ML17059A355

Eals.

ML17059A355
Person / Time
Site:	Nine Mile Point
Issue date:	06/10/1994
From:	AFFILIATION NOT ASSIGNED
To:
Shared Package
ML17059A356	List: ML17059A355 ML18040A132
References
NUDOCS 9407150246
Download: ML17059A355 (940)

Text

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EAIL Uo c ~ede Pit'exeat ZO DUI ao OQ Plant Specific EAL DA RO'4 Guideline Ul>

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Nine Mile Point Unit 2 6/1 0/94 Operations Support Services, Inc.

233 Water Street 2nd Floor Plymouth, MA 02360

Plant Specific EA~uideline (A,H,S)

Nine Mh .wint Unit 2 ICg: AU1 Any unplanned release of gaseous or Ilquld radloactlvlty to the environment that exceeds two times the radlologlcal Technical Speclflcatlons for 60 minutes or longer.

Op. Mode 51(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) 5AII Applicability EU1.1 AU1.2

< valid reading on one or more of the folhwing monitors that exceeds the "value shown Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates with a release duration of 60 minutes or longer in excess of iRRREK isis; 2aha5hmn".

~~hrm Vote: If the monitor readingsm~ sustained for longer than 60 minutes and the required assessments cannot be completed within this period, then the declaration must be made based on the valid reading.

Bases

Plant Specific EA~uideline (A,H,S)

Nine Mii .~oint Unit 2 Re term Unplanned", as used in this context, indudes any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions (e. g., minimum dilution hw, maximum discharge fhw, alarm setpoints, etc.) on the applicable permit. Valid means that a radiation monitor reading has been confirmed by the operators to be correct.

degradation in the Jnplanned releases in excess of two times the site technical specifications that continue for 60 minutes or bnger represent an uncontrolled situation and hence, a potential evel of safety. The final integrated dose (which is very hw in the Unusual Event emergency class) is not the primary concern here; it is the degradation in plant control implied by the fact that a release of 4 times T/S for 30 minutes does not exceed he release was not isolated within 60 minutes. Therefore, it is not intended that the release be averaged over 60 minutes. For example,

'his 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 sill likely exceed 60 minutes.

to demonstrate Monitor indications ~should'alculated on the basis of the methodohgy of the site Offsite Dose Calculation Manual (ODCM compliance with 10CFR20 a~10CFR50 Appendix I requirements. Annual average meteorology shoukkha hused ukoca4lowocL Ihftalarmmdmnh hr lhft lhhd.fnfuulgn mme.'~ualnfftiy.aai hmmmIftghnignl laftgirhalignmfhagtiYily.mlmmlimihaHtngi fnzmdad.

fiauu Qmrftflahrmmhginthr lhftDiaha! BufiaihnHgnilzing Sxshm.

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Plant Specific EA~ uideline (A,H,S)

Nine Ml> . ntUnit2 ICN'U2 Unexpected Increase In plant radlatlon or airborne concentration.

Op. Mode Applicability 01(PwrOps) 02(HSB) 03(HSD) 04(CSD) 05(Refuel) 06(Defuel) ~ All iU2.1 AU2-4

~ ~tgjagtt direct area radiation monitor readinge >399fjam fhR alarm aQiat afhaahhhhrmltinufrummmzunfrull&arm~

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0 Bases

Plant Specific EP'uideline (A,H,S)

Nine Mh oint Unit 2 Valid means that a radiation monitor reading has been confirmed by the operators to be correct.

Oil of the above events tend to have long lead times relative to potential for radiohgical release outside the site boundary, thus impact to public health and safety is very low.

In light of Reactor Cavity Seal failure, incidents at two different PWRs and loss of water in the Spent Fuel Pit/Fuel Transfer Canal at a BWR all occurring since 1984, explicit coverage of these types of events via EALs ¹1 and ¹2 is appropriate given their potential for increased doses to plant staff. Classification as an Unusual Event is warranted as a precursor to a more serious event.

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hhR2rftauirmritmlzatlgnuf aahhuiif the fuftlhtt:aammuztrttrfttI.

EALL~applies to plants with licensed dry storage of older irradiated spent fuel to address degradation of this spent fuel.

EAL¹4~

safety of the plant. ~tittnrtf ~ ~fjgtign htrfth ingr~ing tn p JEBEL" lQYQI2 2KR8%%9fQQIQiR KR QQEIQQllK RQi Qnft dKRdk QYQC QQQQRl

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addresses unplanned increases in in-plant radiation levels that represent a degradation in the control of radioactive material, and represent a potential degradation in the level of

~gg~ g inuitiitiftgf This EAL escalates to an Alert per IC AA3, if the increases imPair thft ftttrftfrtf safe oPeration. Qgly guhrtgfttf 53hl rhtfiagR glaaH1SIlfttrfttf in thLtKhl. KQ?M unnfKQRRE SrtftrgMKrfKIREthn ffufrtrt ingrrtQrtfRE KHfitttrtQgrnE mfiathnhuthtfhtthrittibr mmd 1K fiaiaathtalarmadvent.

Plant Specific EP'ideline (A,H,S)

Nine Mlt nt Unit 2 ICN: AA1 Any unplanned release of gaseous or Ilquld radloactlvlty to the environment that exceeds 200 times radlologlcal Technical Speclficatlc for 15 minutes or longer.

Op. Mode Appllcablllty Q1 (Pwr Ops) Q 2 (HSB) Q3 (HSD) Q4 (CSD) Q5 (Refuel) Q 6 (Defuel) g AII LA1.1 AA1.2 5 valid reading on one or more of the following ggg+lft monitors that exceeds the "value Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates in excess of riterion and indicates the need to assess the release with dote: If the monitor readings~~ sustained for hnger than 15 minutes and the required assessments cannot be completed within this period, then the declaration must be made

>ased on the valid reading.

iA1.3 AA1.4 Bases

Plant Specific EP Nine Mh

'ideline nt Unit (A,H,S) 2 Valid means that a radiation monitor reading has been confirmed by the operators to be correct.

This event escalates from the Unusual Event by escalating the magnitude of the release by a factor of 100. Prorating the 500 mR/yr criterion 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.

Monitor indications shouklhe ~ calculated on the basis of the methodology of who~~ Annual average meteorology 4touldka ~ used wher+4iowed.

Plant Specific EP uideline (A,H,S)

Nine Mii.. nt Unit 2 ICff: AA2 Major damage to Irradiated fuel or loss of water level that has or will result In the uncovering of Irradiated fuel outside the reactor vess Op. Mode Qt Applicability (PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) ~ All iA2.1 AA2.2 ltttmlttggIgr report of ~~ observation of irradiated fuel uncovered.

8.0 QQZ 8.0 B9Z WA2.3 Bases

Plant Specific EP'ideline (A,H,S)

Nine Mii . 0 nt Vnit 2 This IC applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage, which is discussed in NUMARC IC AU2, Unexpected increase in Plant Radiation or Airborne Concentration. NUREG-0818, 'Emergency Action Levels for Light Water Reactors," forms the basis for these EALs. 'hmham EALs ~j18ftfi by the specific area where irradiated fuel is located such as reactor cavity, reactor vessel, or spent fuel pooL There is time available to take corrective actions, and there is little potential for substantial fuel damage. In addition, NUREG/CR-4982, "Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82," July 1987, indicates that even if corrective actions are not taken, no prompt fatalities are predicted, and that risk of injury is low. In addition, NRC Information Notice No. 9048, "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 behw 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.

Licensees may wish to reevaluate whether Emergency Action Levels specified in the emergency plan and procedures governing decayed fuel handling activities appropriately focus on concern for onsite workers and Kr-85 releases in areas where decayed spent fuel accidents could occur, for example, the spent fuel pool working floor. Furthermore, licensees may wish to determine if emergency plans and corresponding implementing procedures address the means for limiting radiohgical exposures of onsite personnel who are in other areas of the plant. Among other things, moving onsite personnel away from the plume and shutting of building air intakes downwind from the source may be appropriate."

Thus, an Alert Classification for this event is appropriate. Escalation, if appropriate, would occur via Abnormal Rad IeveURadiological Effluent or Emergency Director judgement.

Plant Specific E~uideline (A,H,S)

Nine Mt ..'Oint Unit 2 ICy: AA3 Release of radioactive material or Increases In radlatlon levels wlthln the faclllty that Impedes operation of systems required to malnta safe operations or to establish or malntaln cold shutdown.

Op. Mode Appllcablllty Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII iA3.1 AA3.2 lalid radiation monitor reading greater than 15 mR/hr in fbi; vali+st~pecitic) radiation monitor readings greater than (site-specific) QE!Z~ in areas requiring infrequent access to maintain plant safety functions.

Bases

Plant Specific E/" uideline (A,H,S)

Nine Nit . nt Vnlt2 Valid means that a radiation monitor reading has been confirmed by the operators to be correct. Qaly ~gngQd fLBQLQQdjngQ~ QQQQJdQI'Qd ia ihQ Rhl fa aYai{}fiaQQQQQmL ftQ1QIQQIIGK IlQQiara&mduahmgamaIQzQigffhmuummLQIjiQIhalmhthat hrjQfjy.Q jMII'.

This IG addresses increased radiation levels that impede necessary access to operating stations, or other areas containing equipment that must be operated manually, in order to maintain safe operation or perform a safe shutdown. It is this impaired ability to operate the plant that results in the actual or potential substantial degradation of the level of safety of the plant. The cause and/or magnitude of the increase in radiation levels is not a concern of this IC. The Emergency Director must consider the source or cause of the increased radiation levels and determine if any other IG 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, an SAE or GE may be indicated by the fission product barrier matrix ICs.

These EALs could result in declaration of an Alert at oaawait NMP-2due to a radioactivity release or radiation shine resulting from a major accident at the-This is appropriate if the increase impairs operations at the operating unit.

This IG is not meant to apply to increases in the containment chmo radiation monitors as these are events which are addressed in the fission product barrier matrix ICs. Nor is it intended to apply to anticipated temporary increases due to planned events (e. g., incore detector movement, radwaste container movement, deplete resin transfers, etc.)

NMP-2 abnormal operating procedures, emergency operating procedures, the 10GFR50 Appendix R analysis, identifying areas containing safe shutdown equipment.

Areas requiring continuous occupancy gQQJ~ include the control room an

~oraoca central ~ QQ~~ security alarm station. 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 NUREG4737, "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, as a 30 day duration implies an event potentially more significant than an Alert.

ased on QbrtgrmQi radiation levels which result in exposure control measures intended to maintain doses within normal occupational exposure guidelines and limits (i. e., 10CFR20), and in doing so, will impede necessary access.

10

Plant Specific E~ uideline (A,H,S)

Nine Nit . 0 nt Unit 2 ION: AS1 Boundary dose resulting from an actual or lmmlnent release of gaseous radloactlvlty exceeds 100 mR Whole Body or 500 mR Child Thyroid for the actual or proJected duration of the release.

Op. Mode Appllcablllty 01(PwrOps) 02(HSB) 03(HSD) 04(CSD) 05(Refuel) Q6(Defuel) ~ All

'LS1.1

< valid reading on one or more of the following monitors that exceeds or is expected to exceed he above criterion and idicates the need to assess the release with (SPDS only) dote: lf the monitor reading(s) is sustained for longer than 15 minutes and the required assessments cannot be completed within this period, then the declaration must be made

)ased on the valid reading.

WS1.3

/alid dose assessment capability indicates dose consequences greater than 100 mR J92f.

vholo4ody or 500 mRcM4QDE thyroid.

AS1.4 Field survey results indicate site boundary dose rates exceeding 100 mR/hr ~ expected to continue for more than one hour; or analyses of field survey samples indicate 4@dQQQ thyroid dose commitment of 500 mR for one hour of inhalation.

Bases 11

Plant Specific E'ideline Nine Mli . nt Unit 2 (A,H,S)

Valid means that a radiation monitor reading has been confirmed by the operators to be correct.

The 100 mR If'ntegrated dose in this initiating condition is based on the proposed 10CFR20 annual average population exposure. This value also provides a desirable gradient (one order of description. The 500 mR integrated4@444yroid thyroid.

~ fbggjff dose was established in consideration of the 1$ ratio of the EPA Protective Action Guidelines for ~

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 whotakody and gZ based on a site boundaty dose of 100 mR/hour eholahody TEDE or 500 mR/hour QQ~IILgjtichilcL4hyceid, whichever is more limiting (depends on source term assumptions).

The FSAR source terms applicable to each monitored pathway should be used in conjunction with annual average meteorology in determining indications for the monitors on that pathway.

12

Plant Specific EA uideline (A,H,S)

Nine Mlle ..tt Unit 2 ICy AG1 Boundary dose resulting from an actual or Imminent release of gaseous radloactlvlty exceeds 1000 mR Whole Body or 5000 mR Child Thyroid for the actual or projected duration of the release.

Op. Mode Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) gAII Appllcablllty KG1.1 AQ4 '4

't valid reading on one or more of the following monitors that exceeds or is expected to exceed he above criterion and 2@m2mf

~~+ (SPDS only) gj~(SPDS only)

Vote: If the monitor reading(s) is sustained for longer than 15 minutes and the required assessments cannot be completed within this period, then the declaration must be made iased on the valid reading.

AG1.3 Valid dose assessment capability indicates dose consequences

~halo4~ or 5000 mR QDEcM4 thyroid.

greater than 1000 mR ~ AG1.4 Field survey results indicate site boundary dose rates exceeding 1000 mR/hr IEQf. expected to continue for more than one hour; or analyses of field survey samples indicate cbildgZ thyroid dose commitment of 5000 mR for one hour of inhalation.

Bases 13

Plant Specific EAI+uideline (A,H,S)

Nine Mile Pc iiit Unit 2 Valid means that a radiation monitor reading has been confirmed by the operators to be correct.

The 1000 mR ~mhotahody and the 5000 mR QQf chiM 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 3EfKwhoh4odyor 5 rem QZcNd 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 in the initiating condition since it gives the most accurate dose assessment. Actual meteorology (including forecasts) should be used whenever possible.

d based on site boundary doses for either whoiahc4y TEDE orMkLthyroicl ~yfgigwhichever is more limiting (depends on source term assumptions(s).

The FSAR source terms applicable to each monitored pathway should be used in conjunction with annual average meteorology in determining indications for the monitors on that pathway.

14

Plant Specific EA uideline (A,H,S)

Nine Mile ..c Unit 2 ICg: HU1 Natural and destructive phenomena affectin the protected area.

Op. Mode Appllcablllty Q1(PwrOps) 02(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) gAII lU1.1 HU1.2 Report by plant personnel of tornado striking within protected area.

tU1.3 HU1.4

'assessment by the control room that an event Vehicle crash into plant structures or systems within protected UihaMtta has occurred. area boundary.

HU1.5 HU1.6 Report by plant personnel of an unanticipated explosion within protected area boundary Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.

resulting in visible damage to permanent structure or equipment.

HU1.7 15

Plant Specific EA uideline (A,H,S)

Nine Mlle c...it Unit 2 Bases The protected area boundary is~ical'hat part within the security isolation zone and is defined in the site security plan.

F~ EAL ¹~1 , NMP-2 seismic instrumentation actuates at 0.01 g. Damage may be caused to some portions of the site, but should not affect ability of safety functions to operate. Method of detection caaho hbased on instrumentation, validated by a reliable sourc 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."

EAL ¹EJ12 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.

EAL ¹Jjgf 3 allows for the control room to determine that an event has occurred and take appropriate action based on personal assessment as opposed to verification (i. e., an earthquake is felt but does not register on any plant-specific instrumentation, etc.).

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'i 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.

'f'tructures For EAL ¹~5, 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, if applicable.

EAL ¹~6 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 via HU2 and HU3. 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. Escalation of the emergency dassification is based on potential damage done by missiles generated by the failure or by the radiological releases These latter events would be dassified by the radiological ICs or fission product barrier ICs.

EAL ¹~7 covers events.

e precursors of more serious 16

Plant Specific EAluideline (A,H,S)

Nine Mile Pt...it Unit 2 IQy: HU2 Fire wlthln protected area boundary not extinguished within 15 minutes of detection.

Op. Mode Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) RAII Applicability lU2.1 fire in buildings or areas contiguous to any of the followin~~cif+areas not extinguished within 15 minutes of control room notification 4acm:

Bases 17

l 0

Plant Specific EA uideline (A,H,S)

NineMile i.itUnlt2 nie purpose of this IC is to address the magnitude and extent of fires that may be potentially significant precursors to damage to safety systems. This exdudes such items as fires within administration buildings, waste-basket fires, and other small fires of no safety consequence. This IC applies to buildings and areas that are contiguous or immediately adjacent to plant vital areas.

Escalation to a higher emergency class is by IC HA2, "Fire Affecting the Operability of Plant Safety Systems Required for the Current Operating Mode".

18

Plant Specific EAIuideline (A,H,S)

Nine Mlle P~..it Unit 2 Icy: HU3 Release of toxic or flammable gases deemed detrimental to safe operation of the plant.

Op. Mode Q1(pwrOps) Q2(HSB) Q3(HSD) Q4(GSD) Q5 (Refuel) Q6(Defuel) %All Applicability

<U3.1 HU3.2 disport or detection of toxic or flammable gases that could enter within the site Report by local, county or state officials for potential evacuation of site trea boundary in amounts thatch ggghf affect~ personnel based on offsite event.

operation of the plant.

Bases 19

Plant Specific EAIuideline (A,H,S)

Nine Mile Pt..it Unit 2 This IG is based on releases in concentrations within the site boundaty 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 ls as determined from the DOT Evacuation Tables for Selected Hazardous Materials, in the DOT Emergency Response Guide for Hazardous Materials.

20

Plant Specific EAIuideline (A,H,S)

Nine Mile P~..it Unit 2 icy: HU4 Confirmed security event which Indicates a potential degradatlon ln the level of safety of the plant.

Q1(Pwrops) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) AAII lU4.1 HU4.2 lomb device discovered within plant protected area andltttt outside the fgjjttittjttgplant vital Other security events as determined from~~pecitio)

~rea~ Qggtirj~ Safeguards Contingency Plan.

Bases

Plant Specific EA uideline (A,H,S)

Nine Mlle .,t Unit 2 this EAL is based o . 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. The plant protected area boundary istypically that part within the security soiation zone and is defined in the (site-specific) security plan.

22

Plant Specific EAuideline (A,H,S)

Nine Mlle Pt..t Unit 2 ICN: HU5 Other condltlons exlstlng which In the judgement of the Emergency Director warrant declaration of an Unusual Event.

Op. Mode Applicability Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) ~ All fU5.1

)ther conditions exist which in the judgement of the Emergency Director indicate a potential legradation of the level of safety of the plant.

Bases 23

Plant Specific EA uideline (A,H,S)

Nine Mlle t..t Unit 2

%is EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the "mergency Director to fall under the Unusual Event emergency class.

From a broad perspective, one area that may warrant Emergency Director judgement is related to likely or actual breakdown of site specific event mitigating actions. Examples to consider

'nclude 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.

It is also intended that the Emergency Directors judgement not be limited by any list of events as defined here or as augmented by the site. This list is provided solely as examples for consideration and it is recognized that actual events may not always follow a pre-conceived description.

24

Plant Specific EAIuideline (A,H,S)

Nine Mile Pc..t Unit 2

!CD: HA1 Natural and destructive phenomena affecting the plant vital area.

Op. Mode Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) gAII Appllcablllty 3A1.1 HA1.2 (sit~ci~ QQ mph strike within the protected Tornado or4igh area boundary.

~i~ winds greater than indicates seismic event greater than~~

3A1.3 HA1.4 he folhwing plant structures:

HA1.5 HA1.6 Vehicle crash affecting plant vital areas.

HA1.?

25

Plant Specific EA uideline (A,H,S)

Nine Mile c..r Unit 2 Bases Each of these EALs is intended to address 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 dassification. No attempt is made in this EAL to assess the actual magnitude of the damage. Escalation to a higher emergency dass, if appropriate, will be based on System Malfunction, Fission Product Barrier Degradation, Abnormal Rad Releases/Radiohgical Effluent, or Emergency Director Judgement ICs.

EAL ¹~1 should be based o+sit~pociflc) FSAR design basis of MZ~. Seismic events of this magnitude can cause damage to safety functions.

EAL ¹~2 should be based o+alt~cific) FSAR design basis of mph. Wind hads of this magnitude can cause damage to safety functions.

9Q EAL ¹~3 should specify~~pociflc) structures containing systems and functions required for ~ ~ gggzgfigft EAL ¹55i 4 EAL tlal." ~ - ~ -

¹~5 is intended to address such items as plane or helicopter crash, or on some sites, train crash, oc barge crash

'i -"" >> "' "i into a plant vital area.

include all areas containing safety-related equipment, their controls, and their power supplies. This EAL is, therefore, consistent with the definition of an ALERT in that if missiles have damaged or penetrated areas containing safety-related equipment the potential exists for substantial degradation of the level of safety of the plant.

EAL gM~covers These EALs can also be g precursors of more serious events. In particular, sites subject to severe weather as defined in the NUMARC station blackout initiatives, should include an EAL based on activation of the severe weather mitigation procedures (e. g., precautionary shutdowns, diesel testing, staff call-outs, etc.).

26

Plant Specific EA uideline (A,H,S)

Nine Mile t ..it Unit 2 fog: HA2 Fire or explosion affectin the operablllty of plant safety systems required to establish or malntaln safe shutdown.

Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) gAII SA2.1 rhe folhwing conditions exist:

i. Fire or exphsion in any of th areas:

AND

a. Affected system parameter indications show degraded performance or plant personnel

.eport visible damage to permanent structures or equipment within the structures or equipment within the specified area.

Bases 27

III~ I'"" ""' '" '

Plant Specific EA i'

Nine Mile

' '.t'~

uideline (A,H,S)

...( Unit 2

i~EM"'"'onsulted for equipment and plant areas required for the applicable mode. This will make it easier to determine the fire or explosion is potentially affecting one or more redundant trains of Ni

afety systems. Escalation to a higher emergency class, if appropriate, will be based on System Malfunction, Fission Product Barrier Degradation, Abnormal Rad Releases/Radiological

=ffluent, or Emergency Director Judgement ICs.

With regard to explosions, only those exphsions 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

-tructures and materials. The inclusion of a "report of visible damage" should not be interpreted as mandating a lengthy damage assessment prior to dassification. 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 declaration of an Alert and the activation of the TSC will provide the Emergency Director with the resources needed to perform these damage assessments. The Emergency Director also needs to consider any security aspects of the explosions, if applicable.

28

Plant Specific EAIuideline (A,H,S)

Nine I@lie Pt..it Unit 2 ICy: HA3 Release of toxic or flammable gases wlthln a faclllty structure which Jeopardizes operation of systems required to malntaln safe operat or to establish or malntaln cold shutdown.

3p. Mode Gi(pwrops) G2(HSB) 03(HSD) D4(CSD) Q5(Refuel) Q6(Defuei) QAII Applicability IA3.1 HA3.2 leport or detection of toxic gases within a lltfLfgllttttf jttft facility structure in concentrations Report or detection of flammable gases within a ~ttilmiftg structure in concentrations that sat will be life threatening to plant personnel: will~ the safe operation of the plant:

Bases 29

Plant Specific EA uideline (A,H,S)

Nine Mile t.. it Unit 2 This IC h based on gases that have entered a plant structure&fectlng the safe operation of the plant. This IC applies to buildings and areas contiguous to plant vital areas or other significant buildings or areas (i. e., Service Water Pump house). The intent of this IC 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. Escalation to a higher emergency class, if appropriate, will be based on System Malfunction, Fission Product Barrier Degradation, Abnormal Rad Releases/Radiological Effluent, or Emergency Director Judgement ICs.

30

Plant Specific EAluideline (A,H,S)

Nine Mile Pc ..'Unit 2 ICy HA4 Security event ln a plant protected area.

Op. Mode Q1(Pwrops) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) ~ All Appllcablllty lA4.1 HA4,2 ntrusion into plant protected area by akosga4orco yn a~gZyl. Other security events as determined from~~cific)

Rgffffl~ Safeguards Contingency Plan.

Bases 31

Plant Specific EAuideline (A,H,S)

Nine Mlle P~ .it Unit 2 This dass of security events represents an escalated threat to plant safety above that contained in the Unusual Event. For the purposes of this IG, this event to a Site Area Emergency.

32

4 ~

Plant Specific EA uideline (A,H,S)

Nine Mile Pi .it Unit 2 ICy: HA5 Control room evacuation has been lnltlated.

Op. Mode Q1(pwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) ~ All Appllcablllty 3A5.1

~

=ntry into for control room evacuation.

Bases 33

Plant Specific EAIO:uideline (A,H,S)

Nine Mlle Pi..it Unit 2 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.

34

Plant Specific EAIuideline (A,H,S)

Nine Mile Pi.it Unit 2 ICg: HA6 Other conditions exlstlng which In the Judgement of the Emergency Director warrant declaration of an Alert.

Op. Mode Appllcablllty Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) ~ All

)A6.1 7ther conditions existing which in the judgement of the Emergency Director indicate that plant efety systems may be degraded and that increased monitoring of plant functions is varranted.

Bases 35

Plant Specific EAiO'Uideline (A,H,S)

Nine Mile Pi..it Unit 2 This EAL is intended to address 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.

36

Plant Specific EAluideline (A,H,S)

Nine Mile Pi. it Unit 2 lCN: HS1 Security event In a plant vital area.

Op. Mode Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII Appllcablllty

)S1.1 HS1.2 ntrusion into ~gjj~jpg plant vital arear'y akosti korea ~~gg Other security events as determined from~~ecitic)

Gguirjl~ Safeguards Contingency Plan Bases 37

Plant Specific EAIO'Uideline (A,H,S)

Nine Mlle Pi .it Unit 2 This class of security events represents an escalated threat to plant safety above that contained in the Alert IG in that akostilaforce m ad~g, has progressed from the protected area to the vital area.

38

Plant Specific EA uideline (A,H,S)

NlneMlle ~..<Unlt2 fC¹: HS2 PW4 a a el a aRN Op. Mode Applicability Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) %All

)S2.1 ihe following conditions exist:

t. Control room evacuation has been initiated.

cannot be established per~acpecTiic)-

ROC 04UCIR within si~cifio) ~ minutes.

Bases 39

Plant Specific EA uideline (A,H,S)

Nine Mile oint Unit 2 Expeditious transfer of safety systems has not occurred but fission product barrier damage may not yet be Indicated.~it~cifg time for transfer ia based on analysis or assessments as to how quicMy control must be reestablished without core uncovering and/or core damage. This time shoukl not exceed 15 minutes. In cokl 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 maintaining critical safety functions and thereby assuring fission product barrier integrity. Escalation of this event, if appropriate, would be by Fission Product Barrier Degradation, Abnormal Rad Releases/Radiological Effluent, or Emergency Director Judgement ICs.

5dhraartagt tn arntmfiannf tha tuhlia anti aafa aslant anaratiana.~ aztrnLmtiathm arimarilxrtnihaabilitlr h rnairttainiharaaahzin a mahd.auuhtian. IharahrL itia artnrunriata trt

~tha Gannet E6Lartihaiitftmuhasizaathanaaffhr amaulinunhanmatrglling thauiant frgmnirhhhtha QantrulHaun.

40

Plant Specific EA uideline (A,H,S)

Nine Mile .rtt Unit 2 ICg HS3 Other condltlons which ln the judgement of the Emergency Director warrant declaration of Site Area Emergency.

Op. Mode Appltcablllty Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Deluel) ~ All RS3.1 3ther conditions which in the judgement of the Emergency Director warrant declaration of Site brea Emergency.

Bases 41

Plant Specific EA uideline (A,H,S)

Nine Mile ..tt Unit 2 This EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the

=mergency Director to fall under the emergency class description for Site Area Emergency.

42

Plant Specific EA uideline (A,H,S)

Nine Mlle ~.hatt Unit 2 ICy: HG1 Security event resulting ln loss of ablllty to reach and maintain cold shutdown.

Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) gAll Appllcablllty 4G1.1 HG1.2 oss of ~~ysical control c4Qgm the control room due to security event.

uuaf almhal patrol af fhaaunuh ~hhmn muahililx dm h.mmuily. mumt.

Bases 43

Plant Specific EA uideline (A,H,S)

Nine Mile ..it Unit 2 This IC encompasses conditions under which a hostile force has taken physical control of vital area required to reach and maintain safe shutdown. Ihft~gftrahtrfrhrhr~glgbQy >

ahuhlmaihftrmhr xdmaiahmammliaa ItmrftfmfhhEhLhmhmnmttdififtrliumhmtahaartfuhnt aziul fatmhdhlhfrmalrrtl catmmf ftmuhCrtddfttframmh.

44

Plant Specific EA uideline (A,H,S)

Nine Mlle ..it Unit 2 fCy. HG2 Other conditions existing which ln the Judgement of the Emergency Director warrant declaration of General Emergency.

Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) ~ All 3G2.1 7ther conditions existing which in the judgement of the Emergency Director indicate: (t) tctual or imminent substantial core degradation with potential for loss of containment, or (2) etential for uncontrolled radio nuclide releases. These releases can reasonably be expected o exceed EPA PAG plume exposure levels outside the site boundary.

Bases 45

Plant Specific EA uideline (A,H,S)

Nine Mile ..t Unit 2 this EAL is intended to address 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 General Emergency dass.

46

Plant Specific EAf+uideline (A,H,S)

Nine Mile Pt...it Unit 2 fC¹: SU~ Loss of all offslte power to essential busses for greater than 15 minutes.

Op. Mode Applicability Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel)- Q6(Defuel) %All 3U'f.1 rhe folhwing conditions exist:

s. Loss of power t~i~pocTiic) for lreater than 15 minutes.

AND x At least~~pocTiic) tttftt emergency generators are supplying power to emergency buses; Bases 47

Plant Specific EA uideline (A,H,S)

NlneMile t. tUnit2 Prolonged hss of AG power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete hss of AC power (station blackout). Fifteen minutes was selected as a threshold to exdude transient or momentary power losses.

48

Plant Specific EA uideline (A,H,S)

Nine Mile ..it Unit 2 iC¹: SU2 Inablllty to reach required shutdown wlthln Technical Speclflcatlon Llmlts.

Op. Mode Appllcablllty S1(PwrOps) R2(HSB) ~ 3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII S U2.1 Plant is not brought to required operating mode withiQei~ocific) Technical Specifications LCO Action Statement Time.

Bases 49

Plant Specific EA uideline (A,H,S)

Nine Mile i. it Unit 2 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) nonwmergency events. The plant is within its safety envehpe 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 System malfunction Hazards, or Fission Product Barrier Degradation ICs.

50

Plant Specific EA uideline (A,H,S)

NineMlle t... tUnlt2 ICg: SU3 Unplanned loss of most or all safety system annunclatlon or lndlcatlon In the control room for greater than 15 minutes.

8 1 (Pwr Ops) ~ 2 (HSB) ~ 3 (HSD) Q 4 (CSD) 0 5 (Refuel) 0 6 (Defuel) 0 All IU3.1 he folhwing conditions exist:

u Loss of annunciatore for greater than 15 minutes.

AND AND

. In the opinion of the Shift Supervisor, the loss of the annunciators or indicators requires ncreased surveillance to safely operate the un.

AND L Annunciator or indicator loss does not result from planned action.

Bases 51

Plant Specific EA uideline (A,H,S)

Nine Mile ..it Unit 2 This IC and its associated EAL are intended to recognize the dNiculty 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'oss of annunciators or indicator excludes scheduled maintenance and testing activities.

'Compensatory non-alarming indications: in this context includes computer based information such as SPDS specific plant design and subsequent retrofits.

Mhmfha gfaahhrthamf fthm,ihhr lm.

h ~l Ib 'Skl This should Include all computer systems available for this use depending on

't is not intended that plant personnel perform a detailed count of instrumentation lost but fha use gf thw4~ judgement hg fha 5455gggudm m fha threshold for determining the severity of the plant conditions. This judgement 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 unit~.

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 difficultyassociated 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 hss will be reported via 10CFR50.72. If the shutdown is not in compliance with the Technical Specification action, the Unusual Event is based on SU2, Inability to Reach Required Shutdown Within Technical Specification Limits."

(Site-speci~ Annunciators ~cator for this EAL must include those identified in the Abnormal Operating procedures, in the Emergency Operating Procedures, and in other EALs (e. 9.,

area, process, and/or effluent rad monitors, etc.).

(Sitaepecfflc) Annunciators oc4adicaW 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, no IC is indicated 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.

52

Plant Specific EA uideline (A,H,S)

Nine Mile ..t Unit 2 ICttf: SU4 Fuel clad degradatlon.

Op. Mode Applicability Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) ~ All 3U4.1 SU4.2 coolant sample activity Bases 53

Plant Specific EA uideline (A,H,S)

Nine Mile ..i Unit 2 This IC is included as an Unusual Event because it is considered to be a potential degradation in the level of safety of the plant and a potential precursor of more serious problems.

EAL ¹~1 addresses g~pecIIQ~ radiation monitor reading , that provide indication of fuel clad integrity. ~

EAL ¹Qfl4g addresses coolant samples exceeding coolant technical specifications for Iodine spike. Escaiatiori of this IC to the Alert level is via the fission product barrier degradation monitoring ICs.

54

Plant Specific EA uideline (A,H,S)

Nine Mile t...it Unit 2 ICg SU5 RCS leakage.

Op. Mode Applicability ~ 1(PwrOps) 52(HSB) ~ 3(HSD) D4(CSD) 05(Refuel) Q6(Defuel) QAII SU5.1 Either of the following conditions exist:

a. Unidentified togpm

~~gghmlg~ leakage greater than OR

b. identified cgggfgZ~ gy2gm g ~ leakage greater than 25 gpm.

Bases 55

0 Plant Specific EA Uideline (A,H,S)

Nine Mlle a...it Vnit2 This IC is included as an Unusual Event because it 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 and pressure boundary leakage was selected as it is observable with normal control room indications. Lesser values must generally be determined through time~nsuming surveillance test (e. g., mass balances). The EAL for identified leakage is set at a higher value due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage. In either case, escalation of this IC to the Alert level is via Fission Product Barrier Degradation ICs or IC SA3, "Inability to Maintain Plant in Cold Shutdown.'nly operating modes in which there is fuel in the reactor coolant system and the system Is pressurized are specified.

56

Plant Specific EA uideline (A,H,S)

Nine Mlle .it Unit 2 iC¹: SU6 Unplanned loss of all onslte or offslte communications capabllltles.

Op. Mode Appllcablllty Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) %All 3U6.1

=ither of the following conditions exist:

Urdu l'b"'l' '

ability to perform routine operations:

3R

>. Loss of all ~hLlttlitt)ftlttgoffsite communications capability:

Bases 57

Plant Specific EA uideline (A,H,S)

Nine Mile .. t Unit2 the purpose of this IC and its associated EALs is to recognize a hss of communications capability that either defeats the plant operations staff ability to perform routine tasks necessary. for slant operations or the ability to communicate problems with offsite authorities. The hss of offsite communications ability is expected to be significantly more comprehensive than the condition addressed. by 10CFR50.72.

Z talkies).

i" '-'"- '"- '"'- --'- l -' t'.', ""'-- "" '- " .

'W'"adios/walkie Xl and dedicated EPP phone systems. 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.).

58

Plant Specific EA uideline (A,H,S)

Nine Mile ..it Unit 2 ICy: SU7 Unplanned loss of required DC power during cold shutdown or refueling mode for greater than 15 minutes.

Op. Mode Applfcablllty Q1(PwrOps) Q2(HSB) Q3(HSD) g4(CSD) ~ 5(Refuel) Q6(Defuel) QAII 3U7.1

@hot ~ of the following conditions exist:

'BKxh

>us voltage indications AND

i. Failure to restore power to at least one acquired DC bus within 15 minutes from the time of Oss.

Bases 59

Plant Specific EA iideline (A,H,S)

Nine Mile i.. Unit 2 The purpose of this IC and its associated EALs is to recognize a hss of DC power compromising the ability to monitor and control the removal of decay heat during cokl 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.

Unplanned is Included in this IC and EAL to preclude the declaration of an emergency as a result of planned maintenance activities. Routinely plants will perform maintenance on a train related basis during shutdown periods. It is intended that the loss of the operating (operable) train is to be considered. If this hss results in the inability to maintain cold shutdown, the escalation to an Alert will be per SA3 "Inability to Maintain Plant in Cold Shutdown."

Pi~pec~ ~ bus voltage shoul4bo h based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value should incorporate a margin of at least 15 minutes of operation before the onset of inability to operate those loads. This voltage is usually near the minimum voltage selected when battery sizing is performed. Typically the value for the entire battery set is approximately 105 volts per cell. For a 56 string battery set the minimum voltage is typically 1.81 volts per cell.

60

Plant Specific EA uideline (A,H,S)

Nine Mile ~;.it Unit 2 ICy. SA1 Loss of all offslte power and loss of all onslte AC power to essential busses during cold shutdown or refueling mode.

l Op. Mode Appllcablllty Q1(PwrOps) Q2(HSB) Q3(HSD) ~ 4(CSD) ~ 5(Refuel) 56(Defuel) QAII 3A1.1 All of the following conditions exist:

~. Loss of power tQcitezpocific) transformers AND x Failure of~te.specific)

AND

" Failure to restore power to Meastene.

anorgcncykuc within 15 minutes from the time of loss of both offsite and onsite AC power.

Bases 61

Plant Specific EA uideline (A,H,S)

Nine Mile ~..it Unit 2 Loss of all AG power compromises all plant safety systems requiring electric power i When in cold shutdown, refueling, or defueled mode the event can be class Tiied as an Alert, because of the significantly reduced decay heat, lower temperature and pressure, 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 Judgement ICs. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

62

Plant Specific EAIuideline (A,H,S)

Nine Mile Pt. t Unit 2 lCN: SA2 Y " " P y reactorscram h h w h r a Reactor Protection system setpolnt has been exceeded aad~ manual scram Op. Mode Appllcablllty ~ 1(PwrOps) ~ 2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII lA2.1 Bases 63

Plant Specific EA>>ideline (A,H,S)

Nine Mile i... Unit 2 This condition indicates failure of the automatic anrf rnanttat protection system to scram the reactor tuiha atrtant zdgghurttultrrfaa iha raauturitalnurnada attb~tjua[. This condition is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient and thus the plant safety has been compromised, and design limits of the fuel may have been exceeded. An Alert is indicated because conditions exist that lead to potential loss of fuel clad or RCS.

A manual scram is any set of actions uttahhtttun:~ mudamitah ur hBO.

EbiaEanrfiamrltinuEBLh~rahtanmauifiualitr roufifiarftunIuiaautriatttitr tfttIinathauuitrlitIundmuiihtttfhxtbauanariuhamtaaaauuiitttf tuhuilinumahr raautuia. Iba failmuf atrtuinatiu 5fiatiunufaiaautuimamfulhntttfhx2ruumhi~initiatiunautiunauhhhuanharauJdlxtalranat tharaauhzmntruluunauiarfuaanut uuaaa uutantialiuuufaithtr fttal ulatfur fKR hunIfarbq. fhuatraafaituiatummauufitiunain BEEhara fafinaif nuthx tbamatttnuaufuriTiualihrhufhxthaatraiiabiamaruinufmhuiitiuality.tbauttna1luuuilIi~amuRiauuira alamTiiuatiunufanhhd furmnditiunainmhhhthaqautur la in faut ahuttfumaaamultuf tbaaurammmL Jfiatbauuntintrarfuritiua!Itx trntfaruunditIunamuirinua iaautur auiainnhhb uumta 1ha uutantial tbiaat tuBSQur fuel ularf intagrity. ltia ahuirnuurtant tu nuta that tba failtria uf tbaraautui urutautiunairabun iuinitiata an autumatiuaurarnrfuaa nut infar auttral ui uutantlal laituraauf uthtr mstttrnanuriait.in anrfufitattlf.auDtutrraur iufiaaiun rrutftrutharriar dauralfatiun. Iha HER aanraa nuuthttr aafatx funutlun btrt tu initaiia raautur mrama. Iharttfurrt.unua tha r mbzhaahtanzzaaattlkmammttrL faihiaaintbaBERmhmuanftatranuuiantaafttthiinuaut. Ifintrnttdiatamanualautiunatumamtba iaautur arazzmfulfulluninurauuunitiunufan atrturnatiumamfailtra.thtrahnutbittat tuaitbttiulantaafatxurMhnurudtrut Jnttturity. ttiatatftutbaautuntatiuauramfahrh Ihhrftt~uniauunmatantmithtbaubilumtMuf maitinu a~trratam uunaartratbrauiaamfiuatiuna

{Hah; an't faiturauf anatituinatiu azam signal tuinitiataamuuauftd auram nuuld haimmttdiatabr fullunatfhxanuuaratur initiatatf rnaiml mam. Ibarttfura 1ha Ehl. zurdinu iniha IB abail unlxmakarttfaranua iuiha faiturauf irnrnttdiatamanual aurainal 64

Plant Specific EA Uideline (A,H,S)

Nine Mile ..it Unit 2 ling SA3 lnablllty to maintain plant ln cold shutdown.

Op. Mode Q1(Pwrops) Q2(HSB) Q3(HSD) ~ 4(CSD) 55(Refuel) Q6(Defuel) QAII Appllcablllty SA3.1 The following conditions exist:

OR Bases 65

Plant Specific EA ~sideline (A,H,S)

Nine Mlle i.. Unit 2 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 woukl be via Abnormal Rad Levels/Radiohgical Effluent or Emergency Director Judgement ICs.

