the core.

FPB Loss/Potential Loss:

N/A Mode Applicability:

Power operation, startup/hot standby Basis:

Under the conditions of this EAL, the efforts to bring the reactor subcritical have been unsuccessful and, as a result, the reactor is producing more heat than the maximum decay heat load for which the safety systems were designed.

An extreme challenge to the ability to cool the core is. indicated when RPV water level cannot be restored and maintained above the Hinimum Steam Cooling RPV Water Level. This RPV water level is used in the EOPs to define the lowest RPV water level in a failure-to-scram event above which adequate core cooling can be maintained without sufficient steam cooling flow. This situation could be precursor for a core melt sequence.

In this situation, core degradation can occur rapidly For this reason, the General Emergency declaration is intended to be anticipatory of the loss of two fission product barriers and a loss of a third thus permitting the maximum offsite 'otential intervention time..

An immediate manual scram is any set of actions by the reactor operator at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor subcritical including manual scram push buttons, ARI and mode switch.

Hay 1998 Page 30 EPHP-EPP-0102 N

Rev 02

2.2.3 (Cont)

PEG Reference(s):

SG2.1 Basis Reference(s):

1. N2-EOP-C5, Level/Power Control 3.0 PRIMARY CONTAINMENT PC The primary containment structure is a pressure suppression system.

It forms a fission product barrier designed to limit the release of radioactive fission products generated from any postulated accident so as to preclude exceeding offsite exposure limits.

The primary containment structure is a low leakage pressure suppression system housing the reactor pressure vessel (RPV), the reactor coolant recirculation piping and other branch connections of the reactor primary system. The primary containment is equipped with isolation valves for most systems which penetrate the containment boundary. These valves automatically actuate to isolate systems under emergency conditions.

There are four primary containment parameters which are indicative of conditions which may pose a threat to primary containment integrity or indicate degradation of RPV or reactor fuel integrity.

Primar Containment Pressure: Excessive primary containment pressure is also indicative of either primary system leaks into containment or loss of containment cooling function. Primary containment pressures at or above specified limits pose a direct threat to primary containment integrity and the pressure suppression function.

Su ression Pool Tem erature: Excessive suppression pool water temperatures can result in a loss of the pressure suppression capability of containment and thus be indicative of severely degraded RPV and containment conditions.

Combustible Gas Concentrations: The existence of combustible gas concentrations in containment pose a severe threat to containment integrity and are indicative of severely degraded reactor core and/or RPV conditions.

Containment Isolation Status: The existence of an unisolable steam line break outside containment constitutes a loss of containment integrity as well as a loss of RCS boundary. Should a loss of fuel cladding integrity occur, the potential for release of large amounts of radioactive materials to the environment exists.

Hay 1998 Page 31 EPHP-EPP-0102 Rev 02

3.1 Containment Pressure 3.1.1 Alert Primary containment pressure cannot. be maintained < 1.68 psig due to coolant leakage NUMARC IC:

N/A FPB Loss/Potential Loss:

RCS loss Node Applicability:

Power operation, startup/hot standby, hot shutdown Basis:

The drywell pressure value is the drywell high pressure scram setpoint and is indicative of a LOCA event. The term "cannot be maintained below" is intended to be consistent with the conditions specified in the Primary Containment Control EOP indicative of a high energy release into containment for which normal containment cooling systems are insufficient.

PEG Reference(s):

RCS2.1 Basis Reference(s):

1. N2-0P-97, annunciator 603401 3.1.2 Site Area Emer enc Primary containment pressure cannot be maintained < 1.68 psig AND Coolant activity > 300 pCi/gm NUMARC IC:

N/A FPB Loss/Potential Loss:

Fuel clad loss, RCS loss May 1998 Page 32 EPMP-EPP-0102 Rev 02

3.1.2

~ ~ (Cont)

Node Applicability:

Power operation, startup/hot standby, hot shutdown Basis:

The drywell pressure value is the drywell high pressure scram setpoint and is indicative of a LOCA event. The term "cannot be maintained below" is intended to be consistent with the conditions specified in the Primary Containment Control EOP indicative of a high energy release into containment for which normal containment cooling systems are insufficient.

Elevated reactor coolant activity represents a potential degradation in the level of safety of the, plant and a potential precursor of more serious problems. This amount of coolant activity is well above that expected for iodine spikes and corresponds to about 2% to 5% fuel clad damage. When reactor coolant activity reaches this level, significant clad heating has occurred and thus the'uel clad barrier is considered lost.

The combination of these conditions represents a loss of two fission product barriers and, therefore, declaration of a Site Area Emergency is warranted.

Reference(s):

FC1. 1 RCS2.1

'EG Basis Reference(s):

1. N2-0P-97, annunciator 603401

2. General Electric NED0-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions 3.1.3 General Emer enc Primary containment venting is required due to PCPL NUMARC IC:

N/A FPB Loss/Potential Loss:

Fuel clad loss, RCS loss, containment loss Hay 1998 Page 33 EPHP-EPP-0102 Rev 02

3.1.3 (Cont)

Mode Applicability:

Power operation, startup/hot standby,,hot shutdown Basis:

Loss of primary containment is indicated when proximity to the Primary Containment Pressure Limit (PCPL) requires venting irrespective of the offsite radioactivity release rate. To reach the PCPL, primary containment pressure must exceed that predicted in any plant design basis accident analysis. A loss of the RCS barrier must have occurred with a potential loss of the fuel clad barrier.

PEG Reference(s):

PC1.3 PC2.2 Basis Reference(s):

1. N2-EOP-PC, Primary Containment Control 3.2 Su ression Pool Tem erature 3.2.1 Site Area Emer enc RPV pressure and suppression pool temperature cannot be maintained

( HCTL (non-ATWS)

NUMARC IC:

Complete loss of function needed to achieve or maintain hot 'shutdown with reactor coolant >200 F.

FPB Loss/Potential Loss:

N/A Mode Applicability:

Power operation, startup/hot standby, hot shutdown Basis:

This EAL addresses complete loss of functions, including ultimate heat sink, required for hot shutdown with the reactor at pressure and temperature, Under these conditions, there is an actual major failure of a system intended for protection of the public. Thus, declaration of a Site Area Emergency is warranted.

May 1998 Page 34 EPMP-EPP-0102 4

Rev 02

3.2.1 (Cont)

Functions required for hot shutdown consist of the ability to achieve reactor shutdown and to discharge decay heat energy from the reactor to the ultimate heat sink. Inability to remove decay heat energy is reflected in an increase in suppression pool temperature. Elevated suppression pool temperature is addressed by the Heat Capacity Temperature Limit (HCTL). The HCTL is a function of RPV pressure and suppression pool temperature. If RPV pressure and suppression pool temperature cannot be maintained below the HCTL, the ultimate heat sink is threatened and declaration of a Site Area Emergency is warranted.

PEG Reference(s):

SS4.1 Basis Reference(s):

1. USAR, Revision 2, Section 9B.2

2. USAR, Revision 2, Section 98.4.3

3. N2-EOP-PC, Primary Containment Control 3.3 Combustible Gas Concentration 3.3. 1 Site Area Emer enc

> 4% Hz exists in DW or suppression chamber NUNARC IC:

N/A FPB Loss/Potential Loss:

Fuel clad loss, RCS loss Node Applicability:

All Basis:

4% hydrogen concentration is the lowest hydrogen concentration which, in the presence of sufficient oxygen, can support upward flame propagation. This hydrogen concentration is generally considered the lower boundary of the range in which localized deflagrations may occur. To generate such a concentration of combustible gas, loss of both the fuel clad and RCS barriers must have occurred. Therefore, declaration of a Site Area Emergency is warranted.

Hay 1998 Page 35 EPHP-EPP-0102 Rev 02

3.3.1 (Cont)

If hydrogen concentrations increase in conjunction with the presence of oxygen to global deflagration levels (i.e. ~ 6% hydrogen and > 5%

oxygen), venting of the containment irrespective of the offsite radioactive release rate would be required by EOPs and declaration of a General Emergency required.

PEG Reference(s):

SS5.2 Basis Reference(s):

1. N2-EOP-PC, Primary Containment Control, Revision 5 3.3.2 General Emer enc Primary containment venting is required due to combustible gas concentrations NUNARC IC:

N/A FPB Loss/Potential Loss:

Fuel clad loss, RCS loss, Containment loss Node Applicability:

All Basis:

6% hydrogen concentration in the presence of 5% oxygen concentration is the lowest concentration at which a deflagration inside of the primary containment could occur. When hydrogen and oxygen concentrations reach or exceed combustible limits, imminent loss of the containment barrier exists. To generate such levels of combustible gas, loss of the fuel clad and RCS barriers must have occurred. Venting of the containment irrespective of the offsite radioactive release rate is required by EOPs for this condition.

PEG Reference(s):

PC1.4 PC2.2 Basis Reference(s):

1. NZ-EOP-PC, Primary Containment Control Hay 1998 Page 36 EPMP-EPP-0102 Rev 02

3.4 Containment Isolation Status 3.4.1 Site Area Emer enc Hain Steam Line, RCIC Steam Line or Reactor Water Clean-up isolation failure AND A release pathway, outside normal process system flowpaths from the unisolable system, exists outside primary containment.

NUMARC IC:

N/A FPB Loss/Potential Loss:

RCS loss, Containment loss Mode Applicability:

Power operation, startup/hot standby, hot shutdown Basis:

The conditions of this EAL include required containment isolation failures allowing a flow path to the environment. A release pathway outside primary containment exists when steam flow is not prevented by downstream isolations. In the case of a failure of both isolation valves to close but in which no downstream flowpath exists, declaration under this EAL would got be required. The conditions of this EAL represent the loss of both the RCS barrier and the primary containment barrier and thus justifies declaration of a Site Area Emergency.

PEG Reference(s):

PC2.1 Basis Reference(s):

None Hay 1998 Page 37 EPHP-EPP-0102 Rev 02

3.4.2 General Emer enc Main Steam Line, RCIC steam line or Reactor Water Clean-up isolation failure AND A release pathway, outside normal process system flowpaths from the unisolable system, exists outside primary containment AND any:

~ Coolant activity > 300 pCi/gm I-131 equivalent

~ RPV water level < top of active fuel

~ DW radiation > 3100 R/hr NUNARC IC:

N/A FPB Loss/Potential Loss:

Fuel clad loss/potential loss, RCS loss, Containment loss Node Applicability:

'I Power operation, startup/hot standby, hot shutdown Basis:

The 'conditions of this EAL include required containment isolation failures allowing a flow path to the environment. A release pathway outside primary containment exists when steam flow is not prevented by downstream isolations. In the case of a failure of both isolation valves to close but in which no downstream flowpath exists, declaration under this EAL would not be required. Containment isolation failures which result in a release pathway outside primary containment are the basis for declaration of Site Area Emergency in EAL 3.4. 1.

When isolation failures are accompanied by elevated coolant activity, RPV water level below TAF, or high drywell radiation, declaration of a General Emergency is appropriate due to loss of the primary containment barrier, RCS barrier, and loss or potential loss of the fuel clad barrier.

May 1998 Page 38 EPHP-EPP-0102 Rev 02

3.4.2 (Cont)

Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This amount of coolant activity is well above that expected for iodine spikes and corresponds to about 2% to 5% fuel clad damage. When reactor coolant activity reaches this level, significant clad heating has occurred and thus the fuel clad barrier is considered lost.

The RPV water level used in this EAL is the top of active fuel (TAF).

This value corresponds to the level which is used in EOPs to indicate challenge to core cooling and loss of the fuel clad barrier. This is the minimum water level to assure core cooling without further degradation of the clad. Severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured if RPV water level is not maintained above TAF.

The drywell radiation reading is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell. The reading is calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 pCi/gm dose equivalent I-131 into the drywell atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations allowed within Technical Specifications (including iodine spiking) and are therefore indicative of fuel damage (approximately 2% 5% clad failure depending on core inventory and RCS volume).

It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. Drywell High Range Radiation Monitors are installed in the following locations:

2CEC*Pnl880D: DBMS 2RMS*RE1B/D RHS*RUZ1B RHS*RUZ1D 2CEC*Pnl880B: ORMS 2RMS*RE1A/C RMS*RUZIA RHS*RUZ 1 C PEG Reference(s):

PC2.1 and FC1.1 PC2.1- and FC2.1 PC2.1 and FC3.1 Hay 1998 Page 39 EPMP-EPP-0102 Rev 02

3.4.2 (Cont)

Basis Reference(s):

1. General Electric NED0-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions

2. N2-EOP-RPV, RPV Control

3. NUREG-1253 Technical Specifications Nine Hile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7.5-1

4. Calculation PR-C-24-0, Rev. 4 4.0 SECONDARY CONTAINHENT SC The secondary containment is comprised of the reactor building and associated ventilation, isolation and effluent systems. The secondary containment serves as an effective fission product barrier and is designed to minimize any ground level release of radioactive materials which might result from a serious accident. 1 The reactor building provides secondary containment during reactor operation and serves as primary containment when the reactor is shutdown and the drywell is open, as during refueling. Because the secondary containment is an integral part of the complete containment system, conditions which pose a threat to vital equipment located in the secondary containment are classifiable as emergencies.

There are two secondary containment parameters which are indicative of conditions which may pose a threat to secondary containment integrity or equipment located in secondary containment or are indicative of a direct release by a primary system into secondary containment:

Secondar Containment Tem eratures: Abnormally high secondary containment area temperatures can also pose a threat to the operability of vital equipment located inside secondary containment including RPV water level instrumentation. High area temperatures may lim'it personnel accessibility to vital areas.

High area temperatures may also be indicative of either primary system discharges into secondary containment or fires.

Secondar Containment Area Radiation Levels: Abnormally high area radiation levels in secondary containment, although not necessarily posing a threat to equipment operability, may pose a

. thr eat to personnel safety and the ability to operate vital equipment due to a lack of accessibility. Abnormally high area radiation levels may also be the result of a primary system discharging into the secondary containment and be indicative of precursors to significant radioactivity release to the environment.

Hay 1998 Page 40 EPHP-EPP-0102 Rev 02

4.1 Reactor Bui1 din Tem erature 4.1.1

~ Site Area

~

Emer enc Primary system is discharging inside RB AND RB area temperatures are > 212'F in more than one area, N2-EOP-SC NUNARC IC:

N/A FPB Loss/Potential Loss:

RCS loss, Containment loss

,Node Applicability:

Power operation, startup/hot standby, hot shutdown Basis:

The presence of elevated area temperatures in the secondary containment may be indicative of an unisolable primary system leakage outside the primary containment. These conditions represent a loss of the containment barrier and a potential loss of the RCS barrier.

PEG Reference(s):

PC2.3 RCS1.3 Basis Reference(s):

1. NZ-EOP-SC, Secondary Containment Control 4.1.2 General Emer enc Primary system is discharging into RB AND RB area temperatures are > 212'F in more than one area, N2-EOP-SC AND any:

~ Coolant activity > 300 pCi/gm I-131 equivalent

~ RPV water level ( top of active fuel

~ DW radiation > 3100 R/hr May 1998 Page 41 EPMP-EPP-0102 Rev 02

4.1.2 (Cont)

NUNARC IC:

N/A FPB Loss/Potential Loss:

Fuel clad loss/potential loss, RCS loss, Containment loss Node Applicability:

Power operation, startup/hot standby, hot shutdown Basis:

The presence of elevated area temperatures in the secondary containment may be indicative of an unisolable primary system leakage outside the primary containment. These conditions represent a loss of the containment barrier and a potential loss of the RCS barrier.

When secondary containment area temperatures are accompanied by elevated coolant activity, RPV water level below TAF, or high drywell radiation, declaration of a General Emergency is appropriate due to loss of the primary, containment barrier, RCS barrier, and loss or potential loss 'of the fuel clad barrier.

Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This amount of coolant activity is well above that expected for iodine spikes and corresponds to about 2X to 5X fuel clad damage. When reactor coolant activity reaches this level, significant clad heating has occurred and thus the fuel clad barrier is considered lost.

The RPV water level used in this EAL is the top of active fuel (TAF).

This value corresponds to the level which is used in EOPs to indicate challenge to core cooling and loss of the fuel clad barrier. This is the minimum water level to assure core cooling without further degradation of the clad. Severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured if RPV water level is not maintained above TAF.

The drywell radiation reading is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell. The reading is calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 pCi/gm May 1998 Page 42 EPMP-EPP-0102 Rev 02

4.1.2 (Cont) dose equivalent I-131 into the drywell atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations allowed within Technical Specifications (including iodine spiking) and are therefore indicative of fuel damage (approximately 2% - 5N clad failure depending on core inventory and RCS volume).

It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. Drywell High Range Radiation Monitors are installed in the following locations:

2CEC*Pnl880D: ORMS 2RHS*RE1B/D RMS*RUZIB RMS*RUZID 2CEC*Pnl8808: DRHS 2RHS*RE1A/C RHS*RUZ1A RMS*RUZ1C PEG Reference(s):

PC2.3 and FC1.1 PC2.3 and FC2. 1 PC2.3 and FC3. 1 Basis Reference(s):

1. N2-EOP-SC, Secondary Containment Control

2. N2-EOP-RPV, RPV Control

3. General Electric NED0-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions

4. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7.5-1

5. Calculation PR-C-24-0, Rev. 4 May 1998 Page 43 EPMP-EPP-0102 Rev 02

4.2 Reactor Buil din Radi ation Level 4.2.1 Site Area Emer enc Primary system is discharging into the RB AND RB area radiation levels are >8.0 R/hr in more than one area, N2-EOP-SC NUMARC IC:

N/A FPB Loss/Potential Loss:

RCS loss, Containment loss Node Applicability:

Power operation, startup/hot standby, hot shutdown Basis:

The presence of elevated area radiation levels in the secondary containment may be indicative of an unisolable primary system leakage outside the primary containment. These conditions represent a loss of the containment barrier and a potential loss oF the RCS barrier, PEG Reference(s):

PC2.3 RCS1.3 Basis Reference(s):

N2-EOP-SC, Secondary Containment Control 4.2.2 General Emer enc Primary system is discharging into the RB AND RB area radiation levels are >8.0 R/hr in more than one area, N2-EOP-SC AND any:

~ Coolant activity > 300 pCi/gm I-131 equivalent

~ RPV water-l-evel < top of active fuel

~ DW radiation > 3100 R/hr May 1998 Page 44 EPMP-EPP-0102 Rev 02

4.2.2 (Cont)

NUMARC IC'/A FPB Loss/Potential Loss:

Fuel clad loss/potential loss, RCS loss, Containment loss Node Applicability:

Power operation, startup/hot standby, hot shutdown Basis:

The presence of elevated area radiation levels in the secondary containment may be indicative of an unisolable primary system leakage outside the primary containment. These conditions represent a loss of the containment barrier and a potential loss of the RCS barrier.

When secondary containment radiation levels are accompanied by elevated coolant activity, RPV water level below TAF, or high drywell radiation, declaration of a General Emergency is appropriate due to loss of the primary containment barrier, RCS barrier, and loss or potential loss of the fuel clad barrier.

Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This amount of coolant activity is well above that expected for iodine spikes and corresponds to about 2% to 5% fuel clad damage. When reactor coolant activity reaches this level, significant clad heating has occurred and thus the fuel clad barrier is considered lost.

The RPV water level used in this EAL is the top of active fuel (TAF).

This value corresponds to the level which is used in EOPs to indicate challenge to core cooling and loss of the fuel clad barrier. This is the minimum water level to assure core cooling without further degradation of the clad. Severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured if RPV water level is not maintained above TAF.

May 1998 Page 45 EPHP-EPP-0102 Rev 02

4.2.2 (Cont)

The drywell radiation reading is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell. The reading is calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 pCi/gm dose equivalent I-131 into the drywell atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations allowed within Technical Specifications (including iodine spiking) and are therefore indicative of fuel damage (approximately 2% - 5% clad failure depending on core inventory and RCS volume).

It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. Drywell High Range Radiation Monitors are installed in the following locations:

2CEC*Pnl880D: DRHS 2RMS*RE1B/D RHS*RUZ1B RHS*RUZ1D 2CEC*Pnl880B: DRMS 2RMS*REIA/C RMS*RUZ1A RHS*RUZlC PEG Reference(s):

PC2.3 and FC1.1 PC2.3 and FC2. 1 PC2.3 and FC3.1 Basis Reference(s):

1. N2-EOP-SC, Secondary Containment Control

2. N2-EOP-RPV, RPV Control

3. General Electric NED0-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions

4. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7.5-1

5. Calculation PR-C-24-0, Rev. 4 Hay 1998 -Page 46 EPHP-EPP-0102 Rev 02

5.0 RADIOACTIVITY RELEASE Many EALs are based on actual or potential degradation of fission product barriers because of the increased potential for offsite radioactivity release. Degradation of fission product barriers though, is not always apparent via non-radiological symptoms.

Therefore, direct indication of increased radiological effluents are appropriate symptoms for emergency classification.

At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions.

There are two basic indications of radioactivity release rates which warrant emergency classifications.

~ Effluent Monitors: Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits.

~ Dose Pro 'ection and or Environmental Measurements: Projected offsite doses (based on effluent monitor readings) or actual offsite field measurements indicating doses or dose rates above classifiable limits.

5.1 Effluent Monitors 5.1.1 Unusual Event A valid reading from an unplanned release on any monitors Table 3 column "UE" for > 60 min. unless sample analysis can confirm release rates <2 x technical specifications within this time period.

Table 3 Effluent Monitor Classification Thresholds Monitor UE Alert SAE GE Radwaste/Reactor Bl dg.

Vent Effluent 2 x GEMS alarm 200 x GEMS alarm >5.5E6 pCi/s N/A Hain Stack Effluent 2 x GEMS alarm 200 x GEMS alarm N/A N/A Service Water Effluent 2 x ORMS High (red) 200 x ORMS High (red) N/A N/A Liquid RadWaste Effluent 2 x DRHS High (red) 200 x DRHS High (red) N/A N/A Cooling Tower Blowdown 2 x DRHS High (red) 200 x ORMS High (red) N/A N/A May 1998 Page 47 EPHP-EPP-0102 Rev 02

5.1.1 (Cont)

NUHARC IC'ny unplanned release of gaseous or liquid radioactivity to the environment that exceeds two times the radiological Technical Specifications for 60 minutes or longer.

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

Valid means that a radiation monitor reading has been confirmed by the operators to be correct. Unplanned releases in excess of two times the site technical specifications that continue for 60 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of safety. The final integrated dose (which is very low in the Unusual Event emergency class) is not the primary concern; it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes. Therefore, it is not intended that the release be averaged over 60 minutes. For example, a release of 4 times T/S for 30 minutes does not exceed this initiating condition. Further, the Emergency Director should not wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 60 minutes.

The alarm setpoints for the listed monitors are conservatively set to ensure Technical Specification radioactivity release limits are not exceeded. The value shown for each monitor is two times the high alarm setpoint for the Digital Radiation Monitoring System (ORMS).

Instrumentation that may be used to assess this EAL is listed below:

Radwaste/Reactor Building Vent Effluent Honitoring System monitor: 2RMS-CAB180 recorder: 2RHS-RR170/180 annunciator: 851248 Main Stack Effluent Monitoring System monitor: 2RMS-CAB170 recorder: 2RHS-RR170/180 annunciator: 851256 Hay 1998 Page 48 EPMP-EPP-0102 Rev 02

5.1.1 (Cont)

Service Water Effluent Loop A/B Radiation monitor: 2SWP*RE146A/B recorder: 2SWP*RR146A/B annunci ator: 851258 Liquid Effluent Line monitor: LWS-RE206 annunciator: 851258 Cooling Tower Blowdown Line monitor: CWS-RE 157 annunciator: 851258 PEG Reference(s):

AU1.1 Basis Reference(s):

1. N2-0P-79, Radiation Monitoring System

2. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7. 10-1

3. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7.1-1 5.1.2 Al ert A valid reading from an unplanned release on any monitors Table 3 column "Alert" for > 15 min. unless dose assessment can confirm releases are below Table 4 column "Alert" within this time period.

Table 3 Effluent Monitor Classification Thresholds Monitor Alert SAE Radwaste/Reactor Bldg.