"Uncontrolled" means that system temperature increase is not the result of planned actions by the plant staff.

The EAL guidance related to uncontrolled temperature rise is necessary to preserve the anticipatory philosophy of NUREG-0654 for events starting from temperatures much hwer than the cold shutdown temperature limit.

Escalation to the Site Area Emergency is by IC SSS, "Loss of Water Level in the Reactor Vessel that has or will Uncover Fuel in the Reactor Vessel," or by Abnormal Rad Levels/Radiological Effluent ICs.

66

Plant Specific EAle'uideline (A,H,S)

Nine Mlle Pi . t Unit 2 ICy: SA4 Unplanned loss of most or all safety system annunclatlon or Indlcatlon In control room with.either (1) a slgnlfleant transient In progress (2) compensatory non-alarming Indicators are unavailable.

Op. Mode Appllcablllty ~ 1(PwrOps) ~ 2(HSB) S3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII iA4.1 rhe following conditions exist:

t. Loss of annunciatore for greater than 15 minutes.

AND

). ln the opinion of the Shift Supervisor, the loss of the annunciators or indicators requires ncreased surveillance to safely operate the unit(e).

AND Annunciator or indicator loss does not result from planned action.

AND

3. Either of the following:

~ A significant plant transient is in progress OR

~ C are unavailable.

Bases 67

Plant Specific EAIOluideline (A,H,S)

Nine Mile Pi .t Unit 2 This IC and its associated EAL are intended to recognize the dNiculty 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.).

Planned loss of annunciators or indicators included scheduled maintenance and testing activities.

It is not intended that plant personnel perform a detailed count of the instrumentation lost but the use gf 4~~~ judgement threshold for determining the severity of the plant conditions. This judgement is supported by the specific opinion o f the Shift Supervisor that additional operating personnel will be required to provide increased monitoring of system operation to safely operate the unit(s).

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 hss will be reported via 10CFR50.72. If the shutdown is not in compliance with the Technical Specification action, the Unusual Event is based on SU2 "Inability to Reach Required Shutdown Within Technical Specification Limits."

(SI!e-specific) Qnnunciators cu-hdicatore 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.).

"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 10% or greater.

"Compensatory non-alarming indications" in this context Indudes computer based information such as SPDS. This should include all computer systems available for this use depending on specific plant design and subsequent retrofits. 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 IC 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.

68

Plant Specific EA uideline (A,H,S)

Nine Mile .it Unit 2 fc¹: SA5 AC power capablllty to essential busses reduced to a single power source for greater than 15 minutes such that any additional single failure would result ln station blackout.

Op. Mode Appllcablllty 51(PwrOps) 52(HSB) 53(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII 3A5.1 fhe following conditions exist (a and b):

a. Loss of power t+si~pecific) transformers for greater than 15 minutes.

AND x Onsite power capability has been degraded to gaiy. one ggitfLlgll~ittg~iaef)emergency Bases 69

Plant Specific EA uideline (A,H,S)

Nine Mlle .. Unit 2 This IC and the associated EALs are intended to provide an escalation from IC SU1 "Loss of AllOffsite Power to Essential Busses for Greater than 15 Minutes." The condition indicated by this IC 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 all offsite power and hss of onsite emergency dieseis with only one train of emergency busses being badded from the unit main generator, or the loss of onsite emergency diesels with only one train of emergency busses being backfed from offsite power. The subsequent hss of this single power source would escalate the event to a Site Area Emergency in accordance with IC SS1 "Loss of All Offsite and Loss of All Onsite AC Power to Essential Busses."

Example EAL ¹QE5.1b should be expanded to identify the control room indication of the status gf offsite-specific power sources and distribution busses that, if unavailable, establish a single failure vulnerability.

70

Plant Specific EA uideline (A,H,S)

Nine Mile i'nit 2 ICy: SS1 Loss of all offslte power and loss of all onslte AC power to essential busses.

Op. Mode Applicability 51(PwrOps) 52(HSB) 83(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII SS1.1 Loss of all offsite and onsite AG power as indicated by:

a. Loss of power to ~~ecifie) transformers.

AND

>. Failure of~~pecific)

AND

=. Failure to restore power to Meastoao-

~rgoacy4ve within~~ueific) ~ minutes from the time of hss of both offsite and onsite 4,C power.

Bases

Plant Specific EA uideline (A,H,S)

Nine Mile .t Unit 2 Loss of all AC power compromises all plant safety systems requiring electric power i Prolonged loss of all AC.power will cause core uncovering and loss of containment integrity, thus this event can escalate to a General Emergency. Th+sitaspecifg time duration should be selected to exclude transient or momentary power hsses, but should not exceed 15 minutes.

Escalation to General Emergency is via Fisson Product Barrier Degradation or IC SG1, 'Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power."

72

Plant Specific EA Uideline (A,H,S)

NlneMile i .tUnlt2 lC¹: SS2 Failure of Reactor Protection system Instrumentatlon to complete or Initiate an automatic reactor scram once a Reactor Protection sysl setpolnt has been exceeded and manual scram was not successful n n in r ir Op. Mode Appllcablllty ~ 1(PwrOps) ~ 2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII 3S2.1 Bases 73

Plant Specific EA uideline (A,H,S)

Nine Mile .< Unit 2 Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed. A Site Area Emergency is indicated because conditions exist that lead to imminent loss or potential hss of both fuel clad and RCS. Although this IC may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response. Escalation of this event to a General Emergency would be via Fission Product Barrier Degradation or Emergency Director Judgement ICs.

74

Plant Specific EA uideline (A,H,S)

Nine Mile . t Unit 2 iC¹: SS3 Loss of all vltaf DC power.

Op. Mode 51(PwrOps) 52(HSB) g3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII Applicability 3S3.1 r, bus voltage indications for greater than 15 minutes.

Bases 75

Plant Specific EA uideline (A,H,S)

Nine Mile .t Unit 2 Loss of all DC power compromises ability to monitor and control plant safety functions. Prohnged hss 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 Abnormal Rad Levels/Radiological Effluent, Fission Product Barrier Degradation, or Emergency Director Judgement ICs. Fifteen minutes was selected as a threshold to exdude transient or momentary power losses.

76

Plant Specific EA Uideline (A,H,S)

Nine Mile .t Unit 2 ICy: SS4 Complete loss of function needed to achieve or malntaln hot shutdown.

Op. Mode Applicability 51(PwrOps) ~ 2(HSB) ~ 3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII 3S4.1 Bases 77

Plant Specific EA uideline (A,H,S)

Nine Mile .t Unit 2 This EAL addresses complete loss of functions, including ultimate heat sink and reactivity control, required for hot shutdown with the reactor at pressure and temperature. Under these conditions, there is an actual major failure of a system intended for protection of the public. Thus, declaration of a Site Area Emergency is warranted. Escalation to a General Emergency would occur by Abnormal Rad Levels/Radiohgical Effluent, Fission Product Barrier Degradation, or Emergency Director Judgement ICs.

78

Plant Specific EA uideline (A,H,S)

Nine Mile .t Unit 2 ling: SS5 Loss of RPV water level that has or will uncover fuel In the RPV.

51(PwrOps) 52(HSB) ~ 3(HSD) S4(CSD) S5(Refuel) Q6(Defuel) QAII 3S5.1 SS5.2 Bases 79

Plant Specific EA uideline (A,H,S)

Nine Mile i .t Unit 2 Under the conditions specified by this IC, severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured.

'III I h Ih 'll d l~ d5 'hhl azmmiraii~tnia mtnarallxagnaldaradilm her httuadanr rtf fba rangaiamhhhhzalizmf dafiagratignamaxrtmr. Iuganaraiazzba amzaniraiignrtf mahttaiihhgaa. Raartfhttihibahai ahdand BGRharrlaramttaihatramurrad. Ztmrahra.dat;larafhnaf a2hhamEmargarmxianarranfttd.

ifbmdmgan auzaniraiiunalncraaaainmaiuziitIn mdhiha uraaant:artf ntmanitt ghhal daflagraihn huraia {Lh)Pkhttiruganand a Pl mtfgtuQ. mmiiau rtf ilmrttniainmani irraammibnr ref ihaaffaiia radhat;iharaiaaaa rafa frttttidharattttiradhx EZaanddat:laraihnof a Qamal Emargantnt rartttlrad Thus, declaration of a Site Area Emergency is warranted under the conditions specified by the IC. Escalation to a General Emergency is via radiological effluence IC AG1 harrhrlaaa.

80

Plant Specific EA Uideline (A,H,S)

Nine Mile ..it Unit 2 ICO: SS6 lnablllty to monitor a slgnlf leant transient tn progress.

Op. Mode ~ 1(PwrOps) ~ 2(HSB) 53(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII Appllcablllty SS6.1 Wl of the following conditions exist:

s. Loss of~e-spociTIe) annunciators are unavailable.

AND

. Indications needed to inonitor~~cific)

AND

f. Transient in ro ress.

Bases 81

Plant Specific EA uideline (A,H,S)

Nine Mile,tUnit 2 This IC and its associate EAL are intended to recognize 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.

(Q~oci~ 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.).

Compensatory non-alarming indications" in this context indudes computer based information such as SPDS. This should include all computer systems available for this use depending on specific plant design and subsequent retrofits.

'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 10% or greater.

@to.spec~ 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 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.

82

Plant Specific EA uideline (A,H,S)

Nine Mile ..it Unit 2 iC¹: SG1 Prolonged loss of all offslte power and prolonged loss of all onslte AC power.

Op. Mode Applicability R1(Pwrops) ~ 2(HSB) ~ 3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII SG1.1 prolonged loss of all offsite and onsite AG power as indicated by:

a. Loss of power to specific) transformers.

AND x Failure of~~pecific)

. At least one of the folhwing conditions exist:

~ Restoration of ggmttLta at least one emergency bus within~~pocTiic) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is not likely OR

~

(

Bases 83

Plant Specific EA uideline (A,H,S)

Nine Mile ..it Unit 2 Loss of all AC power compromises all plant safety systems requiring electric power i

. Prolonged loss of all AC power will lead to hss of fuel clad, RCS, and containment. Thgai~peciflc) hours to restore AC power can be based on site blackout coping analysis performed in conformance with 10CFR50.63 and Regulatory Guide 1.155, "Station Blackout, as available, with appropriate allowance for offsite emergency response. Although this IC may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response.

This IC 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 hss 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, it is necessary to give the Emergency Director a reasonable idea of how quickly (s)he may need to 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'?

(Refer to Tables 3 and 4 for more information.)

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 cora cooling degradation must be based on fission product barrier monitoring with particular emphasis on Emergency Director judgement as it relates to imminent loss or potential loss of fission product barriers and degraded ability to monitor fission product barriers.

84

Plant Specific EA uideline (A,H,S)

Nine Mlle ..it Unit 2 ICe: SG2 Failure of the Reactor Protection system to complete an automatic scram and manual scram was not successful and there ls Indlcatlott an extreme challenge to the ablllty to cool the core.

Op. Mode g1(PwrOps) 52(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII Appllcablllty 3G2.1 AND Either of the following:

a. (S b.

(s'R Bases 85

Plant Specific EA uideline (A,H,S)

Nine Mile ~..it Unit 2 Automatic and manual scram are not considered successful if action away from the reactor control console is required to scram the reactor.

Under the conditions of this IC and its associated EALs, the efforts to bring the reactor subcritical have been unsuccessful and, as a result, the reactor is maximum decay heat had for which the safety systems were designed. Although there are capabilities away from the reactor control console, such as liquid control in BWRs, the continuing temperate rise indicates that these capabilities are not effective. This situation could be precursor for a core melt sequence.

'tandby producing more heat than the For BWRs, the extreme challenge to the ability to cool the core is intended to mean that the reactor vessel water level is below 2/3 coverage of active fuel Another consideration is the inability to initially remove heat during the early stages of this sequence.

BWRs, ~~peels considerations include inability to remove heat via the main condenser, or via the suppression pool oc4oruc (e. g., due to high pool water temperature).

In the event either of these challenges exist at a time that the reactor has not been brought below the power associated with the safety system design (typically 3 to 5% power) a core melt sequence exists. In this situation, core degradation can occur rapidly For this reason, the General Emergency declaration is intended to be anticipatory of the fission product barrier matrix declaration to permit maximum offsite intervention time.

86

Plant Specific EAl sideline (FPB)

Nine Mile Point Unit 2 BWR FPB tC¹'C1 Barrier: Fuel Claddin Type: Loss

==

Description:==

Primary Coolant Activity Level FC1.1 Coolant activity greater than Bases:

Assessment by the NUMARG EAL Task Force indicates that this amount of coolant activity is well above that expected for iodine spikes and corresponds to about 2% to 5% fuel clad damage. This amount of clad damage indicates significant clad heating and thus the fuel clad barrier is considered hst.

There is no equivalent 'Potential Loss" EAL for this item.

87

Plant Specific EAI jideline (FPB)

Nine Mile Point Unit 2 BWR FPB IC¹'C2 Barrier: Fuel Claddin Type: Loss/Pot. Loss Descript toll: Reactor Vessel Water Level FC2.1 Level less than -bio JEBB Bases:

The "Loss" EA+si~poci~ value corresponds to the level which is used in EOPs to indicate challenge of core cooling.

o~tiuo-fuel. This is the minimum value to assure core cooling without further degradation of the dad. The "Potential Loss" EAL is the same as teh RCS barrier "Loss" EAL 4 below and corresponds to the~~pecifg water level at the top of the active fuel. Thus, this EAL indicates a "Loss" of RCS barrier and a "Potential Loss" of the Fuel. Clad Barrier.

This EAL appropriately escalates the emergency class to a Site Area Emergency.

88

Plant Specific EAl sideline (FPB)

Nine Mile Point Unit 2 BWR FPB IC¹: FC3 Barrier: Fuel Claddin Type: Loss Descrfptloit: Drywell Radiation Monitoring Drywell radiation monitor reading greater tha+sitaepecific) + R/hr.

Bases:

~ is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywelL The assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine reading should be calculated pCi/gm dose equivalent 1-131 into the drywall spRing) allowed within technical specifications and are there fore indicative of fuel damage (approximately 2'/a - 5% dad failure depending on core inventory and RCS volume). This value is higher than that specified for RCS barrier loss EAL ¹3.

Thus, this EAL indicates a loss of both fuel dad barrier and RCS barrier.

Caution: it is important to recognize that in the event the radiation monitor is sensitive to shine from the reactor vessel or piping spurious readings will be present and another indicator of fuel clad damage is necessary.

There is no "Potential Loss EAL associated with this item.

89

Plant Specific EAI sideline (FPB)

Nine Mile Point Unit 2 BWR FPB IC¹: FC4 Baggiest Fuel Claddin Type: Loss Descrlptlotl: Other (Site-Specific) Indications Bases:

Q~ocif+ instrumentation. ~

This EAL is to cover other~Wpec~ indications that may indicate loss or potential hss of the fuel clad barrier, Including indications from containment air monitors or any other I

90

Plant Specific EAI lideline (FPB)

Nine Mile Point Unit 2 BWR FPB ICN: FCS Barrier: Fuel Claddin Type: LossIPot. Loss

==

Description:==

Emergency Director Judgement FC5.1 Any condition in the judgement of the Emergency Director that indicates loss or potential loss of the fuel cladding barrier.

Bases:

This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the fuel clad 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 judgement that the barrier may be considered lost or potentially lost. (See also IC SG1, "Prolonged Loss of All Oifsite Power and Prolonged Loss of All Onsite AC Power", for additional information.)

91

Plant Specific EA sideline (FPB)

Nine Mile Point Unit 2 BWR FPB IC¹: RCS1 Barrfer: RCS Type: Loss

==

Description:==

RCS Leak Rate RCS1.1 Q~pec~ Indicationg of main steam line break:

Bases:

The 'Loss" EAL is based on design basis accident analyses which show that even if MSIV closure occurs within design limits, dose consequences offsite from a "puff release would be in excess of 10 millirem. Thus, this EAL is included for consistency with the Alert emergency classlcation.

92

Plant Specific EAI sideline (FPB)

Nine Mile Point Unit 2 BWR FPB IC¹: RCS1 Barrier: RCS Type: Potential Loss

==

Description:==

RCS Leak Rate RCS1.2 RCS leakage greater than 50 gpm inside the drywell RCS1.3 Bases:

The potential loss of RCS based on leakage is set at a level indicative of a small breach of the RCS but which is well within the makeup capability of normal and emergency high pressure systems. Core uncovery is not a significant concern for a 50 gpm leak, however, break propagation leading to significantly larger loss of inventory is possible.

Many BWRs may be unable to measure an RCS leak of this size because the leak would likely increase drywell pressure above the drywell isolation setpoint. 'The system normally used to monitor leakage is typically isolated as part of the drywell isolation and is therefore unavailable. If primary system leak rate information is unavailable, other indicators of RCS leakage should be used. Potential loss of RCS based on primary system leakage outside the drywall is determined from site-specific hfgzimllm3gh Qptttatittg Lgygh 4acaa in the areas of the main steam line tunnel, rnaia4urbiao-etc., which indicate a direct path from the RCS to areas outside primary containment.

93

Plant Specific EAI jideline (FPB)

Nine Mile Point Unit 2 BWR FPB IC¹: RCS2 Barrier: RCS Type: Loss

==

Description:==

Drywell Pressure RCS2.1 EQIBREKQiginlmlttpressure ~ psig Bases:

The+~ciTQ drywall pressure is based on the drywell high pressure alarm setpoint '

higher value may be used if supporting documentation is provided which indicates the chosen value is less than the pressure which would be reached for a 50 gpm reactor coolant system leak.

There is no 'Potential Loss" EAL corresponding to this item.

94

Plant Specific EAL ideline (FPB)

Nine Mlle Point Unit 2 BWR FPB IC¹: RCS3 Barrier: RCS Type: Loss Descriptlotl: Drywell Radiation Monitoring RCS3.1 Drywell radiation monitor reading greater tha+sft~ocific) 39@R/hr Bases:

Ihlaal<<- '--

dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (I. e., within T/S) into the drywell atmosphere. This that specified for fuel clad barrier EAL ¹3. Thus, this EAL would be indicative of a RCS leak only. If the radiation monitor reading increased to that value reading will be less than-specified by the fuel clad barrier EAL

¹3, fuel damage would also be indicated.

However, if the site-specific physical location of the drywall radiation monitor is such that radiation from a cloud of released RCS gases could not be distinguished from radiation from adjacent piping and components containing elevated reactor coolant activity, this EAL should be omitted and other site -specific indications of RCS leakage substituted.

There is no "Potential Loss" EAL associated with this item.

95

Plant Specific EAl ideline (FPB)

Nine Mile Point Unit 2 BWR FPB IC¹: RCS4 Barrier: RCS Type: Loss

==

Description:==

Reactor Vessel Water Level RCS4.1 Level less than 'BfD Bases:

This "Loss" EAL is the same as "Potential Loss" fuel clad barrier EAL ¹2. Th+sit~cif+ water level corresponds to the level which is used in EOPs to indicate challenge of core cooling.

Depending on the plant this may be top of active fuel or 2/3 coverage of active fuel. This EAL appropriately escalates the emergency dass to a Site Area Emergency. Thus, this EAL indicates a loss of the RCS barrier and a potential toss of the fuel clad barrier.

96

Plant Specific EAI'sideline (FPB)

Hine Mile Point Unit 2 BWR FPB IC¹: RCS5 Barrier: RCS Type: Loss Descrlptloll: Other (site-specific) indications RCS5.1 Bases:

This EAL is to cover other~~cif+ indications that may indicate loss or potential loss of the RCS barrier.

97

Plant Specific EAI sideline (FPB)

Nine Mile Point Unit 2 BWR FPB IC¹: RCS5 Barrier: RCS Type: Potential Loss

==

Description:==

Other (site-specific) indications RCS5.2 Bases:

This EAL is to cover other~~pecTilc) indications that may indicate hss or potential loss of the RCS barrier.

98

Plant Specific EAI >ideline (FPEI)

Njne Mlle pojnt Unit 2 BWR FPB ICC: RCSS Barrier: RC8 Type: Loss/Pot. Loss

==

Description:==

Emergency Director Judgment RCS6.1 Any condition in the judgment of the Emergency Director that indicates loss or potential loss of the RCS barrier Bases:

This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is lost or potentially tost. In addition, the inability to monitor the barrier should also be incorporated in the EAL as a factor in Emergency Director judgement that the barrier may be considered hst or potentially lost. (See also IC SG1, "Prolonged Loss of Offsite Power and Prolonged Loss of AllOnsite AC Power,, for additional information.)

99

Plant Specific EAI ideline (FPB)

Nine Mlle Point Unit 2 BWR FPB ICO: PC1 Barrier: Prima Containment Type: Loss Oescrlptlon: Drywell Pressure PC1.1 PC1.2 Bases:

100

Plant Specific EAI sideline (FPB)

Nine Mile Point Unit 2 BWR FPB IC¹: PC1 Barrier: Prima Containment Type: Potential Loss

==

Description:==

Drywall Pressure PC1.3 PC1.4 Exphsive mixture of 'xists.

Bases:

Existence of an explosive mixture means a hydrogen and oxygen concentration of at least the lower deflagration limitcuaco exists.

101

Plant Specific EAl sideline (FPB)

Nine Mile PoInt Unit 2 BWR FPB 1C¹: PC2 Bar ffat'; Prima Containment Type: Loss Descriptloll: Containment Isolation Valve Status after Containment Isolation Signal PC2.1 bhin2aamlina BQIQ2thtmiim PC2.2 Intentional venting per EOPs:

PC2.3 hurimarymhunhdht;htrrtittututhhh arimarxmafaiuamrtt mirtithttr; Bmtgr BuildIrtahrmImnuttraturmabm thtir mazimum~fmttmtirtg lmh QB Huger Buiirhrtahum Bafiatignlmhahm thttirmaximum~urtttmtint lych Bases:

to the condenser.

~~~ ~ ~ hfttm g ~ g ~ittiiTtgrtot ~ Jggggtt ~ ~

This EAL is intended to cover containment isolation failures allowing a direct flow path to the environment such as failure of both MSIVs to close with open valves downstream to the turbine or uatfttr ggrttfitj~, In addition, the presence of area radiation or temperature alarms indicating unisolable primary system leakage outside the drywell are covered. Also, an intentional venting of primary containment per EOPs to the secondary containment and/or the environment to considered a loss of containment.

There is no 'Potential Loss" EAL associated with this item.

102

Plant Specific EAI sideline (FPB)

Nine Mile Point Unit 2 BWR FPB 1C¹: PC3 Barrier: Prima Containment Type: Potential Loss

==

Description:==

Significant Radioactivity Inventory in Containment PC3.1 Containment radiation monitor reading greater th Bases:

RIKEt " " ""'"" ' "'"'"' "'" """" i" "'"' """'"

of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure into the reactor coolant. Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such sever consequences that it is prudent to treat this as a potential loss of containment, such that a General Emergency declaration is warranted.. NUREG-1228, "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates that such conditions do not exist when the amount of clad damage is less than 20%. a radiation monitor reading corresponding to 20% fuel clad damagshe ~ specified here.

There is no 'Loss EAL associated with this item.

103

Plant Specific EAI'sideline (FPB)

Nine Mile Point Unit 2 BWR FPB )C¹: PC4 Barrier: Prima Containment Type: Potential Loss

==

Description:==

Reactor Vessel Water Level PC4.1 Bases:

The conditions in this potential hss EAL represent imminent melt sequences which, if not corrected, could lead to vessel failure and increased potential for containment failure. In conjunction operating procedures have been ineffective in restoring reactor vessel level

'n fggJ there is not a "success" path. ~

with the level EALs in the fuel and RCS barrier columns, this EAL will result ln the declaration of a General Emergency loss of two barriers and the potential loss of a third. If the emergency Severe accident analysis (e. g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation with the reactor vessel in a significant fraction of the core damage scenarios, and the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide a reasonable period to allow emergency operation procedures to arrest the core melt sequence. Whether or not the procedures willbe effective should be apparent within the time provided. The Emergency Director should make the declaration as soon as it is determined that the procedures have been, or will be ineffective.

There is no "loss" EAL associated with this item.

104

Plant Specific EAI sideline (FPB)

Nine Mile Point Unit 2 BWR FPB IC¹: PC5 Barrier: Prima Containment Type: Loss Descrlptlort: Other (site-specific) indicathns PC5.1 Bases:

This EAL is to cover other~~ocific) indications that may indicate toss or potential hss of the containment barrier.

105

Plant Specific EAI iideline (FPB)

Nine Mile Point Vnlt 2 BWR FPB IC¹: PC5 Bayriey: Prima Containment Type: Potential Loss

==

Description:==

Other (site-specific) indications PC5.2 Bases:

This EAL is to cover other~~pec~ indications that may indicate loss or potential hss of the containment barrier.

106

Plant Specific EAI jideline (FPB)

Nine Mile Point Unit 2 BWR FPB IC¹: PC6 Barrier: Prima Containment Type: Loss/Pot. Loss

==

Description:==

Emergency Director Judgment PC6.1 Any condition in the judgment of the Emergency Director that indicates loss or potential loss of the containment barrier LmaafamtaiamftatindiaatnramaxiadttdL'aamhtttat gt:unttltnftgtmf LQQhmmum Haahf uafatglaimddamaaahljtattina jaitialirzreminmataiamttaturmum Bases:

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 judgement that the barrier may be considered lost or potentially lost. (See also IG SG1,

'Prohnged Loss of All Offsite Power and Prolonged Loss of All Onsite AG Power", for additional information.)

107

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Fission Product Barrier Evaluation Revision 0 N'ra Mohawk Power Co Nine Mile Point Unit 2 Operations Support Services, Inc.

233 water Street 2nd Floor Plymouth, MA 02360

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Evaluation of NMP-2 Fission Product Barrier Emer enc Action Levels The Fission Product Barrier (FPB) degradation category for a BWR plant is illustrated in the following table which is designated "Table 3" in NESP-007, Revision 2.

The Initiating Condition (IC) for each of the four emergency classifications (Unusual Event, Alert, Site Area Emergency, and General Emergency) are designated FUl, FAl, FSl, and FGl, respectively.

Each IC is defined by one or more EALs or combination of EALs which are indicative of a loss or potential loss of one or more of the three fission product barriers. The three fission product barriers are:

~ Fuel Clad (FC)

~ Reactor Coolant System (RCS)

~ Primary Containment (PC)

NESP-007, Revision 2, prescribes example EALs for each of the three fission product barriers. An EAL is defined by one or more plant conditions. For mcaznple, there are five FC barrier example $ W,Ls, six RCS barrier acample EALs, and six PC acample EALs. Each EAL may consist of one or more conditions representing a loss of the barrier and a potential loss of the barrier. Some EALs may have only loss conditions, others only potential loss conditions, some have both loss and potential loss conditions. Each EAL is given a sequential number in Table 3. In the foQowing list under the column labeled "NESP-007", NUMARC KALa with a defined condition (i. e., labeled as needing "site-specific" input in Table 3) are identified with a "yes",

and those without a defined condition (i. e. labeled "not applicable" in Table 3) are identified with a "no". Similarly, EAL conditions applicable to NMP-2 are identified with a yes/no under the column labeled "NMP-2".

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 NUMARC NMP-2

~Brrf r EAL ¹ Loss ~P mPi Lum FC Yes No Yes(FC1.1) No Yes Yes Yes(FC2.1) Yes(FC2.1)

Yes No Yes(FC3.1) No Yes Yes Yes(FC4.1) No Yes Yes Yes(FC5.1) Yes(FC5.1)

RCS la Yes Yes Yes(RCS1, 1) Yes(RCS1.2) lb No Yes No Yes(RCS1.3) 2 Yes No Yes(RCS2.1) No 3 Yes No Yes(RCS3.1) No 4 Yes No Yes(RCS4.1) No 5 Yes Yes No No 6 Yes Yes Yes(RCS6.1) Yes(RCS6.1)

PC la Yes Yes No Yes (PC1.3) lb Yes Yes No Yes (PC1.4) 2a Yes No Yes(PC2.1) No 2b Yes No Yes(PC2.2) No 2c Yes No Yes(PC2.3) No 3 No Yes No Yes(PC3.1)

No Yes No Yes(PC4.1) 5 Yes Yes No No 6 Yes Yes Yes(PC6.1) Yes(PC6.1)

Based on the classiQcation key given at the beginning of Table 3, the number of example EALs, and the number of loss and potential loss conditions, the set of conditions that can yield a given emergency classiQcation can be computed.

The maximum, theoretically possible set of conditions that can yield an Unusual Event classification is given in column 1 of Table A. These consist of the PC loss and PC potential loss conditions.

The maximum, theoretically possible set of conditions that can yield an Alert classification is given in column 1 of Table B. These consist of FC loss and potential loss conditions, and RCS loss and potential loss conditions.

The maximum, theoretically possible set of conditions that can yield a Site Area Emergency classification is given in column 1 of Table C.

These consist of any of the following conditions:

~ Loss of FC and RCS, or

~ Potential loss of FC and RCS, or

~ Potential loss of FC or RCS

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 BIld Loss of another barrier The third set of conditions listed above can be represented by the following conditions to eliminate reference to "loss of another barrier":

~ Potential loss of FC and loss of RCS, or

~ Potential loss of FC and loss of PC, or

~ Potential loss of RCS and loss of FC, or

~ Potential loss of RCS and loss of PC The maximum, theoretically possible set of conditions that can yield a General Emergency classification is given in column 1 of Table D.

These consist of the following conditions:

~ Loss of any two barriers, and

~ Potential loss of a third These conditions can be represented by the following conditions to correlate barrier loss and potential loss to the three specific barriers:

~ Loss of FC and loss of RCS and potential loss of PC, or

~ Loss of RCS and loss of PC and potential loss of FC, or

~ Loss of PC and loss of FC and potential loss of RCS Since the EAL conditions are listed numerically in Table 3, Tables A through D utilize a similar numbering system which is modifled by letter abbreviations to define each set of conditions. For example, condition "FCl-loss" corresponds to a loss of the Fuel Clad barrier due to primary coolant activity level greater than the site-specific value.

Similarly, "RCSlb-pot. loss" corresponds to a potential loss of the Reactor Coolant System barrier due to unisolable primary system leakage outside the drywell, and so on.

An evaluation of each condition or set of conditions listed in Tables A through D is made to determine if it properly defines the appropriate threshold for the classification. If a condition or set of conditions is appropriate, a comment reflecting this conclusion is recorded in the "Remarks" column. If a condition or set of conditions is determined to be inappropriate, it is lined out and the reason for this conclusion is similarly recorded in the "Remarks" column. Where additional space is required to complete comments, the comments are recorded by

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 number in Appendix 1 of this document. The numbers of the comments are recorded in the "Remarks" column with the associated condition or set of conditions to which they apply.

A summary of the results of the Qssion product barrier evaluation is presented in Appendix 2.

RECOGNITION CATEGORY F FISSION PRODUCT BARRIER DEGRADATION INITIATINGCONDITION MATRIXTABLE S BWR UNUSUAL EVENT SITE AREA EMERGENCY GENERAL EMERGENCY FU1 Any loss or any Any loss or any Loss of both fuel clad FG1 Loss of any two potential loss of tentfal loss of either and RCS barriers containment el clad or RCS. OR AND Potentfal loss of both Potential loss of third Op. Modes: Op. Modes: fuel clad and RCS barrfer.

Power operaUon Hot Power operation Hot OR Standby/Startup Standby/Startup Potentfal loss of either Op. Modes:

(BWR) (BWR) fuel clad or RCS, and Power operation Hot Hot Shutdown Hot Shutdown loss of any additional Standby/Startup barrfer. (BWR)

Hot Shutdown Op. Modes:

Power operation Hot Standby/Startup (BWR)

Hot Shutdown NOTES:

l. Although the logic used for these fnfUatfng condfUons appears overly complex, ft ls necessary to reflect the following consfderations:

~ The fuel clad barrfer and the RCS barrier are weighted more heavily than the containment barrier (see Sections 3.4 and 3.8 for more fnformaUon on this point). Unusual Event ICs associated with RCS and Fuel Clad barriers are addressed under System MalfuncUon ICs.

~ At the Site Area Emergency level. there must be some ability to dynamically assess how far present condfUons are for General Emergency.

For example. ifFuel Clad barrfer and RCS ~er "Loss" EALs existed, this would indicate to the Emergency Dfrector that, in additional to offslte dose assessments. conUnual assessments of radioactive inventory and containment integrity must be focused on. If, on the other hand. both Fuel Clad barrier and RCS barrier "PotenUal Loss" EALs existed, the Emergency Director would have more assurance that there was no fmmedfate need to escalate to a General Emergency.

~ 'Ihe ability to escalate to higher emergency classes as an event gets worse must be maintained. For example, RCS leakage steadily increasing would represent an increasing risk to public health and safety.

2. Ffssfon Product Barrier ICs must be capable of addressing event dynamfcs. 'Ihus, the EAL Reference Tables 3 and 4 state that IMMINENT(L e., within I to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) loss or potenUal loss should result fn a classfflcaUon as ffthe affected threshold(s) are already exceeded, partfcularly for the hfgher emergency classes.

O

0 (8

RECOGNITION CATEGORY F (0 INITIATINGCONDITION MATRIXTABLE 8 BWR O

Fuel Clad Barrier Ezample EALs'otential Loss z

1. Prim lant A tivl Coolant activity greater than (site-specfAc) value Not Applicable

2. R t V Wa Lev Level less than (site-speciAc) value Level less than (site-spcciAc) value d at o onitorln Drywell radiation monitor reading greater than (site-speciAc) Not Applicable R fhr

4. t A d (site-specfAc) as applfcable (site-specfAc) as applicable t J d Any condition ln the]udgment of the Emergency Director that fndfcates loss or potential loss of the fuel clad barrier Determine which combfnatlon of the three barriers are lost or have a potential loss and use the following key to classify the event. Also. an event for multiple events could occur which result in the conclusion that exceeding the loss or potential loss thresholds fs imminent (l. e.,

within I to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />). In this imminent loss situation, use]udgment and classffy as ffthe thresholds are exceeded.

RECOGNITION CATEGORY F CO 0)

INITIATINGCONDITION MATRIXTABLE 3 BWR RCS Barrier Example EAL8' O Potential Loss

1. Lt~~~kRa g (site-speciQc) fndfcatfon of main steam line break RCS leakage greater than 50 gpm fnsfde the drywell OR unisolatble prfmary system leakage outside drywell as indfcated by area temperature or area radiation alarm Pressure greater than (site-speclQc) psfg Not applicable 1 daf o lo Dvgarell radiation monitor reading greater than (site-specfQc) Not applicable R fhr

4. to Vessel Water 1 Level less than (site-speciQc) value Not applicable (site-speciQc) as applicable (site-speci Qc) as applicable tor Jud en Any condition fn the]udgment of the Emergency Director that K indicates loss or potential loss of the RCS barrier O

p

RECOGNITION CATEGORY F Co 0)

INITIATINGCONDITION MATRIXTABLE S BWR (0 Primary Containment Barrier Example EALs' O Rapid unexplained decrease following fnftfal increase (site-specfffc) psfg and increasing OR OR Drywell pressure response not consistent with LOCA conditions explosive mixture exists

2. ta solatfo Valv a ta e t o Failure of both values fn any one lfne to dose and downstream Not applfcable pathway to the environment exists OR Intentional venting per EOPs Not applicable OR Unfsolable prfmary system leakage outsfde chywell as indicated Not applfcable by area temperature or area radiation alarm R O

Not applicable Containment radiation monitor reading greater than (site-specific)

R/hr f O

4. a Vessel Wat e Not applfcable Reactor vessel water level less than (site-specific) value and the 9 maxtmum core uncover time limit fs fn the unsafe region (sfte-specific) as applfcable (site-specfffc) as applicable K

tor Jud Q

Any condftfon fn the judgment of the Emergency Director that p indicates loss or potential loss of the containment barrier O

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table A BWR Fission Product Barrier Unusual Events NESP-007 Remarks Loss or pot. 1oss of PC Qua-less Condition not supported in PEG.

PC4b-1ess Condition not supported in PEG.

Pea-less 21 DCRb-less PCQe-1ess 2 PCS-1ess Condition not supported in PEG.

QCG-1ess Subsumed in "Judgment" EAL.

3 3,25 4,26 5,27 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

10

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table B BWR Fission Product Barrier Alerts NESP-007 Remarks Loss or pot. loss of PC s s t u>> e e 8%4oss">>s '>> ~

g e "sp~'$"'"'::""'"'4'.@r+44l" ~'3".CoolQnt'.8ctivitv"@~" ~ >@4:+r "'+'s~~"""~'p ~>>". S u e S s

8 'CQ-less FCS-less Subsumed tn "Judgment" EAL.

8 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Loss or pot. loss of RCS RCS~ess 6 RC34-less 8 RCSS-less Condition not supported in PEG.

RCSGMss Subsumed in "Judgment" EAL.

15 23 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table C BWR Fission Product Barrier Site Area Emergencies NESP-007 Remarks Loss of FC and RCS 16 I":-i!i::::-.:ii'"':::::::!xiii::,i:::::::::::::,:-::---::--'-o--:::ii"':

18 8

Condition not supported in PEG.

,'-.": ii"':-:i ii,:--!iiii ~iii ":ii( Subsumed in "Judgment" EAL.

'!.',:::,.ll""'i"":l"""I~i

8 8

9

':lid'""""

Condition not supported in PEG.

Subsumed in "Jud ent" EAL.

FW%"-'-'-'.:i':" t *,:,::i::: ':::: .."-::":"- -"-'-"::: iii" 1i-::: ""'di 'll" iii:: -;:-:"

19 10 ll Condition not supported in PEG.

Subsumed in "Judgment" EAL.

24 24 24 8

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

12 12 12 12 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Pot. loss of PC and RCS 8

ess 8 12

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table C HWR Fission Product Barrier Site Area Emergencies MES P-007 Remarks Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Subsumed ln "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Pot. loss of FC and loss of RCS 8

8 8

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported ln PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Subsumed ln "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Pot. loss of FC and loss of PC Condition not supported in PEG.

Condition not supported ln PEG.

8 13

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table C HWR Fission Product Barrier Site Area Emergencies NESP-007 Remarks 8

8 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported fn PEG.

Condition not supported fn PEG.

Condition not supported fn PEG.

Condition not supported in PEG.

Subsumed fn "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Pot. loss of RCS and loss of FC 19 8

19 20 12 23 8

19 24 12 Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D HWR Fission Product Barrier General Emergencies MES P-007 Remarks Loss of FC+ loss of RCS+ pot. loss of PC

"": ""'""'-kpCti-: pesoreiti toss Ifrtp 'rES"i count sii sniienitpnrsepsdsnsre>bWpl"':::gIj'"" '::i'l::@

'-" "'-"'-"".'.O'CIS-:yet.

I Condition not supported In PEG.

Subsumed in "Judgment" EAL.

25 25 26 27 ess Condition not supported in PEG.

Subsumed in "Judgment" EAL.

25 25 26 27 Condition not supported in PEG.

Subsumed in "Judgment" EAL, 25 25 26 27 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

16

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D BWR Fission Product Barrier General Emergencies NESP-007 Remarks Condition not supported in PEG.

Subsumed in "Judgment" EAL.

25 25 26 27 CondiUon not supported in PEG.

Subsumed in "Judgment" EAL.

25 25 26 27 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

25 25 26 27 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

25 25 26 27 CondiUon not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

CondiUon not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL 17

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D BWR Fission Product Barrier General Emergencies NESP-007 Remarks Subsumed in "Judgment" EAL Condition not supported in PEG.

Subsumed in "Judgment" EAL.

25 25 26 27 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

25 25 26 27 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

25 25 26 27 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

25 25 26 27 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

18

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D HWR Fission Product Barrier General Emergencies NESP-007 Remarks Subsumed in "Judgment" EAL Subsumed in "Judgment" EAL.-

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

25 25 26 27 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

25 25 26 27 CondiUon not supported in PEG.

Subsumed in "Judgment" EAL.

25 25 26 27 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

25 25 26 27 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Condition not supported in PEG.

CondiUon not supported in PEG.

CondiUon not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

19

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D BWR Fission Product Barrier General Emergencies NESP-007 Remarks Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

12 12 12 12 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

12 12 12 12 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

12 25 26 27 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

25 12 12 12 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

20

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D BWR Fission Product Barrier General Emergencies NESP-007 Remarks Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Loss of RCS + loss of PC + pot. loss of PC CondiUon not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

CondiUon not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

28 CondiUon not supported in PEG.

Subsumed in "Judgment" EAL.

28 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

25 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

CondiUon not supported in PEG.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

CondiUon not supported in PEG.

Condition not supported in PEG, Condition not supported in PEG.

21

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D BWR Fission Product Barrier General Emergencies NESP-007 Remarks 28 Condition not supported in PEG.

Subsumed in "Judgment" EAL 28 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

25 Condition not supported in PEG.

Subsumed in "Judgment" EAL Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

28 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

28 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

25 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

22

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D BWR Fission Product Barrier General Emergencies NESP-007 Remarks Subsumed 'in "Judgment" EAL.

Condition not supported in PEG.

Condition not supported ln PEG.

Condition not supported fn PEG.

Condition not supported in PEG.

CondfUon not supported ln PEG.

Condition not supported in PEG.

28 CondlUon not supported in PEG.

Subsumed in "Judgment" EAL.

28 Condition not supported fn PEG.

Subsumed in "Judgment" EAL.

25 Condition not supported fn PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

CondlUon not supported in PEG.

Condition not supported fn PEG.

Subsumed fn "Judgment" EAL.

Condition not supported ln PEG.

Subsumed in "Judgment" EAL.

CondiUon not supported in PEG.

CondfUon not supported ln PEG.

CondiUon not supported fn PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

CondlUon not supported ln PEG.

Condition not supported ln PEG.

Condition not supported fn PEG.

Condition not supported fn PEG.

Condition not supported in PEG.

Condition not supported fn PEG.

Condition not supported in PEG.

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D SWR Fission Product Barrier General Emergencies NESP-007 Remarks Condition not supported in PEG.

Condition not supported fn PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported fn PEG.

Condition not supported in PEG.

Condition not supported fn PEG.

Condition not supported in PEG.

Condftfon not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condftfon not supported in PEG.

Subsumed fn "Judgment" EAL, Subsumed in "Judgment" EAL.

Condition not supported fn PEG.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported fn PEG.

Condition not supported fn PEG.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Loss of PC + loss of FC+ pot. loss of RCS Condition not supported fn PEG.

Condition not supported fn PEG.

Condition not supported fn PEG.

Condition not supported fn PEG.

24

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL EvaluaUon, Rev. 0 Table D BWR Fission Product Barrier General Emergencies NESP-007 Remarks CondlUon not supported ln PEG.

CondlUon not supported in PEG.

Condition not supported ln PEG.

not supported ln PEG. 'ondition Condition not supported ln PEG.

Condition not supported in PEG.

CondlUon not supported in PEG.

Condition not supported ln PEG.

CondlUon not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

CondlUon not supported ln PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

CondiUon not supported ln PEG.

Condition not supported ln PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported ln PEG.

Condition not supported in PEG.

Condition not supported ln PEG.

Condition not supported ln PEG.

Condition not supported in PEG.

Condition not supported ln PEG.

Condition not supported ln PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

25

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D BWR Fission Product Barrier General Emergencies NESP-007 Remarks

""*""" *'PC'-"-l"""'"-'-*-"~i""'!ilia Condition not supported in PEG.

Si'll I!i!i i @tiki"" " '"

'0 ii&Alhi""'dpi" 29 i".'1 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

PCS:3688:, >$+%FC2-."Toss """ i'"' ' "" '""i28'"'-'f8il to~fsol :

eamWc'An 'FAF'N '0 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

24, 28 24, 28 Condition not supported in PEG.

Subsumed in "Judgment" EAL 12 12 Condition not supported in PEG.

Subsumed in "Judgment" EAL 22 22 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

22, 22 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

22 22 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

22 22 26

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D SWR Fission Product Barrier General Emergencies NESP-007 Remarks i'3 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

12 12 CondiUon not supported in PEG.

Subsumed in "Judgment" EAL.

I,::- '"""":llilia'i'F CondiUon not supported in PEG.

Subsumed in "Judgment" EAL.

'si'eii "e'4"::""'8*,:.'":""':::w:;:.*"i!i:.'-i'i!i):::

32 Condition not supported in PEG.

Subsumed tn "Jud ent" EAL.

34 Condition not supported in PEG.

Subsumed in "Judgment" EAL.

24 24 CondiUon not supported in PEG.

ass Subsumed in "Judgment" EAL.

12 12 Condition not supported in PEG.

Subsumed in "Judgment" EAL Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

CondiUon not supported in PEG.

Condition not supported in PEG.

ess CondiUon not supported in PEG.

CondiUon not supported in PEG.

Condition not supported in PEG.

27

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D HWR Fission Product Barrier General Emergencies NESP-007 Remarks Condition riot supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Condition not supported in PEG, Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL Subsumed in "Judgment" EAL.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Loss of PC + loss of FC + loss of RCS Condition not supported in PEG.

Condition not supported in PEG.

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D HWR Fission Product Barrier General Emergencies

-NESP-007 Remarks Condition not supported fn PEG.

Condition not supported fn PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

CondfUon not supported in PEG.

Condition not supported in PEG.

CondiUon not supported in PEG.

Condition not supported in PEG.

CondfUon not supported fn PEG.

Condition not supported fn PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported fn PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported fn PEG.

CondfUon not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

CondiUon not supported fn PEG.

Condition not supported fn PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported fn PEG.

Condition not supported in PEG.

Condftfon not supported in PEG.

Condition not supported in PEG.

Condition not supported fn PEG.

Condition not supported fn PEG.

Condition not supported fn PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported fn PEG.

29

I OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D HWR Fission Product Barrier General Emergencies NESP-007 Remarks Condition not supported in PEG.

Condition not supported in PEG.

CondiUon not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

CondiUon not supported in PEG.

Condition not supported in PEG, Condition not supported in PEG.

CondiUon not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

CondiUon not supported in PEG.

Condition not supported in PEG.

CondiUon not supported in PEG.

Condition not supported in PEG.

CondiUon not supported in PEG.

CondiUon not supported in PEG.

35 35 35 35 35 Subsumed in "Judgment" EAL.

35 35 35 35 35 Subsumed in "Judgment" EAL.

0 OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D BWR Fission Product Barrier General Emergencies NESP-007 Remarks 35 35 35 35 35 35 24, 28 24, 28 24, 28 24, 28 24, 28 Subsumed in "Judgment" EAL.

35 35 35 35 35 Subsumed in "Judgment" EAL.

25 25 25 25 25 Subsumed in "Judgment" EAL.

25 25 25 25 25 Subsumed in "Judgment" EAL.

25 25 25 25 25 31

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D SWR Fission Product Barrier General Emergencies NESP-007 Remarks Subsumed in "Judgment" EAL.

25 25 25 25 25 Subsumed in "Judgment" EAL.

25 25 25 25 25 Subsumed in "Judgment" EAL.

35 35 35 35 35 Subsumed in "Judgment" EAL.

35 35 35 35 35 Subsumed in "Judgment" EAL.

35 35 35 35 35 Subsumed in "Judgment" EAL.

24, 28 24, 28 24, 28 24, 28 32

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D SWR Fission Product Barrier General Emergencies NESP-007 Remarks 24, 28 Subsumed in "Judgment" EAL.

35 35 35 35 35 Subsumed in "Judgment" EAL.

CondlUon not supported ln PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported ln PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported ln PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported ln PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported ln PEG.

Condition not supported ln PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported ln PEG.

Condition not supported in PEG.

Condition not supported ln PEG.

Condition not supported in PEG.

i.

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Table D BWR Fission Product Barrier General Emergencies NESP-007 Remarks Condition not supported in PEG.

Condition not supported in PEG.

Condition not supported in PEG.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL Subsumed in "Judgment" EAI Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

'ubsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

Subsumed in "Judgment" EAL.

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 AppencHx 1 Fission Product Barrier Remarks Although intentional venting per the EOPs in EAL¹ PC2.2 is a voluntary loss of the primary containment boundary, declaration of an Unusual Event at the Drywell Pressure Limit {DWPL) or combustible gas concentrations requires an emergency response beyond the Unusual Event requirements. Drywell pressure above the scram setpoint is an indication of a loss of the RCS barrier (EAL¹ RCS2.1). Loss of the RCS barrier is always an Alert declaration. It is reasonable to assume that the DWPL and combustible gas concentrations will always be reached with drywell pressure above 3.5 psig. Since the RCS2.1 will always be reached before PC2.2, EAL¹ PC2.2 is unnecessary and can be deleted.

2. Although unisolable primary system leakage outside the drywell as indicated by secondary containment radiation levels at the maximum safe operating level in EAL¹ PC2.3 is a loss of the primary containment, EAL¹ RCS1.3 requires an Alert declaration at the maximum normal operating radiation level. Since RCS1.3 will always be reached before PC2.3, EAL¹ PC2.3 is unnecessary and can be deleted.

3. Although drywell pressure above the DWPL and the presence of combustible gas concentrations is an indication of a potential loss of the primary containment boundary, emergency classiQcation at these limits requires an emergency response beyond the Unusual Event.

Drywell pressure above the scram setpoint is an indication of a loss of the RCS barrier (EAL¹ RCS2.1). Loss of the RCS barrier is always an Alert declaration. It is reasonable to assume that the drywell pressure at the DWPL and combustible gas concentrations will always be reached with drywell pressure above the scram setpoint. Since the RCS2.1 will always be reached before PC1.3 and PC1.4, EAL¹s PC1.3 and PC1.4 are unnecessary and can be deleted.

4. EAL¹ PC3.1 would require an Unusual Event declaration at a containment radiation level which is well in excess of that required for the loss of RCS. Since loss of RCS is an Alert classiQcation, EAL¹ PC3.1 is unnecessary and can be deleted.

5. Entry to the Drywell Flooding EOP is identified in EAL¹ PC4.1 as a condition representing an imminent melt sequence where RPV water level cannot be restored above the top of active fuel. This potential loss EAL requires an Unusual Event declaration. However, EAL¹ FC2.1 requires an Alert declaration when RPV water level is less than the top of active fuel. Since FC2.1 will always be reached before PC4.1, MHt PC4.1 is unnecessary and can be deleted,

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Appendix 1 Fission Product Barrier Remarks

6. A main steam line break inside the primary containment would result in drywell pressure above the scram setpoint and is addressed by EAL¹ RCS2.1. A main steamline break outside primary containment would result in a loss of two Qssion product barriers and is addressed by the combination of conditions requiring a Site Area Emergency.

Therefore, this EAL is unnecessary and can be deleted.

7. EAL¹ FC3.1 and EAL¹ RCS3.1 identify drywell radiation monitor readings requiring an Alert classification. Since the monitor reading in EAL¹ FC3.1 is always greater than that used in EAL¹ RCS3.1, EAL¹ FC3.1 is unnecessary and can be deleted.

8. RPV water level less than TAP is a Site Area Emergency based on EAL¹ SS5.1. Therefore, this portion of the EAL is unnecessary and can be deleted.

9 EAL¹ FC2.1 and EAL¹ RCS4.1 identify RPV water level less than TAF as a condition requiring an emergency classiQcation. Since they are the same condition, the appropriate classiQcation is provided at the Alert level under EAL¹ FC2.1. Therefore, this combination of conditions as a Site Area Emergency classification is unnecessary and can be deleted.

10. EAL¹ FC3.1 and EAL¹ RCS3.1 identify drywell radiation as a condition requiring an emergency classification. since they are the same condition, the appropriate classification is provided at the Alert level under RCS3.1. Therefore, this combination of conditions as a Site Area Emergency classification is unnecessary and can be deleted.

FC3-loss + RCS4-loss is identical to FC2-loss + RCS3-loss. Since these Site Area Emergency conditions are redundant, FC3-loss + RCS4-loss can be deleted.

12. The emergency director has the latitude to declare an emergency classification at any level based on his assessment of combinations of plant conditions. Therefore, any judgement decision involving FC5-loss and another condition is the same as the judgement made for FC5-loss alone and can be deleted.

13. EAL¹ PC2.3 and EAL¹ RCS1.3 (which addresses area temperatures and radiation levels at the maximum safe operating level) are redundant.

Since either condition warrants declaration of a Site Area Emergency by themselves, this EAL combination can be deleted.

1-2

OSSI 92-402A-2-NMP2 NMP2 Fission Product Harrier EAL Evaluation, Rev. 0 Appendix 1 Fission Product Barrier Remarks

14. N/A

15. RCS leakage into the drywell must also result in a high drywell pressure above the scram setpoint. This condition is addressed under

$ MLN RCS2.1. Therefore, this condition is unnecessary and can be deleted.

16. For leaks inside the drywell this combination of conditions is adequately addressed under EAL¹ FC3.1. For leaks outside the drywell with successful containment isolation this combination would be adequately covered under ASl.l. For conditions in which the containment does not sucessfully isolate, a General Emergency would be required.

17. N/A

18. The drywell radiation level given in EAL¹ RCS3.1 is less than the drywell radiation level associated with the coolant activity of EAL¹ FCl.l. Eh'C1.1 coolant activity combined with EAL¹ RCS3.1 is adequately addressed by EAL¹ FC3.1.

19. EAL¹ FC3.1 is based on all of the coolant activity of EAL¹ FCI.1 deposited into the primary containment. Such a condition must result from the loss of the fuel clad and RCS barriers. Therefore, EAL¹ RCS1.1 is unnecessary for the Site Area Emergency condition and can be deleted.

20. RCSla.pot. loss is > 50 gpm drywell leakage. FC4 loss is very high offgas activity. High offgas activity under conditions where steam flow to the main condenser is ongoing (i.e. off gas readings valid) alone is indicative of a MSL faQure to isolate with downstream pathway to the environment. This condition requires declaration of a Site Area Emergency under EAL PC2.1. Therefore, this combination of conditions is unnecessary and can be deleted.

21. FaQure of a steamline to isolate with a direct path to the environment can only occur with the loss of the Primary Containment boundary and the loss of the RCS boundary. By deflnition, this combination of conditions by itself requires declaration of a Site Area Emergency.

Therefore, declaration of the Unusual Event is unnecessary and any Site Area Emergency combination of this condition can be deleted.

1-3

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAI Evaluation, Rev. 0 Appendix 1 Fission Product Barrier Remarks

22. To intentionally vent the primary containment in accordance with the EOPs, two Qssion product barriers must have been lost and a third barrier is about to be lost due to venting. By deQnition, this combination of losses warrants declaration of a General Emergency.

23. The combination of a primary system discharging into secondary containment and secondary containment parameters at the maximum safe operating levels is a loss of two barriers. By deflnition, this requires a Site Area Emergency declaration. EAL¹ PC2.1 is equivalent to this combination of conditions.

24. Offgas monitors are not a reliable indicator of fuel failure under severely degraded conditions in that the system would be isolated and process monitors would not be monitoring an unisolated process stream. High offgas activity under conditions where steam flow to the main condenser is ongoing (i.e. off gas readings valid) alone is indicative of a MSL failure to isolate with downstream pathway to the environment. Therefore this condition requires declaration of a Site Area Emergency under EAL PC2.1.

25. Primary containment pressure at or above design or the presence of combustible gas concentrations each requires venting of the primary containment in accordance with the EOPs. Loss of two fission product barriers must have occurred and it must be assumed that the fuel clad barrier is lost or about to be lost. Therefore, EAGAN PC1.3, EAL¹ PC1.4 or EAL ¹ PC2.2 alone warrants declaration of a General Emergency.

26. According to the NUMARC guidance given in the basis for IC¹ PC3, the level of activity deposited in the primary containment as a result of the condition of EAL¹ PC3.1 warrants declaration of a General Emergency.

27. Drywell Flooding is required when means of restoring and maintaining adequate core cooling cannot be established. This condition is a direct precursor to core melt which warrants declaration of a General Emergency.

28. EAL¹ PC2.1 or EAL PC2.3 is a loss of the RCS and primary containment. EAL¹ FC1.1, FC2.1 and FC3.1 are each losses of the fuel clad. These conditions alone meet the definition of a General Emergency. Therefore, any combinations of these EALs are redundant and can be deleted.

1-4

OSSI 92-402A-2-NMP2 NMP2 Fission Product Barrier EAL Evaluation, Rev. 0 Appendix 1 Fission Product Barrier Remarks

29. This combination of conditions is a subset of the previously listed combination (EAL¹ PC2.1 and EAL¹ FCl.l) and can, therefore, be deleted.

30. This combination of conditions is a subset of the previously listed combination (EAL¹ PC2.1 and EAL¹ FC2.1) and can, therefore, be deleted.

31. This combination of conditions is a subset of the previously listed combination (EAL¹ PC2.1 and EAL¹ FC3.1) and can, therefore, be deleted.

32. The combination of a primary system discharging into secondary containment and secondary containment parameters at the maximum safe operating levels is a loss of two barriers. RPV water level less than the top of active fuel is a potential loss of a third barrier. By definition, this requires a General Emergency declaration.

33. The combination of a primary system discharging into secondary containment and secondary containment parameters at the maximum safe operating levels is a loss of two barriers. Elevated coolant activity is a potential loss of a third barrier. By deAnition, this requires a General Emergency declaration.

34. The combination of a primary system discharging into secondary containment and secondary containment parameters at the maximum safe operating levels is a loss of two barriers. Elevated primary containment radiation is a potential loss of a third barrier. By definition, this requires a General Emergency declaration.

35. EAL ¹PC2.1 or EAL ¹PC2.3 in combination with any of EM@ FCl.l, FC2.1 or FC3.1 has previously been evaluated as justiQcation of General Emergency. Therefore this combination of conditions is redundant and can be deleted.

1-5

OSSI 92-402A-2-NMP 1 NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 Appendix 2 Suaunary of Fission Product Barrier Evaluation The following summarizes the EALs which resulted from the analysis performed of the fission product barrier methodology of NVMARC-007 for NMP-2:

~ Emergency Director Judgement FC1.1-loss FC4.1-loss RCS2.1-loss RCS3.1-loss Emergency Director Judgement

~ FC2.1-loss

~ FC3.1-loss

~ RCS2.1-loss

~ PC2.1-loss

~ PC2.3-loss

~ Emergency Director Judgement

OSSI 92-402A-2-NMP1 NMP1 Fission Product Barrier EAI. Evaluation, Rev. 0 Appendix 2 Suamaaxy of Fission Product Barrier Evaluation

~ PC1.3-pot. loss

~ PC1.4-pot. loss

~ PC3.1-pot. loss

~ PC4.1-pot, loss

~ PC2.1-loss + FC1.1-loss, FC2.1-loss or FC3.1-loss

~ PC2.3-loss + FC1.1-loss, FC2.1-loss or FC3.1-loss Emergency Director Judgement 1-2

OSSI 92<02A-3-BWR BWR EAL Binning Document, Rev. 0 EAL Binnin Document KQY>~>R2 New York Power Authori J; A. EErpatrtck Nuclear Power Plant Indian Point Nuclear Power Plant Unit 3 Ni ara Mohawk Power Co oration Nine Mile Point Unit 1 Nine Mile Point Unit 2 Consolidated Edison Com an Indian Point Station Unit 2 Rochester Gas and Electric Com an R. E. Ginna Nuclear Power Station Operations Support Services, Inc.

233 Water Street 2nd floor Plymouth, MA 02360

OSSI 92<02A-3-BWR BWR EAL Binning Document, Rev. 0 1.0 Reactor Fuel Coolant Activity SU4.2 FC1.1 (Alert)

Off-gas Activity SU4.1 FC4.1 (Alert)

Contaixunent Radiation RCS3.1 (Alert)

FC3.1 (SAE)

PC3.1 (General)

Other Radiation Monitors AU2.4 AA2.1 AA3.1 AA3.2 Refueling Accidents AU2.1 AA2.2 2.0 Reactor Pressure Vessel 2.1 RPV Water Level SU5.1 SS5.1 FC2.1 (SAE)

PC4.1 (General) 2.2 Reactor Power/Reactivity Control SA2.1 SS2.1 SG2.1

OSSI 92<02A-3-BWR BWR EAL Binning Document, Rev. 0 3.0 Contaixxment 8.1 Primary Contaixxment Pressure RCS2.1 (Alert)

FC1.1 + RCS2.1 (SAE)

PC1.3 (General)

PC2.2 (General) 3.2 Suppression Pool Temperature SS4.1 (SAE) I 8.8 Combustible Gas Concentration SS5.2 (SAE)

PC1.4/PC2.2 (General)

Contaixxment Isolation Status PC2.1 (SAE)

PC2.1+ FC1.1 (General)

PC2.1+ FC2.1 (General)

PC2.1+ FC3.1 (General) 4.0 Secondary Containment Reactor Building Temperatures PC2.3 (Temp)/RCS1.3 (SAE)

PC2.3 + FC1.1 (Temp) (General)

.PC2.3 + FC2.1 (Temp) (General)

PC2.3 + FC3;1 (Temp) (General) 4.2 Reactor Building Radiation Levels PC2.3 (Rad)/RCS1.3 (SAE)

PC2.3 + FC1.1 (Rad) (General)

PC2.3 + FC2.1 (Rad) (General)

PC2.3 + FC3.1 (Rad) (General)

OSSI 92-402A-3-BWR BWR EAL Binning Document, Rev. 0 5.0 Radioactivity Release 5.1 Effluent Monitors AU1.1 AA1.1 AS1.1 AG1.1 5.2 Dose Projections/ Environmental Measurements AU1.2 I AA1.2 AS1.3 AS1.4 AG1.3 AG1.4 6.0 Electrical Failures 6.1 'Loss of AC Power Sources SU1.1 SA1.1 SA5.1 SS1.1 SG1.1 6.2 Loss of DC Power Sources SU7.1 SS3.1 7.0 Equipment Failures 7.1 Technical Speci6cation Requirements SU2.1 7.2 System Failures or Control Room Evacuation HU1.6 HA5.1 SA3.1 HS2.1 7.3 Loss of Indications/Alarm/Communication Capability SU3.1 SU6.1 SA4.1 SS6.1

OSSI 92-402A-3-BWR BWR EAL Binning Document, Rev. 0 8.0 Hazards 8.1 Security Threats HU4.1 HU4.2 HA4.1 HA4.2 HS1.1 HS1.2 HG1.1 HG1,2 8.2 Fire or Explosion HU2.1 HA2.1 8.8 Man-Made Events HU1.4 HU1.5