Vent Effluent x GEMS alarm 200 x GEMS alarm >5.5E6 pCi/s N/A Main Stack Effluent 2 x GEMS alarm 200 x GEMS alarm N/A N/A Service Water Effluent 2 x ORMS High (red) 200 x ORMS High (red) N/A N/A Liquid RadWaste Effluent 2 x ORMS High (red) 200 x DRMS High (red) N/A N/A Cooling Tower Blowdown' 2 x DRMS High (red) 200 x ORMS High (red) N/A N/A May 1998 Page 49 EPMP-EPP-0102 Rev 02

5. 1.2 (Cont)

Table 4 Dose Pro 'ection Env. Measurement Classification Thresholds Alert SAE GE TEDE 10 mRem 100 mRem 1000 mRem CDE Thyroid N/A 500 mRem 5000 mRem External exposure rate 10 mRem/hr 100 mRem/hr 1000 mRem/hr Thyroid exposure rate N/A 500 mRem/hr 5000 mRem/hr (for 1 hr. of inhalation)

NUMARC IC:

Any unplanned release of gaseous or liquid radioactivity to the environment that exceeds 200 times radiological Technical Specifications for 15 minutes or longer.

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

Valid means that a radiation monitor reading has been confirmed by the operators to be correct. This event escalates from the Unusual Event by increasing the magnitude of the release by a factor of 100 over the Unusual Event level (i. e., 200 times Technical Specifications).

Prorating the 500 mR/yr basis of the 10CFR20 non-occupational DAC limits for both time (8766 hr/yr) and the 200 multiplier, the associated site boundary dose rate would be 10 mR/hr. The required release duration was reduced to 15 minutes in recognition of the increased severity.

The values for the gaseous effluent radiation monitors are based upon not exceeding 10 mR/hr at the site boundary as a result of the release.

Instrumentation that.-may be used to assess this EAL is listed below:

Hay 1998 Page 50 EPMP-EPP-0102 Rev 02

5.1.2 (Cont)

Radwaste/Reactor Building Yent Effluent Monitoring System monitor: 2RHS-CAB180 recorder: 2RMS-RR170/180 annunciator: 851248 Hain Stack Effluent Monitoring System monitor: 2RMS-CAB170 recorder: 2RMS-RR170/180 annunciator: 851256 Service Water Effluent Loop A/B Radiation monitor: 2SWP*RE146A/B recorder: 2SWP*RR146A/B annunciator: 851258 Liquid RadWaste Effluent Line monitor: LWS-RE206 annunciator: 851258 Cooling Tower Blowdown Line monitor: CWS-RE 157 annunciator; 851258 PEG Reference(s):

AAl,l Basis Reference(s):

1. N2-0P-79, Radiation Monitoring System

2. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7. 10-1

3. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87; Table 3.3.7. 1-1 May 1998 Page 51 EPHP-EPP-0102 Rev 02

5.1.3 Site Area Emer enc A valid reading from an unplanned release on any monitors Table 3 column "SAE" for > 15 min. unless dose assessment can confirm releases are below Table 4 column "SAE" within this time period.

Table 3 Effluent Monitor Classification Thresholds Monitor Alert SAE GE Radwaste/Reactor Bldg.

Vent Effluent 2 x GEMS alarm 200 x GEMS alarm >5.5E6 pCi/s N/A Main Stack Effluent 2 x GEMS alarm 200 x GEMS alarm N/A N/A Service Water Effluent 2 x ORMS High (red) 200 x DRMS High (red) N/A N/A Liquid RadWaste Effluent 2 x DRMS High (red) 200 x ORMS High (red) N/A N/A Cooling Tower Blowdown 2 x DRMS High (red) 200 x ORMS High (red) N/A N/A Table 4 Dose Pro 'ection Env. Measurement Classification Thresholds Alert SAE GE TEDE 10 mRem 100 mRem 1000 mRem CDE Thyroid N/A 500 mRem 5000 mRem External exposure rate 10 mRem/hr 100 mRem/hr 1000 mRem/hr Thyroid exposure rate N/A 500 mRem/hr 5000 mRem/hr (for 1 hr. of inhalation)

NUMARC IC:

Boundary dose resulting from an actual or imminent release of gaseous

'radioactivity exceeds 100 mRem TEDE or 500 mRem CDE Thyroid for the actual or projected duration of the release.

FPB Loss/Potential Loss:

N/A Mode Applicability:

All May 1998 Page 52 EPMP-EPP-0102 Rev 02

5.1.3

~ ~ (Cont)

Basis:

Valid means that a radiation monitor reading has been confirmed by the operators to be correct. The SAE values of Table 5. 1 are based on the boundary dose resulting from an actual or imminent release of gaseous radioactivity that exceeds 100 mR whole body or 500 mR child thyroid for the actual or projected dur ation of the release. The 100 mR integrated dose is based on the proposed 10CFR20 annual average population exposure. The 500 mR integrated child thyroid dose was established in consideration of the 1:5 ratio of the EPA Protective Action Guidelines for whole body thyroid.

These values provide a desirable gradient (one order of magnitude) between the Alert, Site Area Emergency, and General Emergency classifications. It is deemed that exposures less than this limit are not consistent with the Site Area Emergency class description.

Integrated doses are generally not monitored in real-time. In establishing this emergency action level, a duration of one hour is assumed based on site boundary doses for either whole body or child thyroid, whichever is more limiting (depends on source term assumptions).

The FSAR source terms applicable to each monitored pathway are used in determining indications for the monitors on that pathway.

The values are derived from Calculation PR-C-24-X, Rev. 2.

PEG Reference(s):

AS1.1 Basis Reference(s):

1. N2-0P-79, Radiation Monitoring System

2. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No, 50-410, 7/87, Table 3.3.7, 10-1

3. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3,3.7.1-1

4. Calculation PR-C-24-X, Rev. 2 May 1998 Page 53 EPMP-EPP-0102 Rev 02

5.2 Dose Pro 'ections Environmental Heasurements 5.2.1 Unusual Event Confirmed sample analyses for gaseous or liquid release rates > 2 x technical specifications limits for > 60 min.

NUMARC IC:

Any unplanned release of gaseous or liquid radioactivity to the environment that exceeds two times the radiological Technical Specifications for 60 minutes or longer.

FPB Loss/Potential Loss:

N/A Hode Applicability:

Al 1 Basis:

Confirmed sample analyses in excess of two times the site technical specifications that continue for 60 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of safety. The final integrated dose (which is very low in the Unusual Event emergency class) is not the primary concern; it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes. Therefore, it is not intended that the release be averaged over 60 minutes. For example, a release of 4 times T/S for 30 minutes does not exceed this initiating condition.

Further, the Emergency Director should not wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 60 minutes.

PEG Reference(s):

AU1.2 Basis Reference(s):

1. N2-0P-79, Radiation Monitoring System

2. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7. 10-1

3. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7. 1-1 Hay 1998 Page 54 EPHP-EPP-0102 Rev 02

5.2.2 Alert Confirmed sample analyses for gaseous or liquid release rates > 200 x technical specifications limits for > 15 min.

NUMARC IC:

Any unplanned release of gaseous or liquid radioactivity to the environment that exceeds 200 times radiological Technical Specifications for 15 minutes or longer.

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

Confirmed sample analyses in excess of two hundred times the site technical specifications that continue for 15 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of safety. This event escalates from the Unusual Event by increasing the magnitude of the release by a factor of 100 over the Unusual Event level (i. e., 200 times Technical Specifications).

Prorating the 500 mR/yr basis of the 10CFR20 non-occupational HPC limits for both time (8766 hr/yr) and the 200 multiplier, the associated site boundary dose rate would be 10 mR/hr. The required release duration was reduced to 15 minutes in recognition of the increased severity.

PEG Reference(s):

AA1. 2 Basis Reference(s):

1. N2-0P-79, Radiation Honitoring System

2. NUREG-1253 Technical Specifications Nine Hile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7. 10-1

3. NUREG-1253 Technical Specifications Nine Hile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7. 1-1 Hay 1998 Page 55 EPHP-EPP-0102 Rev 02

5.2.3 Alert Dose projections or field surveys resulting from actual or imminent release which indicate doses / dose rates > Table 4 column "Alert" at the site boundary or beyond Table 4 Dose Pro 'ection Env. Measurement Classification Thresholds Alert SAE GE TEDE 10 mRem 100 mRem 1000 mRem CDE Thyroid N/A 500 mRem 5000 mRem External exposure rate 10 mRem/hr 100 mRem/hr 1000 mRem/hr Thyroid exposure rate N/A 500 mRem/hr 5000 mRem/hr (for 1 hr. of inhalation)

NUMARC IC:

Any unplanned release of gaseous or liquid radioactivity to the environment that exceeds 200 times radiological Technical Specifications for 15 minutes or longer.

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

Offsite integrated doses in excess of 10 mR TEDE or dose rates in excess of 10 mR/hr TEDE represent an uncontrolled situation and hence, a potential degradation in the level of safety. This event escalates from the Unusual Event by increasing the magnitude of the release by a factor of 100 over the Unusual Event level (i, e., 200 times Technical Specifications). Prorating the 500 mR/yr basis of 10CFR20 for both time (8766 hr/yr) and the 200 multiplier, the associated site boundary dose rate would be 10 mR/hr.

May 1998 Page 56 EPMP-EPP-0102 Rev 02

5.2.3 (Cont)

As previously stated, the 10 mR/hr value is based on a proration of 200 times the 500 mR/yr basis of 10CFR20, rounded down to 10 mR/hr.

Imminent is intended to mean that a release will occur.

PEG Reference(s):

AA1.2 Basis Reference(s):

1. N2-0P-79, Radiation Monitoring System

2. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7.10-1

3. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7. 1-1 5.2.4 Site Area Emer enc Dose projections or field surveys resulting from actual or imminent release which indicate doses / dose rates > Table 4 column "SAE" at the site boundary or beyond Table 4 Dose Pro 'ection Env. Measurement Classification Thresholds Alert SAE GE TEDE 10 mRem 100 mRem 1000 mRem CDE Thyroid N/A 500 mRem 5000 mRem External exposure rate 10 mRem/hr 100 mRem/hr 1000 mRem/hr Thyroid exposure rate N/A 500 mRem/hr 5000 mRem/hr (for 1 hr. of inhalation)

NUMARC IC:

Boundary dose resulting from an actual or imminent release of gaseous radioactivity exceeds 100 mRem TEDE or 500 mRem CDE Thyroid for the actual or projected..duration of the release.

FPB Loss/Potential Loss:

N/A May 1998 Page 57 EPMP-EPP-0102 Rev 02

5.2.4 (Cont)

Node Applicability:

All Basis:

The 100 mR integrated TEDE dose in this EAL is based on the proposed 10CFR20 annual average population exposure. This value also provides a desirable gradient (one order of magnitude) between the Alert, Site Area Emergency, and General Emergency classes. It is deemed that exposures less than this limit are not consistent with the Site Area Emergency class description. The 500 mR integrated CDE thyroid dose was established in consideration of the 1:5 ratio of the EPA Protective Action Guidelines for whole body thyroid. In establishing the dose rate emergency action levels, a duration of one hour is assumed. Therefore, the dose rate EALs are based on a site boundary dose rate of 100 mR/hr TEDE or 500 mR/hr CDE thyroid, whichever is more limiting.

Imminent is intended to mean that a release will occur.

PEG Reference(s):

AS1.3 AS1.4 Basis Reference(s):

1. N2-0P-79, Radiation Monitoring System

2. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7. 10-1

3. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7. 1-1 5.2.5 General Emer enc Dose projections or field surveys resulting from actual or imminent release which indicate doses / dose rates in excess of Table 5.2 column "GE" at the site boundary or beyond Hay 1998 Page,58 EPHP-EPP-0102 Rev 02

5.2.5 (Cont)

Table 5.2 Dose Pro 'ection Env. Measurement Classification Thresholds Alert SAE GE TEDE 10 mRem 100 mRem 1000 mRem CDE Thyroid N/A 500 mRem 5000 mRem External exposure rate 10 mRem/hr 100 mRem/hr 1000 mRem/hr Thyroid exposure rate N/A 500 mRem/hr 5000 mRem/hr (for 1 hr. of inhalation)

NUNARC IC:

Boundary dose resulting from an actual or imminent release of gaseous radioactivity exceeds 1000 mRem TEDE or 5000 mRem CDE Thyroid for the actual or projected duration of the release using actual meteorology.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

The General Emergency values of Table 5.2 are based on the boundary dose resulting from an actual or imminent release of gaseous radioactivity that exceeds 1000 mR TEDE or 5000 mR CDE thyroid for the actual or projected duration of the release. The 1000 mR TEDE and the 5000 mR COE thyroid integrated dose are based on the EPA protective action guidance which indicates that public protective actions are indicated if the dose exceeds 1 rem TEDE or 5 rem CDE thyroid. This is consistent with the emergency class description for a General Emergency. This level constitutes the upper level of the desirable gradient for the Site Area Emergency. Actual meteorology is specifically identified since it gives the most accurate dose assessment. Actual meteorology (including forecasts) should be used whenever possible. In establishing the dose rate emergency action levels, a duration of one hour is assumed. Therefore, the dose rate EALs are based on a-site boundary dose rate of 1000 mR/hr TEDE or 5000 mR/hr CDE thyroid, whichever is more limiting.

Imminent is intended to mean that a release will occur.

May 1998 Page 59 EPMP-EPP-0102 Rev 02

5.2.5 (Cont)

PEG Reference(s):

AG1.3 AG1.4 Basis Reference(s):

1. N2-0P-79, Radiation Monitoring System

2. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7. 10-1

3. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 3.3.7. 1-1 6.0 ELECTRICAL FAILURES Loss of vital plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.

The events of this category have been grouped into the following two loss of electrical power types:

~ Loss of AC Power Sources: This category includes losses of onsite and/or offsite AC power sources including station blackout events.

~ Loss of DC Power Sources: This category involves total losses of vital plant 125 vdc power sources.

6.1 Loss of AC Power Sources 6.1.1 Unusual Event Loss of power for >15 min. to all:

~ Reserve Transformer A

~ Reserve Transformer B

~ Aux Boiler Transformer NUHARC IC:

Loss of all offsite power to essential busses for greater than 15 minutes.

Hay 1998 Page 60 EPHP-EPP-0102 Rev 02

6.1.

~ ~ 1 (Cont)

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

Prolonged loss of all offsite AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete loss of AC power (station blackout). Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

PEG Reference(s):

SU1.1 Basis Reference(s):

1. N2-0P-70, Station Electrical Feed and 115 kv Switchyard

2. N2-0P-100A, Standby Diesel Generators

3. N2-0P-100B, HPCS Diesel Generator 6.1.2 Alert Loss of all emergency bus AC power for >15 min.

NUNARC IC:

Loss of all offsite power and loss of all onsite AC power to essential busses during cold'hutdown, refueling or defueled mode.

FPB Loss/Potential Loss:

N/A Node Applicability:

Cold shutdown, refuel, defuel Hay 1998 Page 61 EPHP-EPP-0102 Rev 02

6.1. 2 (Cont)

Basis:

Loss of all AC power compromises all plant safety systems requiring electric power. This EAL is indicated by:

Loss of power for >15 min. to all:

Reserve Transformer A Reserve Transformer B Aux Boiler Transformer AND failure of all DGs to power any emergency bus AND failure to restore power to 2ENS*SWG101, 2ENS*SWG102 or 2ENS*SW103 in <15 min.

When in cold shutdown, refueling, or defueled mode this event is classified as an Alert. This is because of the significantly reduced decay heat, lower temperature and pressure, thus increasing the time to restore one of the emergency busses, relative to that specified for the Site Area Emergency EAL. Escalating to the Site Area Emergency, if appropriate, is by Abnormal Rad Levels/Radiological Effluent, or Emergency Director Judgment ICs. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

PEG Reference(s):

SA1. 1 Basis Reference(s):

1. N2-0P-70, Station Electrical Feed and 115 kv Switchyard

2. N2-0P-100A, Standby Diesel Generators

3. N2-0P-100B, HPCS Diesel Generator 6.1.3 Available emergency bus AC power reduced to only one of the following sources for >15 min.:

~ Reserve Transformer A

~ Reserve Transformer 8

~ Aux Boiler Transformer

~ 2EGS*EG1

~ 2EGS"EG2

~ 2EGS*EG3 Hay 1998 Page 62 EPHP-EPP-0102 Rev 02

6.1.3 (Cont)

HUNARC IC:

AC power capability to essential busses reduced to a single power source for greater than 15 minutes such that any additional single failure would result in station blackout with reactor coolant >200 F.

FPB Loss/Potential Loss:

N/A Node Applicability:

Power operation, startup/hot standby, hot shutdown Basis:

The condition indicated by this EAL is the degradation of the offsite power with a concurrent failure of one emergency generator to supply power to its emergency busses. Another related condition could be the loss of onsite emergency diesels with only one train of emergency busses being fed from offsite power. The subsequent loss of this single power source would escalate the event to a Site Area Emergency.

PEG Reference(s):

SA5.1 Basis Reference(s):

1. N2-0P-70, Station Electrical Feed and 115 kv Switchyard

2. N2-0P-lOOA, Standby Diesel Generators

3. N2-0P-100B, HPCS Diesel Generator 6.1.4 Site Area Emer enc Loss of all emergency bus AC power for >15 min.

NUNARC IC:

Loss of all offsite power and loss of all onsite AC power to essential busses with reactor coolant >200'F.

FPB Loss/Potential Loss:

H/A Node Applicability:

Power operation, startup/hot standby, hot shutdown Hay 1998 Page 63 EPHP-EPP-0102 Rev 02

6.1. 4 (Cont)

Basis:

Loss of all AC power compromises all plant safety systems requiring electric power. This EAL is indicated by:

Loss of power to Reserve Transformer A, Reserve Transformer B, and Aux Boiler Transformer AND failure of all DGs to power any emergency bus AND failure to restore power to 2ENS*SWG101, 2ENS*SWG102 or 2ENS*SWG103 in

< 15 min.

Prolonged loss of all AC power can cause core uncovery and loss of containment integrity, thus this event can escalate to a General Emergency. The time duration selected, 15 minutes, excludes transient or momentary power losses.

PEG Reference(s):

SS1.1 Basis Reference(s):

1. N2-0P-100A, Standby Diesel Generators

2. N2-0P-1008, HPCS Diesel Generator

3. N2-0P-70, Station Electrical Feed and 115 kv Switchyard

4. N2-0P-72, Standby and Emergency AC Distribution System 6.1.5 General Emer enc Loss of all emergency bus AC power AND either:

Power restoration to any emergency bus is not likely in < 4 hrs OR RPV water level cannot be restored and maintained > top of active fuel NUMARC IC:

Prolonged loss of all offsite power and prolonged loss of all onsite AC power with reactor coolant >200'F.

FPB Loss/Potential Loss:

N/A Node Applicability:

Power operation, startup/hot standby, hot shutdown Hay 1998 Page 64 EPHP-,EPP-0102 Rev 02

6.1.5

~ ~ (Cont)

Basis:

Loss of all AC power compromises all plant safety systems requiring electric power. Prolonged loss of all AC power will lead to loss of fuel clad, RCS, and containment. Although this EAL may be viewed as redundant to the RPV Water Level EALs, its inclusion is necessary to better assure timely recognition an'd emergency response.

This EAL is specified to assure that in the unlikely event of prolonged station blackout, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency occurs as early as is appropriate, based on a reasonable assessment of the event trajectory.

The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of the situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing public protective actions.

In addition, under these conditions, fission product barrier monitoring capability may be degraded, Although it may be difficult to predict when power can be restored, the Emergency Director, should declare a General Emergency based on two major considerations:

1. Are there any present indications that core cooling is already degraded to the point that Loss or Potential Loss of fission product barriers is imminent?

2. If there are no present indications of such core cooling degradation, how likely is it that power can be restored in time to assure that a loss of two barriers with a potential loss of the third barrier can be prevented?

Thus, indication of continuing core cooling degradation must be based on fission product barrier monitoring with particular emphasis on Emergency Director judgment as it relates to imminent loss or potential loss of fission product barriers and degraded ability to monitor fission product barriers.

The time to restore AC power is based on site blackout coping analysis performed in conformance with 10CFR50.63 and Regulatory Guide 1. 155, "Station Blackout", with appropriate allowance for offsite emergency response.

The terminology of "cannot be restored and maintained" is intended to be consistent with the interpretation that:

Hay 1998 Page 65 EPHP-EPP-0102 Rev 02

6.1.5 (Cont)

"The value of the identified parameter(s) is/is not able to be returned to above/below specified limits. This determination includes making an evaluation that considers both current and future systems performance in relation to the current value and trend of the parameter(s). Neither implies that the parameter must actually exceed the limit before the classification is made nor that the classification must be made before the limit is reached. Does not imply any specific time interval but does not permit prolonged operation beyond a limit without making the specified classification."

This definition would require the emergency classification be made prior to water level dropping below TAF if, based on an evaluation of the current trend of RPV water level and in consideration of current and future injection system performance, that RPV water level will not likely be restored and maintained above TAF. This definition however, also provides the latitude, based ont hat same evaluation, not to declare the SAE for those situations in which the RPV water level transiently drops below TAF in the process of RPV water level restoration.

PEG Reference(s).:

SG1.1 Basis Reference(s):

l. N2-0P-74A, Emergency DC Distribution

2. N2-0P-748, HPCS 125 vdc System

3. N2-0P-70, Station Electrical Feed and 115 kv Switchyard

4. N2-EOP-RPV, RPV Control

5. Nine Nile Point Unit 2 SBO Study, GENE-770-04-02-1290 dated 9/93 Rev 2 Page 74 EPNP-HAPP-0102 Hay 1998 Page 66 Rev 02

6.2 Loss of DC Power Sources 6.2.1 Unusual Event

( 105 vdc on 2BYS*SWGZA and B for >15 min.

NUHARC IC:

Unplanned loss of required DC power during cold shutdown or refueling mode for greater than 15 minutes.

FPB Loss/Potential Loss:

N/A Mode Applicability:

Cold shutdown, Refuel Basis:

The purpose of this EAL is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations. This EAL is intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss.

The bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value incorporates a margin of at least 15 minutes of operation before the onset of inability to operate loads.

PEG Reference(s):

SU7.1 Basis Reference(s):

1. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No, 50-410, Amendment 5, Article 4.8.2.1.d.2

2. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, Basis 3/4.8. 1-3, pg.

B3/4 8-2

3. Operations Technology BYS/BWS, Plant DC Electrical Distribution System May 1998 Page 67 EPMP-EPP-0102 Rev 02

6.2.2 Site Area Emer enc

< 105 vdc on 2BYS*SWG2A and B for > 15 min.

NUNARC IC:

Loss of vital DC power with reactor coolant >200'F.

FPB Loss/Potential Loss:,

N/A Node Applicability:

Power operation, startup/hot standby, hot shutdown Basis:

Loss of all DC power compromises ability to monitor and control plant safety functions. Prolonged loss of all DC power will cause core uncovering and loss of containment integrity when there is significant decay heat and sensible heat in the reactor system. Escalation to a General Emergency would occur by other EAL categories. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

The bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value incorporates a margin of at least 15 minutes of operation before the onset of inability to operate loads.

PEG Reference(s):

SS3.1 Basis Reference(s):

1. NUREG-1253 Technical Specifications Nine Hile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, Amendment 5, Article 4.8.2.1.d.2

2. NUREG-1253 Technical Specifications Nine Hile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Basis 3/4.8. 1-3, pg. B3/4 8-2

3. Operations Technology BYS/BWS, Plant DC Electrical Distribution System Hay 1998 Page 68 EPHP-EPP-0102 Rev 02

7.0 E UIPHENT FAILURES Numerous plant system related equipment failure events which warrant emergency classification, based upon their potential to pose actual or potential threats to plant safety, have been identified in this category.

The events of this category have been grouped into the following event types:

~ Technical S ecifications: Only one EAL falls un'der this event type related to the failure of the plant to be brought to the required plant operating condition required by technical specifications.

~ S stem Failures or Control Room Evacuation: This category includes events which are indicative of losses of operability of safety systems such as ECCS, isolation functions, Control Room habitability or cold and hot shutdown capabilities.

Loss of Indication Alarm or Communication Ca abilit : Certain events which degrade the plant operators ability to effectively assess plant conditions or communicate with essential personnel within or external to the plant warrant emergency classification.