=HU3.1 HU3.2 HA1.5 HA3.1 HA3.2 8.4 'atural Events HU1.1

.HU1.2 HU1.3 HU1.7 HA1.1 HA1.2 HA1.3 HA1.7 9.0 Other HU5.1 PC6.1 HA6.1 FC5.1 RCS6.1 HS3.1 PC1.1 PC1.2 HG2.1

NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT NUCLEAR STATION EMERGENCY PLANNING MAINTENANCE PROCEDURE EPMP-EPP-0102 REVISION 00 UNIT 2 EMERGENCY CLASSIFICATION TECHNICAL BASIS TECHNICAL SPECIFICATION REQUIRED Approved by:

J. H. Mueller Plant nager U i 2 Date Effective Date:

NOT TO BE USED AFTER SUBJECT TO PERIODIC REVIEW

LIST OF EFFECTIVE PAGES

~PN . JCC N ~PN . ~CI N ~PN. ~Ch N.

Coversheet . 22 ~ ~ ~ ~ 47 .

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=10 .~ -

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15 . 40 . 65 .

16 . 41 66 .

17 ~ ~ ~ ~ 42 . 67 18 . 43 ~ ~ ~ ~ 68 .

19 . 44 . 69 .

20 . 45 . 70 .

21 46 . 71 Hay 1994 Page i EPHP-EPP-0102 Rev 00

LIST OF EFFECTIVE PAGES (Cont)

~PN . ~CN N ~PN . ~CN N ~PN . ~CN N 72 . 97 ~ ~ 0 ~

73 . 98 ~ o ~ ~

74 . 99 75 ~ ~ ~ ~ 100 .

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81 106 .

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Nay 1994 Page ii EPMP-EPP-0102 Rev 00

TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE . 1 2.0 PRIMARY RESPONSIBILITY 1 3.0 PROCEDURE . 1 3.1 Emergency Preparedness Group . . . . . . 1 3.2 Responsible Procedure Owner for Technical Basis References . 1 4.0 DEFINITIONS .

5.0 REFERENCES

AND COMMITMENTS ~ ~ ~ 2 6.0 RECORD REVIEW AND DISPOSITION . 2 ATTACHMENT 1: UNIT 2 EMERGENCY ACTION LEVEL TECHNICAL Basis....... 3 INT ODUCTION PURPOSE . ~ ~ ~ 3 DISCUSSION ~ ~ ~ 3 NA IINI AL BA I 1.0 REACTOR FUEL 8 2.0 REACTOR PRESSURE VESSEL (RPV) ~ ~ ~ ~ 23 3.0 PRIMARY CONTAINMENT (PC) 29 4.0 SECONDARY CONTAINMENT (SC) 38 5.0 RADIOACTIVITY RELEASE . 44 6.0 ELECTRICAL FAILURES . ~ ~ ~ ~ 56

7. 0 E(UIPMENT FAILURES 63 8.0 HAZARDS . ~ ~ ~ ~ 74 9.0 OTHER . 94 ATTACHMENT 2: WORD LIST/DEFINITIONS . 101 May 1994 Page iii EPMP-EPP-0102 Rev 00

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'EPIP-EPP-02, Emergency Action Level Matrix/Unit, and Attachment 1 of this procedure, as needed.

4. 0 DEFINITIONS See Attachment 2.

May 1994 Page 1 EPMP-EPP-0102 Rev 00

5.0 REFERENCES

AND COMMITMENTS 5.1 Licensee Documentation None 5.2 Standards Re ulations and Codes NUHARC 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.

5.4 Su lemental References Nine Nile Point Unit 2 Plant-Specific EAL Guideline 5.5 Commitments None 6.0 ECO D REVIEW AND DISPOSITION None Nay 1994 Page 2 EPNP-EPP-0102 Rev 00

ATTACHMENT 1 UNIT 2 EMERGENCY ACTION LEVEL TECHNICAL BASIS URPOSE 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.

DISCUSSIO 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 NHP-2 plant interpretation of the NUHARC methodology for developing EALs.

Hay '1994 Page 3 EPHP-EPP-0102 Rev 00

STTACHMtNT 1 (C t)

DISCUSS IO (Cont)

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.

C. imar Containment PC  : The primary containment is comprised of the drywell, suppression 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.

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 May 1994 Page 4 EPMP-EPP-0102 Rev 00

ATTACHHENT I (Cont).

DISCUSSION (Cont)

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 NHP-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.

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 Hay 1994 Page 5 EPHP-EPP-0102 Rev 00

ATTACHMENT I (C t)

DISCUSSION (Cont)

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 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).

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 Nay 1994 Page 6 EPNP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) dtdtddtltd tt tl The identified operating modes are defined as follows:

Power 0 eratio s 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.

~tdt td Node switch is usually in SHUTDOWN and reactor coolant temperature is >200 'F.

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 8140 'F.

~ef cled RPV contains no irradiated fuel.

May 1994 Page 7 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 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 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.

~pf-will A It: A ~ ith 1 t titty, yf 1fi1 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 Accide ts: 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 1994 Page 8 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont)

Coolant Activit l.l.l E~E Coolant activity > 0.2 pCi/gm I-131 equivalent or >100/Ebar pCi/gm 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.

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 1994 Page 9 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 1.1.2 ~1ert Coolant activity > 300 pCi/gm 1-131 equivalent Node Applicability:

Power operation, 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 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 May 1994 Page 10 EPMP-EPP-0102 Rev 00

STTA IINENT 1 (C t) 1.2 Off- as Activit 1.2.1 Unusual Event Valid offgas radiation high alarm (at DRMS red).

Node Applicability:

All 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 ORMS 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 ODCM.

PEG Reference(s):

SU4.1 Basis Reference(s):

1. NUREG-1253 Technical Specifications Nine Mile 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 Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Article 3.4.5.a and b

4. NUREG-1253 Technical Specifications Nine Mile 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 May 1994 Page ll EPMP-EPP-0102 Rev 00

ETTA IIIIENT 1 (C 1.3 Containment Radiation 1.3.1 alert Drywell area radiation ~ 41 R/hr 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.

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 2RHS*RElB/D RHS*RUZ1B RHS*RUZ1D 2CEC*Pnl880B: DRMS 2RMS*REIA/C RMS*RUZIA RMS*RUZlC PEG Reference(s):

RCS3.1 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 Hay 1994 Page 12 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 1.3.2 Site Area Emer enc Drywell area radiation > 3100 R/hr Node Applicability:

Power operation, startup/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 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.

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 2RHS*RE1B/D RMS*RUZ1B RMS*RUZ1D 2CEC*Pnl880B: DRMS 2RMS*RElA/C RMS*RUZ1A RMS*RUZlC 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 May 1994 Page 13 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 1.3.3 General Emer enc Drywell area radiation > 5.2E6 R/hr Node Applicability:

Power operation, startup/hot standby, 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 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 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,'his 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 2RHS*RE1B/D RHS*RUZ1B RHS*RUZ1D 2CEC*Pnl880B: ORMS 2RMS*RE1A/C RMS*RUZ1A RHS*RUZlC 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 1994 Page 14 EPHP-EPP-0102 Rev 00

CTTNCIINENT I EC tt 1.4 Other Radiation Monitors 1.4.1 Unusual Event Any sustained ARH reading > 100 x DRMS high radiation alarm (red) or offscale high (DETECTOR SATURATION) resulting from an uncontrolled process Mode Appl i cabi 1 i ty:

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. EPIP-EPP-13, pg. 84

3. Calculation PR-C-25-1 Hay 1994 Page 15 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 1.4.2 A1ert 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 1.1 Table 1.1 Re e oor d on'tors ARM RMSlll, RB 354'est of Spent Fuel Pool ARM RHS112, RB 354'ast of Spent Fuel Pool 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 via Emergency Director judgment in EAL Category 9.0.

The basis for the reactor building ventilation monitor setpoint is a spent fuel handling accident (isolation setpoint) and is, therefor e, appropriate for this EAL. Technical Specification requires isolation at < 2.36 E-3 pCi/cc).

Hay 1994 Page 16 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 1.4.2 (Cont)

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 Hay 1994 Page 17 EPHP-EPP-0102 Rev 00

NTTACIIIIENT I EC t) 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)

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 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 Alelt.

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 EAL 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 planned events (e. g., radwaste container movement, depleted resin transfers, etc.).

Hay 1994 Page 18 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) ~

1.4.3 (Cont)

PEG Reference(s):

AA3. 1 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 R/hr in any areas, Table 1.2 AND Access is required for safe operation or shutdown Table 1.2 Plant Sa et Fu ction 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 Node Applicability:

All 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 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 NMPC 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.

Hay 1994 Page 19 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 1.4.4 (Cont) I 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 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 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'ile Code NHP31027 "Exposure Guidelines For Unusual/Accident Conditions", Revision 1, 3/18/93 1.5 Refue in 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 Node Applicability:

All Hay 1994 Page 20 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 1.5.1 (Cont)

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.

Node Applicability:

All May 1994 Page 21 EPMP-EPP-0102 Rev 00

ATTACHMENT t (C t) 1.5.2 (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 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.

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 May 1994 Page 22 EPMP-EPP-0102 Rev 00

NTTA NNTNT I IC tt 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:

APPW I I: NPII status of adequate core cooling, I 11 and dt tly I tdt tt 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.

eactor Power React'vit Contro : 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 Level N.l.l Unidentified drywell leakage > 10 gpm OR Reactor coolant to drywell identified leakage > 25 gpm Node Applicability:

Power operation, startup/hot standby, 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 the unidentified drywell leakage was selected because it is observable with normal Control Room indications. Smaller values must be determined through time-consuming surveillance tests (e. g., mass balances). 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.

Hay 1994 Page 23 EPHP-EPP-0102 Rev 00

STTACHNENT 1 lC tl 2.1.1 (Cont)

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 > -14 in. (TAF)

Node Applicability:

Power Operation, Startup/Hot Standby, 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 if RPV water level is not maintained > TAF.

Uncovery of the fuel irrespective of the event that causes fuel uncovery is justification alone for declaring a Site Area Emergency.

This includes e'vents 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.

PEG Reference(s):

SS5.1 FC2.1 Nay 1994 Page 24 EPHP-EPP-0102 Rev 00

STTACIINEIIT 1 (C t) 2.1.2 (Cont)

Basis Reference(s):

1. N2-EOP-RPV, RPV Control 2.1.3 General Emer enc Primary Containment Flooding required Node Applicability:

Power operation, startup/hot standby, hot shutdown Basis:

The condition in this EAL represents imminent melt sequences which, not corrected, could lead to RPV failure and increased potential for if 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 All immediate manual scrams fail to shut down the reactor Node Applicability:

Power operation, startup/hot standby May 1994 Page 25 EPHP-EPP-0102 Rev 00

A~TTACAAIA I TC tt 2.2.1 (Cont)

Basis:

This condition indicates failure of the automatic and/or manual protection system to scram the reactor to the extent which precludes the reactor being made sub-critical. It is the continued criticality under conditions requiring a reactor scram which poses the potential threat to RCS or fuel clad integrity. This condition is more than a potential degradation of a safety system. A front line automatic protection system did not function in response to a plant transient, and thus plant safety has been compromised and design limits of the fuel may be exceeded. An Alert is indicated because conditions exist that lead to a potential loss of the fuel clad barrier or the RCS barrier.

An immediate 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.

PEG Reference(s):

SA2.1 Basis Reference(s):

,1. N2-EOP-RPV, RPV Control, Section RL 2.2.2 Site Area Emer enc All immediate manual scrams fail to shut down the reactor AND Boron injection is required Node Applicability:

Power operation, startup/hot standby Basis:

This condition indicates failure of the automatic and/or manual protection system to scram the reactor to the extent which precludes

,the reactor being made subcritical. Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed. A Site Area Emergency is indicated because conditions exist that lead to imminent loss or potential loss of both fuel clad and primary containment.

Nay 1994 Page 26 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 2.2.2 (Cont)

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 2.2.3 General Emer enc All immediate manual scrams fail to shut down the reactor AND RPV water level cannot be restored and maintained > -39 in.

Node 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.

Hay 1994 Page 27 EPHP-EPP-0102 Rev 00

ATTACHMENT I (Cont) 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 (-39 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. 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 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. N2-EOP-C5, Level/Power Control 2.2.4 Gene a Emer enc All immediate manual scrams fail to shut down the reactor AND Suppression pool temperature and RPV pressure cannot be maintained

< HCTL Node 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.

May 1994 Page 28 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 2.2.4 (Cont)

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 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 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. N2-EOP-C5, Level/Power Control

.2. N2-EOP-PC, Primary Containment Control 3.0 PRINARY 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.

May 1994 Page 29 EPMP-EPP-0102 Rev 00

ATTACHNENT 1 (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.

~ rimar 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.

Co t inme t Isolat o 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 Containment Pressure 3.1.1 Al ert Primary containment pressure'cannot be maintained < 1.68 psig due to coolant leakage 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.

Nay 1994 Page 30 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 3.1.1 (Cont)

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 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 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 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 May 1994 Page 31 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 3.1. 2 (Cont)

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 Ge eral Emer enc Primary containment venting is required due to PCPL Node 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 S te rea Emer enc RPV pressure and suppression pool temperature cannot be maintained

( HCTL (non-ATWS)

Node Appl icabil ity:

Power operation, startup/hot standby, hot shutdown Nay 1994 Page 32 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont).

3.2.1 (Cont)

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 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):

'-.l. USAR, Revision 2, Section 9B.2

2. 'SAR, Revision 2, Section 9B.4.3

3. N2-EOP-PC, Primary Containment Control 3.3 Combustible Gas Concentration 3.3.1 Site Area Emer enc

> 4A Hz exists in DM or suppression chamber Node Applicability:

All May 1994 Page 33 EPHP-EPP-0102 Rev 00

ATTACHHENT 1 (Cont) 3.3.1 (Cont)

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 h 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.

PEG Reference(s):

SS5.2 Basis Reference(s):

1. N2-EOP-PC, Primary Containment Control, Revision 5 3.3."2 Genera Emer enc Primary containment venting is required due to combustible gas concentrations 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 concentr ations 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.

Hay 1994 Page 34 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 3.3.2 (Cont)

PEG Reference(s):

PC1.4 PC2.2 Basis Reference(s):

1. N2-EOP-PC, Primary Containment Control 3.4 Containment Isolatio Status 3.4.1 Site Area Emer enc Main Steam Line, RCIC steam line or Reactor Water Clean-up isolation failure resulting in a release pathway outside primary containment.

Node 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 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 May 1994 Page 35 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 3.4.2 Gene al Emer enc Hain Steam Line, RCIC steam line or Reactor Mater Clean-up isolation failure resulting in a release pathway outside primary containment AND any:

~ Coolant activity > 300 pCi/gm I-131 equivalent

~ RPV water level < -14 in. (TAF)

~ DW radiation > 3100 R/hr Node 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 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.

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 SX 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.

Hay 1994 Page 36 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 3.4.2 (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 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 Monito} s are installed in the following locations:

2CEC*Pnl880D: DRHS 2RHS*RE1B/D RHS*RUZ1B RHS*RUZ1D 2CEC*pnl880B: ORMS 2RHS*RElA/C RHS*RUZlA RHS*RUZ1C PEG Reference(s):

PC2.1 and FCl.l PC2.1 and FC2.1 PC2.1 and FC3.1 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 Mile 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 May 1994 Page 37 EPHP-EPP-0102 Rev 00

NTIACNIIENT I (C I) 4.0 SECOND RY CONTAINNENT 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 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 Conta nment 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 Radiatio 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.

4.1 Reactor Buildin Tem erature 4.1.1 Site Area Emer enc Primary system is discharging outside PC AND RB area temperatures are > maximum safe operating levels in two or more areas, N2-EOP-SC Node Applicability:

.Power operation, startup/hot standby, hot shutdown May 1994 Page 38 EPMP-EPP-0102 Rev 00

CTTACIIMIMT I TC tt 4.1.1 (Cont)

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. N2-EOP-SC, Secondary Containment Control

2. N2-EOP-SC

4. 1.2 General Emer enc Primary system is discharging outside PC AND RB area temperatures are > maximum safe operating levels in two or more areas, N2-EOP-SC AND any:

~ Coolant activity > 300 pCi/gm I-131 equivalent

~ RPV water level < -14 in. (TAF)

~ DW radiation > 3100 R/hr 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.

May 1994 Page 39 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 4.1.2 (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. 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).

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*RElA/C RHS*RUZlA RHS*RUZ1C PEG Reference(s):

PC2.3 and FCl.l PC2.3 and FC2.1 PC2.3 and FC3.1 May 1994 Page 40 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 4.1.2 (Cont)

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 4.2 Reacto Bui din ad ation Leve 4.2.1 Site rea Emer enc Primary system is discharging outside PC AND RB area radiation levels are > maximum safe operating levels in two or more areas, N2-EOP-SC 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 May 1994 Page 41 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 4.2.1 (Cont)

Basis Reference(s):

N2-EOP-SC, Secondary Containment Control 4.2.2 Ge e a er e c Primary system is discharging outside PC AND RB area radiation levels are > maximum safe operating levels in two areas, or'ore N2-EOP-SC AND any:

~ Coolant activity > 300 pCi/gm I-131 equivalent

~ RPV water level < -14 in. (TAF)

~ DW radiation > 3100 R/hr 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 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.

May 1994 Page 42 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 4.2.2 (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 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 are installed in the following locations:

2CEC*Pnl880D: DRMS 2RMS*RElB/D RHS*RUZlB RHS*RUZ1D 2CEC*Pnl880B: DRHS 2RHS*RE1A/C RHS*RUZIA RMS*RUZ1C PEG Reference(s):

PC2.3 and FC1.1 PC2.3 and FC2.1 PC2.3 and FC3.1 May 1994 Page 43 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 4.2.2 (Cont)

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

6. N2-EOP-SC 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 precur sors 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.

E fluent 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.

May 1994 Page 44 EPMP-EPP-0102 Rev 00

ETT CMMEET I TC tt 5.1 Effluent Monitors E.l.l ~EE A valid reading on any monitors Table 5. 1 column "UE" for > 60 min.

Table 5.1 Effluent Monitor Classification Thresholds Monitor UE Al ert 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 DRMS High (red) 200 x DRMS High (red) N/A N/A Liquid RadWaste Effluent 2 x DRMS High (red) 200 x DRMS High (red) N/A N/A Cooling Tower Blowdown 2 x DRMS High (red) 200 x DRMS High (red) N/A 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.

May 1994 Page 45 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 5.1.1 (Cont)

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 (DRHS).

Instrumentation that may be used to assess this EAL is listed below:

Radwaste/Reactor Building Vent Effluent Monitoring System monitor: 2RMS-CAB180 recorder: 2RHS-RR170/180 annunciator: 851248 Main Stack Effluent Monitoring System monitor: 2RHS-CAB170 recorder: 2RHS-RR170/180 annunciator: 851256 Service Water Effluent Loop A/B Radiation monitor: 2SWP*RE146A/B recorder: 2SWP*RR146A/B annunciator: 851258 Liquid Effluent Line monitor: LWS-RE206 annunciator: 851258 Cooling Tower Blowdown Line monitor: CWS-RE 157 annunci ator: 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 May 1994 Page 46 EPMP-EPP-0102 Rev 00

CTTACIIIITIIT I CC tl 5.1.2 alert A valid reading on any monitors Table 5. 1 column "Alert" for > 15 min.

Table 5.1 Effluent Monitor Classification Thresholds Monitor UE Alert SAE GE Radwaste/Reactor Bldg.

Vent Effluent 2 x GEMS alarm 200 x GEMS alarm ~5.5E6 pCi/s N/A Hain Stack Effluent 2 x GENS alarm 200 x GEMS alarm N/A N/A Service Water Effluent 2 x DRHS High (red) 200 x DRMS 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 ORMS High (red) 200 x DRHS High (red) N/A 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:

Radwaste/Reactor Building Vent Effluent Monitoring System monitor: 2RMS-CAB180 recorder: 2RHS-RR170/180 annunciator: 851248 Hay 1994 Page 47 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont)

5. 1. 2 (Cont)

Main 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):

AA1.1 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 May 1994 Page 48 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 5.1.3 Site Area Emer enc A valid reading on any monitors Table 5. 1 column "SAE" for > 15 min.

Table 5.1 Effluent Monitor Classification Thresholds Monitor UE Al ert 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 DRMS High (red) 200 x DRMS High (red) N/A N/A Liquid RadWaste Effluent 2 x DRMS High (red) 200 x DRMS High (red) N/A N/A Cooling Tower Blowdown 2 x ORMS High (red) 200 x ORMS High (red) N/A N/A Node Applicability:

All 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 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 PR-C-24-X, Rev. 2.

May 1994 Page 49 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 5.1.3 (Cont)

PEG Reference(s):

AS1.1 Reference(s):

1. N2-0P-79, Radiation Monitoring System 2'asis2. 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

5. Dose P o ections Environmenta easureme ts 5:2:1 ~Ul E Confirmed sample analyses 'for gaseous or liquid release rates > 2 x technical specifications limits for > 60 min.

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 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.

May 1994 Page 50 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 5.2.1 (Cont)

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 5.2.2 Alert Confirmed sample analyses for gaseous or liquid release rates > 200 x technical specifications limits for > 15 min.

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 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 Hay 1994 Page 51 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 5.2.2 (Cont)

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 5.2.3 Alert Dose projections or field surveys which indicate doses / dose rates >

Table 5.2 column "Alert" at the site boundary or beyond Table 5.2 Dose Pro 'ection Env. Measurement Classification Thresholds Alert SAE GE TEDE 10 mR 100 mR 1000 mR CDE Thyroid N/A 500 mR 5000 mR TEDE rate 10 mR/hr 100 mR/hr 1000 mR/hr CDE Thyroid rate N/A 500 mR/hr 5000 mR/hr 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 basis of 10CFR20 for both time (8766 hr/yr) and the 200 multiplier, the associated site boundary dose rate would be 10 mR/hr. h 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.

May 1994 Page 52 EPMP-EPP-0102 Rev 00

BITACHIIENT I (C t) 5.2.3 (Cont)

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 which indicate doses / dose rates >

Table 5.2 column SAE" at the site boundary or beyond Table 5.2 Dose Pro 'ection Env. Measurement Classification Thresholds Alert SAE GE

'TEDE 10 mR 100 mR 1000 mR CDE Thyroid N/A 500 mR 5000 mR TEDE rate 10 mR/hr 100 mR/hr 1000 mR/hr CDE Thyroid rate N/A 500 mR/hr 5000 mR/hr Node Applicability:

All May 1994 Page 53 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 5.'2.4 (Cont)

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.

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 which indicate doses / dose rates in excess of Table 5.2 column "GE" at the site boundary or beyond Table 5.2 Dose ro 'ection Env. Measurement Classification Thresholds Alert SAE GE TEDE 10 mR 100 mR 1000 mR CDE Thyroid N/A 500 mR 5000 mR TEDE rate 10 mR/hr 100 mR/hr 1000 mR/hr CDE Thyroid rate N/A 500 mR/hr 5000 mR/hr Node Applicability:

All Hay 1994 Page 54 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 5.2.5 (Cont)

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 CDE 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 description CDE for thyroid. This is consistent with the emergency class 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.

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 May 1994 Page 55 EPMP-EPP-0102 Rev 00

ATTACHMENT I tC tj 6.0 ELECTRIC L 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 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.

Nay 1994 Page 56 EPHP-EPP-0102 Rev 00

ATTACHHENT 1 (Cont) 6.1.1 (Cont)

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 Al ert Loss of all emergency bus AC power for >15 min.

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 for >15 min. to all:

~ Reserve Transformer A

~ Reserve Transformer B

~ Aux Boiler Transformer 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.

Hay 1994 Page 57 EPHP-EPP-0102 Rev 00

STIA IINENT 1 (C t) 6.1. 2 (Cont)

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 Alert Available emergency bus AC power reduced to only one of the following sources for >15 min.:

Reserve Transformer A Reserve 'Transformer B Aux Boiler Transformer 2EGS*EG1 2EGS*EG2 2EGS*EG3 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.

Hay 1994 Page 58 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 6.1.3 (Cont)

PEG Reference(s):

SA5.1 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 6.1.4 Site Area Emer enc Loss of all emergency bus AC power for >15 min.

.Node Applicability:

Power operation, startup/hot standby, 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 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 May 1994 Page 59 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 6.1. 4 (Cont)

Basis Reference(s):

1. N2-0P-100A, Standby Diesel Generators

2. N2-0P-100B, 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 Gene al Emer e Loss of all emergency bus AC power AND either:

Power cannot be restored to any emergency bus in < 2 hr s OR RPV water level cannot be restored and maintained ) -14 in. (TAF)

Node Applicability:

Power operation, startup/hot standby, 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.

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:

May 1994 Page 60 EPMP-EPP-0102 Rev 00

ATTACHHENT 1 (Cont) 6.1. 5 (Cont)

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 imminent7

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.

PEG Reference(s):

SG1.1 Basis Reference(s):

1. N2-0P-74A, Emergency DC Distribution

2. N2-0P-74B, HPCS 125 vdc System

3. N2-0P-70, Station Electrical Feed and 115 kv Switchyard

4. N2-EOP-RPV, RPV Control 6.2 Loss of DC Power Sources 6.2.1 Unusual Event

< 105 vdc on 2BYS*BAT2A and B for >15 min.

Node Applicability:

Cold shutdown, Refuel Hay 1994 Page 61 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 6.2.1 (Cont)

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 oper ation 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 6.2.2 Site Area Emer enc

< 105 vdc on 2BYS*BAT2A and B for > 15 min.

Node Applicability:

Power operation, startup/hot standby, hot shutdown May 1994 Page 62 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 6.2.2 (Cont)

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 Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, Amendment 5, Article 4.8.2.l.d.2

2. NUREG-1253 Technical Specifications Nine Mile 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 7.0 E UIPMENT 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 under this event type related to the failure of the plant to be brought to the required plant operating condition required by technical specifications.

May 1994 Page 63 EPMP-EPP-0102 Rev 00

ATTACHMENT I (Cont) 7.0 (Cont)

~ 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 Tec n ca S ecifications 7.1.1 Unusual Event Plant is not brought to required operating mode within Technical Specifications LCO Action Statement Time Node Applicability:

Power operation, startup/hot standby, hot shutdown 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 May 1994 Page 64 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont)

'.1.1

~ ~ (Cont)

Basis Reference(s):

1. NUREG-1253 Technical Specifications Nine Mile 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 Node Applicability:

Power operation, startup/hot standby, hot shutdown 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 Control Room evacuation Node Applicability:

All May 1994 Page 65 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 7.2.2 (Cont)

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):

HAS.1 Basis Reference(s):

1. N2-0P-78, Remote Shutdown System, Section H.2.0 7.2.3 Alert Reactor coolant temperature cannot be maintained < 200 'F 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.

This EAL does not apply during hydrostatic testing.

PEG Reference(s):

SA3.1 May 1994 Page 66 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) .

7.2.3 (Cont)

Basis Reference(s):

1. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Amendment 26, Article 3.4.9.2

2. NUREG-1253 Technical Specifications Nine Mile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, Table 1.2 7.2.4 Site Area Emer enc Control Room evacuation AND Control of core cooling systems cannot be established in g 15 min.

Node 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 Lette} 88-17, "Loss of Decay Heat Removal"

2. N2-0P-78, Remote Shutdown System, Section H.2.0

3. NMP-2 FSAR Section 9B.8.2.2, Safe Shutdown Scenario, pg. 9B.8-5a, May 1994 Page 67 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 7.3 Loss of Indications Alarm Communication Ca abilit 7.3.1 ~UE 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 Node 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 1994 Page 68 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 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):

1. USAR Figure 1.2-15, Control Room layout

2. N2-0P-91A, Process Computer

3. N2-0P-91B, Safety Parameter Display System (SPDS)

May 1994 Page 69 EPMP-EPP-0102 Rev 00

ATTACHHENT 1 (Cont) 7.3.2 U~1 Loss of all communications capability affecting the ability to either:

Perform routine onsite operations OR Notify offsite agencies or personnel Node Applicability:

Al 1 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 SPC system M/CC system PP/PA system Hand-Held Portable radio 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.).

Hay 1994 Page 70 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 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 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 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.

Hay 1994 Page 71 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 7.3.3 (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 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 25X thermal power change, ECCS injections, or thermal power oscillations of 10X 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):

1. USAR Figure 1.2-15, Control Room layout

2. N2-0P-91A, Process Computer

3. N2-0P-91B, Safety Parameter Display System (SPDS)

May 1994 Page 72 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 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 Node Applicability:

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 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 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.

May 1994 Page 73 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont).

7.3.4 (Cont)

PEG Reference(s):

SS6.1 I

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 P

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.

Man-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 1994 Page 74 EPHP-EPP-0102 Rev 00

ETIACAIIENT I EC I) 8.1 Securit Threats 8.1.1 Unusual Event Bomb device or other indication of attempted sabotage discovered within plant Protected Area Node Applicability:

All Basis:

This EAL is based on the Nine Mile 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.

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.

8.1.2 Alert Intrusion into plant Protected Area by an adversary Node Applicability:

All May 1994 Page 75 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 8.1. 2 (Cont)

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-1 and NMP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also see SEW 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. SN Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89 8.1.3 Site Area Emer enc Intrusion into a plant security vital area by an adversary Node 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.

May 1994 Page 76 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 8.1.3 (Cont)

PEG Reference(s):

HS1.1 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:

Loss of plant control from the Control Room AND Loss of r emote shutdown capability 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 May 1994 Page 77 EPMP-EPP-0102 Rev 00

ATTACHHENT 1 (Cont) 8.2 Fire or Ex losion 8.2. 1 Unusual Event Confirmed fire in any plant area, Table 8.2 or Table 8.3, not extinguished in g 15 min. of Control Room notification Table 8.2 Plant Areas Service Building 115 KV Switchyard 345 KV Switchyard Table 8.3 Plant Vita Areas South Aux. Bay North Aug. Bay RadWaste Building Reactor Building Turbine Building

'Diesel Generator Building Screenwell Building/Service Water Pump Bays Condensate Storage Tanks Building Standby Gas Treatment Building Control Building Unit 2 Security Building Node Applicability:

Al l 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 Hay 1994 Page 78 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 8.2.1 (Cont)

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, Table 8.2 or Table 8.3, which results in damage to plant equipment or structures needed for safe plant operation Table 8.2 Plant Areas Service Building 115 KV Switchyard 345 KV Switchyard Table 8.3 lant Vital Areas 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 Control Building Unit 2 Security Building Node Applicability:

All Basis:

The listed areas contain functions and systems required for the safe shutdown of the plant. The NMP-2 safe shutdown analysis was consulted for equipment and plant areas required for the applicable mode.

May 1994 Page 79 EPMP-EPP-0102 Rev 00

ATTACHHENT 1 (Cont) 8.2.2 (Cont)

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. N2-0P-47, Fire Detection

2. USAR, Figure 9B.6-1

3. USAR, Section 9B

4. NUREG 0737,Section II.B.2-2 8.3 -lid E 8.3.1 Unusual Event Vehicle crash into or projectile which impacts plant structures or systems within Protected Area boundary Node Applicability:

All Basis:

The Protected Area boundary is within the security isolation zone and is defined in the site security plan. NHP-1 and NHP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also, refer to S8W Drawing No. 12187-SK-032483-25; Issue No. 1, Site Facilities Layout Status as of 8/1/89.

Hay 1994 Page 80 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 8.3.1 (Cont)

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.

PEG Reference(s):

HU1.4 Basis Reference(s):

l. USAR Figure 1.2-2 Station Arrangement

2. SEW Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89 S.3.2

.Report by plant personnel of an explosion within Protected Area boundary resulting in visible damage to permanent structures or equipment 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.

May 1994 Page 81 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 8.3.2 (Cont)

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 securit'y aspects of the explosion.

PEG Reference(s):

HU1.5 Basis Reference(s):

l. USAR Figure 1.2-2 Station Arrangement

2. SLW 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 oper ation OR report by local, county or state officials for potential evacuation of site personnel based on offsite event 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 1994 Page 82 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 8.3.3 (Cont)

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 8.3.4 alert Vehicle crash or projectile impact which precludes personnel access to or damages equipment in plant vital areas, Table 8.3 Table 8.3 Plant Vital Areas 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 Control Building Unit 2 Security Building Node Applicability:

All May 1994 Page 83 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 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.

NMP-1 and NMP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also see SSW 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. SKW 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 8.3, in concentrations that will be life threatening to plant personnel or preclude access to equipment needed for safe plant operation Table 8.3 Plant Vital Areas 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 Control Building Unit 2 Security Building May 1994 Page 84 EPMP-EPP-0102 Rev 00

ATTACHNENT 1 (Cont) 8.3.5 (Cont)

Rode Applicability:

All Basis:

This EAL is based on gases that have entered a plant structur e 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):

1. 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 by any operator AND either:

NNP-2 seismic instrumentation actuated OR confirmation of earthquake received on NHP-I or JAFNPP seismic instrumentation Hay 1994 Page 85 EPHP-EPP-0102 Rev 00

ATTACNIIENT I EC tt 8.4. 1 (Cont)

Node Applicability:

All Basis:

NHP-2 seismic instrumentation actuates at 0.01 g causing:

Power to remote acceleration sensor units Activation of MRSl recorders EVENT alarm light on PWRSl 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. Methods of detection can be based on instrumentation validated by a reliable source, operator assessment, or indication received from NMP-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):

HU1.1 Basis Reference(s):

1. N2-0P-90, Seismic Monitoring

2. NUREG-1253 Technical Speci-fications 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 Hay 1994 Page 86 EPMP-EPP-0102 Rev 00

ATTACHHENT 1 (Cont) 8.4.2 Unusual Event Report by plant personnel of tornado striking within plant Protected Area boundary Node 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 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.

NHP-1 and NHP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also see SEW 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 Hay 1994 Page 87 EPHP-EPP-0102 Rev 00

ATTACHHENT 1 (Cont) 8.4.3 Unusual Event Assessment by Control Room personnel that a natural event has occurred precluding access to a plant .vital area, Table 8.4 Table 8.4 Plant Vital Areas 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 Control Building Unit 2 Security Building Node Appl icabi 1 ity:

All Basis:

This EAL allows for the Control Room to determine that an event has occurred and take appropriate action based on personal assessment as opposed to verification (i. e., an earthquake is felt but does not register on any plant-specific instrumentation, etc.).

NHP-1 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):

HU1.3 Hay 1994 Page 88 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 8.4.3 (Cont)

Basis Reference(s):

1. USAR Figure 1.2-1

2. SKW 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 8.4.4 Unusual Event Lake water level > 248 ft OR intake water level < 237 ft 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 8.4.8 Alert Earthquake felt in plant by any operator AND NMP-2 seismic instrumentation indicates > 0.075 g Node Applicability:

All May 1994 Page 89 EPMP-EPP-0102 Rev 00

ATTACHHENT 1 (Cont) 8.4.5 (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.

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):

HAl. 1 Basis Reference(s):

1.'2-0P-90, Seismic Honitoring

2. NUREG-1253 Technical Specifications Nine Hile Point Nuclear Stations, Unit No. 2, Docket No. 50-410, 7/87, article 3.3.7.2 Hay 1994 Page 90 EPHP-EPP-0102 Rev 00

)TTACHNENT ) )C t) 8.4.6 Alert Sustained winds > 90 mph OR Tornado strikes a plant vital area, Table 8.4 Table 8.4 Plant Vital Areas 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 Control Building Unit 2 Security Building 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 basis of 90 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 SEW Drawing No. 12187-SK-032483-25, Issue No.

1, Site Facilities Layout Status as of 8/1/89.

PEG Reference(s):

HA1.2 Hay 1994 Page 91 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 8.4.6 (Cont)

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 8.4.7 Alert Assessment by the Control Room personnel that a natural event has resulted in damage to equipment needed for safe plant operation, Table 8.4 Table 8.4 Plant Vital Areas 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 Control Building Unit 2 Security Building 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.

May 1994 Page 92 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 8.4.7 (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 8.4.8 Alert Lake water level > 254 ft OR Intake water level < 233 ft 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 intake bay water level which provides adequate submergence to the service water pumps.

PEG Reference(s):

HA1.7 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 May 1994 Page 93 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 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.

E.l.l ~ll 1 1 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.

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 Unusual Event emergency class.

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 May 1994 Page 94 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) ~