Under this event type are losses of annunciators and/or communication equipment.

7. 1 Technical S ecifications 7.1.1 Unusual Event Plant is not brought to required operating mode within Technical Specifications LCO Action Statement Time

'e NUMARC IC:

Inability to reach required shutdown within Technical Specification Limits.

FPB Loss/Potential Loss:

N/A Node Applicability:

Power operation, startup/hot standby, hot shutdown Hay 1998 Page 69 EPMP- EPP-0102 Rev 02

7.1. 1 (Cont)

Basis:

Limiting Conditions of Operation (LCOs) require the plant to be brought to a required shutdown mode when the Technical Specification required configuration cannot be restored. Depending on the circumstances, this may or may not be an emergency or precursor to a more severe condition. In any case, the initiation of plant shutdown required by the site Technical Specification requires a one hour report under 10CFR50.72 (b) non-emergency events. The plant is within its safety envelope when being shut down within the allowable action statement time in the Technical Specifications. An immediate Notification of an Unusual Event is required when the plant is not brought to the required operating mode within the allowable action statement time in the Technical Specifications. Declaration of an Unusual Event is based on the time at which the LCO-specified action statement time period elapses under the site Technical Specifications and is not related to how long a condition may have existed. Other required Technical Specification shutdowns that involve precursors to more serious events are addressed by other EALs.

PEG Reference(s):

SU2.1 Basis Reference(s):

1. NUREG-1253 Technical Specifications Nine Nile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, article 3.0.3 7.2 S stem Failures or Control Room Evacuation 7.2. 1 Unusual Event Report of main turbine failure resulting in casing penetration or damage to turbine seals or generator seals NUMARC IC:

Natural and destructive phenomena affecting the protected area.

FPB Loss/Potential Loss:

N/A Mode Applicability:

Power operation, startup/hot standby, hot shutdown Hay 1998 Page 70 EPMP-EPP-0102 Rev 02

7.2.1

~ ~ (Cont)

Basis:

This EAL is intended to address main turbine rotating component failures of sufficient magnitude to cause observable damage to the turbine casing or to the seals of the turbine generator. Of major concern is the potential for leakage of combustible fluids (lubricating oils) and gases (hydrogen cooling) to the plant environs.

Actual fires and flammable gas build up are appropriately classified through other EALs. This EAL is consistent with the definition of an Unusual Event while maintaining the anticipatory nature desired and recognizing the risk to non-safety related equipment.

PEG Reference(s):

HU1.6 Basis Reference(s):

None 7.2.2 Alert Entry into N2-0P-78, "Remote Shutdown System" NUNARC IC:

Control room evacuation has been initiated.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

With the Control Room evacuated, additional support, monitoring and direction through the Technical Support Center and/or other Emergency Operations Center is necessary. Inability to establish plant control from outside the Control Room will escalate this event to a Site Area Emergency.

PEG Reference(s):

HA5. 1 Basis Reference(s):

1. N2-0P-78, Remote Shutdown System, Section H.2.0 Hay 1998 Page 71 EPHP-EPP-0102 Rev 02

7.2.3 Alert Reactor coolant temperature cannot be maintained < 200 F NUNARC IC:

Inability to maintain plant in cold shutdown.

FPB Loss/Potential Loss:

N/A Node Applicability:

Cold shutdown, refuel Basis:

This EAL addresses complete loss of functions required for core cooling during refueling and cold shutdown modes. Escalation to Site Area Emergency or General Emergency would be through other EALs.

A reactor coolant temperature increase that approaches or exceeds the cold shutdown technical specification limit warrants declaration of an Alert irrespective of the availability of technical specification required functions to maintain cold shutdown. The concern of this EAL is the loss of'bility to maintain the plant in cold shutdown which is defined by reactor coolant temperature and not the operability of equipment which supports removal of heat from the reactor.

This EAL does not apply during hydrostatic testing.

PEG Reference(s):

SA3.1 Basis Reference(s):

1. NUREG-1253 Technical Specifications Nine Nile Point Nuclear Stations, Unit No. 2, Oocket No. 50-410, 7/87, Amendment 26, Article 3.4.9.2

2. NUREG-1253 Technical Specifications Nine Nile Point Nuclear Stations, Unit No. 2, Oocket No. 50-410, 7/87, Table 1.2 May 1998 Page 72 EPHP-EPP-0102 Rev 02

7.2.4

~ ~ Site Area

~

Emer enc Entry into N2-0P-78, "Remote Shutdown System".

ANO Plant control cannot be established per N2-0P-78, "Remote Shutdown System" in < 15 min.

NUNARC IC:

Control room evacuation has been initiated and plant control cannot be established.

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

This EAL indicates that expeditious transfer of control of safety systems has not occurred. The time interval for transfer is based on analysis or assessments as to how quickly control must be reestablished without core uncovering and/or core damage. In cold shutdown and refueling modes, operator concern is directed toward maintaining core cooling such as is discussed in Generic Letter 88-17, "Loss of Decay Heat Removal." In power operation , hot standby, and hot shutdown modes, operator concern is primarily directed toward monitoring and controlling plant parameters dictated by the EOPs and thereby assuring fission product barrier integrity.

PEG Reference(s):

HS2.1 Basis Reference(s):

1. Generic Letter 88-17, "Loss of Decay Heat Removal"

2. N2-0P-78, Remote Shutdown System, Section H.2.0

3. NMP-2 FSAR Section 98.8.2.2, Safe Shutdown Scenario, pg. 9B.8-5a, May 1998 Page 73 EPMP-EPP-0102 Rev 02

7.3 Loss of Indications Alarm Communication Ca abilit 7.3.1 Unusual Event Unplanned loss of annunciators or indicators on any of the following panels for > 15 min.:

~ 2CEC*PNL601

~ 2CEC*PNL602

~ 2CEC*PNL603

~ 2CEC*PNL852

~ 2CEC*PNL851 AND Increased surveillance is required for safe plant operation NUMARC IC:

Unplanned loss of most or all safety system annunciation or indication in the control room for greater than 15 minutes.

FPB Loss/Potential Loss:

N/A Mode Applicability:

Power operation, startup/hot standby, hot shutdown Basis:

This EAL recognizes the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment. Recognition of the availability of computer based indication equipment is considered (SPDS, plant computer, etc.).

"Unplanned" loss of annunciators or indicators excludes scheduled maintenance and testing activities.

It is not intended that plant personnel perform a detailed count of instrumentation lost but the use of judgment by the Shift Supervisor as the threshold for determining the severity of the plant conditions.

This judgment is supported by the specific opinion of the Shift Supervisor that additional operating personnel will be required to provide increased monitoring of system operation to safely operate the plant.

May 1998 Page 74 EPMP-EPP-0102 Rev 02

7.3.1

~ ~ (Cont)

It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptable power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by their specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10CFR50.72. If the shutdown is not in compliance with the Technical Specification action, the Unusual Event is based on EAL 7. 1. 1, Inability to Reach Required Shutdown Within Technical Specification Limits.

Annunciators or indicators for this EAL must include those identified in the Abnormal Operating procedures, in the Emergency Operating Procedures, and in other EALs (e. g., area, process, and/or effluent rad monitors, etc.).

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Due to the limited number of safety systems in operation during cold shutdown, refueling, and defueled modes, this EAL is not applicable during these modes of operation.

This Unusual Event will be escalated to an Alert if a transient is in progress during the loss of annunciation or indication.

PEG Reference(s):

SU3.1 Basis Reference(s):

l. USAR Figure 1.2-15, Control Room layout

2. N2-0P-91A, Process Computer

3. N2-0P-91B, Safety Parameter Display System (SPDS)

May 1998 Page 75 EPMP-EPP-0102 Rev 02

7.3.2 Unusual Event Loss of all communications capability affecting the ability to either:

Perform routine onsite operations OR Notify offsite agencies or personnel NUMARC IC:

Unplanned loss of all onsite or offsite communications capabilities.

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

The purpose of this EAL is to recognize a loss of communications capability that either defeats the plant operations staff ability to" perform routine tasks necessary for plant operations or the ability to communicate problems with offsite authorities. The loss of offsite communications ability is expected to be significantly more comprehensive than the condition addressed by 10CFR50.72.

The onsite communications loss must encompass the loss of all means of routine communications, Table 7. 1.

Table 7.1 Communications S stems

~Sstem Onsite Offsite Dial telephones x SPC system x M/CC system X.

PP/PA system x Hand-Held Portable radio x Red phone to USNRC-Bethesda Black phone to USNRC-King of Prussia Black phone direct to JAFNPP PBX REGS Health physics-network and FTS 2000 UHF radios The offsite communications loss must encompass the loss of all means of communications with offsite authorities, Table 7. 1. This EAL is intended to be used only when extraordinary means are being utilized to make communications possible (relaying of information from radio transmissions, individuals being sent to offsite locations, etc.).

May 1998 Page 76 EPMP-EPP-0102 Rev 02'

7.3.2

~ ~ (Cont)

PEG Reference(s):

SU6.1 Basis Reference(s):

1. N2-0P-76, Plant Communications 7.3.3 Alert Unplanned loss of annunciators or indicators on any of the following panels for > 15 min.:

~ 2CEC*PNL601

~ 2CEC*PNL602

~ 2CEC*PNL603

~ 2CEC*PNL852

~ 2CEC*PNL851 AND increased surveillance is required for safe plant operation AND either:

Plant transient in progress OR Plant computer and SPDS are unavailable NUMARC IC:

Unplanned loss of most or all safety system annunciation or indication in control room with either (1) a significant transient in progress, or (2) compensatory non-alarming indicators are unavailable.

FPB Loss/Potential Loss:

N/A Mode Applicability:

Power operation, startup/hot standby, hot shutdown Basis:

This EAL recognizes the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or i'ndication equipment during a transient. Recognition of the availability of computer based indication equipment is considered (SPDS, plant computer, etc.).

"Unplanned" loss of annunciators or indicators does not include scheduled maintenance and testing activities.

Hay 1998 Page 77 EPMP-EPP-0102 Rev 02

7.3.3 (Cont)

It is not intended that plant personnel perform a detailed count of the instrumentation lost but the use of the value as a judgment by the shift supervisor as the threshold for determining the severity of the plant conditions. This judgment is supported by the specific opinion of the Shift Supervisor that additional operating personnel will be required to provide increased monitoring of system operation to safely operate the plant.

It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptable power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10CFR50.72.

Annunciators or indicators for this EAL includes those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures, and in other EALs (e. g., area, process, and/or effluent rad monitors, etc.).

"Significant transient" includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, ECCS injections, or thermal power oscillations of 10N or greater.

If both a major portion of the annunciation system and all computer monitoring are unavailable to the extent that the additional operating personnel are required.to monitor indications, the Alert is required.

Due to the limited number of safety systems in operation during cold shutdown, refueling and defueled modes, no EAL is indicated during these modes of operation.

This Alert will be escalated to a Site Area Emergency if the operating crew cannot monitor the transient in progress.

PEG Reference(s):

SA4.1 Basis Reference(s):

l. USAR Figure 1.2-15, Control Room layout

2. N2-0P-91A, Process Computer

3. N2-0P-918, Safety Parameter Display System (SPDS)

May 1998 Page 78 EPMP-EPP-0102 Rev 02

7.3.4

~ ~ Site Area Emer enc Loss of annunciators or indicators on any of the following panels:

~ 2CEC*PNL601

~ 2CEC*PNL602

~ 2CEC*PNL603

~ 2CEC*PNL852

~ 2CEC*PNL851 AND plant computer and SPDS are unavailable AND indications to monitor all RPV and primary containment EOP parameters are lost AND plant transient is in progress NUHARC IC:

Inability to monitor a significant transient in progress.

FPB Loss/Potential Loss:

N/A Node Applicabi,lity:

Power operation, startup/hot standby, hot shutdown Basis:

This EAL recognizes the inability of the Control Room staff to monitor the plant response to a transient., A Site Area Emergency is considered to exist if the Control Room staff cannot monitor safety functions needed for protection of the public.

Annunciators for this EAL should be limited to include those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures, and in other EALs (e. g., rad monitors, etc.).

"Significant transient" includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25X thermal power change, ECCS injections, or thermal power oscillations of lOX or greater.

Indications needed to monitor safety functions necessary for protection of the public must include Control Room indications, computer generated indications and dedicated annunciation capability.

The specific indications should be those used to determine such functions as the ability to shut down the reactor, maintain the core cooled and in a eoolable geometry, to remove heat from the core, to maintain the reactor coolant system intact, and to maintain containment intact.

Nay 1998 Page 79 EPMP-EPP-0102 Rev 02

7.3.4 (Cont)

"Planned" actions are excluded from the is EAL since the loss of instrumentation of this magnitude is of such significance during a transient that the cause of the loss is not an ameliorating factor.

PEG Reference(s):

SS6.1 Basis Reference(s):

1. N2-EOP-PC, Primary Containment Control

2. N2-EOP-RPV, RPV Control

3. N2-0P-91A, Process Computer

4. N2-0P-91B, Safety Parameter Display System (SPDS)

5. USAR Figure 1.2-15, Control Room layout 8.0 HAZARDS Hazards are those non-plant system related events which can directly or indirectly impact plant operation or reactor plant and personnel safety.

The events of this category have been grouped into the following types:

Securit Threats: This category includes unauthorized entry attempts into the Protected Area as well as bomb threats and sabotage attempts. Also addressed are actual security compromises threatening loss of physical control of the plant.

Fire or Ex losion: Fires can pose significant hazards to personnel and reactor safety. Appropriate for classification are fires within the site Protected Area or which may affect operability of vital equipment.

Han-made Events: Han-made events are those non-naturally occurring events which can cause damage to plant facilities such as aircraft crashes, missile impacts, toxic or flammable gas leaks or explosions from whatever source.

Natural Events: Events such as hurricanes, earthquakes or tornadoes which have potential to cause damage to plant structures or equipment significant enough to threaten personnel or plant safety.

Hay 1998 Page 80 EPHP-EPP-0102 Rev 02

8.1 Securit Threats 8.l. 1 Unusual Event Bomb device or other indication of attempted sabotage discovered within plant Protected Area OR Any security event which represents a potential degradation in the level of safety of the plant.

NUNARC IC:

Confirmed security event which indicates a potential degradation in the level of safety of the plant.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This EAL is based on the Nine Nile Point Nuclear Station Physical Security and Safeguards Contingency Plans. Security events which do not represent at least a potential degradation in the level of safety of the plant, are reported under 10CFR73.71 or in some cases under 10CFR50.72.

I The plant Protected Area boundary is within the security isolation zone and is defined in the security plan.

PEG Reference(s):

HU4.1 HU4.2 Basis Reference(s):

1. Nine Mile Point Nuclear Station Physical Security and Safeguards Contingency Plans.

Nay 1998 Page 81 EPNP-EPP-0102 Rev 02

8.1.2 Alert Intrusion into plant Protected Area by an adversary OR Any security event which represents an actual substantial degradation of the level of safety of the plant.

NUHARC IC:

Security event in a plant protected area.

FPB Loss/Potential Loss:

N/A Node Applicability:

Al 1 Basis:

This class of security events represents an escalated threat to plant safety above that contained in the Unusual Event. For the purposes of this EAL, the intrusion by an adversary inside the Protected Area boundary can be considered a significant security threat. Intrusion into a vital area by an adversary will escalate this event to a Site Area Emergency.

NMP-I and NHP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also see S&W Drawing No. 12187-SK-032483-25, Issue No.

1, Site Facilities Layout Status as of 8/1/89.

PEG Reference(s):

HA4.1 HA4.2 Basis Reference(s):

1. Nine Nile Point Nuclear Station Physical Security and Safeguards Contingency Plans

2. S8W Drawing No. 12187-SK-032483-25, Issue No. 1,'Site Facilities Layout Status as of 8/1/89 Hay 1998 Page 82 EPHP-EPP-0102 Rev 02

8.1.3 Site Area Emer enc Intrusion into a plant security vital area by an adversary OR Any security event which represents actual or likely failures of plant systems needed to protect the public.

NUMARC IC:

Security event in a plant vital area.

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

This class of security events represents an escalated threat to plant safety above that contained in the Alert in that an adversary has progressed from the Protected Area to the vital area.

PEG Reference(s):

HSl. 1 HS1.2 Basis Reference(s):

1. Nine Mile Point Nuclear Station Physical Security and Safeguards Contingency Plans 8.1.4 General Emer enc Security event which results in either:

Loss of plant control from the Control Room OR Loss of remote shutdown capability NUMARC IC:

Security event resulting in loss of ability to reach and maintain cold shutdown.

May 1998 Page 83 EPHP-EPP-0102 Rev 02

1'.1.4 (Cont)

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This EAL encompasses conditions under which unauthorized personnel have taken physical control of vital areas required to reach and maintain safe shutdown.

PEG Reference(s):

HGl. 1 HGl.'

Basis Reference(s):

None 8.2 Fire or Ex losion 8.2.1 Unusual Event Confirmed fire in or contiguous to any plant area, Table 5 or Table 6, not extinguished in < 15 min. of Control Room notification I

Table 5 Plant Areas

~ Service Building

~ 115 KV Switchyard

~ 345 KV Switchyard Table 6 Plant Vital Areas Reactor Building Control Room Diesel Generator Engine and Board Rooms Standby Switchgear and Battery Rooms HPCS Switchgear and Battery Rooms Remote Shutdown Rooms Control Building HVAC Rooms Service Water Pump Rooms Electrical Protection Assembly Room PGCC Relay Room Hay 1998 Page 84 EPHP-EPP-0102 I Rev 02.

8.2.1 (Cont)

NUMARC IC:

Fire within protected area boundary not extinguished within 15 minutes of detection.

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

The purpose of this EAL is to address the magnitude and extent of fires that may be potentially significant precursors to damage to safety systems, This excludes such items as fires within administration buildings, waste-basket fires, and other small fires of no safety consequence.

PEG Reference(s):

HU2. 1 Basis Reference(s):

1. Nine Mile Point Nuclear Station Physical Security and Safeguards Contingency Plans

2. NUREG 0737, Section II.B.2-2 8.2.2 Alert Fire or explosion in any plant area which results in damage to plant equipment or structures needed for safe plant operation, Table 5 or Table 6 Table 5 Plant Areas Service Building 115 KV Switchyard 345 KV Switchyard May 1998 Page 85 EPMP-EPP-0102 Rev, 02

8.2.2 (Cont)

Table 6 Plant Vital Areas Reactor Building Control Room Diesel Generator Engine and Board Rooms Standby Switchgear and Battery Rooms HPCS Switchgear and Battery Rooms Remote Shutdown Rooms Control Building HVAC Rooms Service Water Pump Rooms Electrical Protection Assembly Room PGCC Relay Room NUNARC IC:

Fire or explosion affecting the operability of plant safety systems required to establish or maintain safe shutdown.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

The listed areas contain functions and systems required for the safe shutdown of the plant. The NNP-2 safe shutdown analysis was consulted for equipment and plant areas required for the applicable mode.

With regard to explosions, only those explosions of sufficient force to damage permanent structures or equipment required for safe operation within the identified plant areas should be considered. As used here, an explosion is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts significant energy to nearby structures and materials. No attempt is made in this EAL to assess the actual magnitude of the damage. The declaration of an Alert and the activation of the TSC will provide the Emergency Director with the resources needed to perform damage assessments. The Emergency Director also needs to consider any security aspects of the explosions.

PEG Reference(s):

HA2.1 Nay 1998 Page 86 EPNP-EPP-0102 Rev 02

8.2.2 (Cont)

Basis Reference(s):

1. N2-0P-47, Fire Detection

2. USAR, Figure 98.6-1

3. USAR, Section 98

4. NUREG 0737, Section, II.B.2-2 8.3 Man-Made Events 8.3. 1 Unusual Event Vehicle crash into or projectile which impacts plant structures or systems within Protected Area boundary NUMARC IC:

Natural and destructive phenomena affecting the protected area.

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

The Protected Area boundary is within the security isolation zone and is defined in the site security plan. NMP-1 and NMP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also, refer to S&W Drawing No. 12187-SK-032483-25, Issue No, 1, Site Facilities Layout Status as of 8/1/89.

This EAL addresses such items as plane, helicopter, train, car, truck, or barge crash, or impact of other projectiles that may potentially damage plant structures containing functions and systems required for safe shutdown of the plant. If the crash is confirmed to affect a plant vital area, the event may be escalated to Alert.

For the purpose of this EAL, a plant structure is any permanent building or structure which houses plant process / support systems and equipment. Administrative buildings, support buildings/trailers or other non plant operations related structures are not intended to be included here.

May 1998 Page 87 EPHP-EPP-0102 Rev 02

8.3.1 (Cont)

PEG Reference(s):

HU1.4 Basis Reference(s):

l. USAR Figure 1.2-2 Station Arrangement

2. S&W Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89 8.3.2 Unusual Event Report by plant personnel of an explosion within Protected Area boundary resulting in visible damage to permanent structures or equipment NUMARC IC:

Natural and destructive phenomena affecting the protected area.

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

The Protected Area boundary is within the security isolation zone and is defined in the site security plan. NMP-1 and NMP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also, refer to S&W Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89.

For this EAL, only those explosions of sufficient force to damage permanent structures or equipment within the Protected Area should be considered. As used here, an explosion is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts significant energy to near by structures and materials. No attempt is made in this EAL to assess the actual magnitude of the damage. The occurrence of the explosion with reports of evidence of damage (e. g., deformation, scorching) is sufficient for declaration. The Emergency Director also needs to consider any security aspects of the explosion.

May 1998 Page 88 EPMP-EPP-0102 Rev 02

8.3.2 (Cont)

PEG Reference(s):

HU1.5 Basis Reference(s):

1. USAR Figure 1.2-2 Station Arrangement

2. S8W Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89 8.3.3 Unusual Event Report or detection of a release of toxic or flammable gases that could enter or have entered within the Protected Area boundary in amounts that could affect the health of plant personnel or safe plant operation OR Report by local, county or state officials for potential evacuation of site personnel based on offsite event NUNARC IC:

Release of toxic or flammable gases deemed detrimental to safe operation of the plant.

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

This EAL is based on releases in concentrations within the site boundary that will affect the health of plant personnel or affecting the safe operation of the plant with the plant being within the evacuation area of an offsite event (i. e., tanker truck accident releasing toxic gases, etc.). The evacuation area is as determined from the DOT Evacuation Tables for Selected Hazardous Materials, in the DOT Emergency Response Guide for Hazardous Materials.

May 1998 Page 89 EPMP-EPP-0102 Rev 02

8.3.3 (Cont)

NMP-I and NHP-2 share no common safety systems, but their respective Protected Area boundaries share common borders in some places.

Therefor e it is possible that a toxic or flammable gas incident happening on one site could affect the other site.

Should an explosion occur within a specified plant area, an Alert would be declared based on EAL 8.2.2 PEG Reference(s):

HU3. 1 HU3.2 Basis Reference(s):

None 8.3.4 Alert Vehicle crash or projectile impact which precludes personnel access to or damages equipment in plant vital areas, Table 6.

Table 6 Plant Vital Areas Reactor Building Control Room Diesel Generator Engine and Board Rooms Standby Switchgear and Battery Rooms HPCS Switchgear and Battery Rooms Remote Shutdown Rooms Control Building HVAC Rooms Service Water Pump Rooms Electrical Protection Assembly Room PGCC Relay Room NUMARC IC:

Natural and destructive. phenomena affecting the plant vital area.

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Hay 1998 Page 90 EPHP-EPP-0102 Rev 02

8.3.4

~ ~ (Cont)

~

Basis:

This EAL addresses events that may have resulted in a plant vital area being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage.

NHP-I and NHP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also see S&W Drawing No. 12187-SK-032483-25, Issue No.

1, Site Facilities Layout Status as of 8/1/89.

This EAL addresses such items as plane, helicopter, train, car, truck; or barge crash, or impact of 'other projectiles into a plant vital area.