9.1.2 V usual 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.

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 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.

PEG Refer ence(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.

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 Alert emergency class.

May 1994 Page 95 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 9.1.3 (Cont)

PEG Reference(s):

HA6.1 Basis Reference(s):

None 9.1.4 Al ert 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.

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 Hay 1994 Page 96 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 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.

Rode 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.

PEG Reference(s):

HS3.1 Basis Reference(s):

None 9.1.6 Site rea 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 OR Loss or potential loss of either fuel clad or RCS barrier in conjunction with a loss of containment Loss of containment indicators may include:

~ Inconsistent or unexpected LOCA response

~ Rapid unexplained decrease following initial increase in containment pressure Node Applicability:

Power operations, Startup/Hot standby, Hot Shutdown Hay 1994 Page 97 EPMP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 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 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.

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.

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 be consistent with the General Emergency classification description.

Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the site boundary.

May 1994 Page 98 EPMP-EPP-0102 Rev 00

STIACIINENT 1 (C t) 9.1.7 (Cont)

PEG Reference(s):

HG2.1 Basis Reference(s):

None 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.

Loss of containment indicators may include:

~ Inconsistent or unexpected LOCA response Rapid unexplained decrease following initial increase in containment pressure 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 May 1994 Page 99 EPHP-EPP-0102 Rev 00

ATTACHMENT 1 (Cont) 9.1. 8 (Cont)

Basis Reference(s):

None May 1994 Page 100 EPMP-EPP-0102 Rev 00

TTACHMENT 2 WORD LIST/DEFINITIONS May 1994 Page 101 EPMP-EPP-0102 Rev 00

ATTACHMENT 2 (Cont) ~

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.

de 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.

~vail abie 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.

May 1994 Page 102 EPMP-EPP-0102 Rev 00

ATTACHNENT 2 (Cont)

Can Cannot be restored above below The value of the identified parameter(s) is/is not able to be returned to above/below specified limits after having passed those 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). Does not imply any specific time interval but does not permit prolonged operation beyond a limit without taking the specified action.

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 maki,ng an evaluation that considers both current and future restoration capabilities.

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 Confirmatio 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.

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.

Decrease To become progressively less in size, amount, number, or intensity.

~0fschar e Removal of a fluid/gas from a volume or system.

May 1994 Page 103 EPHP-EPP-0102 Rev 00

STTACHNENT (C t)

~0r 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 limit, measure, or degree.

E st 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.

Hay 1994 Page 104 EPHP-EPP-0102 Rev 00

STTACIIMENT 2 (C t)

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.

~Iitiate 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 Tluid into a volume or vessel.

Intrusion The act of entering without authorization Loss Failure of operability or lack of access to.

Naintain Take action, as necessary, to keep the value of the specified parameter within the applicable limits.

Nay 1994 Page 105 EPHP-EPP-0102 Rev 00

STTACIIMfNT 2 (C ti Naximum 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.

Nonitor 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.

~eeeabl 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 ste 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.

Nay 1994 Page 106 EPHP-EPP-0102 Rev 00

ATTACHMENT 2 (Cont)

Remove To change the location or position of.

e ort To describe as being in a specific state.

lLeeuf re 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."

May 1994 Page 107 EPMP-EPP-0102 Rev 00

ATTACHHENT 2 (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 ress on 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 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 U controlled An evolution lacking control but is not the result of operator action.

~U1a ned Not as an expected result of deliberate action.

Until Indicates that the associated prescribed action is to proceed only so long as the identified condition does not exist.

Hay 1994 Page 108 EPHP-EPP-0102 Rev 00

ATTACHHENT 2 (Cont) .

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.

Valid Supported or corroborated on a sound basis.

ent To open 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."

Iti 11 Any plant area which contains vital equipment.

Hay 1994 Page 109 EPHP-EPP-0102 Rev 00

IL

~ 'I 0-,

NIKP-2 Emergency Action Levels Category 1.0 Reactor Fuel Category 2.0 Reactor Pressure Vessel Category 3.0 Primary Containment Category 4.0 Secondary Containment Category 5.0 Radioactivity Release Category 6.0 Elecrtical Failures Category 7.0 Equipment Failures Category 8.0 Hazards Category 9.0 Other 6/20/94

Category 1.0 Reactor Fuel 1.0 Reactor Fuel 1.0 Reactor Fuel 1.1 Coolant Activity 12 Off-gas Activity 1.1.1 Unusual Event 1.2.1 Unusual Event Coolant activity > 0.2 pCi/gm I-131 equivalent or >100 Valid ofFgas radiation high alarm (at > DRMS red) for Ebar pCi/gm >15 min.

1.1.2 Alert Coolant activity > 800 pCi/gm I-131 equivalent Power operation, startup/hot standby, hot shutdown 1-2

e Category 1.0 Reactor Fuel 1.0 Reactor Fuel 1.0 Reactor Fuel 18 Contaimnent Radiation 1.4 Other Radiation Monitors 1.3.1 Alert 1.4.1 Unusual Event Drywell area radiation > 41 R/hr Any sustained ARM reading > 100 x DRMS high radiation alarm (red) or offscale high (DETECTOR Power operation, startup/hot standby, hot shutdown SATURATION) resulting from an uncontrolled process ALL 1.8.2 Site Area Emergency Drywell area radiation 2 8100 R/hr 1.4.2 Alert Power operation, startup/hot standby, hot shutdown Valid reading on Rx Bldg. above Refueling Floor Radiation Monitor 2HVR*HE14A or B, Gaseous Radiation Monitors (Channel 1) isolation 1.8.8 General Emergency OR Any sustained refuel Qoor rad monitor > 8.0 R/hr, Drywell area radiation t 5.2E6 R/hr Table 1.1 Power operation, startup/hot standby, hot shutdown 1.4.8 Alert Sustained area radiation levels ) 15 mR/hr in either:

Control Room OR Central Alarm Station and Secondary Alarm Station 1-3

Category 1.0 Reactor Fuel 1.0 Reactor Fuel 1.0 Reactor Fuel 1.4 Other Radiation Monitors 1.6 Refueling Accidents 1.4.4 Alert 1.6.1 Unusual Event Sustained area radiation levels > 8 R/hr in any areas, Spent fuel pooV reactor cavity water level cannot be Table 1.2 restored and maintained above the spent fuel pool low AND water level alarm Access is required for safe operation or shutdown Table 1.1 Refuel Floor Rad Monitors 1.6.2 Alert ARM RMS111, RB 354'est Spent Fuel Pool Imminent or report of actual observation of irradiated ARM RMS112, RB 354'ast of Spent Fuel Pool fuel uncovered Table 1.2 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 Bay Condensate Storage Tanks Building Standby Gas Treatment Building

Category 2.0 Reactor Pressure Vessel 2.0 Reactor Pressure Vessel 2.0 Reactor Pressure Vessel 2.1 RPV Water Level 2.2 Reactor Power / Reactivity Control 2.1.1 Unusual Event 2%1 Alert Unidentified drywell leakage > 10 gpm Allimmediate manual scrams fail to shut down the OR reactor Reactor coolant to drywell identified leakage > 25 gpm Power operation, startup/hot standby Power operation, startup/hot standby, hot shutdown 2.2.2 Site Area Emergency 2.1.2 Site Area Emergency Allimmediate manual scrams fail to shut down the RPV water level cannot be restored and maintained reactor

> -14 in. (TAF) AND Boron injection is required Power Operation, Startup/Hot Standby, Hot Shutdown, Cold Shutdown, Refuel Power operation, startup/hot standby 2.1.8 General Emergency 2.2.8 General Emergency Primary Containment Flooding required Allimmediate manual scrams fail to shut down the reactor Power operation, startup/hot standby, hot shutdown AND RPV water level cannot be restored and maintained

> -89 in.

Power operat'ion, startup/hot standby

Category 2.0 Reactor Pressure Vessel 2.0 Reactor Pressure Vessel 2.2 Reactor Power / Reactivity Control 2.2.4 General Emergency Allimmediate manual scrams fail to shut down the reactor AND Suppression pool temperature and RPV pressure cannot be maintained c HCTL Poujer operation, startup )hot standby

Category 8.0 Primary Containment 3.0 Primary Containment 8.0 Primary Containment 3.1 Containment Pressure 3.2 Suppression Pool Temperature 3.1.1 Alert 3.2.1 Site Area Emergency Primary containment pressure cannot be maintained RPV pressure and suppression pool temperature

< 1.68 psig due to coolant leakage cannot be maintained < HCTL (non-ASS)

Power operation, startup/hot standby, hot shutdown Power operation, startup/hot standby, hot shutdown 8.1.2 Site Area Emergency Primary containment pressure cannot be maintained

< 1.68 psig AND Coolant activity > 800 pCi/gm Power operation, startup/hot standby, hot shutdown 8.1.8 General Emergency Primary containment venting is required due to PCPL Power operation, startup/hot standby, hot shutdown 3-1

Category 8.0 Primary Containment 8.0 Primary Containment 8.0 Primary Containment 3.8 Combustible Gas Concentration 8.4 Containment Isolation Status 3.8.1 Site Area Emergency 8.4.1 Site Area Emergency

> 4% H2 exists in DW or suppression chamber Main Steam Line, RCIC steam line or RWCU isolation failure resulting in a release pathway outside primary containment.

Power operation, startup/hot standby, hot shutdown 8.8.2 General Emergency Primary containment venting is required due to 3.4.2 General Emergency combustible gas concentrations Main Steam Line, RCIC steam line or RWCU isolation failure resulting in a release pathway outside primary containment AND any:

~ Coolant activity > 800 p,Ci/gm I-131 equivalent

~ RPV water level < -14 in. (TAF)

~ )

DW radiation 8100 R/hr Power operation, startup/hot standby, hot shutdown 3-2

Category 4.0 Secondary Containment 4.0 Secondary Containment 4.0 Secondary Containment 4.1 Reactor Building Temperature 4.2 Reactor Building Radiation Level 4.1.1 Site Area Emergency 4.2.1 Site Area Emergency Primary system is discharging outside PC Primary system is discharging outside PC AND AND RB area temperatures are > maximum safe operating RB area radiation levels are > maximum safe levels in two or more areas, N2-EOP-SC,RR operating levels in two or more areas, N2-EOP-SC,RR Power operation, startuplhot standby, hot shutdown Power operation, startup/hot standby, hot shutdown 4.12 General Emergency 4.2.2 General Emergency Primary system is discharging outside PC Primary system is discharging outside PC AND AND RB area temperatures are > maximum safe operating RB area radiation levels are > maximum safe levels in two or more areas, N2-EOP-SC,RR operating levels in two or more areas, N2-EOP-SC,RR AND any: AND any:

~ Coolant activity > 800 pCi/gm I-131 ~ Coolant activity > 800 pCi/gm I-131 equivalent equivalent

~ (

RPV water level -14 in. (TAF) ~ (

RPV water level -14 in. (TAF)

~ DW radiation > 8100 R/hr ~ DW radiation > 8100 R/hr Power operation, startup/hot standby, hot shutdown Power operation, startup/hot standby, hot shutdown

Category 6.0 Radioactivity Release /Area Radiation 6.0 Radioactivity Release /Area Radiation 6.0 Radioactivity Release /Area Radiation 5.1 EQluent Monitors 6.2 Dose Projections/ Environmental Measurements/ Release Rates 6.1.1 Unusual Event 6.2.1 Unusual Event A valid reading on any monitors Table 5.1 column "UE" for > 60 min. Confirmed sample analyses for gaseous or liquid release rates > 2 x technical specifications limits for >

60 min.

6.1.2 Alert A valid reading on any monitors Table 5.1 column 5.2.2 Alert "Alert"for > 15 min.

Confirmed sample analyses for gaseous or liquid release rates > 200 x technical specifications limits for

> 15 min.

5.1.3 Site Area Emergency A valid reading on any monitors Table 5.1 column "SAE" for > 15 min. 6.2.3 Alert Dose projections or field surveys which indicate doses /

dose rates > Table 5.2 column "Alert" at the site boundary or beyond.

5-1

Category 6.0 Radioactivity Release /Area Radiation 5.0 Radioactivity Release /Area Radiation 6.2 Dose Projections/ Environmental Measurements/ Release Rates 6.2.4 Site Area Emergency Dose projections or field surveys which indicate doses /

dose rates > Table 5.2 column "SAE" at the site boundary or beyond.

6.2.6 General Emergency Dose projections or field surveys which indicate doses /

dose rates > Table 5.2 column "GE" at the site boundary or beyond.

e Category 6.0 Radioactivity Release /Area Radiation Table 5.1 Effluent Monitor Classification Thresholds Monitor RadWaste/Reactor Bldg. Vent Effluent 2 x GEMS alarm 200 x GEMS alarm  %.5E6 p,Ci/s N/A Main Stack Effluent 2 x GEMS alarm 200 x GEMS alarm N/A N/A Service Water Effluent 2 x DRMS High (red) 200 x DRMS High (red) N/A N/A Liquid Rad Waste Effluent 2xDRMSHigh(red) 200xDRMS High(red) N/A N/A Cooling Tower Blowdown 2 x DRMS High (red) 200 x DRMS High (red) N/A N/A Table 5.2 Dose Pro ection / Env. Measurement Classification Thresholds GE TEDE 10 mR 100 mR 1000 mR CDE Thyroid N/A 500 mR 5000 mR TEDE rate 10 mR/hr 100 mR/hr 1000 mR/hr CDE Th oid rate N/A 500 mR/hr 5000 mR/hr 5-3

Electrical Failures 6.0 Electrical Failures 6.0 Electrical Failures 6.1 Loss of AC Power Sources 6.1 Loss of AC Power Sources 6.1.1 Unusual Event 6.1.4 Site Area Emergency Loss of power for >15 min. to all: Loss of all emergency bus AC power for >15 min.

~ Reserve Transformer A

~ Reserve Transformer B Power operation, startup/hot standby, hot shutdown

~ AuxiliaryBoiler Transformer 6.1.5 General Emergency Loss of all emergency bus AC power 6.1.2 Alert AND either:

Power cannot be restored to any emergency bus in Loss of all emergency bus AC power for >15 min. <2 hrs OR Cold shutdown, refuel RPV water level cannot be restored and maintained > -14 in. (TAF) 6.1.8 Alert Power operation, startupIhot standby, hot shutdown Available emergency bus AC power reduced to only'ne of the following sources for >15 min.:

~ Reserve Transformer A

~ Reserve Transformer B

~ AuxiliaryBoiler Transformer

~ 2EGS*EG1

~ 2EGS*EG2

~ 2EGS*EG3 Power operation, st'artup (hot standby, hot shutdown

Category 6.0 Electrical Failures 6.0 Electrical Failures 6.2 Loss of DC Power Sources 6.2.1 Unusual Event

< 105 vdc on 2BYS*BAT2Aand B for >15 min.

Cold shutdown, Refuel 6.2,2 Site Area Emergency

< 105 vdc on 2BYS*BAT2Aand B for > 15 min.

Power operation, startuplhot standby, hot shutdown

Equipment Failures 7.0 Equipment Failures 7.0 Equipment Failures V.1 Technical Speci6cationXRequirements V.2 System Failures or Control Room Evacuation 7.1.1 Unusual Event V.2.1 Unusual Event Plant is not brought to required operating mode within Technical Specifications LCO Action Statement Time. Report of main turbine failure resulting in casing penetration or damage to turbine seals or generator Power operation, startup (hot standby, hot shutdown seals.

Power operation, startup )hot standby, hot shutdown 7.2.2 Alert Control Room evacuation 7.2.3 Alert Reactor coolant temperature cannot be maintained (

200 'F Cold shutdown, refuel 7-1

Category V.O Equipment Failures 7.0 Equipment Failures V.O Equipment Failures 7.2 System Failures or Control Room 7.8 Loss of Indications/Alarm/Communication Evacuation Capability 7.2.4 Site Area Emergency V.8.1 Unusual Event Control Room evacuation Unplanned loss of annunciators or indicators on any of AND the following panels for > 15 min.:

Control of core cooling systems cannot be established in 5 15 min. ~ 2CEC*PNL601

~ 2CEC*PNL602

~ 2CEC*PNL603

~ 2CEC*PNL852

~ 2CEC*PNL851 Power operation, startup/hot standby, hot shutdown 7.8.2 Unusual Event Loss of all communications capability affecting the ability to either:

Perform routine onsite operations OR Notify oQsite agencies or personnel 7-2

Category 7.0 Equipment Failures 7.0 Equipment Failures 7.0 Equipment Failures V.8 Loss of Indications/Alarm/Communication 7.8 Loss of Indications/Alarm/Communication Capability Capability V.8.8 Alert 7.8.4 Site Area Emergency Unplanned loss of annunciators or indicators on any of Loss of annunciators or indicators on any of the the following panels for > 15 min.: following panels:

~ 2CEC*PNL601 ~ 2CEC*PNL601

~ 2CEC*PNL602 ~ 2CEC*PNL602

~ 2CEC*PNL603 2CEC*PNL603

~ 2CEC*PNL852 ~ 2CEC*PNL852

~ 2CEC*PNL851 ~ 2CEC*PNL851 AND AND Increased surveillance is required for safe plant Plant computer and SPDS are unavailable operation AND AND either: Indications to monitor all RPV and primary Plant transient in progress containment EOP parameters are lost OR AND Plant computer and SPDS are unavailable Plant transient is in progress Power operation, startup/hot standby, hot shutdown Power operation, startup/hot standby, hot shutdown 7-3

Category 8.0 Hazards 8.0 Hazards 8.0 Hazards 8.1 Security Threats 8.2 Fire or Exylosion 8.1.1 Unusual Event 8.2.1 Unusual Event Bomb device or other indication of attempted sabotage Confirmed fire in any plant area, Table 8.2 or Table discovered within plant Protected Area 8.3, not extinguished in < 15 min. of Control Room notification 8.12 Alert 8.2.2 Alert Intrusion into plant Protected Area by an adversary Fire or explosion in any plant area, Table 8.2 or Table 8.3, which results in damage to plant equipment or structures needed for safe plant operation 8.1.3 Site Area Emergency Intrusion into a plant security vital area by an adversary 8.1.4 General Emergency Security event which results in:

Loss of plant control from the Control Room AND Loss of remote shutdown capability 8-1

Category 8.0 Hazards 8.0 Hazards 8.0 Hazards 8.8 Man-Made Events 8.3 Man-Made Events 8.3.1 Unusual Event 8.8.4 Alert Vehicle crash into or projectile which impacts plant Vehicle crash or projectile impact which precludes structures or systems within Protected Area boundary personnel access to or damages equipment in plant vital areas, Table 8.3 8.3.2 Unusual Event Report by plant personnel of an explosion within 8.3.5 Alert Protected Area boundary resulting in visible damage to permanent structures or equipment Report or detection of toxic or flammable gases within a plant vital area, Table 8.3, in concentrations that will be life threatening to plant personnel or preclude access to equipment needed for safe plant operation 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 ofBcials for potential evacuation of site personnel based on offsite event 8-2

Category 8.0 Hazards 8.0 Hazards 8.0 Hazards 8.4 Natural Events 8.4 Natural Events 8.4.1 Unusual Event 8.4.4 Unusual Event Earthquake felt in plant by any operator Lake water level > 248 ft AND either: OR NMP-2 seismic instrumentation actuated Intake water level < 237 ft OR Conarmation of earthquake received on NMP-1 or JAFNPP seismic instrumentation 8.4.5 Alert Earthquake felt in plant by any operator AND 8.4.2 Unusual Event NMP-2 seismic instrumentation indicates ) 0.075 g Report by plant personnel of tornado striking within All plant Protected Area boundary 8.4.6 Alert Sustained winds ) 90 mph 8.4.3 Unusual Event OR Tornado strikes a plant vital area, Table 8.3 Assessment by Control Room personnel that a natural event has occurred precluding access to a plant vital area, Table 8.3 8-3

Category 8.0 Hazards 8.0 Hazards 8.4 Natural Events 8.4.7 Alert Assessment by the Control Room personnel that a natural event has resulted in damage to equipment needed for safe plant operation, Table 8.3 8.4.8 Alert Lake water level ) 264 ft OR Intake water level c 288 ft 8-4

Hazards Table 8.2 Plant Areas

~ Service Building

~ 115 KV Switchyard

~ 345 KV Switchyard Table 8.3 Plant Vital Areas

~ 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

~ Control Building

~ Unit 2 Security Building 8-5

Category 9.0 Other 9.0 Other 9.0 Other 9.1.1 Unusual Event 9.1.4 Alert Any event, as determined by the Shift Supervisor or Any event, as determined by the Shift Supervisor or Site Emergency Director, that could lead to or has led Site Emergency Director, that could lead or has led to to a potential degradation of the level of safety of the a loss or potential loss of either fuel clad or RCS plant. barrier.

Power operation, startuplhot standby, hot shutdown 9.1d Unusual Event 9.1.6 Site Area Emergency Any event, as determined by the Shift Supervisor or As determined by the Shift Supervisor or Site Site Emergency Director, that could lead to or has led Emergency Director, events are in progress which to a loss or potential loss of containment. indicate actual or likely failures of plant systems needed to protect the public. Any releases are not Power operation, startup Ihot standby, hot shutdown expected to result in exposures which exceed EPA PAGs.

9.1.3 Alert Any event, as determined by the ShiR Supervisor or Site Emergency Director, that could cause or has caused actual substantial degradation of the level of safety of the plant.

Category 9.0 Other 9.0 Other 9.0 Other 9.1.6 Site Area Emergency 9.1.8 Geheral Emergency Any event, as determined by the ShiR Supervisor or Any event, as determined by the Shift Supervisor or Site Emergency Director, that could lead or has led to Site Emergency Director, that could lead or has led to either: a loss of any two fission product barriers and loss or Loss or potential loss of both fuel clad and RCS potential loss of the third.

barrier OR Loss of contaizunent indicators may include:

Loss or potential loss of either fuel clad or RCS Inconsistent or unexpected LOCA response barrier in conjunction with a loss of containment ~ Rapid unexplained decrease following initial increase in containment pressure Loss of containment indicators may include:

~ Inconsistent or unexpected LOCA response Power operation, startupfhot standby, hot shutdown

~ Rapid unexplained decrease following initial increase in containment pressure Power operation, startup/hot standby, hot shutdown 9.1.V General Emergency As determined by the ShiR 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.

OSSI 93-402A-10-NMP2 NMP-2 EAL VeriQcatton & Validation Report, Rev. 0 Erner enc Action Level V==-'cation R Validation Re ort Revision 0 New York Power Authori J. A, Fitzpatrick Nuclear Power Plant Indian Point Nuclear Power Plant Unit 3 Ni ara Mohawk Power Co oration Nine Mile Point Unit 1 Nine Mile Point Unit 2 Consolidated Edison Com an Indian Point Station Unit 2 Rochester Gas and Electric Com an R. E. Ginna Nuclear Power Station Operations Support Services, Inc.

233 Water Street 2r.d Floor Plymouth, MA 02360

e I

OSSI 93-402A-10-N M P2 NMP-2 EAL VeriQcatfon & Validation Report, Rev. 0 Table of Contents Section ~Pa e 1 Introduction ...................,....,......,..... ~ ~ ~ t ~ ~ otto ~ ~ otooooeootoesososooooo ~ o ~ o ~ oooottotoottoto 1

2. P reparations ~ ~o~ oo ~ ~~ ~~~~~~o~~s~o~~~~ eoo ~~~~~ eoo ~ ~ ~ oooo ~ ~ ~ ~ etto ~ ~ et ~ e ~ e ~ ~ s ~ oooto ~~ ettoo ~ otoo ~ e ~ ootoooottooteoo ~ 1

3. Process ~ oo oeoooosoooooooooooooooeoooooo 4 3 .1 Verification Activities,.......,........,.................................,.........,............,...4 32~ Valid a tion Activities ~..... ~ .. ~ ..,....,........,... .,.......,..................,............. 5

~

4 Comment Resolution ....,......,......,........,...................................................................9 5 References ... ~ ~ ~ ~o~ ooooeotooo ~ eeoooooeo ~ ooooosoooooo 9 EAL Verlfjcat;o s Checklists 1 1 EAL Verification Comment Database .. ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ o ~ o ~ ~ o ~ o ~ ~ ~ ~ ~ o ~ o ~~~o~o~~o~~~ to~

EAL Validaif.in~~

io ~n <n~-curios

> arios 3 1 EALVamationp S, Q .~own~n p ia i~r C4c s i-c.

~r Sh 4-EALValidation L.-.c: cise Checklists...........................................................................5-1 nA1 131 A EAL Validatict ..0:...",:cnt s1 Database .................................,.........................................6-1

OSSI 93-402A-'.* - '",! ." NMP-2 EAL Verificatio & Validation Report, Rev. 0

1. Introduc'n The verificat'~:". process was performed to ensure the NMP-2 Emergency Action Level." -'L';) and classification procedures are written correctly and are technica!i,.:.: ...'cct. The NMP-2 EAL verification was conducted prior to the EAL valic';:;"..'on exercises. Verification activities were completed according to Rc"c:.ence l.

The validation piocess was performed to ensure that the NMP-2 EALs and classification procedures are usable and operationally correct, and to ensure that response!,'i en.ei.gency response organization personnel are able to arrive at co:-.'.-.=!,ct t .'Lerpretations of EALs under varying conditions. The NMP-2 EAL ...:..;...;;,.; exercises were conducted on October 7, 1993 at the Nine Mile Poi'".:..;.'.ining Center NMP-2 control room simulator. Validation activities wci oi..:piet.ed according to Reference 2.

The NMP-2 E!4. vc..ification/validation was one of six verification/validations conducted by CSS! at each of the six participating plants in the NYPA EAL Upgrade Pre)".".~.

2. P~re areal"",.s Mr. C, K, W." .", ": ("SR?) was assigned EAL verification and validation team leader. For.. '.'., ~-;,;;.ication, he was responsible for:

~ Det.'"-'.:-".;-;.g the extent to which the EAL documentation is verii::*...

~ Scl".,".:.:..:g '.=.".,m members to conduct EAL verification reviews.

OSSI 93-402A- "': NMP-2 EAL Verification & Validation Report, Rev. 0

~ Providing appropriate source documents so team members can con;!::ct vciification reviews.

~ Coo:.;..'.-.nt;.-:< resolutions to any verification review comments.

~ Coor ".:--",.!:.:g update of EAL program documentation consistent with !.i.:: r=solution of verification review comments.

~ Deter:~iining the extent to which each selected EAL is validated.

For EAL valid.:o>>, Mr'. Walker was responsible for:

~ Selec".. ~,<< 'c:-.m members to participate as validation exercise ob"-e.", .:. s;hand as emergency response organization personnel du in;", EA!. validation exercises.

~ Prepar::;< .. validation exercise test plan and schedule. (EALs select d I;:r validation are documented on the Validation Summary Sheet which served as the validation test plan.)

~ Ob'..":,;...'.g appropriate scenarios to test emergency response org:;r .=:;t.i.", classification activities while using the EALs.

~ Coor;i::-~at.:;ig resolutions to validation exercise comments.

~ Coor l nat,ing update of EAL program documentation consistent wi'..'.. !,".c r soIiition of validation exercise comments.

Mr. J. P. St.-" ...:,2=::.-"i) was assigned to the verification team and was responsible .~;

~ Beco:>,."g 'anliliar with appropriate verification source documents and thc NMP-2 EALs to be verified.

~ Pe:.'"..r....::." assigned EAL verification reviews.

OSSI 93-402P.-: NMP-2 EAL Ver1Qcatton & Valfdation Report, Rev. 0

~ Complcf.ing verification checklists for technical accuracy and wr!.'.".:". c,.rrcctness reviews.

~ As.". '< n thc preparation of resolutions to verification review cori!i e!! ts.

Mr. M. C. D,.:.; ( <<.";:S.') and Mr. J. Toothaker (NMPC) were assigned EAL validation exe! cise observers, They were responsible for:

~ Beco'..ing familiar with appropriate NMP-2 EAL development do.,:..-..; ..'.s and the EALs to be validated.

~ OL.- ..: emergency response organization participants using the

.EAl.." ..v.'ii.c responding to simulated emergency events.

Comp.'.i:".g the validation exercise checklists

~ Ass.',', r!g in the preparation of resolutions to validation exercise co a a ~ is Several me:"...'.::". l.h" NMP-2 operations training staff were also assigned to the valid ...:!;. t.o play the role of emergency response organization positions. Th:".': n,.mcs and titles are listed on the EAL Validation Summary sheets (Atta"..l.:.'.c:.t 4). They were responsible for:

~ Bcc".: ..',:p fa!:..iliar with the EALs to be validated.

~ Us.: '. '.'" <<P Ls while responding to simulated emergency events.

~ Con':.'.e~ing the validation exercise checklists.

OSSI 93-402h-! "-"':i.-? NMP-2 EAI. VeriQcation & Validation Report, Rev. 0

3. Proces~

3.1 Verificaf1on Activities The technic",1:.;.""...:..","..y and written correctness of the upgraded EALs were verified thrc;- .'. f.'..-i -'op reviews which addressed the following EAL attributes:

Wri ten Correctness Hum" n cn<inccring factors of the EAL Writer's Guide Fo.- ..'...,..:.":::.ance and terminology consistent, to the extent pos..ihlc. a...on< BWR and PWR plants involved in the NYPA EAL Up~ -..-;cfc .: oject EAL c:f:-ii'AL terminology is clear and well defined Technical Accurac Tc.".!.:;:::; . ".,o:".-l',;leteness and appropriateness for each classification lcv".'o'...'"":

.'.: "';nssification upgrade only when there is an increased th1'::::;-...!-.:.c health and safety Log ",al =.r)<:ession in classification for combinations of multiple events Con".~tcncy of'ALs, to the extent possible, among BWR and PWR p11ny le( ~ i I i ab

~

The EALs v; .'.: .",."::.".*.cd in terms of the evaluation criteria embodied in the checklists f:. 'c"h:::cal accuracy and written correctness (Attachments 1 and 2 of Rc" -..: '1. KAL verification reviews for technical accuracy and written corr:;:! ..".".. a~ere accomplished by a comparative Table-Top evaluation c: ';,h('";:Giving:

OSSI 93-402A- 'MP-2 EAL VeriAcation & Validation Report, Rev. 0

~ '.

Wri!'cn , cctness of the EALs including human factors guidance of '.;. '... >>.rater's guide.

~ Tec!~nical accuracy of the EALs compared to the EAL Technical Bas .. 1'.=,:- is ion Product Barrier Evaluation, Plant-Specific EAL Gu - i'..;:. F.h.L Binning Document, and NUMARC NESP-007, Rev.'.'"= '..'.".c!uding NUMARC/NRC Questions and Answers).

~ Co...pat.i! ..!y of the EALs with the plant.

~ Nu:.;"ri:,".': v'.1 pcs, quantitative and calculated information.

The Walk-th.'" n.c'.!'.od of verification was performed during EAL validation where nece."":",::.'. fcrences to equipment, indications and instrument" '-.n w";" checked against control room hardware as represented; t'.;;.. ':>...i1ator control room.

Verification r vic>>,"s were performed using the applicable sections of the EAL verifical.:.".n c1"..".cklists (Attachment 1). All discrepancies were documented 'n ':".', Co::ament Forms in the EAL Verification Comment Database. A 'i ~.'o.:t o! this database is provided in Attachment 2.

3.2 Valid~" .D '.;.'.'i:i."s The usabilit~ and c- "rational correctness of the upgraded EALs were validated t!." u~>'...:".,c;.>>ation of emergency response organization personnel respondin<:, =..'....; ',=":.. emergency events using the $ UiLs. The group of EALs select:.. '.'......:;".:".ion were sufficiently representative to test that the EALs posse'.;" ~h~,;",.: n< attributes:

Usabili'~ai f 'ii "'1 'iqpqs Eca '-" oi .) ~

e 'deriding

OSSI 93-402P - ':-.'i NMP-2 EAL VeriAcation & Validation Report, Rev. 0

~ Ease of place-keeping

~ Ea.";; of;: ".,:.:!-:ng and declassifying

~Oerati", -'...',-.:-,c" ~,.sess Pol..':"..; = classification upgrade only when there is an increased thre.".i: t.o ".::blic health and safety Teel:;;i"::. "o:...pleteness and appropriateness for each classification ieq pnl P. i':"."'; '- ~,:".ssion in classification for combinations of multiple ev...

EALs not sei'. ".;; .".::.;.;-.!idation were compared to the validation checklist criteria at L>;.e co..".li sion of the validation exercises.

The EALs w..:." ~;~.:..'.!ated in terms of the evaluation criteria embodied in the checklist for ".,A!.;.'., Iuation. EAL validation exercises were conducted using the Table-T-"". ! ..".'.!~ad and the Simulator method. Scenarios were developed "....:.:.". t:".c performance of the Table-Top and Simulator methods of'- .:: ...;.:: (h.t.tachment 3). The scenarios provided the means for validatic" ': ...:-. observers to view emergency response organization personnel c"...'.8'.i.".g the EALs for proper emergency classification.

In the clas.;.;". >>"..::.":..hers of the EAL validation team were introduced to the upgrac!,, *I.e team leader. Classification categories and subcatego..: ',:.scussed as were the technical basis for individual EAL conditions. ..-..I '.o familiarize all validation team participants with the conten'. '...;:..".Ls and their relationship to the existing classificai.i; " .::.".s. Members of the validation team were also briefed on the vali;i '.;.:..:cess described in Reference 2. Copies of the upgraded EALs were n:.;!c i .!l.ble to team members during the validation exercises.

The EAL v"-." ':lan is given on the EAL Validation Summary Sheets in Attach:. ...h EAL validation scenario, the following activities were perf.

OSSI 93 402 s NMP-2 EAL Verillcation & Validation Report, Rev. 0

1. Tl.. vali.! ..'.Ion team members assumed the emergency response o: .,."..'.;les they were expected to fulfillduring an actual

2. Tl'". '.'":., 'cr described the initial plant conditions.

3. Wl~:-.;i '..c:.gcncy response organization personnel were familiar w".r::-..::!.i.;.! ~"lent conditions, the team leader announced the start of t1-."..: ..a:.; c:;crcise and described changes in key plant p",." -:-': !!'".r the Table-Top method) or he instructed the s'...,.'.."..i instructor to place the simulator in RUN (for the S "."!".:.cr i're'.bod).

4. Th ~". c'<..'.cy response organization personnel described the ac'...;..s Lhey would perform (for the Table-Top method) or they

n. =--.-'.'"..'-.".>>roprlate plant controls in the simulator as needed to .:<..'" "...".anging plant conditions (for the Simulator 5, Tl"..., .;.,;, "." response organization personnel consulted the up:.:.'.  ! '...~'.I'.s according to Emergency Plan procedures and made a;"...'; ..siiications.

6. T.'- - ".".:. " "-rvers occasionaly asked questions of the emergency rc, . '.-..'::stion personnel during the exercise.

7 P7~ "..".t conditions were reached, the validation team

the exercise and held a post scenario briefing

~ ~

~ "".:;i .-...":.ibers jointly discussed problems and comments

!:"...< the exercise.

OSSI 93-402"-'MP-2 EAL Veriilcation & Validation Report, Rev. 0

~ ".'".".:n n~c>>bcrs jointly completed the EAL Validation Exercise

~"""..1.'-'-t~ (h.ttachment 5).

.:.sons for noted problems and comments were

')i;." cases, portions or all of the exercise were

..." ...:"=! to gain a better perspective of noted problems c~n'. CC':lnlellts.

The validatic " '." ">> >",".dci ensured the following information was recorded on each V,",.': ....:".e Checldist:

t Va" .".. n ".am member names and titles EA'. !: .-".! ion number of EALs validated Sce:"": s c'...-~ripe,ion Vai-',.'.,: n,. '.had Following e" '....-'",.::::;.rio briefing, team members compared observatio.", ';..'..cd if any problems and comments noted thus far required n" ". :"!'.~e test plan to achieve validation objectives.

When all v-" '". '.,"r". scs were completed, the team leader, with the assistance ~

t..; nic nbers, consolidated all exercise problems and comments '

p~ .y problem and comment recorded on the EAL

-'hecklists.

~ R- .'!"ms and comments in the EAL Validation Comment

"';-.!:out of this database is provided in Attachment 6.

~ R- ...;.;mment numbers on the EAL Validation Exercise C1~

OSSI 93-402A-l ~-.'P2 NMP-2 EAL VeriQcation & Validation Report, Rev. 0

4. Commen', T'...~~Dition Mr. Walker and Mr. Daus evaluated each verification and validation comment recorded in the EAL comment databases. They reviewed the comment discrepancies and determined the accuracy of the discrepancy. Reference materials in EP.'c~ e1opnient were used to identify the scope of the discrepancy and to prepare appropriate solutions.

They prepared reso!utions to the discrepancies, determined the impact the final resolutio..". i:"ie on EAL Program documentation, determined the impact the final icsolutions have on the plant, and identified any required follow-up act.'."...,.

Results of the ver.*ication and validation comment resolution process were documented in Uie !.AL Verification Comment Database (Attachment 2) and the EAL Valida'.ion C'om:..ent Database (Attachment 6), respectively.

5. Refere~ '..""

l. OSSI 92-~,-, ". ",-"., ".:;;ergency Action Level Verifcation, Revision 0

2. OSSI 92-4 "".- .. H:..ergency Action Level Validation, Revision 0

OSSI 93-402A-10-N YP2 NMP-2 EAL Verification & Validation Report, Rev. 0 Attachment 1 T';.KL Veri6cation Chechlists

OSSI 92-402A-6-NMP2 EAI, Verification Procedure, Rev. 0 Attachment 1 Technical Accuracy Plant: Nine Mle Point 2 Date: 9 20 93 Verifier: J. P. Stal EAL U ade Pro ect En 'ne r name title Yes No NA

1. Plant-specific EAL Guideline (PEG) comparison to NESP-007, Revision 2, including NRC reviewed questions and answers:

1.1 Does each NESP-007 initiating condition have a corresponding PEG initiating condition that reflects the meaning of the NESP-007 IC'? ~ Q Q 1.2 Does the operating mode applicability of each PEG initiating condition agree with the NESP-007 operating mode applicabilityV Q ~ Q 1.3 Is each PEG EAL derived from a corresponding NESP-007 example EAL applicable to plant speciflc design' ~ Q Q 1.4 Do PEG EALs reflect the intent of the NESP-007 example EALs'? Q ~ Q 1.5 Does the PEG EAL bases reflect the intent of the NESP-007 EAL bases which are applicable to plant specific design' Q ~ Q

OSSI 92-402A-6-NMP2 EAI. Verification Procedure, Rev. 0 Attachment I Technical Accuracy Yes No NA 1.6 Are the PEG EALs complete and appropriate (i. e., is additional information needed, should any information be deleted)' Q ~ Q 1.7 Is each applicable PEG fission product barrier EAL properly considered in the Qssion product barrier evaluation for this plant' ~ Q Q 1-2

OSSI 92-402A-6-NMP2 EAL Verificatio Procedure, Rev. 0 Attachment 1 Technical Accuracy Yes No NA

2. EAL Technical Basis (TB) comparison to the Plant-Specific EAL Guideline (PEG), Fission Product Barrier Evaluation (FPBE) and EAL Binning Document:

2.1 Does the set of TB categories and subcategories satisfactorily reflect the set of PEG initiating conditions as defined in the EAL Binning Document' ~ 0 0 2.2 Is each TB EAL derived from one or more corresponding PEG EALs as defined by the FPBE and EAL Binning Document' 0 r 0 2.3 Do TB EALs reflect the intent of the PEG EALs from which they are derived' 0 ~ 0 2.4 Does the operating mode applicability of each TB EAL agree with the corresponding PEG EAL operating mode applicabilityV 0 ~ 0 2.5 Does the TB EAL bases reflect the intent of the PEG EAL bases and FPBE'? ~ 0 0 2.6 Are the references listed for each TB EAL appropriate and consistent with the PEG:

~ PEG Reference(s)'? 0 r 0

~ Basis Reference(s)'? ~ 0 0 1-3

OSSI 92-402A-6-NMP2 EAL VeriQcation Procedure. Rev. 0 Attachment 1 Technical Accuracy Yes No NA 2.7 Are the TB EALs complete and appropriate (i. e., is additional information needed, should any information be deleted)V 2.7 Does each "Remark" in Tables A through D of the fission product barrier evaluation for this plant satisfactorily explain the reason a PEG EAL or combination of PEG EALs is not needed for event classification'? ~ 0 0 2.8 Are the resultant fission product barrier evaluation EALs for this plant properly addressed in the TB at the appropriate classification level:

~ Unusual Event' ~ 0 0

~ Alert' ~ 0 0

~ Site Area Emergency' ~ 0 0

~ General Emergency'? 0 ~ 0 2.9 Does the potential exist for classification upgrade only when there is an increased threat to public health and safety' 0 0 2.10 Is there a logical progression in classiQcation for combinations of multiple events within a category' ~ 0 0 1-4

0 OSSI 92-402A-6-NMP2 EAL Verification Procedure, Rev. 0 Attachment 1 Technical Accuracy Yes No NA

3. EAL comparison to the EAL Technical Basis (TB):

3.1 Does the set of EAL categories and subcategories agree with the TB categories and subcategories, respectivelyV ~ Q Q 3.2 Is each EAL condition derived from a corresponding TB EAL condition' ~ Q Q 3.3 Does the operating mode applicability of each EAL agree with the corresponding TB EAL operating mode applicabilityV S Q Q

4. EAL comparison to the plant Control Room (Simulator):

4.1 Are as-labeled designations used to identify specific components, alarms, controls, and instruments to the extent practicable' ~ Q Q 4.2 Is each EAL adequately supported by plant instruments, approved instructions, or other appropriate sources of information' ~ Q Q 1-5

OSSI S2-402A-6-NMP2 EAI. Verification Procedure, Rev. 0 Attachment 1 Technical Accuracy Yes No NA 4.3 Where EAL conditions specify numerical values, are the units of measurement the same as those presented on the respective plant panel instruments, approved instructions, or other sources of information' 5 0 Q 4.4 Where EAL conditions specify numerical values, are the values expressed to a precision consistent with the accuracy and precision of the respective instrumentation' a ~ a All discrepancies have been recorded on EAL Comment Forms and forwarded to the Verification Team Leader.

Signature: Date: 9 20 93 1-6

OSSI 92-402A-6-NMP2 EAL VeriQcation Procedure, Rev. 0 Attachment 2 Written Correctness Plant: Nine Mile Point 2 Date: 9 20 93 VeriQer: J. P. Stal EAL U rade Pro ect En ineer name title Yes No NA

l. EAL Organization:

1.1 Is each EAL assigned to one of nine categories'? ~ a a 1.2 Is each subcategory clearly associated with its category' S CI C3

2. EAL Identification:

2.1 Is each EAL identified with a unique three digit number whose first digit corresponds to the category number, second digit the subcategory number, and third digit the EAL sequence numbers ~ a o 2.2 Do EAL sequence numbers increase in magnitude as classifications change from Unusual Event, to Alert, to Site Area Emergency, and to General Emergency' ~ 0 Cl 2-1

OSSI 92-402A-6-NMP2 EAI. Verification Procedure, Rev. 0 Attachment 2 Written Correctness Yes No NA 2.3 Where an EAL condition does not exist in a category/subcategory for a given emergency classification, has "NA" been entered in place of the EAL identification number' ~ Q Q

3. EAL Length and Content:

3.1 Is each EAL clear and concise' Q ~ Q 3.2 Have verbs and articles been deleted from EALs where technical accuracy and reading clarity permit' 3.3 Are EALs consisting of multiple conditions formatted such that each condition and its relationship to other conditions are easily understood' 5 Q Q 3.4 Is wording and abbreviations/ acronyms used in the EALs consistent with the deQnitions provided in Attachments 1 and 2 of the EAL Writer's Guide' ~ Q Q 3.5 Are EAL conditions expressed quantifiably where possible? ~ Q Q 3.6 Where used, do limit modiQers (<, >, s, >)

simplify presentation of EAL conditions' S Q Q 2-2

OSSI 92-402A-6-NMP2 EAL Verification Procedure, Rev. 0 Attachment 2 Written Correctness Yes No NA 3.7 Are annunciator setpoints not given in EALs when the setpoint is common operator knowledge or the setpoint is subject to frequent adjustment (e. g., area radiation monitor alarm setpoints, offgas radiation monitor alarms, etc.)'? 5 Q Q

4. Use of Logic Terms:

4.1 When an EAL must express a combination of two conditions, are the conditions joined by the logic term AND'? 5 Q Q 4.2 When an EAL must express an alternate combination of two conditions, are the conditions joined by the logic term OR' ~ Q Q 4.3 Is the use of AND and OR within the same EAL avoided where possible'? ~ Q Q 4.4 Is each EAL condition clear and concise' ~ Q Q 2-3

OSSI 92-402A-6-NMP2 EAL Verification Procedure, Rev. 0 Attachment 2 Written Correctness Yes ~N NA

5. Presentation of information in tables:

5.1 Is each table presented in a rectangular enclosure with a table number and title printed above the table entries' 8 0 0 5.2 Are column headings with applicable engineering units provided for tables with multiple columns of information' S 0 0 5.3 Where vertical lines separate columns of information, is readability improved' 5.4 If an entry is not required in a table cell, is the abbreviation "N/A" used' ~ 0 0

6. Mechanics of style:

6.1 Is the use of hyphens minimized, and ~no used to break words between lines' ~ 0 0 6.2 Is punctuation used only as necessary to aid reading and prevent misunderstandingV ~ 0 0 6.3 Are parentheses used to enclose location information in EALs and to visually separate supplemental/qualifying information from the primary information being stated' ~ 0 0 2-4

OSSI 92-402A-6-NMP2 EAL Verification Procedure, Rev. 0 Attachment 2 Written Correctness Yes No NA 6.4 Is word usage consistent among the EALs'? ~ Q Q 6.5 Are numbers in the EALs printed in Arabic numerals' ~ Q Q 6.6 .Are EAL limits specified in such a way that addition and subtraction by the user is

.avoided' 8 Q Q 6.7 Are EAL limits expressed to a precision consistent with the intent of the EAL as specified in the TB and PEG'? a Q Q

7. EAL format:

7.l Are three or more multiple items (systems, plant conditions, etc.) for which there is no preference or priority arranged in a list format with each item prefaced by a bullets ~ Q Q 7.2 Are EAL limit values, value modiQers and value engineering units printed in bold print' Q ~ Q All discrepancies have been recorded on EAL Comment Forms and forwarded to the Verification Team Leader.

Signature D t: 9//20 93 2-5

OSSI 92-402A-6-NMP2 EAL Verification Procedure, Rev. 0 Attachment 3 Inter-Plant EAL Comparison Plants: J. A. FitzPatrlck Date: 9 20 93 Nine Mile 1 Nine Mile 2 VeriQer: J. P. Stale EAL Pro ect En ineer name title Yes No NA

1. Within the constraints of BWR and PWR plant design, is each plant type EALs composed of the same categories' ~ Q Q

2. Within the constraints of BWR and PWR plant design, is each plant type EALs category composed of the same subcategories'? ~ Q Q

3. Within the constraints of BWR and PWR plant design, does the operating mode applicability of each EAL the same for each plant' S Q Q 4, Where individual plant design permits, are the condition(s) of each EAL the same for each plant? Q ~ Q 3-1

OSSI 92-402A-6-NMP2 EAL Verification Procedure, Rev. 0 Attachment 3 Inter-Plant EAL Comparison Yes No NA

5. Where individual plant design permits, are the limit value(s) of each EAL condition the same for each plant' 0 ~ 0

6. Within the constraints of BWR and PWR plant design, is EAL word usage the same for each plant' S 0 0 All discrepancies have been recorded on EAL Comment Forms and forwarded to the VeriQcation Team Leader.

Signature: Date: ~920 93 3-2

0 OSSI 93-402A-10-NMP2 NMP-2 EAL VerIAcatlon & Validation Report, Rev. 0 Attachment 2 EA." V"Hfication Comment Database 2-1

~ ~ . ~ ~

a

~ ~ a ~ ~

Record No. 5 Date 9/20/93 Name M. C. Daus Orlglnatlng Site Site Appllcablllty JAF O JAF O IP-3 O NMP-2 O Generic BWR 8 General NMP-1 Impact O IP-2 O NMP-1 O Ginna O Generic PWR NMP-2 O NUMARC-007 O Procedure O Verification O Training O Hardware g EAL El Technical Bases O Validation O Deviation O None Cat. PC Ic¹ 2 No. 2 Emer. Class. LOSS Comment (verification question 2.3): EAL 3.4.2 is declared when H2/02 exceed combustible limits. PEG EAL PC2.2 requires declaration when they cannot be determined to be below comubustible limits. It is not clear if EAL 3.4.2 addresses the latter condition.

Consider "Primary containment venting is required due to H2 and 02 concentrations limits".

t combustible C!nncirfor ovnlanatinn in tho hacic that inrlinatoc that tho ovictinn uinrdinn onr nmnaccoc iashon Resolution Explain in TB what is meant by combustible gas concentrations. PEGs are ok.

Changed EAL to state "Primary containment venting is required due to combustible gas concentrations".

status 0 Open 0 Resolved/Awaiting Disposition Qe Closed Record No. 7 Date 9/20/93 Name M. C. Daus Orlglnatlng Site Site Appllcablllty JAF O JAF O IP-3 O NMP-2 O Generic BWR 8 General NMP-1 Impact O IP-2 O NMP-1 O Ginna O Generic PWR NMP-2 O NUMARC-007 O Procedure O Verification O Training O Hardware IHI EAL O Technical Bases O Validation O Deviation O None cat. System Malf. No. 1 Emer. Class. UnuSual EVent Comment (verification question 3.2): EAL matrix and TB 7.3.1 is missing condition that EPIC is available.

. Resolution Added justification in PEG for the reason this condition is not required. See PWR verification comments for specific resolution.

status 0 Open 0 Resolved/Awaiting Disposition OI Closed

Record No. 8 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability JAF Cl JAF C3 IP-3 Q NMP-2 P Generic BWR 8 General NMP-1 Impact CI IP-2 0 NMP-1 CI Ginna C3 Generic PWR NMP-2 CI NUMARC-007 0 Procedure 0 Verification Cl Training 0 Hardware g EAL rHI Technical Bases CI Validation Q Deviation 0 None ca t. Hazards Ic¹ 3 No. 1 Emer. Class. UnuSual Event Comment (verification question 2.3): PEG HU3.1 refers to protecting safe operation of the plant. EAL 8.3.3 only addresses personnel protection.

Resolution The concern for safe plant operation has been added to the EALs.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed Record No. 10 Date 9/20/93 Name M. C. Daus rlglnatlng Site Site Applicability JAF P JAF Q IP-3 P NMP-2 Cl Generic BWR 8 General NMP-1 Impact Q IP-2 P NMP-1 Cl Ginna P Generic PWR NMP-2 Q NUMARC-007 C3 Procedure C3 Verification Cl Training P Hardware g EAL H Technical Bases Q Validation 0 Deviation 0 None cat. Barrier IC¹ ** No. ** Emer. Class.

Comment (verification question 1): 9.0 category refers to loss and potential loss of barriers. NESP-007 provides clear definition of these conditions in the FPB tables, but the EAL matrix never makes a distinction between a loss or potential loss. This could present a problem regarding interpretation of loss and potential barrier losses.

Resolution Check for this in validation.

10/22/93 This was checked during validation and was not observed to be a problem status 0 Open 0 Resolved/Awaiting Disposition Oe Closed

lRecord No. 11 orlglnatlng Site JAF NMP-1 NMP-2 cat ~ N/A Impact 0 NUMARC-007 g EAL lc¹ Date 9/2Q/93 0 Procedure Name Site Appllcablllty P JAF Q IP-3 C3 Technical Bases M. C. Daus Q NMP-2 Q Generic BWR E General 0 IP-2 0 NMP-1 0 Ginna 0 Generic PWR 0 Verification 0 Training CJ Hardware 0 Validation 0 Deviation Q None Emer. Class.

Comment (verification question 7.2): EAL matrix 5.2.3, 6.1.4, 6.2.2, the EAL numbers should be in bold print.

Resolution EAL numbers have been properly embolded.

status 0 Open 0 Resolved/Awaiting Disposition OI Closed Record No. 13 Date 9/2Q/93 Name M. C. DauS orlglnatlng Site Site Appllcablllty JAF Q JAF Q IP-3 P NMP-2 8 Generic BWR Q General NMP-1 Impact Q IP-2 P NMP-1 P Ginna P Generic PWR NMP-2 Cl NUMARC-007 0 Procedure 0 Verification P Training 0 Hardware g EAL Q Technical Bases Q Validation Q Deviation Q None ca t ~ Hazards Ic¹ 1 No. ** Emer. Class. UnuSual EVent Comment (verification question 2.4): PEG IC HU1, HA3 operating mode is unchecked. It should be "All".

Resolution Checked "All"for HU1, HA3 operating mode applicability in the PEG.

status 0 Open 0 Resolved/Awaiting Disposition OI Closed

Record No. 15 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability JAF 0 JAF P IP-3 Q NMP-2 mj Generic BWR P General NMP-1 Impact 0 IP-2 0 NMP-1 Q Ginna C] Generic PWR NMP-2 Q NUMARC-007 CI Procedure 0 Verification 0 Training CI Hardware IHI EAL 0 Technical Bases 0 Validation 0 Deviation 0 None cat. System Malf. Icg 2 No. ** Emer. Class. Unusual Event Comment (verification question 2.4): PEG IC HU2 operating mode includes hot shutdown but the TB EAL 7.1.1 only includes power operations and hot standby. Is this intentional or should the TB include hot shutdown?

Resolution EAL 7.1.1 should include hot shutdown. Changed TBs to include hot shutdown.

10/9 need to change matrices.

Mnto that thic ic alen a RWR FAI ualirlatinn nnmmont status 0 Open 0 Resolved/Awaiting Disposition 0+ Closed Record No. 16 Date 9/2Q/93 Name M. C. DauS rlglnatlng Site Site Appllcablllty JAF 0 JAF Q IP-3 C] NMP-2 [3 Generic BWR E General NMP-1 Impact Cl IP-2 Q NMP-1 Q Ginna CJ Generic PWR NMP-2 Q NUMARC-007 D Procedure D Verification 0 Training Q Hardware gl EAL Q Technical Bases Q Validation Q Deviation C3 None ca t. System Malf. Ict 1 No. ** Emer. Class. Alert Comment (verification question 2.4): PEG IC SA1 operating mode includes defuel but the TB EAL 6.1.2 only includes cold shutdown and refuel. Is this intentional or should the TB include defuel?

Resolution The IC specifically states that the loss of power is applicable to cold shutdown and refueling modes.

Therefore, NESP-007 operating mode applicability should not list defueled.

Changed PEG SA1 to exclude defueled mode. Added statement to PEG basis: "Note that Defuel mnrfo ic nnt annlinahlo tn thic It . honaiico tho IC'. ic cnonifirallu writton fnr nnlA chi>trlnwn anrl rohiol status 0 Open 0 Resolved/Awaiting Disposition Oi Closed

Record No. 17 Date 9/2Q/93 Name M. C. Daus Orlglnatlng Site Site Applicability JAF OJAF OIP-3 ONMP-2 EGeneric BWR OGeneral NMP-1 OIP-2 ONMP-1 OGinna OGeneric PWR Impact NMP-2 0 NUMARC-QQ7 0 Procedure 0 Verification 0 Training 0 Hardware HEAL OTechnical Bases OValidation ODeviation ONone cat. System Malf. Ic¹ 2 No. ** Emer. Class. Site Al'ea Comment (verification question 2.4): PEG IC SS2 operating mode is power operations only, but the TB EAL 2.2.2 includes startup/hot standby. Is this intentional?

Resolution This EAL is concerned with ATWS conditions in a BWR. Power operation mode does not encompass all of the plant conditions where an ATWS would be of concern in a BWR, therefore, it is appropriate to expand this EAL to include startup/hot standby mode.

C'.honnorl RWR PFA IC'. c'c'9 tn inning carlo ctort>>n/hnt ctonrlhv mnrto onrl orlrlorl ohnvo ovnbntotinn tn status OOpen OResolved/Awaiting Disposition 0+Closed Record No. 18 Date 9/2Q/93 Name M. C. Daus rlglnatlng Site Site Appllcab Illty JAF OJAF OIP-3 ONMP-2 ILGeneric BWR OGeneral NMP-1 Impact 0 IP-2 0 NMP-1 OGinna 0 Generic PWR NMP-2 ONUMARC-QQ7 0 Procedure OVerification OTraining 0 Hardware HEAL 0 Technical Bases 0 Validation 0 Deviation 0 None cat. System Malf. lc¹ 2 No. ** Emer. Class. General Comment (verification question 2.4): PEG IC SG2 operating mode is power operations only, but the TB EAL 2.2.3 includes startup/hot standby. Is this intentional?

Resolution This EAL is concerned with ATWS conditions in a BWR. Power operation mode does not encompass all of the plant conditions where an ATWS would be of concern in a BWR, therefore, it is appropriate to expand this EAL to include startup/hot standby mode.

l .honnorl RWR PC(~ IC'. c'C~P tn inning carlo ctorti >n/hnt ctonrlhv mnrlo onrl orlrtorl ohnvo ovnbntotinn tn status 0 Open 0 Resolved/Awaiting Disposition 0 Closed

Record No. 22 Date 9/2Q/93 Name M. C. DauS Orlglnatlng Site Site Applicability NMP-2 Q JAF 0 IP-3 0 NMP-2 0 Generic BWR 8 General NMP-1 Impact HIP-2 QNMP-1 QGinna CIGeneric PWR 0 NUMARC-007 Q Procedure 0 Verification Q Training 0 Hardware g EAL 8 Technical Bases I]Validation 0 Deviation 0 None No. **

Ic¹ ** **

cat. Abnorm. Rad. Emer. Glass.

Comment (verification) NMP-1, AU-1 bases: Why no reference listed to NMP-1 Tech Specs? Same for AA-1, AS-1, AG-1, and many others.

Resolution Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications is referenced in each of the above PEG EAL basis discussions.

status 0 Open 0 Resolved/Awaiting Disposition 0 Closed Record No. 23 Date 9/2Q/93 Name M. C. DauS rig inatlng Site Site Applicability NMP-1 OJAF OIP-3 ENMP-2 PGeneric BWR PGeneral NMP-2 Impact 0 IP-2 g NMP-1 0Ginna 0 Generic PWR Q NUMARC-007 Q Procedure Q Verification Q Training C3 Hardware laI EAL IHITechnical Bases OValidation ODeviation ONone cat. Abnorm. Rad. Ic¹ 1 No. ** Emer. Class. Site Area Comment (verification) AS1: Note "laters" here, both units.

Resolution Still waiting for numbers from NMP.

sta~us OOpen OResolved/Awaiting Disposition OClosed

Record No. 27 Date 9/2Q/93 Name M. C. DauS Originating Site Site Applicability NMP-1 OJAF OIP-3 8NMP-2 OGeneric BWR OGeneral NMP-2 Impact O IP-2 HNMP-1 OGinna OGeneric PWR O NUMARC-007 O Procedure O Verification C3Training O Hardware g EAL H Technical Bases O Validation O Deviation O None ca t. System Malf. Ic¹ 7 No. 1 Emer. Glass. UnuSual Event Comment (verification) NMP-1, NMP-2 SU-7.1 (also applies to SS3): It would seem that these two plants might agree on which-105 volts or 106 volts-constitutes loss of DC poweri Resolution Agree, but that's what we have from their data sources.

status OOpen OResolved/Awaiting Disposition OeClosed Record No. 28 Date 9/2Q/93 Name M. C. DauS rlglnatlng Site Site Applicability NMP-1 C3JAF C3IP-3 g NMP-2 C3Generic BWR OGeneral NMP-2 Impact C3 IP-2 8 NMP-1 OGinna O Generic PWR C3 NUMARC-007 C3 Procedure C3 Verification C3Training O Hardware g EAL HTechnical Bases C3 Validation O Deviation C3 None cat. System Malf. lc¹ 2 No. ** Emer. Class. Alert Comment (verification) NMP-1, NMP-2 SA2 bases: "Existence" is misspelled in second paragraph.

Resolution Corrected typo in NMP1, 2. JAF ok.

senatus 0 Open 0 Resolved/Awaiting Disposition 0+ Closed

Record No. 29 Date 9/20/93 Name M. C. DauS Originating Site Site Applicability NMP-2 OJAF HIP-3 ENMP-2 QGenencBWR OGeneral HIP-2 ONMP-1 CIGinna OGeneric PWR Impact C3 NUMARC-007 0 Procedure Q Verification 0 Training Cl Hardware g EAL ISL Technical Bases 0 Validation CI Deviation C3 None c¹ 4 ** Alert cat. System Malf. I No. Emer. Glass.

Comment (verification) NMP-2, SA4: "COLD SHUTDOWN" is checked as applicable, but shouldn't be.

Resolution Corrected NMP2; NMP1 and JAF are ok.

status OOpen 0 Resolved/Awaiting Disposition 0> Closed Record No. 31 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability NMP-1 0 JAF Q IP-3 8 NMP-2 Q Generic BWR Q General NMP-2 Impact Q IP-2 isi NMP-1 P Ginna 0 Generic PWR 0 NUMARC-007 0 Procedure 0 Verification 0 Training C3 Hardware g EAL H Technical Bases 0 Validation 0 Deviation 0 None cat. System Malf. Ic¹ 2 No. ** Emer. Class. Site Area Comment (verification) NMP-1, NMP-2 SS2: NUMARC says this EAL is applicable in Power Operation, but PEG says Power Operation and Hot Standby.

Resolution Deselected hot standby in NMP1, 2. JAF ok.

status OOpen OResolved/Awaiting Disposition OeClosed

Record No. 32 Date 9/20/93 Name M. C. Daus orlglnatlng Site Site Applicability NMP-1 Q JAF Q IP-3 Q NMP-2 g Generic BWR Q General NMP-2 Impact Q IP-2 Q NMP-1 Q Ginna Q Generic PWR JAF Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware g EAL 8 Technical Bases Q Validation Q Deviation Q None cat. System Malf. Ic¹ 2 No. ** Emer. Class. Site Af'ea Comment (verification) NMP-1, NMP-2 SS2 bases: The statement that "the generic guidance would require classification of a SAE for conditions in which the reactor is in fact shut down as a result of the scram signal..." is wrong. Apparently this sentence was copied over from the corresponding Alert.

Resolution Deleted sentence containing the above statement from NMP1, 2 and JAF.

status 0 Open 0 Resolved/Awaiting Disposition 0 Closed Record No. 33 Date 9/20/93 Name M. C. Daus rlglnatlng Site Site Applicability NMP-1 Q JAF Q IP-3 Q NMP-2 8 Generic BWR Q General NMP-2 Impact Q IP-2 Q NMP-1 Q Ginna Q Generic PWR JAF Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware g EAL H Technical Bases Q Validation Q Deviation Q None ca t. System Malf. lc¹ 5 No. ** Emer. Class. Site Al'ea Comment (verification) NMP-1, NMP-2 SS5: Would primary containment Hydrogen concentration above 4%

be better treated as a containment barrier potential breach? Also, loss of water level in the power operation, HSB and HSD conditions is treated as a fuel clad barrier eal for modes 1,2,3,4 & 5. So is it appropriate or necessary to expand SS2 from cold s/d and refueling to all modes?

Resolution It could be treated as a potential containment breach, but hydrogen generation is most directly an indication of prolonged inadequate core cooling. Expanding SS2 mode applicability is not necessary.

status 0 Open 0 Resolved/Awaiting Disposition 0+ Closed

Record No. 35 Date 9/20/93 Name M. C. DauS Orlglnatlng Site Site Appllcablllty NMP-1 C3 JAF P IP-3 P NMP-2 P Generic BWR E General NMP-2 Impact CI IP-2 Q NMP-1 0 Ginna Q Generic PWR JAF 0 NUMARC-007 0 Procedure CI Verification E Training 0 Hardware g EAL H Technical Bases C] Validation 0 Deviation C1 None cat. System Malf. Ic¹ 5 No. ** Emer. Class. Site Area Comment (verification) NMP-1, NMP-2 SS6: Should the EAL state that ALL of the indications needed to monitor plant parameters have to be unavailable? Why not half, or most?

Resolution NESP-007 specifies "most or all" indications where "most" is stated to be approximately 75'/o. But, NESP-007 also states that they do not expect the operator tally up the number of lost indicators.

This EAL is poorly worded in NESP-007. The emphasis needs to be on the need for increased surveillance resulting from whatever number is lost. This is a training issue until NUMARC chooses tn hottor rlofino thin I=AI status 0 Open 0 Resolved/Awaiting Disposition 0 Closed Record No. 36 Date 9/20/93 Name M. C. Daus rlglnatlng Site Site Appllcablllty NMP-1 0 JAF 0 IP-3 0 NMP-2 C3 Generic BWR 8 General NMP-2 Impact C3 IP-2 C] NMP-1 0 Ginna 0 Generic PWR Q NUMARC-007 0 Procedure C3 Verification g Training Q Hardware g EAL 8 Technical Bases 0 Validation Q Deviation Q None cat. System Malf. IC¹ 5 No. ** Emer. Class. Site AI'ea Comment (verification) NMP-1, NMP-2 SG1: Should a statement be added to the bases justifying use of only one parameter, i.e. RPV water can't be maintained above TAF, instead of broader fission product barrier monitoring?

Resolution If the core is covered, adequate core cooling exists no matter what the status of other fission product barriers.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed

Record No. 37 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability NMP-1 OJAF OIP-3 ONMP-2 EGenericBWR DGeneral NMP-2 Impact 0 IP-2 0 NMP-1 Q Ginna 0 Generic PWR JAF 0 Verification 0 NUMARC-007 CI Procedure El Training CI Hardware IL EAL ETechnical Bases QValidation DDeviation CINone Ic¹ No.

• • Site Area cat. System Malf. 5 Emer. Glass.

Comment (verification) NMP-1, NMP-2 FC2.1: Part of basis from NUMARC is missing.

Resolution Added to NMP1, 2 and JAF FC2.1 basis: The "Potential Loss" EAL is the same as the RCS barrier "Loss" EAL 4 below and corresponds to the (site-specific) water level at the top of the active fuel.

Thus, this EAL indicates a "Loss" of RCS barrier and a "Potential Loss" of the Fuel Clad Barrier.

This EAL appropriately escalates the emergency class to a Site Area Emergency.

status OOpen 0 Resolved/Awaiting Disposition 0 Closed Record No. 38 Date 9/20/93 Name M. C. Daus

'originating Site Site Applicability NMP-1 OJAF OIP-3 HNMP-2 I)Generic BWR C3General NMP-2 OIP-2 gNMP-1 OGinna QGenericPWR Impact Q NUMARC-007 P Procedure Q Verification Q Training Q Hardware laL EAL EITechnical Bases QValidation QDeviation ClNone cat. Barrier I c¹ ** No. ** Emer. Class.

Comment (verification) NMP-1, NMP-2 FC3.1, RC3, PC3.1: Note "later" ¹s still needed.

Resolution Still waiting for numbers from NMP.

status OI Open 0 Resolved/Awaiting Disposition 0 Closed

Record No. 39 Date 9/2Q/93 Name M. C. DauS Originating Site Site Applicability NMP-1 OJAF CIIP-3 DNMP-2 HGeneric BWR OGeneral NMP-2 0IP-2 CINMP-1 PGinna CIGeneric PWR Impact 0 NUMARC-007 0 Procedure Cl Verification Cl Training Cl Hardware g EAL ETechnical Bases C3Validation CjDeviation ClNone lc¹ ** No. **

cat. Barrier Emer. Class.

Comment (verification) NMP-1, NMP-2 PC1.1, 1.2 basis statement: I wonder if the NRC will question this addition to the bases...

Resolution This comment refers to the line out of the primary containment pressure decrease following rapid increase. Perhaps the NRC will question this, but they should be more concerned with the BWR EOPs than the EALs because the statement in the basis is the reason the operator is not keyed to respond based on the types of conditions suggested by NUMARC. Changed the PEG to include thoro rnnriitinnc Ilnrior tho ii trtnomont FAI PC'.R 1 status 0 Open 0 Resolved/Awaiting Disposition 0+ Closed Record No. 40 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-1 C3JAF CIIP-3 ONMP-2 HGeneric BWR CJGeneral NMP-2 HIP-2 QNMP-1 PGinna QGeneric PWR Impact JAF C3 NUMARC-007 I3 Procedure Q Verification 0 Training Q Hardware gg EAL HTechnical Bases C3Validation CIDeviation ONone I c¹

• • No. ** **

cat ~ Barrier Emer. Class.

Comment (verification) NMP-1, NMP-2 PC4.1: I would suggest more explanation should be added to the bases as to why we are using primary containment flooding as the criterion instead of the NUMARC criteria.

Resolution Added the following to the end of the second paragraph of NMP1, 2, JAF PEG PC4.1: The requirement for primary containmnent flooding addresses all plant conditions for which adequate core cooling is or is about to be lost. This includes RPV water level cannot be restored and maintained above TAF and RPV flooding conditions cannot be established and maintained. Thus, tho PI=A rnnrlitinn onnnmnoccoc tho Nl IMAAC'. nnnrtitinn r nnnorninn I=IPV wotor lovol onrI tho s~a~us 00pen 0 Resolved/Awaiting Disposition 0 Closed

0 Record No. 41 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability NMP-1 ClJAF OIP-3 C3NMP-2 gGeneric BWR QGeneral NMP-2 HIP-2 QNMP-1 QGinna CIGeneric PWR Impact JAF CJ NUMARC-007 0 Procedure C3 Verification Q Training 0 Hardware g EAL 8 Technical Bases 0 Validation 0 Deviation CI None I C¹ cat. Barrier Emer. Class.

Comment (verification) NMP-1, NMP-2: General comment: The barrier loss/potential loss table on page 3 of the evaluation is confusingly laid out; it would be better to assign a unique identifier to each loss or potential loss condition.

Resolution Agree, should identify in parentheses after each "Yes" the specific PEG EAL number.

status OOpen OResolved/Awaiting Disposition 0+Closed Record No. 42 Date 9/20/93 Name M. C. Daus rlglnatlng Site Site Applicability NMP-2 OJAF HIP-3 ONMP-2 HGeneric BWR CIGeneral Impact C3IP-2 0 NMP-1 QGinna C3Generic PWR 0 NUMARC-007 C3 Procedure Q Verification 0 Training Q Hardware g EAL g Technical Bases OValidation ODeviation 0 None cat. Barrier I C¹ ** No. ** Emer. Class.

Comment (verification) NMP-2, remark f21: Why isn't failure of a steamline to isolate with a direct path to the environment a loss of RCS rather than a potential loss as stated in the remark?

Resolution After review of the remark, I don't know why it is not a loss of RCS as opposed to a potential loss.

Changed remark to state loss of RCS.

sta~us OOpen OResolved/Awaiting Disposition OClosed

Record No. 43 Date 9/20/93 Name M. C. Daus orlglnatlng Site Site Appllcablllty NMP-2 C3 JAF 0 IP-3 Q NMP-2 IIGeneric BWR Q General Impact 0 IP-2 C7 NMP-1 0 Ginna C3 Generic PWR 0 NUMARC-007 0 Procedure C3 Verification Cl Training 0 Hardware g EAL lal Technical Bases P Validation CJ Deviation CJ None cat C¹ ** ** **

~ Barrier I Emer. Class.

Comment (verification) NMP-2 Unusual Event table: PC1a Pot. loss shouldn't reference remark ¹24, and PC1b Pot. loss should reference ¹25.

Resolution Agree, changed to remark 25.

status 0 Open 0 Resolved/Awaiting Disposition Closed Record No. 44 Date 9/20/93 Name M. C. Daus rlglnatlng Site Site Appllcabllity NMP-2 0 JAF 0 IP-3 0 NMP-2 8 Generic BWR Q General Impact CI IP-2 C3 NMP-1 0 Ginna 0 Generic PWR Q NUMARC-007 C3 Procedure 0 Verification C3 Training 0 Hardware IHI EAL IHI Technical Bases C3 Validation Q Deviation 0 None ca t. Barrier lc¹

• No. ** Emer. Class. **

Comment (verification) NMP-2 Unusual Event table: Remark ¹25 doesn't apply to PC3-pot. loss. Should ref.

26.

Resolution Agree, changed to remark 26.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed

Record No. 45 Date 9/20/93 Name M. C. DauS Originating Site Site Appllcablllty NMP-2 O JAF O IP-3 O NMP-2 8 Generic BWR O General Impact O IP-2 0 NMP-1 O Ginna O Generic PWR O NUMARC-007 O Procedure O Verification O Training O Hardware g EAL 8 Technical Bases O Validation O Deviation O None cat. Barrier I C¹ ** Emer. Class.

Comment (verification) NMP-2 Unusual Event table: PC-4 potential loss should reference remark ¹27 vs. 26.

Resolution Agree, changed to remark 27.

status 0 Open 0 Resolved/Awaiting Disposition 0 Closed Record No. 46 Date 9/20/93 Name M. C. Daus rlglnatlng Site Site Appllcablllty NMP-1 O JAF O IP-3 O NMP-2 8 Generic BWR O General NMP-2 Impact O IP-2 O NMP-1 O Ginna O Generic PWR O NUMARC-007 O Procedure O Verification O Training O Hardware g EAL H Technical Bases O Validation O Deviation O None c¹ ** ** **

cat ~ Barrier I Emer. Class.

Comment (verification) Remark ¹15: In most of the containments I'm familiar with, 50 GPM of RCS leakage is not very much and would take a very long time to result in a pressure increase to 1.68 psig, if ever.

Suggest this remark be reexamined.

Resolution Remark ¹15 applies to RCS1a-pot loss (RCS1.2 leakage into the dlywell >50 gpm). The NESP-007 basis for this EAL states in part "Many BWRs may be unable to measure an RCS leak of this size because the leak would likely increase drywell pressure above the drywell isolation setpoint". Measurement of leakage into the drywell for NMP1 is very limited. It just does not seem wnrth it tn irlontifu on FAI hocorl nn o uoru orhitroru nnmhor Sn nnml whon onu ciihctontiol omnnnt status 0 Open 0 Resolved/Awaiting Disposition Oe Closed

Record No. 47 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-2 0 JAF Q IP-3 Q NMP-2 H Generic BWR 0 General Impact 0 IP-2 0 NMP-1 0 Ginna 0 Generic PWR 0 NUMARC-007 0 Procedure 0 Verification CJ Training Q Hardware IL EAL 8 Technical Bases 0 Validation 0 Deviation 0 None ca t. Barrier I C¹ ** No. ** Emer. Class.

Comment (verification) NMP2 SAE table: First line (FC1 loss + RCS1a loss): I don't understand how remark 16 applies to this. If you have high coolant activity and a steam line break inside OR outside the drywell, it doesn't mean that you will have a dose at the site boundary of 100 mr/hr (AS1.1). If this was actually supposed to be remark 17, it still doesn't seem to work. FC3.1 is high drywell radiation indicating a LOCA with fuel damage. RCS 1a could be a main steam line break inside OR outside nnntolnrnnnt Resolution If the leak is outside the primary containment, this combination would require declaration of a General Emergency if the leak was not isolated. If the leak were isolated, the resulting puff release would require escalation to the SAE only if the exposure at the site boundary reached the level given in AS1.1. Otherwise, the release only justifies an Alert condition. Remark ¹16 has been nhonnori tn roflont tho ohnuo ctotornontc status 0 Open 0 Resolved/Awaiting Disposition 0>> Closed Record No. 48 Date 9/2Q/93 Name M. C. Daus rlglnatlng Site Site Applicability NMP-2 Q JAF 0 IP-3 0 NMP-2 8 Generic BWR Cl General Impact Cl IP-2 0 NMP-1 0 Ginna 0 Generic PWR Q NUMARC-007 0 Procedure 0 Verification C3 Training C3 Hardware g EAL g Technical Bases Q Validation Q Deviation C3 None Ic¹ ** * **

ca t. Barrier No. Emer. Class.

Comment (verification) NMP2 SAE table: Second line (FC1 loss + RCS2 loss): Remark ¹17 states that this condition is adequately covered by FC3.1. FC3.1 is based on all of the coolant activity of FC1.1 being dumped into primary containment. Can we really be sure that ALLthe coolant activity is in the drywell if drywell pressure is >3.5 psig?

Resolution Revised FPBEs to make Remark ¹17 N/A making FC1 loss + RCS2 loss an EAL. Added new EAL to Binning document and EAL TB. Added EAL to matrices.

status 0 Open 0 Resolved/Awaiting Disposition 0>> Closed

Record No. 49 Date 9/20/93 Name M. C. Daus Orlglnatlng Site Site Applicability NMP-2 0 JAF 0 IP-3 D NMP-2 8 Generic BWR Q General Impact G IP-2 0 NMP-1 0 Ginna Q Generic PWR 0 NUMARC-007 0 Procedure 0 Verification 0 Training 0 Hardware g EAL IHL Technical Bases 0 Validation D Deviation I3 None cat. Barrier I C¹ ** ** Emer. Class.

Comment (verification) NMP2 SAE table: FC2 loss + RCS1a loss is shown as a SAE, while referencing remark ¹8. But remark ¹8 says this EAL is unnecessary because it's covered by SS5.1. So why is this combination shown as a SAE?

Resolution Remark ¹8 should state that "... this portion of the EAL is unnecessary and can be deleted". "this portion" refers to RCS1a loss. Changed JAF, NMP1,2 FPBEs to state "... the RCS1a loss portion of the EAL is unnecessary and can be deleted".

status 0 Open 0 Resolved/Awaiting Disposition OI Closed Record No. 50 Date 9/20/93 Name M. C. Daus riglnatlng Site Site Appllcablllty NMP-2 0 JAF 0 IP-3 Q NMP-2 g Generic BWR Cl General Impact 0 IP-2 P NMP-1 Q Ginna Q Generic PWR 0 NUMARC-007 0 Procedure C3 Verification Q Training P Hardware g EAL lal Technical Bases 0 Validation Cl Deviation C] None cat. Barrier IC¹ * ** Emer. Class.

Comment (verification) NMP2 SAE table: FC4 loss + RCS1a loss: With respect to remark ¹24, I don' understand why the system and process monitors would not be in operation just because there is a main steam line break, especially if the break is outside the D/W. I would suggest more justification is needed here.

Resolution Remark ¹24 presumes that a valid gp 1 isolation signal in RCS1a loss would result in isolation of the steam lines. With no flow through offgas, the offgas monitors are not a good indication of fuel failure.

status 0 Open 0 Resolved/Awaiting Disposition OI Closed

Record No. 51 Date 9/2Q/93 Name M. C. DauS originating Site Site Applicability NMP-2 P JAF Cl IP-3 P NMP-2 IHI Genenc BWR P General Impact C] IP-2 0 NMP-1 0 Ginna Cl Generic PWR 0 NUMARC-007 0 Procedure 0 Verification 0 Training 0 Hardware g EAL 8 Technical Bases 0 Validation Q Deviation 0 None IC¹ ** ** **

cat. Barrier N Emer. Class.

Comment (verification) NMP2 SAE page 13: The condition FC4-pot. loss + RCS1a pot. loss is repeated five times in a row.

Resolution Corrected BWR FPBEs by listing correct RCS potential losses with FC4-pot. loss.

status 0 Open 0 Resolved/Awaiting Disposition Oo Closed Record No. 52 Date 9/2Q/93 Name M. C. DauS rlginatlng Site Site Applicability NMP-2 CI JAF Cl IP-3 Q NMP-2 8 Generic BWR 0 General Impact CI IP-2 0 NMP-1 0 Ginna 0 Generic PWR Q NUMARC-007 CI Procedure 0 Verification 0 Training Q Hardware g EAL H Technical Bases Q Validation 0 Deviation I7 None ca t. Bal'I'lel' C¹ ** No. ** Emer. Class. **

Comment (verification) NMP2 SAE page 14: For RCS1a pot. loss + FC4 loss: See ¹27 above.

Resolution This set of conditions is deleted because of Remark ¹20. Remark ¹20 has been clarified as follows:

"RCS1a pot. loss is > 50 gpm in the drywell. FC4 loss is very high offgas activity. High offgas activity under conditions in which steam flow to the main condenser is ongoing (i. e., high offgas readings valid) alone is indicative of a MSL failure to isolate with downstream pathway to the onvirnnmont Thic rnnditinn roniiiroc rlorhrotinn nf o Rito Aroo l=mornonrv>>nidor I=AI c <<cinn status 0 Open 0 Resolved/Awaiting Disposition OI Closed

Record No. 53 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability NMP-2 0 JAF 0 IP-3 P NMP-2 IIGeneric BWR Q General Impact Q IP-2 0 NMP-1 [3 Ginna 0 Generic PWR 0 NUMARC-007 0 Procedure CJ Verification C3 Training C3 Hardware IjEAL 8 Technical Bases 0 Validation 0 Deviation C3 None c¹ ** ** **

ca t. Barrier I Emer. Class.

Comment (verification) NMP2 SAE page 14: For RCS 1b-pot.loss + FC4-loss: See ¹31 above.

Resolution Expanded discussion in Remark ¹24 which justifies the deletion of these conditions. "Offgas monitors are not a reliable indicator of fuel failure under severely degraded conditions in that the system would be isolated and the process monitors would not be monitoring an unisolated process stream. High offgas activity under conditions in which steam flow to the main condenser is ongoing li o hinh nHnoc rooriinnc Molirib olnno ic inriirotiuo nf o moinctoom lino foil>>ro tn icnloto with status 0 Open 0 Resolved/Awaiting Disposition Oe Closed Record No. 54 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability NMP-2 C3 JAF Q IP-3 Cl NMP-2 H Generic BWR 0 General impact P IP-2 0 NMP-1 CI Ginna P Genenc PWR Q NUMARC-007 0 Procedure 0 Verification C3 Training Q Hardware g EAL IITechnical Bases CI Validation Q Deviation 0 None ca t. Barrier I c¹ ** ** Emer. Class.

Comment (verification) NMP2 SAE page 15: RCS6-pot. loss + FC1 loss should say: "Subsumed in

'Judgement EAL.'"

Resolution Changed JAF FPBE Remarks on page 15 for this combination of EALs to "Subsumed in

'Judgement'AL".

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed

Record No. 55 Date 9/2p/93 Name M. C. Daus Originating Site Site Applicability NMP-2 0 JAF C3 IP-3 Q NMP-2 H Genenc BWR 0 General Impact 0 IP-2 0 NMP-1 0 Ginna 0 Generic PWR 0 NUMARC-007 0 Procedure 0 Verification Cl Training 0 Hardware 8 EAL 8 Technical Bases 0 Validation 0 Deviation P None ca t. Barrier I c¹ No. Emer. Class.

Comment (verification) NMP2 SAE page 15: The combination of RCS1a-pot. loss + PC2b-loss references remark ¹22. Why not simply state "Not supported in PEG?"

Resolution "Condition not supported in PEG" is not appropriate because these conditions are supported in the PEGs.

.. status O Open O Resolved/Awaiting Disposition Oe Closed Record No. 56 Date 9/2P/93 Name M. C. DauS Originating Site Site Applicability NMP-2 Cj JAF Q IP-3 Q NMP-2 H Generic BWR C] General Impact Cl IP-2 0 NMP-1 Q Ginna 0 Generic PWR 0 NUMARC-007 Q Procedure Bases Q Verification 0 Training 0 Hardware g EAL 8 Technical 0 Validation Cl Deviation 0 None

• • ** Emer. Class.

cat. Barrier I C¹ Comment (verification) NMP2 GE page 17: For FC2-Loss + RCS4-Loss + PC1b-loss, remark ¹25 would be more appropriate than 14, since it states that PC1b-loss by itself constitutes a GE. Similarly, remarks 26 and 27 would be better for the two combinations that follow this one.

Resolution Page 17 only refers to PC1b-pot loss, not PC1b-loss. Changed JAF FPBE from Remark ¹14 to ¹25; Remark ¹26 and ¹27 are applied to the two combinations that follow this one. Since Remark ¹14 is no longer in use, N/A has been entered for this remark in the FPBE.

status O Open O Resolved/Awaiting Disposition Oe Closed

Record No. 57 Date 9/20/93 Name M. C. DauS Originating Site Site Applicability NMP-2 0 JAF C3IP-3 P NMP-2 g Generic BWR QGeneral HIP-2 QNMP-1 PGinna PGeneric PWR Impact C3 NUMARC-007 Cl Procedure 0 Verification 0 Training C3 Hardware g EAL 8Technical Bases 0 Validation Q Deviation 0 None cat. Barrier Ic¹ No.

'* Emer. Class.

Comment (verification) NMP2 GE table: Throughout the GE table, only one combination of RCS3-loss is shown in conjunction with loss of FC and PC, that where it is combined with PC-1 a potential loss. I don't understand why. What happened to PC-1b, PC-3, 4, 5, 6 pot. loss?

Resolution Don't know what happened appear to also be missing FC1-loss+ RCS4-loss+ PC1a-pot. loss each place the RCS4-loss follows the RCS2-loss. Added missing general emergency conditions including loss of all three barriers to the BWR FPBEs. Evaluation of the added conditions indicated that all were previously identified as a General Emergency, subsumed in the Judgement EALs, or tho nnnriitinn woc nnt c<<nnnrtori in tho PI=( c status 0 Open 0 Resolved/Awaiting Disposition 0 Closed Record No. 58 Date 9/20/93 Name M. C. Daus rlginatlng Site Site Applicability NMP-2 OJAF CIIP-3 QNMP-2 HGeneric BWR QGenerai Impact C3 IP-2 0 NMP-1 0Ginna 0 Generic PWR Q NUMARC-007 CI Procedure Q Verification 0 Training 0 Hardware IHI EAL H Technical Bases 0 Validation Q Deviation Q None cat. Barrier I c¹ N Emer. Class.

Comment (verification) NMP2 GE table: For combinations of LOSS OF RCS, LOSS OF PC, POT. LOSS OF FC only FC 2, 4 and 5 are listed in the combinations. Why not FC1 and FC3, both of which are admittedly not supported in the PEG?

Resolution FC1 and FC3 potential losses are not included in the Table because they are not supported in NESP-007. This fact is identified in the Table on page 3 of the FPBEs.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed

Record No. 59 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability NMP-2 0 JAF 0 IP-3 0 NMP-2 IJ Generic BWR CJ General Impact 0 IP-2 Q NMP-1 CJ Ginna 0 Generic PWR 0 NUMARC-007 0 Procedure 0 Verification 0 Training Q Hardware g EAL 8 Technical Bases Q Validation D Deviation Q None ca t. Barrier I C¹ ** No. Emer. Class.

Comment (verification) NMP2 GE: Remark ¹28 has a typo, should be EAL ¹FC1.1, not EAL 1.1.

Resolution Changed to EAL¹ FC1.1 in FPBE.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed Record No. 60 Date 9/20/93 Name M. C. Daus riginatlng Site Site Applicability NMP-2 0 JAF 0 IP-3 Q NMP-2 H Generic BWR 0 General Impact C3 IP-2 C3 NMP-1 0 Ginna Q Generic PWR 0 NUMARC-007 0 Procedure C] Verification Q Training Q Hardware g EAL g Technical Bases Cl Validation Q Deviation 0 None c¹ ** No. ** Emer. Class.

ca t. Barrier I Comment (verification) NMP2 GE table, page 25: Remark ¹28 does not apply to the combination of PC2a loss and FC2 loss. A new remark should make reference to RPV WL<TAF as a loss of fuel clad. A similar comments applies for the combination of PC2a loss and FC3 loss.

Resolution Revised Remark ¹28 to state that FC2-loss and FC3-loss are each losses of the fuel clad.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed

Record No. 61 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability NMP-2 0 JAF C3 IP-3 0 NMP-2 H Generic BWR Q General Impact P IP-2 Q NMP-1 P Ginna Q Genenc PWR 0 NUMARC-007 0 Procedure CI Verification 0 Training Q Hardware 8 EAL IHL Technical Bases C3 Validation 0 Deviation C3 None

• • ** Emer. Class.

cat. Barrier I C¹ Comment (verification) NMP2 GE table, page 25: Remark 28 does not appear to apply to the combinations of PC2a loss + FC4 loss + RCS1a/b pot. loss.

Resolution Revised Remarks ¹24 and ¹28 to explain the conditions in which use of the offgas air ejector setpoint would not be valid for emergency declaration.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed Record No. 62 Date 9/20/93 Name M. C. Daus Originating Site Site Ap pllcabillty NMP-1 Q JAF Q IP-3 Q NMP-2 mj Generic BWR Q General NMP-2 Impact Q IP-2 P NMP-1 P Ginna Q Generic PWR 0 NUMARC-007 Q Procedure CI Verification Q Training 0 Hardware g EAL El Technical Bases 0 Validation Q Deviation 0 None ca t. Barrier I c¹ Emer. Class.

Comment (verification) Throughout the table of LOSS OF PC + LOSS OF FC + POT. LOSS OF RCS, RCS2, 3, and 4 conditions are not listed. Admittedly they are not supported in the PEG.

Resolution These potential losses are not included in the Table because they are not supported in NESP-007.

This fact is identified in the Table on page 3 of the FPBEs.

status 0 Open 0 Resolved/Awaiting Disposition 0 Closed

Record No. 63 Date 9/2Q/93 Name M. C. DauS Orlglnatlng Site Site Appllcablllty NMP-2 Q JAF Q IP-3 Q NMP-2 g Generic BWR Q General Impact Q IP-2 Q NMP-1 Q Ginna Q Generic PWR Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware IL EAL g Technical Bases Q Validation Q Deviation Q None c¹ ** No. ** Class.

ca t. Barrier I Emer.

Comment (verification) NMP2 GE table, page 20 and later: Remark ¹8 applies to SAEs and it would seem should not be referenced in the GE table. These should be reevaluated.

Resolution Agree. Where PC2a-loss or PC2c-loss is used in the GE table, Remark ¹28 is applied. Where PC2c-loss is used in the GE table, Remark ¹25 is applied.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed Record No. 64 Date 9/2Q/93 Name M. C. Daus rlglnating Site Site Appllcab illty NMP-2 Q JAF Q IP-3 Q NMP-2 g Generic BWR Q General Q IP-2 Q NMP-1 Q Ginna Q Generic PWR Impact Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware g EAL g Technical Bases Q Validation Q Deviation Q None IC¹ ** No.

ca t. Barrier Emer. Class.

Comment (verification) NMP2 GE table, page 25: Remark ¹22 states that PC2b-loss should be a GE all by itself. So why isn't it listed as one.

Resolution PC2b-loss appears in the EALs under venting for PCPL and H2/02 at or above combustible limits.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed

Record No. 65 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-2 QJAF tlIP-3 QNMP-2 HGeneric BWR QGeneral C]IP-2 ONMP-1 OGinna DGeneric PWR Impact 0 NUMARC-007 C3 Procedure 0 Verification CI Training CJ Hardware SEAL 0 8 Technical Bases Validation 0 Deviation 0 None

• • No. **

cat. Barrier lc¹ Emer. Class.

Comment (verification) NMP2 GE table, page 25 and 26: Remarks ¹25, 26 and 27 are referenced in a lot of combinations they don't apply to.

Resolution Deleted Remarks ¹25, ¹26, and ¹27 from combinations with EAL PC2b-loss because intentional venting alone is reason to declare a General Emergency.

status OOpen O Resolved/Awaiting Disposition Oi Closed Record No. 66 Date 9/2Q/93 Name M. C. DauS Originating Site Site Applicability NMP-1 QJAF HIP-3 PNMP-2 EGenericBWR QGeneral NMP-2 Impact 0 IP-2 Q NMP-1 P Ginna C3 Generic PWR

,C3 NUMARC-007 C3 Procedure Q Verification 0 Training 0 Hardware g EAL 8 Technical Bases Cl Validation 0 Deviation 0 None cat ~ Barrier lc¹ ** Emer. Class.

Comment (verification) With respect to AU2.4, listed in the" Reactor Fuel" category, other things than fuel degradation could cause a hundredfold increase in area radiation monitors. Same for AA3.1 and AA3.2. Suggest these three EALs belong in the "Equipment Failures" category...?

Resolution Almost all EALs could be grouped under "Equipment Failures" since equipment failures generally contribute to the seriousness of an event and lead to emergency classifications..Validation evaluation of EALs should indicate if these PEG EALs are properly categorized.

senatus OOpen OResolved/Awaiting Disposition O>Closed

Record No. 67 Date 9/2Q/93 Name M. C. DauS Originating Site Site Applicability NMP-1 P JAF C3IP-3 QNMP-2 HGeneric BWR PGeneral NMP-2 0IP-2 CINMP-1 C3Ginna OGeneric PWR Impact JAF 0 NUMARC-007 CJ Procedure CI Verification C3 Training C3 Hardware 8 EAL HTechnical Bases 0 Validation P Deviation I] None cat. Barrier I C¹ ** Emer. Class.

Comment (verification) AA2.1 is duplicated, in 1.4 and 1.5.

Resolution Deleted AA2.1 from subcategory 1.5.

status 0 Open 0 Resolved/Awaiting Disposition 0+ Closed Record No. 68 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-2 0 JAF 0IP-3 GNMP-2 g Generic BWR C3General Impact HIP-2 QNMP-1 QGinna P Generic PWR

0 NUMARC-007 Q Procedure 0 Verification 0 Training P Hardware g EAL 8 Technical Bases OValidation 0 Deviation ONone cat. Barrier I C¹ Emer. Class.

Comment (verification) NMP-2 PEG, EALAS1.3: Says1000mr/hr, should say100mr/hr. NMP1 PEG isOK.

Resolution Changed NMP-2 PEG EAL AS1.3 to 100 mr/hr.

senatus OOpen OResolved/Awaiting Disposition OClosed

Record No. 69 Date 9/20/93 Name M. C. Daus Orlglnatlng Site Site Applicability NMP-1 CIJAF OIP-3 ONMP-2 8Generic BWR OGeneral NMP-2 0IP-2 QNMP-1 PGinna PGeneric PWR Impact JAF 0 NUMARC-007 0 Procedure 0 Verification CI Training p Hardware III EAL IHITechnical Bases 0 Validation Q Deviation 0 None I C¹

• • ** Emer. Class.

cat. Barrier

• 'omment (verification) Considering that sometimes the emergency Coordinator may not be able to distinguish between fire and explosion, and considering the close association of HU1.5 and HU2.1, consider combining the "fire" and "Man-made events" into one category.

Resolution Fire category will be expanded to be fire/explosions and not combined with man-made events.

status 0 Open 0 Resolved/Awaiting Disposition 0 Closed Record No. 70 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability NMP-1 QJAF OIP-3 ONMP-2 SGeneric BWR ClGeneral NMP-2 0IP-2 C3NMP-1 OGinna OGeneric PWR Impact JAF Q Verification Q Training Q Hardware Q NUMARC-007 P Procedure 5g EAL g Technical Bases Q Validation C] Deviation 0 None I C¹

• • ** Emer. Class.

cat. Barrier Comment (verification) In section 2.0, Reactor Vessel, SS5.1 and FC2.1 are redundant EALs (both are RPV WL(TAF)..

Resolution It is possible for Fission Product Barrier EALs to be redundant with event based EALs.

status 0 Open 0 Resolved/Awaiting Disposition 0 Closed

Record No. 71 Date 9/20/93 Name M. C. Daus Orlglnatlng SIte Site Appllcablllty NMP-1 Q JAF P IP-3 P NMP-2 8 Generic BWR Q General NMP-2 Impact CJ IP-2 CI NMP-1 CI Ginna 0 Generic PWR JAF Q NUMARC-007 CJ Procedure 0 Verification 0 Training C3 Hardware SEAL 8 Technical Bases C] Validation CI Deviation 0 None Ic¹ ** No. **

ca t. Barrier Emer. Class.

Comment (verification) RCS3.1 is indicative of an RCS leak only, i.e. no fuel damage. So I suggest that the Reactor Fuel bin is not the appropriate place for this EAL. Maybe the "Reactor Pressure Vessel" category should be made into "Reactor Pressure Vessel and Steam Systems."