PEG Reference(s):

HA1. 5 Basis Reference(s):

1. USAR Figure 1.2-2 Station Arrangement

2. S&W Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89

3. NUREG 0737, Section II.B.2-2 8.3.5 Alert Confirmed report or detection of toxic or flammable gases within a plant vital area, Table 6, in concentrations that will be life threatening to plant personnel or preclude access to equipment needed for safe plant operation Table 6 Plant Vital Areas Reactor Building Control Room Diesel Generator Engine and Board Rooms Standby Switchgear and Battery Rooms HPCS Switchgear and Battery Rooms Remote Shutdown Rooms Control Building HVAC Rooms Service Water Pump Rooms Electrical Protection Assembly Room PGCC Relay Room May 1998 Page 91 EPHP-EPP-0102 Rev 02

8.3.5 (Cont)

NUHARC IC:

Release of toxic or flammable gases within a facility structure which jeopardizes operation of systems required to maintain safe operations or to establish or maintain cold shutdown.

FPB Loss/Potential Loss:

N/A Hode Applicability:

All Basis:

This EAL is based on gases that have entered a plant structure precluding access to equipment necessary for the safe operation of the plant. This EAL applies to buildings and areas contiguous to plant vital areas or other significant buildings or areas. The intent of this EAL is not to include buildings (i. e., warehouses) or other areas that are not contiguous or immediately adjacent to plant vital areas. It is appropriate that increased monitoring be done to ascertain whether consequential damage has occurred.

PEG Reference(s):

HA3.1 HA3.2 Basis Reference(s):

l. USAR Figure 1.2-2 Station Arrangement

2. NUREG 0737, Section II.B.2-2 8.4 Natural Events 8.4. 1 Unusual Event Earthquake felt in plant based upon a consensus of Control Room Operators on duty.

AND either:

NHP-2 seismic instrumentation actuated OR confirmation of earthquake received on NHP-1 or JAFNPP seismic instrumentation Hay 1998 Page 92 EPHP-EPP-0102 Rev 02

8.4. 1 (Cont)

NUHARC IC:

Natural and destructive phenomena affecting the protected area.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

NHP-2 seismic instrumentation actuates at 0.01 g causing:

~ Power to remote acceleration sensor units

~ Activation of HRS1 recorders

~ EVENT alarm light on PWRS1 to light

~ Annunciator 842121 on panel 2CEC-PNL842 to be received

~ EVENT INDICATOR on PWRSl to turn from black to white Damage to some portions of the site may occur but it should not affect ability of safety functions to operate. Hethods of detection can be based on instrumentation validated by a reliable source, operator assessment, or indication received from NHP-1 or JAFNPP instrumentation. As defined in the EPRI-sponsored "Guidelines for Nuclear Plant Response to an Earthquake", dated October 1989, a "felt earthquake" is:

"An earthquake of sufficient intensity such that: (a) the inventory ground motion is felt at the nuclear plant site and recognized as an earthquake based on a consensus of Control Room operators on duty at the time, and (b) for plants with operable seismic instrumentation, the seismic switches of the plant are activated. For most plants with seismic instrumentation , the seismic switches are set at an acceleration of about 0.01 g."

PEG Reference(s):

HUI. I Hay 1998 Page 93 EPHP-EPP-0102 Rev 02

8.4.1 (Cont)

Basis Reference(s):

1. N2-0P-90, Seismic Monitoring

2. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, article 3.3.7.2

3. EPRI document, "Guidelines for Nuclear Plant Response to an Earthquake", dated October 1989 8.4.2 Unusual Event Report by plant personnel of tornado striking within plant Protected Area boundary NUMARC IC:

Natural and destructive phenomena affecting the protected area.

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

This EAL is based on the assumption that a tornado striking (touching down) within the Protected Area boundary may have potentially damaged plant structures containing functio'ns or systems required for safe shutdown of the plant. If such damage is confirmed visually or by other in-plant indications, the event may be escalated to Alert.

NMP-1 and NMP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also see S&W Drawing No. 12187-SK-032483-25, Issue No.

1, Site Facilities Layout Status as of 8/1/89, PEG Reference(s):

HU1.2 Basis Reference(s):

1. USAR Figure 1.2-1

2. S&W Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89 May 1998 Page 94 EPMP-EPP-0102 Rev 02

8.4.3 Unusual Event Lake water level > 248 ft OR intake water level < 237 ft NUNARC IC:

Natural and destructive phenomena affecting the protected area.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This covers high and low lake water level conditions that could be precursors of more serious events. The high lake level is based upon the maximum attainable uncontrolled lake water level as specified in the FSAR. The low level is based on intake water level and corresponds to the design minimum lake level.

PEG Reference(s):

HU1.7 Basis Reference(s):

1. FSAR Section 2.4. 1.2 and 2.4. 11.2 8.4.4 Alert Earthquake felt in plant based upon a consensus of Control Room Operators on duty AND NMP-2 seismic instrumentation indicates > 0.075 g NUNARC IC:

Natural and destructive phenomena affecting the plant vital area.

FPB Loss/Potential Loss:

N/A May 1998 Page 95 EPMP-EPP-0102 Rev 02

8.4.4 (Cont)

Mode Applicability:

All Basis:

This EAL addresses events that may have resulted in a plant vital area being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage.

This EAL is based on the FSAR design operating basis earthquake of 0.075 g. Seismic events of this magnitude can cause damage to plant safety functions.

PEG Reference(s):

HA1. 1 Basis Reference(s):

1. N2-0P-90, Seismic Monitoring

2. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, article 3.3.7.2 8.4.5 Alert Sustained winds > 90 mph OR Tornado strikes a plant vital area, Table 6 Table 6 Plant Vital Areas Reactor Building Control Room Diesel Generator Engine and Board Rooms Standby Switchgear and Battery Rooms HPCS Switchgear arid Battery Rooms Remote Shutdown Rooms Control Building HVAC Rooms Service Water Pump Rooms Electrical Protection Assembly Room PGCC Relay Room NUMARC IC:

Natural and destructive phenomena affecting the plant vital area.

Hay 1998 Page 96 EPHP-EPP-0102 Rev 02

4.8.5 (Cont)

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

This EAL addresses events that may have resulted in a plant vital area being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage.

This EAL is based on the FSAR design basis of 90 mph. Wind loads of this magnitude can cause damage to safety functions.

NMP-1 and NMP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also see S&W Drawing No. 12187-SK-032483-25, Issue No.

1, Site Facilities Layout Status as of 8/1/89.

PEG Reference(s):

HA1.2 Basis Reference(s):

1. FSAR 3.3, Wind and Tornado Loadings, Amendment 26

2. FSAR Table 1.3-7, Amendment 4

3. NUREG 0737, Section II.B.2-2 May 1998 Page 97 EPMP-EPP-0102 Rev 02

8.4. 6 Al ert Any natural event which results in a report of visible structural damage or assessment by Control Room personnel of actual damage to equipment needed for safe plant operation, Table 6.

Table 6 Plant Vital Areas Reactor Building Control Room Diesel Generator Engine and Board Rooms Standby Switchgear and Battery Rooms HPCS Switchgear and Battery Rooms Remote Shutdown Rooms Control Building HVAC Rooms Service Water Pump Rooms "

Electrical Protection Assembly Room PGCC Relay Room NUHARC IC:

Natural and destructive phenomena affecting the plant vital area.

FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This EAL addresses events that may have resulted in a plant vital area being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage.

This EAL specifies areas in which structures containing systems and functions required for safe shutdown of the plant are located.

.Hay 1998 Page 98 EPHP-EPP-0102 Rev 02

8.4.6

~ ~ (Cont)

PEG Reference(s):

HA1.3 Basis Reference(s):

1. USAR Figure 1.2-2 Station Arrangement

2. NUREG 0737, Section II.B.2-2 8.4.7 Alert Lake water level > 254 ft OR Intake water level < 233 ft NUMARC IC:

Natural and destructive phenomena affecting the plant vital area.

FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

This EAL addresses events that may have resulted in a plant vital area being subjected to levels beyond design limits, and thus damage may be assumed to have occurred to plant safety systems. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage.

This EAL covers high and low lake water level conditions that exceed levels which threaten vital equipment. The high lake level is based upon the maximum probable flood level. The low forebay water level corresponds to the minimum intake bay water level which provides adequate submergence to the service water pumps.

PEG Reference(s):

HA1. 7 Hay 1998 Page 99 EPHP-EPP-0102 Rev 02

8.4.7 (Cont)

Basis Reference(s):

1. FSAR Section 2.4.5.2

2. FSAR Section 2.4.1.1

3. FSAR Section 9.2.5.3.1 9.0 OTHER The EALs defined in categories 1.0 through 8.0 specify the predetermined symptoms or events which are indicative of emergency or potential emergency conditions, and which warrant classification.

While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Shift Supervisor or Site Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria, based upon their judgment.

9.1.1 Unusual Event Any event, as determined by the Shift Supervisor or Site Emergency Director, that could lead to or has led to a potential degradation of the level of safety of the plant.

NUMARC IC:

Emergency Director judgement FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Site Emergency Director to fall under the Unusual Event emergency class.

May 1998 Page 100 EPHP-EPP-0102 Rev 02

9.1.1 (Cont)

From a broad perspective, one area that may warrant Site Emergency Director judgment is related to likely or actual breakdown of site specific event mitigating actions. Examples to consider include inadequate emergency response procedures, transient response either unexpected or not understood, failure or unavailability of emergency systems during an accident in excess of that assumed in accident analysis, or insufficient availability of equipment and/or support personnel. Another example to consider would be exceeding a plant safety limit as defined in Technical Specifications.

PEG Reference(s):

HU5.1 Basis Reference(s):

None

9. 1. 2 Unusual Event Any event, as determined by the Shift Supervisor or Site Emergency Director, that could lead to or has led to a loss or potential loss of containment. (Attachment 2)

Loss of containment indicators may include a rapid unexplained decrease following initial increase in containment pressure.

NUMARC IC:

N/A FPB Loss/Potential Loss:

Containment loss/potential loss Mode Applicability:

Power operations, Startup/Hot standby, Hot Shutdown Basis:

This EAL addresses any other factors that are to be used by the Site Emergency Director in determining whether the containment barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also- be incorporated in this EAL as a factor in Emergency Director judgment that the barrier may be considered lost or potentially lost.

Nay 1998 Page 101 EPMP-EPP-0102 Rev 02

PEG Reference(s):

PC6.1 Basis Reference(s):

None 9.1.3 Alert Any event, as determined by the Shift Supervisor or Site Emergency Director, that could cause or has caused actual substantial degradation of the level of safety of the plant.

NUMARC IC:

Emergency Director judgement FPB Loss/Potential Loss:

N/A Mode Applicability:

All Basis:

This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Site Emergency Director to fall under the Alert emergency class.

PEG Reference(s):

HA6.1 Basis Reference(s):

None 9.1.4 Alert Any event, as 'determined by the Shift Supervisor or Site Emergency Director, that could lead or has led to a loss or potential loss of either fuel clad or-RCS barrier . (Attachment 2)

NUMARC IC:

N/A May 1998 Page 102 EPHP-EPP-0102 Rev 02

9.1.4 (Cont)

FPB Loss/Potential Loss:

Loss or potential loss of either fuel clad or RCS barrier Node Applicability:

Power operations, Startup/Hot standby, Hot Shutdown Basis:

This EAL addresses any other factors that are to be used by the Site Emergency Director in determining whether the fuel clad or RCS barriers are lost or potentially lost, In addition, the inability to monitor the barriers should also be considered in this EAL as a factor in Emergency Director judgment that the barriers may be considered lost or potentially lost.

PEG Reference(s):

FC5.1 RCS6.1 Basis Reference(s):

None 9.1.5 Site Area Emer enc As determined by the Shift Supervisor or Site Emergency Director, events are in progress which indicate actual or likely failures of plant systems needed to protect the public. Any releases are not expected to result in exposures which exceed EPA PAGs.

NONARC IC:

Emergency Director judgement FPB Loss/Potential Loss:

N/A Node Applicability:

All Basis:

This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Site Emergency Director to fall under the emergency class description for Site Area Emergency.

Hay 1998 Page 103 EPNP-EPP-0102 Rev 02

9. 1. 5 (Cont)

PEG Reference(s):

HS3.1 Basis Reference(s):

None

9. 1.6 Site Area Emer enc Any event, as determined by the Shift Supervisor or Site Emergency Director, that could lead or has led to either:

Loss or potential loss of both fuel clad and RCS barrier (Attachment 2)

OR Loss or potential loss of either fuel clad or RCS barrier in conjunction with a loss of containment (Attachment 2)

Loss of containment indicators may include a rapid unexplained decrease following initial increase in containment pressure NUNARC IC:

N/A FPB Loss/Potential Loss:

Loss or potential loss of both fuel clad and RCS barriers OR Loss or potential loss of either fuel clad or RCS barrier in conjunction with a loss of containment Node Applicability:

Power operations, Startup/Hot standby, Hot Shutdown Basis:

This EAL addresses unanticipated conditions affecting fission product barriers which are not addressed explicitly elsewhere. Declaration of an emergency is warranted because conditions exist which are believed by the Site Emergency Director to fall under the emergency class description for Site Area Emergency.

Rapid unexplained loss of pressure (i. e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of containment integrity. Drywell pressure should increase as a result of mass and energy release into containment from a LOCA. Thus, drywell pressure not increasing under these conditions indicates a loss of containment integrity.

May 1998 Page, 104 EPHP-EPP-0102 Rev 02

9.1.6

~ ~ (Cont)

PEG Reference(s):

FC5.1 RCS6.1 PC6.1 PC1. 1 PC1.2 Basis Reference(s):

None 9.1.7 General Emer enc As determined by the Shift Supervisor or Site Emergency Director, events are in progress which indicate actual or imminent core damage and the potential for a large release of radioactive material in excess of EPA PAGs outside the site boundary.

NUMARC IC:

Emergency Director judgement FPB Loss/Potential Loss:

N/A Mode Applicability:

Al 1 Basis:

This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Site Emergency Director to be consistent with the General Emergency classification description.

Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the site boundary.

PEG Reference(s):

HG2.1 Basis Reference(s):

None Hay 1998 Page 105 EPMP-EPP-0102 Rev 02

9. 1.8 General Emer enc Any event, as determined by the Shift Supervisor or Site Emergency Director, that could lead or has led to a loss of any two fission product barriers and loss or potential loss of the third. (Attachment 2)

Loss of containment indicators may include a rapid unexplained decrease following initial increase in containment pressure NUNARC IC:

N/A FPB Loss/Potential Loss:

Loss of any two fission product barriers and loss or potential loss of the third Node Applicability:

Power operations, Startup/Hot standby, Hot Shutdown Basis:

This EAL addresses unanticipated conditions affecting fission product barriers which are not addressed explicitly elsewhere. Declaration of an emergency is warranted because conditions exist which are believed by the Site Emergency Director to fall under the emergency class description for the General Emergency class.

Rapid unexplained loss of pressure (i. e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of containment integrity. Drywell pressure should increase as a result of mass and energy release into containment from a LOCA. Thus, drywell pressure not increasing under these conditions indicates a loss of containment integrity.

PEG Reference(s):

FC5.1 RCS6.1 PC6.1 PC1.1 PC1.2 Basis Reference(s):,.

None Hay 1998 Page 106 EPNP-EPP-0102 Rev 02

ATTACHlVIENT 2 FISSION PRODUCT BARRIER LOSS Bc POTENTIAL LOSS INDICATORS May 1998 Page 107 EPMP-EPP-0102 Rev 02

Fission Product Barrier Loss/Potential Loss Matrix (Those thresholds for which loss or potential is determined to be imminent, classify as though the threshold(s) has been exceeded)

Fuel Cladding Potential Loss

~ RPV water level cannot be restored and maintained > top of active fuel

~ Emergency Director Judgment Loss

~ RPV water level cannot be restored and maintained > top of active fuel

~ Coolant activity > 300 pCi/gm I-131 equivalent

~ Drywell radiation > 3100 R/hr

~ Emergency Director Judgment RCS Potential Loss

~ RCS leakage greater than 50 gpm inside the drywell

~ Primary system is discharging into RB AND RB area radiation levels are >8.0 R/hr. in more than one area, N2-EOP-SC

~ Primary system is discharging into RB AND RB area temperatures are >212'F in more than one area, N2-EOP-SC

~ Emergency Director Judgment Loss RPV water level cannot be restored and maintained > top of active fuel

~ Primary containment pressure cannot be maintained < 1.68 psig due to coolant leakage

~ Drywell radiation > 41 R/hr

~ Emergency Director Judgment May 1998 Page 108 EPMP-EPP-0102 Rev 02

Fission Product Barrier Loss/Potential Loss Hatrix (Those thresholds for which loss or potential is determined to be imminent, classify as though the threshold(s) has been exceeded)

Containment Potential Loss

~ Drywell radiation > 5.2E6 R/hr

~ Emergency Director Judgment Loss

~ Primary containment venting is required due to PCPL

~ Primary containment venting is required due to combustible gas concentrations

~ Hain Steam Line, RCIC steam line or RWCU isolation failure resulting in a release pathway outside containment

~ Primary system is discharging into RB AND RB area radiation levels are >8.0 R/hr. in more than one area, N2-EOP-SC

~ Primary system is discharging into RB AND RB area temperatures are >212'F in more than one area, N2-EOP-SC

~ Emergency Director Judgment Loss of containment indication may include rapid unexplained decrease following initial increase in containment pressure Hay 1998 Page 109 EPHP-EPP-0102 Rev 02

ATTACHMENT 3 WORD LIST/DEFINITIONS Hay 1998 Page 110 EPHP-EPP-0102 Rev 02

ATTACHMENT 3 (Cont)

Actuate To put into operation; to move to action; commonly used to refer to automated, multi-faceted operations. "Actuate ECCS".

~Adversar As applied to security EALs, an individual whose intent is to commit sabotage, disrupt Station operations or otherwise commit a crime on station property.

Ade uate Core Coolin Heat removal from the reactor sufficient to prevent rupturing the fuel clad.

Alert Events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

Available The state or condition of being ready and able to be used (placed into operation) to accomplish the stated (or implied) action or function. As applied to a system, this requires the operability of necessary support systems (electrical power supplies, cooling water, lubrication, etc.).

Can Cannot be determined The current value or status of an identified parameter relative to that specified can/cannot be ascertained using all available indications (direct and indirect, singly or in combination).

Can Cannot be maintained above below The value of the identified parameter(s) is/is not able to be kept above

/below specified limits. This determination includes making an evaluation that considers both current-and future system performance in relation to the current value and trend of the parameter(s). Neither implies that the parameter must actually exceed the limit before the action is taken nor that the action must be taken before the limit is reached.

Nay 1998 Page ill EPHP-EPP-0102 Rev 02

ATTACHMENT 3 (Cont)

Can Cannot be restored and maintained above below The value of the identified parameter(s) is/is not able to be returned to above/below specified limits. This determination includes making an evaluation that considers both current and future systems performances in relation to the current value and trend of the parameter(s). Neither implies that the parameter must actually exceed the limit before that classification is made nor that the classification must be made before the limit is reached.

This does not imply any specific time interval but does not permit prolonged operation beyond a limit without making the specified classification.

As applied to loss of electrical power sources (ex.: Power cannot be restored to any vital bus in < 4 hrs) the specified power source cannot be returned to service within the specified time. This determination includes making an evaluation that considers both current and future restoration capabilities.

This implies that the declaration should be made as soon as the determination is made that the power source cannot be restored within the specified time.

Close To position a valve or damper so as to prevent flow of the process fluid.

To make an electrical connection to supply power.

Confirm Confirmation To validate, through visual observation or physical inspection, that an assumed condition is as expected or required, without taking action to alter the "as found" configuration.

Conti uous Being in actual contact; touching along a boundary or at a point.

Control Take action, as necessary, to maintain the value of a specified parameter within applicable limits; to fix or adjust the time, amount, or rate of; to regulate or restrict.

Deer ease To become progressively less in size, amount, number, or intensity.

~Diachar e Removal of a fluid/gas from a volume or system.

Nay 1998 Page 112 EPHP-EPP-0102 Rev 02

ATTACHMENT 3 (Cont)

~Dr el 1 That component of the BWR primary containment which houses the RPV and associated piping.

Enter To go into.

Establish To perform actions necessary to meet a stated condition. "Establish communication with the Control Room."

Evacuate To remove the contents of; to remove personnel from an area.

Exceeds To go or be beyond a stated or implied l.imit, measure, or degree.

Exist To have being with respect to understood limitations or conditions.

Failure A state of inability to perform a normal function.

General Emer enc Events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.

May 1998 Page 113 EPMP-EPP-0102 Rev 02

ATTACHMENT 3 (Cont)

Logic term which indicates that taking the action prescribed is contingent upon the current existence of the stated condition(s). If the identified conditions do not exist, the prescribed action is not to be taken and execution of operator actions must proceed promptly in accordance with subsequent instructions.

Increase To become progressively greater in size, amount, number or intensity.

Indicate To point out or point to; to display the value of a process variable; to be a sign or symbol.

Initiate The act of placing equipment or a system into service, either manually or automatically. Activation of a function or protective feature (i.e. initiate a manual scram).

~In 'ection The act of forcing a fluid into a volume or vessel.

Intrusion The act of entering without authorization Loss Failure of operability or lack of access to.

Haintain Take action, as necessary, to keep the value of the specified parameter within the applicable limits.

Hay 1998 Page 114 EPHP-EPP-0102 Rev 02

ATTACHMENT 3 (Cont)

Maximum Safe 0 eratin arameter The highest value of the identified operating parameter beyond which, required personnel access or continued operation of equipment important to safety cannot be assured.

~one tor Observe and evaluate at a frequency sufficient to remain apprised of the value, trend, and rate of change of the specified parameter.

got~if To give notice of or report the occurrence of; to make known to; to inform specified personnel; to advise; to communicate; to contact; to relay.

~0en To position a valve or damper so as to allow flow of the process fluid.

To break an electrical connection which removes a power supply from an electrical device.

To make available for entry or passage by turning back, removing, or clearing away.

~Ocr abl e Able to perform it's intended function Perform To carry out an action; to accomplish; to affect; to reach an objective.

Primar Containment The airtight volume immediately adjacent to and surrounding the RPV, consisting of the drywell and wetwell in a BWR plant.

Primar S stem The pipes,. valves, and other equipment which connect directly to the RPV or reactor coolant system such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

Hay 1998 Page 115 EPHP-EPP-0102 Rev 02

ATTACHHENT 3 (Cont)

Remove To change the location or position of.

~Re ort To describe as being in a specific state.

~Re uire To demand as necessary or essential.

Restore Take the appropriate action requires to return the value of an identified parameter to within applicable limits.

Rise Describes an increase in a parameter as the result of an operator or automatic action.

~Sam le To perform an analysis on a specified media to determine its properties.

Scram To take action to cause shutdown of the reactor by rapidly inserting a control rod or control rods (BWR).

Secondar Containment The airtight volume immediately adjacent to or surrounding the primary containment in a BWR plant.

Shut down To perform operations necessary to cause equipment to cease or suspend operation; to stop. "Shut down unnecessary equipment."

Hay 1998 Page 116 EPHP-EPP-0102 Rev 02

ATTACHMENT 3 (Cont)

Shutdown As applied to the BWR reactor, subcritical with reactor power below the heating range.

Site Area Emer enc Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels except near the site boundary.

Su ression ool The volume of water in a BWR plant intended to condense steam discharged from a primary system break inside the drywell.

Sustained Prolonged. Not intermittent or of transitory nature Transient Events of off-normal nature such as; scrams, runbacks involving >25% thermal power changes, ECCS injections or thermal power oscillations >10%.

To de-energize a pump or fan motor; to position a breaker so as to interrupt or prevent the flow of current in the associated circuit; to manually activate a semi-automatic feature.

Unavailable Not able to perform it's intended function Uncontrolled An evolution lacking control but is not the result of operator action.

~Un 1 armed Not as an expected result of,deliberate~~action.

Hay 1998 Page 117 EPMP-EPP-0102 Rev 02

ATTACHMENT 3 (Cont)

Until Indicates that the associated prescribed action is to proceed only so long as the identified condition does not exist.

Unusual Event Events are in process or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems oc'curs.

Valid Supported or corroborated on a sound basis.

Vent To'pen an effluent (exhaust) flowpath from an enclosed volume; to reduce pressure in an enclosed volume.

~Veri f To confirm a condition and take action to establish that condition if required. "Verify reactor trip."

Vital Area Any plant area which contains vital equipment.

LO:9d OZ lllf'6.

I H01938-03.hl3338 Hay 1998 Page 118- EPMP-EPP-0102 Rev 02