Resolution Despite the fact that NUMARC says this rad level is indicative of reactor coolant in the drywell with tech spec level of activity, the source of activity is due to exposure to irradiated fuel in the RPV. As such, this EAL is indicative of the status of Reactor Fuel.

status 0 Open 0 Resolved/Awaiting Disposition 0+ Closed Record No. 72 Date 9/20/93 Name M. C. Daus rlglnatlng Site Site Ap plica b ill ty NMP-1 0 JAF 0 IP-3 0 NMP-2 8 Generic BWR 0 General NMP-2 Impact Q IP-2 0 NMP-1 0 Ginna 0 Generic PWR 0 NUMARC-007 0 Procedure 0 Verification Q Training Q Hardware gg EAL g Technical Bases C] Validation Q Deviation Q None ca t. Barrier IC¹ ** No.

• Emer. Class.

• 'omment (verification) NMP-1, 2, NUE 1.1 1: The stated basis for this doesn't read much like the PEG,

~

although it seems OK...

Resolut ton Agree.

status 0 Open 0 Resolved/Awaiting Disposition 0 Closed

Record No. 73 Date 9/2Q/93 Name M. C. Daus Orlglnatlng Site Site Applicability NMP-2 0 JAF Q IP-3 H NMP-2 Q Generic BWR C3 General Impact I7 IP-2 H NMP-1 0 Ginna [3 Generic PWR 0 NUMARC-007 OProcedure OVerification C3Training CIHardware H EAL H Technical Bases C3 Validation 0 Deviation Q None cat. Barrier Ic¹ ** N o.

Emer. Class.

Comment (verification) NMP-2, NUE 1.2.1: The EAL states 15 minutes, but there's no mention of 15 minutes in the PEG.

Resolution NMP-2 PEG EAL is based on the offgas radiation alarm setpoint. The setpoint for NMP-2, unlike NMP-1, includes a 15 minute time delay.

status Q Open Q Resolved/Awaiting Disposition Qe Closed Record No. 74 Date 9/2Q/93 Name M. C. Daus riglnatlng Site Site Applicability NMP-1 0 JAF P IP-3 H NMP-2 P Generic BWR C3 General NMP-2 Impact 0 IP-2 H NMP-1 CJ Ginna 0 Generic PWR C3 NUMARC-007 Q Procedure Q Verification Q Training C3 Hardware HEAL H Technical Bases 0 Validation C3 Deviation Q None cat. Barrier lc¹ ** ** Emer. Class.

Comment (verification) NMP-1, 2, NUE 1.2.2: I don't see where it says in the PEG that 10 times the DRMS alarm setpoint is equivalent to 300 pCi/CC l-131.

Resolution Added discussion in EAL TB basis to PEG EAL basis for NMP1, 2. JAF is ok.

status Q Open Q Resolved/Awaiting Disposition Qe Closed

Record No. 75 Date 9/20/93 Name M. C. Daus Orig lnatlng Site Site Applicability NMP-1 Q JAF Q IP-3 Q NMP-2 g Genenc BWR Q General NMP-2 Impact Q IP-2 Q NMP-1 Q Ginna Q Generic PWR Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware SEAL H Technical Bases Q Validation Q Deviation Q None ca t. Barrier I c¹ ** ** Emer. Class.

Comment (verification) NMP-1, 2, ALERT 1.4.4: The second paragraph of the basis is redundant with the first.

A suggestion: Since the referenced NMPC memo may not be immediately available to anyone reading the Tech Basis, a brief explanation might be appropriate.

Resolution Deleted second paragraph of EAL TB basis for NMP1, 2 and JAF.

status 0 Open 0 Resolved/Awaiting Disposition 0+ Closed Record No. 76 Date 9/20/93 Name M. C. Daus rlginatlng Site Site Appllcablllty NMP-1 Q JAF Q IP-3 8 NMP-2 Q Generic BWR Q General NMP-2 Impact Q IP-2 8 NMP-1 Q Ginna Q Generic PWR Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware 8 EAL g Technical Bases Q Validation Q Deviation Q None ca t. Barrier I C¹

• • No. ** Emer. Class.

Comment (verification) NMP-1, 2, ALERT 1.5.2: PEG reference of AU2.2 is cited. Should be AA2.2.

Resolution Changed NMP 1, 2 EAL TB 1.5.2 reference to AA2.2, JAF ok.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed

Record No. 77 Date 9/20/93 Name M. C. Daus Orlglnatlng Site Site Appllcabllity NMP-1 0 JAF 0 IP-3 0 NMP-2 g Generic BWR 0 General NMP-2 Impact 0IP-2 0NMP-1 0Ginna 0Generic PWR 0 NUMARC-007 0 Procedure 0 Verification 0 Training 0 Hardware g EAL 8 Technical Bases 0 Validation 0 Deviation 0 None cat. Barrier I c¹ '* No. '* Emer. Class.

Comment (verification) NMP-1, 2, EALs 2.2.1, 2.2.2, 2.2.3, 2.2.4: These EALs state, "any manual scram which fails to shut down the reactor." But the PEG states, "Any manual scram or automatic scram followed by a manual scram which fails. ."~

Resolution These are one in the same since operating procedures require that any automatic scram be followed by one or more manual scram attempts.

status OOpen OResolved/Awaiting Disposition OeClosed Record No. 78 Date 9/20/93 Name M. C. Daus Orlglnatlng Site Site Appllcablllty NMP-1 0JAF 0IP-3 0NMP-2 8 Generic BWR 0General NMP-2 Impact 0 IP-2 0 NMP-1 0Ginna 0 Generic PWR 0 NUMARC-007 0 Procedure 0 Verification 0 Training 0 Hardware 5g EAL 8 Technical Bases 0 Validation 0 Deviation 0 None cat. Barrier lc¹ ** No. ** Emer. Class.

Comment (verification) NMP-1, 2: Noted that PC2.2 is referenced for GEs 3.2.2 and 3.4.2.

Resolution It should be because these are conditions requiring intentional venting per EOPs.

status OOpen 0 Resolved/Awaiting Disposition 0 Closed

Record No. 79 Date 9/20/93 Name M. C. Daus Orlginatlng Site Site Appllcablllty NMP-1 Q JAF Q IP-3 Q NMP-2 8 Genenc BWR Q General NMP-2 Q IP-2 Q NMP-1 Q Ginna Q Generic PWR Impact Q NUMARC-007 Q Procedure Q Verification Q Training O Hardware g EAL H Technical Bases Q Validation Q Deviation Q None ca t. Barrier IC¹ * ** Emer. Class.

Comment (verification) NMP-1, 2: PEG section RCS 1.3 is not referenced for EAL 4.1.1 in the binning document, but is referenced in the Tech Basis for 4.1.1.

Resolution Changed binning document 4.1 from PC2.3 (SAE) to "PC2.3 or RCS1.3 (Temp) (SAE)". Changed binning document 4.2 from PC2.3 (SAE) to "PC2.3 or RCS1.3 (Rad) (SAE)".

status 0 Open 0 Resolved/Awaiting Disposition 0 Closed Record No. 80 Date 9/20/93 Name M. C. Daus rlginatlng Site Site Appllcablllty NMP-1 Q JAF Q IP-3 8 NMP-2 Q Generic BWR Q General NMP-2 Impact Q IP-2 Q NMP-1 Q Ginna Q Generic PWR Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware 5g EAL g Technical Bases Q Validation Q Deviation Q None ca t. Barrier I C¹ ** N

• • Emer. Class. **

Comment (verification) NMP-2, EAL 4.1.2: The word "temperature" is misspelled in the description of the EAL.

Resolution Corrected spelling in NMP 2. NMP1 and JAF ok.

status 0 Open CI Resolved/Awaiting Disposition Oe Closed

Record No. 81 Date 9/20/93 Name M. C. DauS Originating Site Site Applicability NMP-2 CIJAF OIP-3 HNMP-2 QGenericBWR OGeneral OIP-2 ONMP-1 QGinna OGeneric PWR Impact Q NUMARC-007 0 Procedure 0 Verification Q Training C1 Hardware g EAL IHITechnical Bases 0 Validation C3 Deviation Q None Ic¹ ** ** **

cat. Barrier N Emer. Class.

Comment (verification) NMP2: For the Effluent Monitor Classification Threshold Table of EAL 5.1.1: At the Alert level, the PEG calls for 200xDRMS setpoint for RW/RxBldg Vent Effl. Mon. and the main stack effluent monitor. But EAL 5.1.1 says "lateV'or both.

Resolution Still waiting for numbers from NMP.

status Oe Open 0 Resolved/Awaiting Disposition 0 Closed Record No. 82 Date 9/20/93 Name M. C. Daus rlginating Site Site Applicability NMP-1 ClJAF HIP-3 P NMP-2 PGeneric BWR gGeneral NMP-2 Q IP-2 P NMP-1 P Ginna Q Generic PWR Impact C3 NUMARC-007 0 Procedure 0 Verification C3 Training Q Hardware g EAL HTechnical Bases 0 Validation Q Deviation ClNone ca t ~ Barrier I C¹ ** N

• • Emer. Class.

Comment (verification) NMP1, 2: For EAL 6.1.1, there is no mention of the PEG statement that at least two emergency generators are supplying power to emergency buses.

Resolution Availability of DGs is unnecessary in this EAL because, if they are unavailable, a higher emergency classification would be declared due to EAL 6.1.2.

senatus 0 Open 0 Resolved/Awaiting Disposition Oe Closed

Record No. 83 Date 9/20/93 Name M. C. DauS Originating Site Site Applicability NMP-2 C3 JAF CJ IP-3 HNMP-2 OGeneric BWR C3General Impact 0IP-2 DNMP-1 C3Ginna OGeneric PWR 0 NUMARC-007 0 Procedure Cl Verification CI Training C3 Hardware g EAL 8 Technical Bases C3 Validation 0 Deviation C3 None cat. Barrier Ic¹ ** Emer. Class.

Comment (verification) NMP2: For EAL6.2.1,the EALspecifies<112.5VDCon 2BYS*BAT2C. There is no mention of 112.5 VDC in the PEG.

Resolution NMP-2 PEG EAL SU7.1 states "<1 05 vdc bus voltage indications on 125 vdc batteries 2BYS*BAT2Aand B, and <112.5 vdc on 125 vdc battery 2BYS*BAT2C".

.status OOpen OResolved/Awaiting Disposition OIClosed Record No. 84 Date 9/20/93 Name M. C. Daus rlglnatlng Site Site Applicability NMP-1 OJAF CJIP-3 ONMP-2 OGeneric BWR HGeneral NMP-2 Impact C3IP-2 P NMP-1 QGinna OGeneric PWR C3 NUMARC-007 0 Procedure 0 Verification 0 Training CI Hardware g EAL IHITechnical Bases C3Validation ClDeviation C3None cat. Barrier I c¹ ** ** Emer. Class. **

Comment (verification) NMP1, 2: For EAL 7.3.4, see comment ¹1 3 above.

Resolution NESP-007 specifies "most or all" indications where "most" is stated to be approximately 75%. But, NESP-007 also states that they do not expect the operator tally up the number of lost indicators.

This EAL is poorly worded in NESP-007. The emphasis needs to be on the need for increased surveillance resulting from whatever number is lost. This is a training issue until NUMARC chooses tn hottor rlofino thic FAI status 0 Open 0 Resolved/Awaiting Disposition OI Closed

Record No. 85 Date 9/20/93 Name M. C. Daus Orlglnatlng Site Site Appllcablllty NMP-1 0 JAF C3 IP-3 C3 NMP-2 Q Generic BWR g General NMP-2 Impact 0 IP-2 CI NMP-1 CJ Ginna Q Generic PWR 0 NUMARC-007 0 Procedure 0 Verification 0 Training Cl Hardware g EAL IHL Technical Bases 0 Validation Q Deviation C1 None cat. Barrier I c¹ ** No. ** Emer. Class.

Comment (verification) NMP1, 2: EAL 8.1.2 references PEG HA4.2, but there is no mention in the EAL of "other" security events. Similar comment for EAL 8.1.3.

Resolution Since there is no defined "other" security event for this example EAL, this condition is addressed under the Judgement EALs.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed Record No. 86 Date 9/20/93 Name M. C. Daus rlglnatlng Site Site Appllcablllty NMP-2 0 JAF Q IP-3 Q NMP-2 Q Generic BWR 8 General Impact P IP-2 Q NMP-1 Q Ginna P Generic PWR P NUMARC-007 Q Procedure C3 Verification 0 Training Q Hardware 5g EAL IHI Technical Bases C] Validation Cl Deviation 0 None ca t. Barrier Ic¹ ** No. ** Emer. Class. **

Comment (verification) NMP2: In EAL 8.2.2, the list of affected areas does not match that in the PEG.

Resolution Updated PEG to use list of areas in HU in HA2.1.

status 0 Open 0 Resolved/Awaiting Disposition Oi Closed

Record No. 87 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability NMP-1 Cl JAF 0 IP-3 C] NMP-.2 0 Generic BWR 8 General NMP-2 Impact C3 IP-2 0 NMP-1 P Ginna CI Generic PWR C3 NUMARC-007 0 Procedure CI Verification 0 Training Cl Hardware g EAL 8 Technical Bases Q Validation 0 Deviation Cl None cat. Barrier ic¹ ** No. Emer. Class.

Comment (verification) NMP1, 2: EAL 8.3.5 references PEG HA3.2, but makes no reference to flammable gas.

Resolution Added reference to flammable gases in wording of EAL.

status .0 Open,O Resolved/Awaiting Disposition Oi Closed Record No. 88 Date 9/20/93 Name M. C. Daus rig inatlng Site Site Applicability NMP-2 0 JAF 0 IP-3 C3 NMP-2 C3 Generic BWR IIIGeneral C3 IP-2 0 NMP-1 Cl Ginna Q Generic PWR Impact 0 NUMARC-007 C3 Procedure C3 Verification Q Training CJ Hardware g EAL IHL Technical Bases C1 Validation CI Deviation Q None No. **

ca t. Barrier IC¹

• Comment (verification) NMP2: For EAL 8.4.3, the Reactor Building is not included in the list of Plant Vital Areas. Same for 8.4.6, 8.4.7.

Resolution Reactor Building is included in Table 8.4.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed

OSSI 93-402A-10-NMP2 NMP-2 EAL VerlQcatlon & Valldatlon Report, Rev. 0 Attachment 3 L<'ALVaMation Scenarios 3-1

OSSI 92-402A-7A-NMP2 EAI Validation Scenarios, Rev. 0 Attachment 2 VaHdation Exercise Scenario Checklist No.:

Plant: NMP-2 Simulator: ~ Table-Top: Cl Scenario ¹1 Scenario Description(s):

Initial Conditions: Reactor power 100%; HPCS DG OOS in seven day LCO.

With the plant at 100% power, main turbine pressure controllers fail low initiating a turbine trip and reactor scram signal.

All rods fully insert.

Reserve and aux boiler station transformers fail to energize when the generator trips (UE 6.1.1). DGs start and energize emergency busses.

RCIC steam line ruptures due to pressure spike and RCIC isolation valves faQ to isolate.

Emergency RPV depressurization due to secondary containment maximum safe operating temp"=.ature values (SAE 4.1.1).

Bomb explosions in the switchyard and the two DGs (UE 8.1.1, Alert 8.2.2) cause loss of RPV injection sources, (105 vdc on all batteries (SAE 6.2.2),

and loss of offsit ~~~ve: (UE 6.1.1); loss of annunciators and indicators and increased survei!!",:-::.. with transient in progress (Alert 7.3.3). [if explosion were to have occu'..-~ v:hile in cold shutdown/refuel, battery loss per UE 6.2.1 and loss of '":! '. ~'.= per 6.1.2.]

RPV water decreases <TAI (SAE 2,1.1); with primary system discharging outside primary co;".'.".'.;".-;.cnt and RB temperatures above MSO levels in two or more areas (GK 4.'.",'.

Some fuel damage cc"'.. s with core uncovery. RB ARMs increase above MSO values in more tK-;;.. 'o areas; with primary system discharging outside primary contain:....... ("-, ~.T: 4.2.1, SAE 7.3.4, GE 4.2.2).

2-1

OSSI 92-402A-7A-N.""~2 EAL Validation Scenarios, Rev. 0 A;."-.:.=. c"..t 2 VaIMation Exercise Scenario Checklist No.:

Plant. NMP-'? Simulator: ~ Table-Top: 0 Scenario ¹2 Scenario Descri".". '.- '".,'.:

Initial Conditions: Reactor power 100% for past 3 months, small steam leak from one turbi "," '.!".rottle valve, plant to shutdown tomorrow With the plant at I "0% power and a small steam leak on turbine throttle valve, a dropped co.-.":o.'-. - od results in fuel clad failure.

Reactor scrams; ~~.'. Is n ar HCUs exceed 100 times alarm setpoint (UE 1.4.1).

Offgas activity i ~.";- (UE 1.2.1).

c e,".

One MSL fails to (3 Z 3.5.1).

Offsite radioactivi',; r"1case increases to the General Emergency level (UE 5.1.1, Alert 1.2.2 "';;: =.l.2, SAE 5.1.3, GE 5.1.4).

Drywell radiation = "..-! i. gs increase (Alert 1.3.1).

Coolant sample r..:..'. -300 pCi/gm) support high offsite radiactivity readings (UE l.l.......'<- i.1.2, GE 3.5.2).

Emergency RPV '..;"-..":..i;.ation is required.

2-2

OSSI 92-402A-7A-NMP2 EAI. Validatfon Scenarios, Rev. 0 A:".::= '.="."- t 2 VaHdation Exercise Scenario Checklist No.:

Plant NMP->>, Simulator: ~ Table-Top: Cl Scenario 0 3 Scenario Descris..".:.= -.

'.-,,'nitial Condition.::;.

":."tor power 60%, return to power delayed with feedwater heate ~

.";..s. HPCS out of service with bearing replacement; due back in 4 hc.."

Earthquake cause".;. 's";.ic activity alarms at JAFNPP and NMP-1/2 (UE 8.4. 1).

Small loca into dg~>>~'!, unidentiQed leakage >10 gpm (UE 3.1.1).

Drywell pressure > "".r'. setpoint (Alert 3.2.1).

Multiple failures c" '..- ~.~lection systems RPV water level d"."...:-;.:-."s < TAF (SAE 2.1.1) 2-3

OSSI 92-402A-7A-NMP'-? EAL ValfdaUon Scenarios, Rev. 0 P ..t."-. -..h=;."..". t 2 Validation Exercise Scenario Checklist No.:

Plant: NMP- -. Simulator: S Table-Top: Cl Scenario ¹4 Scenario Descrip':;..","):

Initial Condition..: .. ""..!or power 100%, no equipment OOS.

Condensate head"='; .-.. s resulting in a loss of feed.

When the reactor s".:".-,n;s on low RPV water level, several control rods fail to insert (Alert 2.2.1). 'reactor power remains above 5%.

RCIC and HPCS do:"..".-t r main operable.

Various other fail ~r";s!~a.ic one low pressure ECCS pump for RPV makeup.

Boron injection i,:: ..=:: "-., (.".-.AE 2.2.2).

SRV operation h ='= .";;:-".:cssion pool and leak from suppression pool causes water lev..! '" .-: .. e~.".e. RPV pressure and suppression pool temperature can.". '..=."..;:aintained below the HCTL (SAE 3.3.1, GE 2.2.3);

RPV water level c":".;. ':>e maintained below MSCRWL (GE 2.2.4).

Hydrogen concen:'"...'".. in t'e suppression chamber reaches 4% (SAE 3.4. 1).

2-4

OSSI 92-402A-7A-NMP2 EAL Valfdation Scenarios, Rev. 0 J'tt". ':....e. t 2 Validation Exercise Scenario Checklist No.:

Plant: NMP-2 Simulator: S Table-Top: Cl Scenario 8 5 Scenario Descript'"..".!~):

Initial Conditions: '." ..ctor power 75%, a shutdown is in progress for a drywell entry to icc:;";.:;"-. c!"-ritified leakage, wetwell is deinerted, drywell deinertion in pr <;, ; ..."!'.;!g~ell oxygen concentration 10%. No equipment OOS.

Loss of offsite pow..-.. o.."u-s (UE 6.1.1).

Reactor fails to scr". -'..."-0 rods out (Alert 2.2.1).

When RPS fuses p~" ':!. all rods fully insert.

All but one DG <a! .":-.::I-." ',-"Jert 6.1.3).

Remaining DG txil;.". ',.=~" r'.1.4).

Major LOCA occurs, "'.>V water level cannot be restored and maintained above TAF-(GE 6.1.",'.

After 20 minutes, c:".e '- returned to operation; available injection cannot restore RPV water l"= -.;.',~eve TAF.

Drywell hydrol::- .~r'. '"..".K 3.4.1).

Primary Conta.'."..;.* ...'..."..' is required (GE 3.1.2).

Hydrogen in drywell >6% (GE 3.4.2).

Primary contain';:c:-:.t .'s vent"..d due to PCPL (GE 3.2.2).

2-5

OSSI 92-402A-7A-N.'.~P2 EAL Validation Scenarios, Rev. 0 Att".: ". ~:.=nt 2 Validation Exercise Scenario Checklist No.:

Plant: Simulator: Cl Table-Top: ~

Scenario 8 6 Scenario Descrip'...-. - ',-',:

After elevated of:"p-.-, ."'.,'s a noted, reactor coolant samples indicate coolant activity >,":,<,',- g,'-131 eq. (UE 1.1.1)

Following react"," ."'".,'..:.-.:-,..d depressurization, coolant samples are taken indicating 390 pCiy -:m i-131 cq. (Alert 1.1.2) 2-6

OSSI 92-402A-7A-NMP2 EAT. Validation Scenarios, Rev. 0 At*"=-.'.; cnt 2 Validation Exercise Scenario Checklist No.:

Plant: M<P-2 Simulator: 0 Table-Top: ~

Scenario ¹7 Scenario Descr'r "..-'."):

Reactor scrams on!.'.x d:g~vell pressure. Drywell radiation levels indicate

[Later] R/hr. (PJer..". ).

Following emerge";."..."..".'V c!@pressurization, drywell radiation levels of

[Later] R/hr are indicated (ShE 1.3.2).

At what level wou!d; ou dcc!are a General Emergency based on drywell radiation levels? (GT 1.3.3) 2-7

OSSI 92-402A-7A-i"'~.".~" EAL Validation Scenarios. Rev. 0 Attachxnent 2 Validation Exercise Scenario Checklist No.:

Plant: renp-." Simulator: Q Table-Top: ~

Scenario 8 8 Scenario Descrip".o". f"-):

A HP Technician pc. farming routine surveys measures Control Room area radiation levels oi ".:..":.2j!'r (Alert 1.4.3).

It is reported that .".n ".~shielded radiography source is in the Relay Room.

General area red ~.tion'eve!s in the relay room are approximately 20 R/hr (Alert 1.4.4) 2-8

OSSI 92-402A-7A-Nl<>2 EAI. Validation Scenarios, Rev. 0 At".".".'....."".t 2 Validation Exercise Scenario Checklist No.:

Plant: Simulator: Cl Table-Top: ~

Scenario 8 9 Scenario Descria"."."!"'efueling operations are in progress and a main steam line plug begins to leak causing the r"f':"! "..< ca.".ty and spent fuel pool level to drop. The SFP low level alarm is r".."ic;cd (UE 1.5.1).

A fuel bund!. i.=. c'. '"e <r~pp'e and in the cattle shute when the refuel floor is evacuated ',.'.:;.; '..:

The refuel floor radiation monitors go offscale high (Alert 1.4.2) 2-9

OSSI 92-402A-7A-NMP2 EAL Valfdatfon Scenarios, Rev. 0 8 t..-..l......"..".t" Validation Exercise Scenario Checklist No.:

Plant. NMP-2 Simulator; Cl Table-Top: ~

Scenario ¹ 10 Scenario Descrii."io "<"'.

Chemistry reports stack effluent analysis indicates that effluents have been approximately 3 '.i::. ".s;"ech. S~ec. allowed limits for the last 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (UE 5.2. 1).

300 times Tec.'s. Spec. for the last 20 minutes (Alert 5.2.2) 2-10

OSSI 92-402A-74-F >.",".P2 EAL Validation Scenarios. Rev. 0 Att,. b...cnt 2 Validation Exercise Scenario Checklist No.:

Plant: NynP 2 Simulator: Cl Table-Top: ~

'cenario ¹ 11 Scenario Descr'-:c.-',".):

Field survey teams rcport ivhole body dose rates at the site boundary of 20 mR/hr (Alert 5.";,.3,'.

200 mR/h (8,':.'..;" " "

Dose projection" .':".".jcate child thyroid doses of 7200 mR (GE 5.2.5) 2-11

OSSI 92-402A-7A-."I;"..? EAI. Validation Scenarios, Rev. 0

" '.:"." -,"-., "..".t ". VA~'dation Exercise Scenario Checklist No.:

Plant: Ni'IP-2 Simulator: Q Table-Top: 0 Scenario 8 12 Scenario Desc The plant has entere..l a 24 LCO action statement at 0700 due to EG operability. At 1800 a p!",nt shutdown is initiated. At 0700 the following day, coolant temple:.ate,'e is still 220 'F while attempting to initiate shut down cooling {:.'=.. '.'=)

Shutdown coo':::< cannot bc established due to a failure of SDC suction valve.

Reactor temper: ..".;:"cannot vc 'educed to 212 F (Alert 7.2.3) 2-12

OSSI 92-402A-7A-i9l; ~? EAL Validation Scenarios, Rev. 0

.. VMdation Exercise Scenario Checklist No.:

Plant: Nvi.p-) Simulator: Q Table-Top: ~

Scenario 8 13 Scenario Descr '".i"..'.'"-"

A tanker carry.'.~< ~,:.=>>., i ~.ia gas overturns on the access road releasing ammonia gas. Th ".lu;.~e caresses onto the site, incapacitating numerous site personnel (UI; 8.... !

The gas then e:"!.,".rs the control room requiring the control room to be evacuated (Ale't ".3.3 8: 7.2.2).

Control of RPV iniection is not acheived after 30 minutes (SAR 7.2.4).

2-13

OSSI 92-402A-7A-Ni~!P2 EAL Validation Scenarios, Rev. 0

~ "".-'".-.""..' VMdation Exercise Scenario Checklist No.:

Plant: NA".P-,'? Simulator: Cl Table-Top: ~

Scenario 8 14 Scenario Descr ':":.'"-':

A severe storm cau,"."" a loss of all telephone systems offsite. No radios respond to att.e:: pL's to c.,!1 cffsite (UE 7.3.2).

Meteorologic" t""" '".;ipchart indicate sustained wind speeds of 95 mph I,

(Alert 8.4.O.

The roof is rir. a."~ o.'f oi the security building (Alert 8.4.7).

2-14

OSSI 92-402A-7h-I'i". - 2 EAL Valfdatfon ScenarloS, Rev. 0

."-. Validation Exercise Scenario Checklist No.:

Plant: NynP 2 Simulator: Q Table-Top: 8 Scenario ¹ 15 Scenario Descr'vt'.on(s):

A bomb threat is r " "". ".d. A search reveals a bomb in the reactor building at a remote shu!.-..'."... ~~ '.. -."! (!.'.-. 8.1.1).

An unauthor"ec':".c vi:!ual is rcco<nized to have scaled the Protected Area fence (Alert 8.! .. '.

The individu.".! i~ '..";.".'.;..d into the reactor building (SAE 8.1.3).

The bomb exp!od s destroying the remote shutdown panel (GE 8.1.4 or Alert 8.2.2).

Instead of t!- .:"."-'-.; '-.. 'Id'ng, a bomb explodes in the Administative Building (U 2-15

OSSI 92-402A-7A-N VP2 EAL Validation Scenarios, Rev. 0 f.';"."-. .. ".;.=+:? Validation Exercise Scenario Checklist No.:

Plant: Simulator: C3 Table-Top: S Scenario 8 16 Scenario Desc '""."!~!.):

A security tru"': ":",". ti",e d; sc! fuel storage tank (UE 8.3.1).

The collisinn 'n~'". ~ '",".'" in the oil tank (Alert 8.3.4).

The spilled ci! c"-"."..hes ~:.re and burns out of control for 30 minutes (UE 8.2.1).

2-16

OSSI 92-402A-7,'. EAL Valtdatfon Scenarios, Rev. 0

..- =."-.: ." Validation Exercise Scenario Checklist No.:

Plant: Simulator: Q Table-Top: ~

Scenario 8 17 Scenario Desc

'he control room operators notice ground motion and that the seismic activity ala".i~ .'."; ." ".",'", '. J.".. i!PP calls and confirms the earthquake (UE 8.4.1).

JAFNPP later ce'.! ~ and says the earthquake was of magnitude O.lg (Alert 8.4.5).

As a result of the earthquake the screenwell building is destroyed (Alert 8.4.7).

2-17

0 OSSI 92-402h.-7A- ~ 2 EAL Validation Scenarios, Rev. 0 Attachment Validation Exercise Scenario Checklist No.:

Plant: N)"t~ -2 Simulator: 0 Table-Top: 0 Gcenario ¹ 18 Scenario Descry ".:",; =',

):

A report to the ccntrcl room states that a tornado has been sighted inside the secur'ty f'e:..ce ',i' +.4.2).

An operate '".-."."'" -'="-.'";. c,".:".not get to the screenwell because of wind and debris ',"..:.',"=..:.

2-18

OSSI 92-402 ~='?.~ -"'~~? EAL Validation Scenarios, Rev. 0 Attach'nt 2 Validation Exercise Scenario Checklist No.:

Plant: Simulator; C3 Table-Top: ~

Scenario ¹ 19 Scenario Descry:,".lo.;.( ..':

Lake flooding has:e,.u'-'..cd in measured lake levels of 248 ft. (UE 8.4.4).

Frizzle ice f".."..:.".'~n !."-s caused t.he intake water level to drop to [Later] ft.

(Alert 8.4.B;.

2-19

OSSI 93-402A-10-NMP2 NMP-2 EAL VerIAcatfon & Validation Report, Rev. 0 Attachment 4 EAL Validation S Sheets 4-1

OSSI 92-402A-7 EAL ValidaUon Procedure, Rev. 0 Attachment 1 VaHdation Sunlmaxy Sheet i i I

4,,').

! ~ Plant +~ 2 IIRIIlC Validation Team Members:

I Ca(6tA.W 5 rst P I Jc k~W 1 ~

ecch(.(' ~ ~

Checklist No.:

~EAL N

~

scrim EAL Rev. No.:

nrl N Checklist No.:

~~N Jim ~T EAL Rev. No.: ~ j ~

I 1I

~

li l

I 8 ~I IiI. o 1t ii Ql et ~ i

~

g. I l.~,z g. Q 8 2- F i, s ~

B Q ~

gi R Q 't Valid. performed and comments recorded; Valida. performed and comments recorded:

Q. (ot>(Z X ~&I.~~

Validati i i Team Leader Date Validation am Leader Date i I

li Checkhst No.: t EAL Rev. No.: Ch cklist No.: ~e( EAL Rev. No.:

Is Iz IU ass I l

~

~

'7~ B Q ~

~@ Q S. II g Q Q Q Q Q Q

Valid. performed and comments recorded: Valida. performed and comments recorded:

I e.-

~

Validatio Team Leader Date Validation earn Leader Date Continuation Sheets Attached:

~ ' I l~

~ I

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 1 VaHdation Suxnmsry Sheet Plant:

XLGI1K t

'I l Validation Team Members: 5~~ e~4~ ) 3t I'

I Checldtst No.:

IL

~ EAL Rev. No.: Checklist No.: ~ EAL Rev. No.: t

'l

3l i

te 'I gl,

'?r< 3 lE Valid. performed and comments recorded: Valida. performed and comments recorded:

LobJaZ ~sf (V l Validatio earn.Leader Date Valtdation earn Leader Date 3 I

t R

Checklist No.:

l AL~N ~i

~ ~T EAL Rev. No.: M~ Checkhst No.:

MBa. Rm. XZ EAL Rev. No.:

s Q t I 5

g a Ii9. a I ~ 3'1 s o 0 6 Valid. performed and comments recorded: Valida. performed and comments recorded: .3 i gofv(q Lb ( ) (ted Validation earn Leader Date Validation earn Leader Date Continuation Sheets Attached:

1-2

OSSI 92-402A-7A-NMP2 EAL Validation Scenarios, Rev. 0, Attachment 1 VaHdation Summaxy Sheet Plant:

Validation Team Members:

Checklist No.: v(tI! t EAL Rev. No.: C Checklist No.: ~II 'AL Rev. No.:

s.

Q 7- I ~

Q Rl /

Q 7. ~.3 C IXI ~I

.'. c Q 9.>.5 Q Q lip Q KI Valid. performed and comments recorded: Valida. performed and comments recorded:

i,t>i~~ r.L,its Validation earn Leader Date VaBdation Team Leader Date Checklist No.: I EAL Rev. No.: ~ Checklist No.: AREALRev. No.:

S S = I! /ALLAN Jim T-T L)QQto Q N /0 7-r L. I!

Q

~

Q o cl Q Q r o Q Q Q Q / Q Q Valid. performed and comments recorded: Valida. performed and comments recorded:

C.. dt i t'( (g~ lo! ( /yg Validation earn Leader Date Validation T m Leader Date Continuation Sheets Attached:

1-3

OSSI 92-402A-7A-NMP2 EAL Validation Scenarios, Rev. 0 Attachment 1 VaHdation Smnmaxy Sheet t c) o ) <<t> Plant:

Validation Team Members:

Checklist No.:

IL

~/6

=

EAL Rev. No.:

e Checklist No.:

IL

~

=

EAL Rev. No.:

e

~

~

M; \

I' Q Q

. (~ o el c et I

I Q Q Cl liP Q Q CI ff! / Q Q tl )

Valid. performed and comments recorded: Valida. performed and comments recorded:

JokL'c2. ( l.c'~3 Validation I

earn Leader Date ValidaUon earn Leader Date Checklist No.:

I"" s EAL Rev. No.: ~ Checldist No.:

~L~N Qm T-T EAL Rev. No.:

8'3 9 Q gl Q Q ~

I

f. z.. I Q Q I Q g3 !l

.9s el Q Q f i

Q 8 Q Q $~

tj Valid. performed and comments recorded: Valida. performed and comments recorded:

I~~~f't Z Validation earn Leader Date UaBdation Team Leader Date )if jII Continuation Sheets Attached:

1-4

OSSI 93-402A-10-NMP2 NMP-2 EAL Verification & Validation Report, Rev. 0 Attachment 5 F.~.T Validation Exercise Checklists 5-1

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 ValMation Exercise Chechlist Date: 10 7 93 Checklist No.: 1 Yes No ~NA

1. When the need for classiQcation was initially recognized, were the EALs easily accessible to the user' ~ Q Q Comments: None.

2. Where plant conditions required emergency classification, did the format and layout of the EALs support easy and rapid classification recognition' Q ~ Q Comments: It would be hei ful to laminate EAL matrix and use erasable markers so ED can mark EALs reach and those about t b de lared.

3. Was classification of any conditions not requiring emergency classification avoided' ~ Q Q Comments: None.

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts' ~ Q Comments: EAL 3.1.1 leak rates ma be better located under RPV. That's where the 're located in Technical S ecifications. Consider movin EAL 3.1.2 to under RPV also.

3-1

OSSI S2-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise Chechlist Date: 10 7 93 Checklist No.: 1 Yes No ~NA

5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized? a o o Comments: None.

6. Did the EALs and required Emergency Plan procedures interface properly' Cl 0 ~

Comments: None.

7. After initial classification, did subsequent classification escalation follow a logical progression in the EM@ when appropriate? ~ 0 CI Comments: None.

8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate' ~ o o Comments: None.

3-2

OSSI 92-402A-7 EAI. ValidatIon Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecMist Date: 10 7 93 Checklist No.: 1 Yes No ~NA 9 Where plant conditions permitted reduction in the level of emergency'response, was down grading of classifications easily recognized using the EALs'? 5 Q Q Comments: None.

10. Are the EALs devoid of any misleading or incorrect information' ~ Q Q Comments: None.

11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures. etc, necessary to effectively evaluate the EALs'? ~ Q Q Comments: None.

12. Did the EALs have adequate detail to be effectively evaluated' Q ~ Q Comments: EAL 1.4.2: s eci channel 1 with 14

13. Are the EALs devoid of excessive detail'? ~ Q Q Comments: EAL 6.1.1: are desi nators for transformers readil understandable'? Is common terminolo used'?

3-3

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 7 93 Checklist No.: 1 Yes No ~NA

14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure'P ~ 0 CI Comments: None.

.15. Additional Comments: None.

3-4

OSSI 92-402A-7 EAL Valfdatfon Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecRHst Date: 10 7 93 Checklist No.: 2 Yes ~N ~NA

1. When the need for classification was initially recognized, were the EALs easily accessible to the user' ~ a o Comments: None.

2. Where plant conditions required emergency classiQcation, did the format and layout of the EALs support easy and rapid classification recognition'? 5 0 C3 Comments: None.

,3. Was classiQcation of any conditions ~no requiring emergency classification avoided' ~ 0 Cl Comments: None.

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts' ~ 0 0 Comments: None.

3-1

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 7 93 Checklist No.: 2 Yes No ~NA

5. Where plant conditions required emergency classiQcation, was the operating mode applicability of the EALs clearly recognized' S Q 0 Comments: None.

6. Did the EALs and required Emergency Plan procedures interface properly' a ~

Comments: None.

7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriate' ~ 0 C3 Comments: None.

8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate' ~ o o Comments: None.

9. Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the F~? 5 0 0 Comments: None.

3-2

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecMist Date: 10 7 93 Checklist No.: 2 Yes No ~NA

10. Are the EALs devoid of any misleading or incorrect information' ~ o o Comments: None.

11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.

necessary to effectively evaluate the EMs'? o r u Comments: EAL 1 3.1 1.3.2 1.3.3: when values are received from NMP should list them in Tables as 2 decimal laces scientioc notation e.

1.76E5 .

12. Did the EALs have adequate detail to be effectively evaluated' 0 ~ 0 Comments: EAL 3.5.1: Pathwa to the environment is nsidered to exist once as outbd PC isola ion because downstream valv s are not leak tested. Would want to consider other indications in the turbine buildin such as visual observation of steam leak ARMs CAMs tc.

3-3

OSSI 92-402A-7 EAI. ValfdatIon Procedure, Rer. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 7 93 Checklist No.: 2 Yes No ~NA

13. Are the EALs devoid of excessive detailV 0 ~ CI Comments: EAL 1.2.1 should read reater than or e ual to DRMS red.

14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure'P ~ a o Comments: None.

15. Additional Comments: None.

3-4

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise ChecMist Date; 10 7 93 Checklist No.: ~

Yes No ~NA

1. When the need for classification was initially recognized, were the EALs easily accessible to the user' Comments: None.

Where plant conditions required emergency classification, did the format and layout of the EALs support easy and rapid classification recognition' r o o Comments: None.

3. Was classification of any conditions not requiring emergency classification avoided' o ~ o Comments: EAL 3.2.1: If 1.68 si is reache due to loss of d ell coolin is Alert declared' A ears so but NESP-007 clearl focuses on loss of RCS and leak into PC. Consider usin wordin for restore and maintain instead of ust maintain.

Restore and maintain would allow the o erator to to reduce ressure. But ou ht to declare if reduction is due to d ell s ra o eration.

annot be maintained <1.68 si due to coolan

~leaks e.

3-1

OSSI 92-402A-7 EAI. Validation Procedure. Rev. 0 Attachment 3 Validation Exercise ChechHst Date: 10 7 93 Checldist No.: ~

Yes No ~NA

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classiAcation efforts' ~ Q Q Comments: None.

5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized? r Q Q Comments: None.

6. Did the EALs and required Emergency Plan procedures interface properly' Q Q ~

Comments: None.

7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriate' ~ Q Q Comments: None.

8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate? ~ Q Q Comments: None.

3-2

OSSI 92-402A-7 EAL Validation Procedure. Rev. 0 Attachment 3 ValMation Exercise ChechHst Date: 10 7 93 'hecklist No.: 3 Yes No ~NA

9. Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the EALs'? ~ Q Q Comments: None.

10. Are the EALs devoid of any misleading or incorrect information' ~ Q Q Comments: None.

11. Did the $ MLs adequately specify controls.

instrumentation, operator aides, procedures, etc, necessary to effectively evaluate the EALs'? ~ Q Q Comments: None.

12. Did the EALs have adequate detail to be effectively evaluated' ~ Q Q Comments: None.

13. Are the EALs devoid of excessive detail' ~ Q Q Comments: None.

3-3

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecMist Date 10 7 93 Checklist No.: 3 Yes No ~NA

14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure'P 5 0 Cl Comments: None.

15. Additional Comments: None.

3-4

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date 10 7 93 CheckBst No.: 4 Y~s ~N ~NA

1. When the need for classification was initially recognized, were the EALs easily accessible to the user'? ~ 0 0 Comments: None.

2. Where plant conditions required emergency classiQcation, did the format and layout of the EALs support easy and rapid classification recognition'? r 0 0 Comments: None.

3. Was classification of any conditions ~n requiring emergency classification avoided' 5 0 0 Comments: None.

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts'? ~ 0 0 Comments: None.

3-1

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise ChechHst Date: 10 7 93 Cimcldist No.: 4 Yes No ~NA

5. Where plant conditions required emergency classiAcation, was the operating mode applicability of the EALs clearly recognized' 5 Q Q Comments: None.

6. Did the EALs and required Emergency Plan procedures interface properly' Q Q ~

Comments: None.

7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriate' 0 Q Q Comments: None.

8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate'? ~ Q Q Comments: None.

9. Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the EALs'? ~ Q Q Comments: None.

3-2

OSSI 92-402A-7 EAL Valfdation Procedure, Rev. 0 Attachment 3 Validation Exercise Chechlist Date 10 7 93 Checklist No.: 4 Yes No ~NA

10. Are the EALs devoid of any misleading or incorrect information' r Q Q Comments: None.

11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.

necessary to effectively evaluate the EALs'? r Q Q Comments: None.

12. Did the EALs have adequate detail to be effectively evaluated' 8 Q Q Comments: None.

13. Are the EALs devoid of excessive detail' ~ Q Q Comments: None.

14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure' ~ Q Q Comments: None.

15. Additional Comments: None.

3-3

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise ChechHst Date: 10 7 93 Checklist No.: ~

Yes No ~NA

l. When the need for classiQcation was initially recognized, were the EALs easily accessible to the users ~ Q Q Comments: None.

2. Where plant conditions required emergency classiQcation, did the format and layout of the

'EALs support easy and rapid classification recognition'? ~ Q Q Comments: None.

3. Was classiAcation of any conditions ~no requiring emergency classification avoided V 8 Q Q Comments: None.

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts' a Q Q Comments: None.

3-1

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise ChechHst Date: 10 7 93 Checklist No.: 5 Yes ~N ~NA

5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognizedV ~ o o Comments: None.

6. Did the EALs and required Emergency Plan procedures interface properly'? a o r Comments: None.

7. After initial classification, did subsequent classiQcation escalation follow a logical progression in the EALs when appropriate' ~ Cl Cl Comments: None.

8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate' ~ 0 0 Comments: None.

9. Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the EALs'? W 0 0 Comments: None.

3-2

OSSI 92-402A-7 EAL Validation Procedure. Rev. 0 Attachment 3 Validation Exercise ChecRHst Date: 10 7 93 CheckBst No.: ~

Yes No ~NA

10. Are the EALs devoid of any misleading or incorrect information'? r o o Comments: None.

ll. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.

.necessary to effectively evaluate the EALs'? ~ 0 0 Comments: None.

12. Did the EALs have adequate detail to be effectively evaluated' r o a Comments: ~Non .

13. Are the EALs devoid of excessive detail' 0 ~ 0 Comments: EAL 6.1.3 should not include 2XTS-XSNl because it cannot be owered from an oA'site ower sources.

14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure' ~ C3 0 Comments: None.

3-3

OSSI 92-402A-7 EAL Validation Procedure. Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 7 93 Checklist No.: 5 Yes No ~NA

15. Additional Comments: None.

3-4

OSSI S2-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 ValMation Exercise Chechlist Date: 10 7 93 Checklist No.: ~

Yes ~N ~NA

1. When the need for classiQcation was initially recognized, were the EALs easily accessible to the user? ~ Q 0 Comments: None.

2. Where plant conditions required emergency classiQcation, did the format and layout of the EALs support easy and rapid classiQcation recognition' ~ 0 0 Comments: None.

3. '-Was classification of any conditions not requiring emergency classification avoided' r o a Comments: None.

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts? 0 ~ 0 Comments: Cate o 6 title "S stem" does not reall At the EALs contained in the Cate o e.

turbine failure control room evacuation .

3-1

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecMist 10 7 93 Checklist No.: 6 Y~s No ~NA

5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized' S 0 0 Comments: None.

6. Did the EALs and required Emergency Plan procedures interface properly' o o ~

Comments: None.

7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriate? ~ 0 Q Comments: None.

8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate' E 0 0 Comments: None.

9. Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the EALs'? ~ Cl 0 3-2

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaIMation Exercise ChecRIist Date 10 7 93 CheekBst No.: 6 Yes No ~NA Comments: Mode a licabili: is it necessa to down rade when the mode shifts from ower o s to cold shutdown'? No. This is a olic issue.

The chan e in mode is not a me hanism to escalate or descalate. NRC will rom l ate in a soon to be issued NUREG that a formal declaration need not be made if one Ands that in the ast a condition re uirin escalation xisted but subse uentl assed.

10. Are the EALs devoid of any misleading or incorrect informationV r o o Comments: None.

11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.

necessary to effectively evaluate the EALsV r o o Comments: None.

12. Did the EALs have adequate detail to be effectively evaluated' o r o Comments: Definitions of classiAcations are needed on EAL matrix at least to hei inte retation of Cate o 9.

13. Are the EALs devoid of excessive detail' ~ o o 3-3

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 7 93 Checklist No.: 6 Yes No ~NA Comments: None.

14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure' ~ o a Comments: None.

15. Additional Comments: Su est chan e Ebar to -E.

3-4

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise Chechlist 10 7 93 ChecMtet No.: 7 Y~e No ~NA

1. When the need for classification was initially recognized, were the EALs easily accessible to the user'? a o a Comments: None.

2. Where plant conditions required emergency

,.classification, did the format and layout of the EALs support easy and rapid classification recognition' ~ Cl 0 Comments: None.

3. Was classification of any conditions ~no requiring emergency classification avoided' 0 ~ CI Comments: If s ra and ressure come down and can therefore maintain less than scram se oint does Alert have o be declared'? Believe should have declared Alert as soon as EOP ste as ed that ermitted use of s ra s o ntrol err,~ur ..

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts'? ~ 0 0 Comments: None.

3-1

OSSI 92-402A-7 EAL Valtdation Procedure, Rev. 0 Attachment 3 Validation Exercise Checklist Date; 10 7 93 Checklist No.: 7 Yes No ~NA

5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized' ~ Q Q Comments: None.

6. Did the EALs and required Emergency Plan procedures interface properly' Q Q ~

Comments: None.

7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriate' r Q Q Comments: None.

8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate' ~ Q Q Comments: None.

3-2

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecMist Datir 10 7 93 Checklist No.: 7 Yes No ~NA

9. Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the EALs'? ~ Q Q Comments: None.

10. Are the EALs devoid of any misleading or incorrect information' ~ Q Q Comments; None.

11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.

necessary to effectively evaluate the EALs'? Q ~ Q Comments: EAL sa s d ell ressure but EOP sa s rima containment ressure.

12. Did the EALs have adequate detail to be effectively evaluated? Q ~ Q Comments: Perha s use d ell area radiation and ut RMS com onent numbers in EAL 1.1.1 1.1.2 1.1.3. Does rad monitor number need o be s ecified'? Possibl .

3-3

OSSI 92-402A-7 EAL Valtdatfon Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date 10 7 9 Checklist No.: 7 Yes No ~NA

13. Are the EALs devoid of excessive detail' r a Comments: Take s stem desi nator 43 out of Table 1.1.

14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure' ~ 0 Cl Comments: None.

15. Additional Comments: None.

34

OSSI 92-402A-7 EAL Valfdatfon Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date 10 7 93 Checldist No.: ~

Yes No ~NA

1. When the need for classification was initially recognized, were the EALs easily accessible to the user? r o o Comments: None.

2. Where plant conditions required emergency classification, did the format and layout of the EALs support easy and rapid classification recognition? r Q Q Comments: None.

3. Was classification of any conditions not requiring emergency classification avoided? r o o Comments: None.

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts? r Q Q r

Comments: None.

3-1

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise ChechHst Date: 10 7 93 Checklist No.: 8 Yes No ~NA

5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized? 5 Q Q Comments; None.

6. Did the EALs and required Emergency Plan procedures interface properly? Q Q ~

Comments: None.

7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriate?

Comments: None.

8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate? 5 Q Q Comments: None.

9. Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the EALs? ~ Q Q Comments: None.

3-2

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise ChecMist Date: 10 7 93 Checklist No.: 8 Yes No ~NA

10. Are the EALs devoid of any misleading or incorrect information' ~ Q Q Comments: None.

ll. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.

necessary to effectively evaluate the EALsV Q S Q Comments: EAL 1.4.4: Would not have arrived declaration of >8R hr unless access to area was re uired. That's correct.

12. Did the EALs have adequate detail to be effectively evaluated? ~ Q Q Comments: None.

13. Are the EALs devoid of excessive detail' S Q Q Commen ts: None.

14. Did the EAL identification scheme adequately support location of the EAL condition within the classification pro'cedure'? ~ Q Q Comments: None.

3-3

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise Checklist Date 10 7 93 Checklist No.: 8 Yes No ~NA

15. Additional Comments: None.

'-4

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise Checklist Date: 10 7 93 Checldist No.: 9 Yes No ~NA

1. When the need for classification was initially recognized, were the EALs easily accessible to the user'? S Q Q Comments: None.

2. Where plant conditions required emergency classification, did the format and layout of the EALs support easy and rapid classification recognition? 5 Q Q Comments: None.

3. Was classilication of any conditions ~no requiring emergency classification avoided? r Q Q Commen ts: None.

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts? ~ Q Q Commen ts: None.

3-1

OSSI 92-402A-7 EAI. Valldatfon Procedure, Rev. 0 Attachment 3 Validation Exercise ChecMist Date: 10 7 93 Checldist No.: ~

Yes No ~NA

5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized' r Q Q Commen ts: None.

6. Did the EALs and required Emergency Plan procedures interface properly?

Commcn ts: None.

7. After initial classification, did subsequent classificai.ion escalation follow a logical progression in the EALs when appropriate' S Q Q Comments: None.

8. Did the E.'ii.s support escalation of emergency classificai.ion when plant conditions indicated that escal:~t.ion was appropriate' S Q Q Commcn t,s: None.

9. Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the EAI.sY ~ Q Q Commcn t s: None.

3-2

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise Checklist DM: ~la IC! Checklist No.: 9 Yes No ~NA

10. Are the EAI.s devoid of any misleading or incorrect information? r Q o Comments: EAL 1.5.2: one SRO would have called Alert wo visual re ort other would have waited for visual.

11. Did-the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.

necessary to effectively evaluate the EALs? r o o Commcnis: None.

12. Did the EALs have adequate detail to be effectively evaluated? o r o Commcn I s: ARMs >100 times set oint which

~set oint nlcrtalarm ,or hi alarm'P Should be "hi".

This is also a roblem with other EALs e.. 5.0

~etc.. J!.:st cut "red" or" ellow" not even in garenUic,"-~'~,

13. Are the I.:KLs devoid of excessive detail? r o o Comments: None.

3-3

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecMist Date: ~10 7(93 Checklist No.: 9 Yes ~N ~NA

14. Did the I;Al. identification scheme adequately support loc;!.ion of the EAL condition within the classificaL:on procedures ~ 0 Cl Commcn!.s: None.

15. Additional Comments: None.

3-4

OSSI 92-402A-7 EAI. Validatfon Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 7 93 Checklist No.: 10 Y~s No ~NA

1. When the need for classification was initially recognized, were the EALs easily accessible to the user'? ~ o u Comments: None.

2. Where plant conditions required emergency classification, did the format and layout of the EALs support easy and rapid classification recognition' r o o Comments: None.

3. Was classification of any conditions not requiring emergency classification avoided' ~ a o Comments: None.

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts' ~ 0 0 Comments: None.

3-1

OSSI 92-402A-7 EAL ValldaUon Procedure, Rev. 0 Attachment 3 VaHdation Exercise Checklist Date: 10 7 93 Checklist No.: 10 Yes No ~NA

5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognizedV r Q Q Comments: None.

6. Did the EALs and required Emergency Plan procedures interface properl+ Q Q ~

Comments: None.

7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriate' r Q Q Comments: None.

8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate' r Q a Comments: None.

9. Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the EALs'? ~ Q Q Comments: None.

3-2

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 7 93 Checklist No.: 10 Yes No ~NA

10. Are the EALs devoid of any misleading or incorrect information' ~ Q Q Comments: None.

11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.

necessary to effectively evaluate the EALs'? Q S Q Comments: Stack and RB RW yen effluent monitor does not o to DRMS. Table 5.1.1 the first instruments will robabl not read low n u h o indi ate UE and should robabl be listed N A in the UE column.

12. Did the EALs have adequate detail to be effectively evaluated' Q Q Comments: None.

13. Are the EALs devoid of excessive detail' ~ Q Q Comments: None.

3-3

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise ChecMist Date: 10 793 Checklist No.: 11 Yes No ~NA

1. When the need for classification was initially recognized, were the EALs easily accessible to the user'? ~ Q Q Comments: None.

2. Where plant conditions required emergency classification, did the format and layout of the FRLs support easy and rapid classiQcation recognition' Q Q Comments: None.

3. Was classification of any conditions ~n requiring emergency classification avoided' Comments: None.

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts' ~ Q Q Comments: None, 3-1

OSSI 92-402A-7 EAI. Valfdatlon Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date 10 7 93 Checklist No.: 11 Yes No ~NA

5. Where plant conditions required emergency classiQcation, was the operating mode applicability of the EALs clearly recognized' ~ Q Q Comments: None.

6. Did the EALs and required Emergency Plan procedures interface properly' Comments: None.

7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriate' ~ Q Q Comments: None.

8, Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate' Q S Q Comments: Table 5.1.1 use 2x and 200x alarm set oint for UE and Alert res ectivel .

3-2

OSSI 92-402A-7 EAL Validatfon Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 7 93 Checklist No.: 11 Yes No ~NA

9. Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the EALs'? ~ o o Comments: None.

10. Are the EALs devoid of any misleading or incorrect information' Comments: EAL 1.4.2 is 10 R hr EAL 1.4.4 is 8 R hr. Should use one value and basis for it.

11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.

necessary to effectively evaluate the EALs'? ~ o o Comments: ~Nne.

12. Did the EALs have adequate detail to be effectively evaluated' r o o Comments: None.

13. Are the EALs devoid of excessive detail' o r o Comments: EAL 1.4.4 re lace one or more with an . Check the use of one or more throu hout the matrix.

3-3

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise Chechlist Date; 10 7 93 Checklist No.: 11 Y~ No ~NA

14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure' ~ o o Comments: None.

15. Additional Comments: None.

3-4

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecMist Date: 10 7 93 Checklist No.: 12 Yes No ~NA

1. When the need for classiAcation was initially recognized, were the EALs easily accessible to the user'? ~ Q Q Comments: None.

2. Where plant conditions required emergency classiQcation, did the format and layout of the EALs support easy and rapid classification recognition'? ~ Q Q Comments: None.

3. Was classification of any conditions not requiring emergency classification avoided? r Q Q Comments: None.

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts? ~ Q Q Comments: None.

3-1

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecRHst Date: 10 7 93 Checklist No,: 12 Y~s No ~NA

5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized' o ~ o Comments: EAL 7.1.1: Mode 3 is blocked out and this EAL is not re uired in this mode. Because of an event'that evolves in time left mode 1 mode 2 and when the LCO time runs out our in mode 3 wher EAL does no a l . P r NESP-007 shoul include mode 3.

6. Did the EALs and required Emergency Plan procedures interface properly' o o ~

Comments: None.

7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriate' r o o Comments: None.

8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate? ~ o o Comments: None.

3-2

OSSI 92-402A-7 EAI. Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 7 93 Checklist No.: 12 Yes ~N ~NA

9. Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the EALs'? r o Q Comments: None.

10. Are the EALs devoid of any misleading or incorrect information Q r Q Comments: EAL 7.2.3: Issue of bein in hot hutdown and an't to below 212 'F. Should onl be a UE. If in cold shutdown but can't sta there then it's an Alert. Should includ this reasonin in Trainin . EAL 7.2.3: Should be usin 200 'F instead of 212 'F.

11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.

necessary to effectively evaluate the EALs'? ~ Q Q Comments: None.

12. Did the EALs have adequate detail to be effectively evaluated' ~ Q Q Comments: None.

3-3

0 OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise Chechlist Date: 10 7 93 Checklist No.: 12 Yes No ~NA

13. Are the EALs devoid of excessive detail? ~ a o Comments: None.

14. Did the EAL identiQcation scheme adequately support location of the EAL condition within the classification procedure' ~ a a Comments: None.

15. Additional Comments: None.

3-4

OSSI 92-402A-7 EAI. Valfdation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecRHst Date; 10 7 93 Checklist No.: 13 Yes No ~NA

1. When the need for classiAcation was initially recognized, were the EALs easily accessible to the user? ~ a o Comments: None.

2. Where plant conditions required emergency classiQcation, did the format and layout of the EALs support easy and rapid classiAcation recognition' o o Comments: None.

3. Was classification of any conditions not requiring emergency classification avoided' ~ a o Comments: None.

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts'? ~ a a Comments: None.

3-1

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 7 9 Checklist No.: 13 Yes No ~NA

5. Where plant conditions required emergency classiQcation, was the operating mode applicability of the EALs clearly recognized' ~ a a Comments: None.

6. Did the EALs and required Emergency Plan procedures interface properl+

Comments: None.

7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriate' ~ 0 Cl Comments: None.

8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriateV r o o Comments: None.

9. Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the EALs'? ~ o o Comments: None.

3-2

OSSI 92-402A-7 EAI. Valfdation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 7 93 Checklist No.: 13 Y~s No ~NA

10. Are the EALs devoid of any misleading or incorrect information? Q ~ Q Comments: Table 8.2 and 8.3 should include "Control Buildin ". Table 8.2 should not include "Coolin Tower" and "Oil Stora e", Take Table 8.3 areas out of Table 8.2 t avoid confusion over iff r n in li t d areas. N th n han e EALs referencin Table 8.2 so that it now references both Table 8.2 and Table 8.3.

EAL 7.2.4 should state cannot be maintained in a of stablished. "Within 15 minute after control room evacuation ade uate core coolin cannot be maintained". No. Use wordin from existin EAL.

ll. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.

necessary to effectively evaluate the EALs? ~ Q Q Comments: None.

12. Did the EALs have adequate detail to be effectively evaluated? Q ~ Q Comments: EAL 8.3.5 should reference "vital" table.

3-3

OSSI 92-402A-7 EAL Valfdatlon Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date 10 7 93 Checklist No.: 13 Y~ ~N ~NA

13. Are the EALs devoid of excessive detail' ~ 0 Cl Comments: None.

14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure? 8 0 0 Comments: None.

15. Additional Comments: None.

3-4

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 7 93 Checklist No.: 14 Yes No ~NA

1. When the need for classification was initially recognized, were the EALs easily accessible to the user? ~ Q Q Comments: None.

2. Where plant conditions required emergency classiQcation, did the format and layout of the les support easy and rapid classification re cognition'? ~ Q Q Comments: None.

3. Was classiQcation of any conditions ~no requiring emergency classification avoided' ~ Q Q Comments: None.

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts? 0 Q Q Comments: None.

3-1

OSSI 92-402A-7 EAL Valfdatfon Procedure, Rev. 0 Attachment 3 VaHdation Exercise Chechlist Date: 10 7 93 Checklist No.: 14 Yes No ~NA

5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized' S Cl 0 Comments: None.

6. Did the EALs and required Emergency Plan procedures interface properly' C3 0 ~

Comments: None.

7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriate' S C3 0 Comments: None.

8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate'? ~ o o Comments: None.

9. Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the EALs'? r a o Comments: None.

3-2

OSSI 92-402A-7 EAI. Validation Procedure, Rev. 0 Attachment 3 ValMation Exercise ChecMist Date: 10 7 93 Checklist No.: 14 Yes No ~NA

10. Are the EALs devoid of any misleading or incorrect information' Q 8 Q Comments: EAL 8.4.6 should include word "sustained" to avoid transient s ike. Don' include a time limit. Met com uter r vides 1 minute avera e r in s.

11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.

necessary to effectively evaluate the EALs'? ~ Q Q Comments: ~Nne.

12. Did the EALs have adequate detail to be effectively evaluated' Q ~ Q Comments: Table 8.3 needs to have "Securi Buildin ".

13. Are the EALs devoid of excessive detail' r Q Q Comments: None.

3-3

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise Checklist Date: 10 7 Checklist No.: 14 Yes No ~NA

14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure' S Q Q Comments: None.

15. Additional Comments: None.

3-4

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise ChecRHst Date 10 7 93 Checklist No.: ~l Y~s No ~NA

1. When the need for classification was initially recognized, were the EALs easily accessible to the user' r Q Q Comments: ~Non .

2. Where plant conditions required emergency classification, did the format and layout of the EALs support easy and rapid classification recognition' Comments: None.

3. Was classification of any conditions not requiring emergency classification avoided' 0 Q Q Comments: None.

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts' S Q Q Comments: None.

3-1

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecMist Date 10 7 93 Checklist No.: 15 Y~e No ~NA

5. Where plant conditions required emergency classiQcation, was the operating mode applicability of the EALs clearly recognized' ~ o o Comments: None.

6. Did the EALs and required Emergency Plan procedures interface properly' Comments: None.

7. After initial classification, did subsequent classiQcation escalation follow a logical progression in the EALs when appropriate' Comments: ~Nne.

8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate' r o o Comments: None.

9 Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the EALs'? ~ o o Comments: None.

3-2

OSSI 92-402A-7 EAL Valfdation Procedure, Rev. 0 Attachment 3 VaHdation Exercise Checklist Date: 10 7 93 Checklist No.: 15 Yes ~N ~NA

10. Are the EALs devoid of any misleading or incorrect information' a ~ o Comments: EAL 8.1.3 needs to state that the areas of concern are those defined b Securi Seems that these areas would all re uir ke card to access. Need to make it clear at SAE level that it is Securi 's call. Is Securi ersonnel knowled cable enou h to know that it ertains to an intrusion. NMP2 will contact Securi to ensure their words are same meanin as EALs.

11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.

necessary to effectively evaluate the EALs'?

Comments: None.

12. Did the EALs have adequate detail to be effectively evaluated'? 0 ~ 0 Comments: EAL 8.1.4: Should be an AND no OR.

Chan e t "Loss of lant contr I from the control room" because an RSP takeover could transfer the control room ca abili to control the lant.

3-3

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 7 93 Checklist No.: 15 Yes ~N ~NA

13. Are the EALs devoid of excessive detail' 8 0 Cl Comments: ~Non .

14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure' ~ 0 0 Comments: None.

15. Additional Comments: None.

3-4

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date; 10 7 93 Checklist No.: 16 Yes No ~NA

1. When the need for classiQcation was initially recognized, were the &MLs easily accessible to the user' 5 Q Q Comments: None.

2. Where plant conditions required emergency classification, did the format and layout of the EALs support easy and rapid classiQcation recognition' ~ Q Q Comments: None.

3. Was classification of any conditions ~no requiring emergency classification avoided' S Q Q Comments: None.

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts' ~ Q Q Comments: None.

3-1

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise Chechlist Date; 10 7 93 Checklist No.: 16 Y~ ~N ~NA

5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized' ~ Q Q Comments: None.

6. Did the EALs and required Emergency Plan procedures interface properly' Q Q ~

Comments: None.

7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriate' ~ Q Q Comments: None.

8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate' ~ Q Q Comments: None.

9. Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the EALs'? ~ Q Q Comments: None.

3-2

OSSI 92-402A-7 EAL Validatfon Procedure, Rev. 0 Attachment 3 VaHdation Exercise Checklist Date: 10 7 93 Checklist No.: 16

~Ye No ~NA

10. Are the EALs devoid of any misleading or incorrect information' ~ o a Comments: ~Nne.

11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.

necessary to effectively evaluate the EALs'? ~ a o Comments: None.

12. Did the EALs have adequate detail to be effectively evaluated' ~ o a Comments: None.

13. Are the EALs devoid of excessive detail' ~ o a Comments: None.

14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure' a o a Comments: To A desk side b side w uld be based. Make them ortable. should be one board.

15. Additional Comments: None.

3-3

(

OSSI 92-402A-7 EAI. VaBdation Procedure, Rev. 0 Attachment 3 Validation Exercise ChechHst Date: 10 7 93 Checklist No.: 17 Yes ~N ~NA

1. When the need for classification was initially recognized, were the EALs easily accessible to the user'? ~ o o Comments: None.

2. Where plant conditions required emergency classification, did the format and layout of the EALs support easy and rapid classification recognition' S Cl C3 Comments: None.

3. Was classification of any conditions not requiring emergency classification avoided' 5 0 Cl Comments: None.

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts' ~ a a Comments: None.

5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized' 5 0 0 Comments: None.

3-1

(

OSSI 92-402A-7 EAL Valfdation Procedure, Rev. 0 Attachment 3 Validation Exercise ChechHst Date: 10 7 93 Checklist No.: 17 Yes No ~NA

6. Did the EALs and required Emergency Plan procedures interface properly?

Comments: ~Non .

7. After initial classification, did subsequent classiQcation escalation follow a logical progression in the EALs when appropriate? ~ CI 0 Comments: None.

8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate? S CI 0 Comments: None.

9. Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the EALs? r o a Comments: None.

10, Are the EALs devoid of any misleading or incorrect information? 5 0 Cl Comments: None.

3-2

OSSI 92-402A-7 EAL Valldatlon Procedure, Rev. 0 Attachment 3 VaHdation Exercise Checldist Date: 10 7 93 Checklist No.: 17 Yes ~N ~NA

11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.

necessary to effectively evaluate the EALs'? S CI 0 Comments: None.

12. Did the EALs have adequate detail to be effectively evaluated' ~ 0 0 Comments: None.

13. Are the EALs devoid of excessive detail'? S 0 0 Comments: None.

14. Did the EAL identiQcation scheme adequately support location of the EAL condition within the classification procedure' ~ o a Comments: None.

15. Additional Comments: None.

3-3

OSSI 92-402A-7 EAL Validatfon Procedure, Rev. 0 Attachment 3 Va1Mation Exercise ChechHst Date: 10 7 93 Checklist No.: ~l Yes ~N ~NA

1. When the need for classification was initially recognized, were the EALs easily accessible to the user'?

Comments: None.

2. Where plant conditions required emergency classification, did the format and layout of the EALs support easy and rapid classification recognition'? ~ Q Q Comments: None.

3. Was classification of any conditions not requiring emergency classification avoided? ~ Q Q Comments: ~Non .

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts'? ~ Q Q Comments: None.

5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized? r Q Q Comments: None.

3-1

L OSSI 92-402A-7 EAL Valldatfon Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 7 93 Checklist No.: 18 Yes No ~NA

6. Did the EALs and required Emergency Plan procedures interface properly' Q 5 Comments: None.

7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriateV r Q Q Comments: None.

8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate' ~ Q Q Comments: None.

9. Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the EALs'? r Q Q Comments: None.

10. Are the EALs devoid of any misleading or incorrect information? ~ Q Q Comments: None.

3-2

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 7 93 Checklist No.: 18 Yes No ~NA

11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.

necessary to effectively evaluate the EALs'? ~ Q Q Comments: None.

12. Did the EALs have adequate detail to be effectively evaluated? ~ Q Q Comments: None.

13. Are the EALs devoid of excessive detail? ~ Q Q Comments: None.

14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure' ~ Q Q Comments: None.

15. Additional Comments: None.

3-3

OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise ChecRHst Date: 10 7 93 Checklist No.: 19 Yes No ~NA 1 ~ When the need for classification was initially recognized, were the EALs easily accessible to the user' r o o Comments: None.

2. Where plant conditions required emergency classiQcation, did the format and layout of the EALs support easy and rapid classification recognition' r o o Comments: None.

3. Was classiQcation of any conditions not requiring emergency classification avoided' r o o Comments: None.

4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts' ~ o o Comments: None.

3-1

OSSI 92-402A-7 EAI. Valfdation Procedure, Rev. 0 Attachment 3 Validation Exercise ChechHst Date: 10 7 93 Checklist No.: 19 Y~s No ~NA

5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized' ~ a a Comments: None.

6. Did the EALs and required Emergency Plan procedures interface properly' 0 0 ~

Comments: None.

7. After initial classification, did subsequent classiQcation escalation follow a logical progression in the EALs when appropriate' ~ o o Comments: None.

8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate' Comments: None.

9. Where plant conditions permitted reduction in the level of emergency response, was down grading of classifications easily recognized using the EALs'? r o o Comments: None.

3-2

OSSI 92-402A-7 EAL Valfdatlon Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 7 93 Checklist No.: '19 Yes No ~NA

10. Are the EALs devoid of any misleading or incorrect information'? Q ~ 0 Comments: Low lake level is 233.1 ft for service water um o erabilit based on USAR level for intake. USAR minimum lake level revents ever reachin minimum intake level. Therefore could o to lak level instead of low intake level'? No.

could have an intake roblem causin low level when lake levels are ok. Can it be measured' Check in simulator. It is on instrument LI-502 on

~ I 6D1.

11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.

necessary to effectively evaluate the EALs'? ~ o a Comments: None

12. Did the EALs have adequate detail to be effectively evaluated' ~ 0 0 Comments: None.

13. Are the EALs devoid of excessive detail? ~ 0 0 Comments: None.

3-3

OSSI 92-402A-7 EAI. Valldatlon Procedure, Rev. 0 Attachment 3 Validation Exercise Checldist Date: 10 7 93 Checklist No.: 19 Yes No ~NA

14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure? r o a Comments: None.

15. Additional Comments: None.

3-4

OSSI 93-402A-10-NMP2 NMP-2 EAL VeriAcatlon & Validation Report, Rev. 0 Attachment 6 FAT. Validation Comment Database 6-1

~ ~ ~ ~ re ~

a

~ ~ a ~ ~

Record No. 18 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-2 Q JAF 0 IP-3 0 NMP-2 C3 Generic BWR H General Impact 0 IP-2 CI NMP-1 Cj Ginna 0 Generic PWR 0 NUMARC-007 0 Procedure I] Verification Q Training Cj Hardware CJ EAL C] Technical Bases CI Validation Cl Deviation C3 None Cat. N/A I C¹

• • No. ** Emer. Class.

Comment General discussion:

mode bar - what is it? what does white vs grey mean?

Try increasing color intensity from UE to GE as means of highlighting changes in classification.

Resolution Mode bar definition is given in the legend at the bottom of the EAL chart.

Increasing color intensity would overuse colors and detract from their effectiveness for separating

--=~EAL categories.

tatus 0 Open 0 Resolved/Awaiting Disposition Oo Closed Record No. 19 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-2 CI JAF Q IP-3 0 NMP-2 C3 Generic BWR 8 General Impact Cl IP-2 0 NMP-1 Q Ginna 0 Generic PWR Q NUMARC-007 C3 Procedure 0 Verification Cl Training Cl Hardware 0 EAL 0 Technical Bases Q Validation Cl Deviation 0 None Cat. N/A I C¹

• • No. ** Emer. Class. '*

"Comment Scenario 1, question 2: It would be helpful to laminate EAL matrix and use erasable markers so ED can mark EALs reached and those about to be declared.

Resolution Agree.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed

0, Record No. 20 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-2 CJ JAF OIP-3 8NMP-2 OGeneric BWR QGeneral Impact PIP-2 PNMP-1 QGinna QGenericPWR C3 NUMARC-007 CI Procedure 0 Verification CJTraining Cl Hardware C3 EAL C3 Technical Bases C3 Validation Q Deviation Q None Cat. N/A I c¹

• • No. ** Emer. Class.

Comment Scenario 1, question 12: EAL 1.4.2: specify channel 1 with 14 A/B.

Resolution Added channel 1 as suggested.

status 0 Open 0 Resolved/Awaiting Disposition OI Closed Record No. 21 Date 9/2Q/93 Name M. C. DauS Originating Site Site Applicability NMP-2 CI JAF Q IP-3 CINMP-2 HGeneric BWR QGeneral Impact P IP-2 QNMP-1 PGinna QGeneric PWR 0 NUMARC-007 C3 Procedure 0 Verification 0 Training D Hardware Q EAL Q Technical Bases Cl Validation Q Deviation Q None cat. N/A I C¹

• • No. ** Emer. Class.

Comment Scenario 1, question 4: EAL 3.1.1 leak rates may be better located under RPV. That's where they'e located in Technical Specifications. Consider moving EAL 3.1.2 to under RPV also.

Resolution Moved EALs as suggested.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed

Record No. 22 Date 9/20/93 Name M. C. DauS Orlglnatlng Site Site Appllcablllty NMP-2 Q JAF Q IP-3 H NMP-2 Q Generic BWR Q General Impact Q IP-2 Q NMP-1 Q Ginna Q Generic PWR Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware Q EAL Q Technical Bases Q Validation Q Deviation Q None ca t. N/A I C¹ ** No. *'mer. Class.

Comment Scenario 1, question 13: EAL 6.1.1: are designators for transformers readily understandable? Is common terminology used?

Resolution Yes.

status 0 Open 0 Resolved/Awaiting Disposition 0 Closed Record No. 23 Date 9/20/93 Name M. C. Daus rlginatlng Site Site Appllcablllty NMP-2 Q JAF Q IP-3 g NMP-2 Q Generic BWR Q General Impact Q IP-2 Q NMP-1 Q Ginna Q Generic PWR Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware Q EAL Q Technical Bases Q Validation Q Deviation Q None ca t. N/A IC¹ ** No. ** Emer. Class.

Comment Scenario 2, question 11: EAL 1.3.1, 1.3.2, 1.3.3: when values are received from NMP, should list them in Tables as 2 decimal places, scientific notation (e. g., 1.76E5).

Resolution Agree.

status Oe Open 0 Resolved/Awaiting Disposition 0 Closed

Record No. 24 Date 9/2Q/93 Name M. C. DauS Orlglnatlng Site Site Appllcablllty NMP-2 Q JAF Q IP-3 C3 NMP-2 Q Generic BWR H General mpact 0 IP-2 Q NMP-1 C] Ginna Q Generic PWR Cl NUMARC-007 Cl Procedure 0 Verification Cl Training Cl Hardware Q EAL Q Technical Bases CJ Validation C] Deviation Q None I C¹

• • No. **

ca t. N/A Emer. Class.

Comment Scenario 2, question 12: EAL 3.5.1: Pathway to the environment is considered to exist once past outbd PC isolation because downstream valves are not leak tested. Would want to consider other indications in the turbine building such as visual observation of steam leak, ARMs, CAMs, etc.

Resolution CKW Agree. Pathway to environment should not be assumed because downstream valves are not leak tested. Other indications should be used to support decision that a pathway to environment exists.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed Record No. 25 Date 9/2Q/93 Name M. C. Daus rlginatlng Site Site Applicability MP-2 0 JAF C] IP-3 IRI NMP-2 Q Generic BWR C3 General C3 IP-2 Q NMP-1 Q Ginna Q Generic PWR Impact 0 NUMARC-007 Q Procedure 0 Verification CI Training Q Hardware 0 EAL Cl Technical Bases Q Validation 0 Deviation 0 None cat. N/A I C¹

• • No. *'mer. Class.

Comment Scenario 2, question 13: EAL 1.2.1 should read greater than or equal to DRMS (red).

Resolution Agree. Changed EAL to "> DRMS red".

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed

Record No. 26 Date 9/2Q/93 Name M. C. DauS Orlglnatlng Site Site Appllcabillty NMP-2 Q JAF Q IP-3 8 NMP-2 C3 Generic BWR 0 General mpact CI IP-2 C] NMP-1 0 Ginna 0 Generic PWR 0 NUMARC-007 0 Procedure Cl Verification Cl Training 0 Hardware Q EAL Q Technical Bases 0 Validation C3 Deviation C3 None cat. N/A I C¹ ** No. ** Emer. Class.

Comment Scenario 5, question 13: EAL 6.1.3 should not include 2XTS-XSN1 because it cannot be powered from any offsite power sources.

Resolution Deleted 2XTS-XSN1.

status 0 Open 0 Resolved/Awaiting Disposition OI Closed Record No. 27 Date 9/2Q/93 rig lnatlng Site Site Applicability NMP-2 0 JAF Q IP-3 0 NMP-2 Q Generic BWR H General Impact 0 IP-2 Q NMP-1 0 Ginna 0 Generic PWR Cl NUMARC-007 C3 Procedure C1 Verification Cl Training 0 Hardware 0 EAL 0 Technical Bases 0 Validation C] Deviation Cl None cat. N/A I C¹

• • No. ** Emer. Class. **

Comment Scenario 6, question 4: Category 7.2 title "System" does not really fit the EALs contained in the Category (e. g., turbine failure, control room evacuation).

Resolution 4

CKW (need to do this for NMP. done for JAF.) Changed title to "System Failures or Control Room Evacuation".

status 0 Open 0 Resolved/Awaiting Disposition 0 Closed

Record No. 28 oate 9/20/93 Name M. C. DauS Orlglnatlng Site Site Appllcablllty NMP-2 0 JAF I7 IP-3 0 NMP-2 C3Generic BWR g General mpact P IP-2 Q NMP-1 P Ginna P Generic PWR 0 NUMARC-007 0 Procedure 0 Verification CI Training 0 Hardware 0 EAL Q Technical Bases CI Validation Q Deviation 0 None Ca t. N/A I C¹ ** No. ** Emer. Class.

Comment Scenario 6, question 12: Definitions of classifications are needed on EAL matrix at least to help interpretation of Category 9.

Resolution Category 9 embodies the definitions for each classification.

status 0 Open 0 Resolved/Awaiting Disposition QI Closed Record No. 29 Date 9/20/93 Name M. C. Daus rlglnatlng Site Site Appllcablllty MP-2 P JAF Q IP-3 P NMP-2 Q Generic BWR 8 General Impact 0 IP-2 0 NMP-1 Q Ginna 0 Generic PWR Cl NUMARC-007 Cl Procedure Cl Verification OTraining CI Hardware Q EAL 0 Technical Bases CI Validation D Deviation Cl None ca t. N/A I C¹

• • No. 'mer. Class.

Comment Scenario 6, question 9: Mode applicability: is it necessary to downgrade when the mode shifts from power ops to cold shutdown?

Resolution No. This is a policy issue. The change in mode is not a mechanism to escalate or descalate. NRC will promolgate in a soon to be issued NUREG that a formal declaration need not be made if one finds that in the past a condition requiring escalation existed but subsequently passed.

status 0 Open 0 Resolved/Awaiting Disposition 0 Closed

Record No. 30 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability NMP-2 P JAF P IP-3 P NMP-2 PGeneric BWR HGeneral Impact P IP-2 P NMP-1 P Ginna P Generic PWR p NUMARC-007 p Procedure p Verification p Training p Hardware P EAL p Technical Bases p Validation p Deviation p None cat. N/A I C¹

• • No. ** **

Emer. Class.

Comment Scenario 6, question 15: Suggest change Ebar to -E.

Resolution Ebar is used by IP2/3 and will be continued here for consistency among EALs.

status O Open O Resolved/Awaiting Disposition OI Closed

.. Record No. 31 Date 9/20/93 Name M. C. Daus riglnatlng Site Site Applicability NMP-2 PJAF PIP-3 8NMP-2 PGenericBWR PGeneral Impact P IP-2 P NMP-1 P Ginna P Generic PWR p NUMARC-007 p Procedure p Verification p Training p Hardware P EAL p Technical Bases p Validation p Deviation p None Cat. N/A I C¹

• • No.

• Emer. Class. '*

Comment Scenario 7, question 13: Take system designator 43 out of Table 1.1.

Resolution Changed as suggested.

sta~us OOpen O Resolved/Awaiting Disposition Oe Closed

Record No. 32 Date 9/20/93 Name M. C. Daus Orlglnatlng Site Site Appllcablllty NMP-2 0 JAF Q IP-3 0 NMP-2 H Generic BWR C3 General mpact C3 IP-2 Q NMP-1 Q Ginna 0 Generic PWR CI NUMARC-007 CI Procedure C3 Verification 0 Training Cl Hardware 0 EAL Cl Technical Bases Q Validation C] Deviation 0 None cat. N/A I C¹ ** No. ** Emer. Class.

Comment Scenario 7, question 3: If spray and pressure comes down and can therefore maintain less than scram setpoint, does Alert have to be declared? Believe should have declared Alert as soon as EOP step passed that permitted use of sprays to control pressure.

Resolution See definition of "cannot be maintained". As long as pressure is below the scram setpoint and pressure can be maintained below the scram setpoint. This EAL does not have to be declared.

Note that "... due to coolant leakage" has been added to this EAL.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed Record No. 33 Date 9/20/93 rlglnatlng Site Site Appllcablllty MP-2 j3 JAF Q IP-3 IHI NMP-2 Q Generic BWR C] General Impact Q IP-2 P NMP-1 P Ginna 0 Generic PWR CI NUMARC-007 Q Procedure Cl Verification CI Training C3 Hardware 0 EAL C3 Technical Bases 0 Validation CJ Deviation 0 None cat. N/A I C¹

• • No

• • Emer. Class. **

Comment Scenario 7, question 12: Perhaps use drywell area radiation and put RMS component numbers in EAL 1.3.1, 1.3.2, 1.3.3. Does rad monitor number need to be specified? Possibly.

Resolution Added word "area".

RMS/rad monitor numbers are not needed.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed

Record No. 34 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability NMP-2 QJAF OIP-3 8 NMP-2 OGeneric BWR ClGeneral C]IP-2 CjNMP-1 QGinna C]Generic PWR Impact 0 NUMARC-007 Q Procedure 0 Verification 0 Training D Hardware D EAL C1Technical Bases C7 Validation C3Deviation Cj None cat. N/A I C¹

• No. Emer. Class.

Comment Scenario 7, question 11: EAL says drywell pressure, but EOP says primary containment pressure.

Resolution Changed to "primary containment pressure" to agree with EOP.

status 0 Open 0 Resolved/Awaiting Disposition 0 Closed Record No. 35 Date 9/20/93 rlglnatlng Site Site Applicability NMP-2 0JAF C3IP-3 ONMP-2 OGeneric BWR HGeneral Impact HIP-2 C7NMP-1 C]Ginna C3Generic PWR CI NUMARC-007 0 Procedure C3 Verification Cl Training 0 Hardware Cj EAL C3 Technical Bases 0 Validation 0 Deviation C3 None Cat. N/A I C¹

• No. ** Emer. Class.

Comment Scenario 8, question 11: EAL 1.4.4: Would not have arrived at declaration of >8R/hr unless access to area was required.

Resolution That's correct.

status OOpen 0 Resolved/Awaiting Disposition 0 Closed

Record No. 36 Date 9/20/93 Originating Site Site Applicability NMP-2 0 JAF Q IP-3 8 NMP-2 Q Generic BWR C] General Impact 0 IP-2 Cl NMP-1 Cl Ginna Q Generic PWR C3 NUMARC-007 CI Procedure Q Verification C1 Training CI Hardware C3 EAL Cl Technical Bases Q Validation D Deviation 0 None cat. N/A I C¹

• • No. ** **

Emer. Class.

Comment Scenario 9, question 12: ARMs >100 times setpoint, which setpoint alert alarm or hi alarm?

Should be "hi". This is also a problem with other EALs (e. g., 5.0, etc.). Just put "red" or "yellow" not even in parentheses.

Resolution Changed to use "DRMS red".

status 0 Open 0 Resolved/Awaiting Disposition 0+ Closed Record No. 37 Date 9/20/93 Name M. C. Daus rlglnatlng Site Site Applicability NMP-2 'Q JAF Q IP-3 C] NMP-2 C3 Generic BWR g General Impact Q IP-2 P NMP-1 C3 Ginna D Generic PWR C3 0

NUMARC-007 C3 Procedure Q Verification IITraining Q Hardware EAL Q Technical Bases 0 Validation C3 Deviation 0 None cat. N/A I C¹

• • No. ** Emer. Class. **

Comment Scenario 9, question 10: EAL 1.5.2: one SRO would have called Alert wo/visual report, other would have waited for visual.

Resolution Declaration is required if bundle was not specifically seen uncovered. This will be covered in training.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed

Record No. 38 Date 9/20/93 Name M. C. Daus Orlglnatlng Site Site Appllcablllty NMP-2 Cl JAF Q IP-3 8 NMP-2 0 Generic BWR 0 General Impact P IP-2 Q NMP-1 Q Ginna Q Generic PWR 0 NUMARC-007 0 Procedure 0 Verification 0 Training C3 Hardware 0 EAL 0 Technical Bases C3 Validation 0 Deviation CI None cat. N/A I C¹ ** No. ** Emer. Class.

Comment Scenario 10, question 11: Stack and RB/RW vent effluent monitor does not go to DRMS. Table 5.1.1 the first two instruments will probably not read low enough to indicate UE and should probably be listed as N/A in the UE column.

Resolution Changed first two entries in Table 5.1.1 to read:

Radwaste/Reactor Bldg Vent Effluent 2 x alarm Main Stack Effluent 2 x alarm

'tatus 0 Open 0 Resolved/Awaiting Disposition OI Closed Record No. 39 Date 9/20/93 Name M. C. DauS Orlglnatlng Site Site Applicability NMP-2 C3 JAF [3 IP-3 Cl NMP-2 8 Generic BWR 0 General Impact 0 IP-2 0 NMP-1 Q Ginna Q Generic PWR 0 NUMARC-007 Q Procedure 0 Verification Q Training Q Hardware 0 EAL CI Technical Bases Cl Validation Q Deviation Cl None cat. N/A I C¹

• • No. *" Emer. Class.

Comment Scenario 11, question 10: EAL 1.4.2 is 10 R/hr, EAL 1.4.4 is 8 R/hr. Should use one value and basis for it. Maybe should change tech basis description to talk about getting staffed up to be able to do actions required with the MSO rad value.

Resolution Agree, but the basis for 8 R/hr has been established by NMPC calculation. Difference from EOP MSO value should be resolved in EOP program.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed

Record No. 40 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability NMP-2 0 JAF Cl IP-3 Q NMP-2 g Generic BWR Q General Q IP-2 Q NMP-1 P Ginna P Generic PWR Impact 0 NUMARC-007 0 Procedure 0 Verification C] Training Q Hardware Q EAL Cl Technical Bases 0 Validation 0 Deviation C] None I C¹

• • No. ** **

cat. N/A Emer. Class.

Comment Scenario 11, question 8: Table 5.1.1 use 2x and 200x alarm setpoint for UE and Alert, respectively.

Resolution Changed as suggested.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed Record No. 41 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability NMP-2 0 JAF 0 IP-3 0 NMP-2 0 Generic BWR 8 General Impact D IP-2 Q NMP-1 C3 Ginna 0 Generic PWR C3 NUMARC-007 0 Procedure C3 Verification Q Training Q Hardware 0 EAL Q Technical Bases C3 Validation 0 Deviation 0 None Cat. N/A I C¹ '* No. ** Emer. Class.

Comment Scenario 11, question 13: EAL 1.4.4 replace one or more with any. Check the use of one or more throughout the matrix.

Resolution Changed as suggested.

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed

Record No. 42 Date 9/20/93 Name M. C. Daus Orlglnatlng Site Site Appllcablllty

.NMP-2 C3 JAF Q IP-3 HNMP-2 OGeneric BWR g General Impact 0 IP-2 Q NMP-1 Q Ginna Q Generic PWR CI NUMARC-007 C]Procedure 0 Verification HTraining 0 Hardware 0 EAL OTechnical Bases C3Validation ClDeviation QNone ca t. N/A I C¹ ** No.

• Emer. Class. **

Comment Scenario 12, question 10: EAL 7.2.3: Issue of being in hot shutdown and can't get to below 212 'F.

Should only be a UE. If in cold shutdown but can't stay there, then it's an Alert. Should include this reasoning in Training. EAL 7.2.3: Should be using 200 'F instead of 212 'F.

Resolution Changed 212 to 200 'F.

status O Open O Resolved/Awaiting Disposition O>> Closed Record No. 43 Date 9/20/93 Orlglnatlng Site Site Appllcablllty NMP-2 OJAF CIIP-3 DNMP-2 OGeneric BWR 8 General OIP-2 QNMP-1 C3Ginna C3Generic PWR Impact CJ NUMARC-007 Cl Procedure Q Verification 0 Training Q Hardware Q EAL C3Technical Bases 0 Validation P Deviation 0 None cat. N/A I C¹

• • No. ** Emer. Class.

Comment Scenario 12, question 5: EAL 7.1.1: Mode 3 is blocked out and this EAL is not required in this mode. Because of an event that evolves in time, left mode 1, mode 2 and when the LCO time runs out your in mode 3 where EAL does not apply. Per NESP-007 should include mode 3.

Resolution Included mode 3.

senatus OOpen OResolved/Awaiting Disposition OiClosed

Record No. 44 Date 9/20/93 Name M. C. Daus Orlglnatlng Site Site Appllcablllty NMP-2 C3 JAF C] IP-3 Q NMP-2 e0impact Q IP-2 NUMARC-007 CI Procedure Q NMP-1 Q Ginna C] Generic BWR 0 Generic PWR C3 Verification 0 Training IHL General 0 Hardware 0 EAL Q Technical Bases Q Validation CI Deviation Q None Cat. N/A I C¹

• • No.

• • Emer. Class.

Comment Scenario 13, question 12: EAL 8.3.5 should reference "vital" table.

Resolution Added "vital" to table. Added to EAL "... needed for safe plant operation".

status 0 Open 0 Resolved/Awaiting Disposition Oe Closed Record No. 45 Date 9/20/93 rlglnating Site Site Appllcablllty NMP-2 Q JAF C1 IP-3 0 NMP-2 CJ Generic BWR g General impact 0 IP-2 Cl NMP-1 Q Ginna 0 Generic PWR C3 NUMARC-007 C3 Procedure C3 Verification Q Training Q Hardware

[7 EAL Q Technical Bases CI Validation Q Deviation Q None cat. N/A l C¹

• • No. ** Emer. Class.

Comment Scenario 13, question 10: Table 8.2 and 8.3 should include "Control Building". Table 8.2 should not include "Cooling Tower" and "Oil Storage". Take Table 8.3 areas out of Table 8.2 to avoid confusion over differences in listed areas. Need to then change EALs referencing Table 8.2 so that it now references both Table 8.2 and Table 8.3.

Resolutfon Changed as suggested.

status 0 Open 0 Resolved/Awaiting Disposition Qe Closed

Record No. 46 Date 9/2Q/93 Name M. C. Daus Orlglnatlng Site Site Applicability NMP-2 QJAF QIP-3 QNMP-2 QGenericBWR HGeneral QIP-2 QNMP-1 QGinna QGeneric PWR Impact Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware Q EAL Q Technical Bases Q Validation Q Deviation Q None cat. N/A I C¹

• • No. ** Emer. Class.

Comment Scenario 13, question 10: EAL 7.2.4 should state cannot be maintained instead of established.

"Within 15 minutes after control room evacuation, adequate core cooling cannot be maintained".

Resolution No. Used wording from existing EAL.

status OOpen O Resolved/Awaiting Disposition 0 Closed Record No. 47 Date 9/2Q/93 Name M. C. Daus rlglnatlng Site Site Applicability NMP-2 Q JAF Q IP-3 QNMP-2 QGeneric BWR HGeneral Impact Q IP-2 Q NMP-1 Q Ginna Q Generic PWR Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware Q EAL Q Technical Bases Q Validation Q Deviation Q None Cat. N/A I C¹ ** No. *'mer. Class. **

Comment Scenario 14, question 10: EAL 8.4.6 should include word "sustained" to avoid transient spike.

Don't include a time limit. Met computer provides 15 minute average readings.

Resolution Added word "sustained".

sta~us OOpen O Resolved/Awaiting Disposition Oe Closed

Record No. 48 Date 9/20/93 Name M. C. DauS Originating Site Site Applicability NMP-2 0 JAF C3 lP-3 II NMP-2 C3 Generic BWR p General mpact D IP-2 Q NMP-1 Q Ginna 0 Generic PWR CJ NUMARC-007 CI Procedure 0 Verification CI Training 0 Hardware Q EAL Q Technical Bases P Validation C3 Deviation P None cat. N/A I C¹

• • No. ** Emer. Class.

Comment Scenario 14, question 12: Table 8.3 needs to have "Security Building".

Resolution Added security building to table.

status 0 Open 0 Resolved/Awaiting Disposition Closed Record No. 49 Date 9/20/93 rlglnatlng Site Site Appllcablllty NMP-2 Impact 0 JAF 0 IP-3 CI NMP-2 0 Generic BWR p IP-2 p NMP-1 p Ginna p Generic PWR I General C3 NUMARC-007 C3 Procedure C] Verification P Training Q Hardware 0 EAL 0 Technical Bases 0 Validation 0 Deviation 0 None cat. N/A I C¹

• • No. ** Emer. Class. **

Comment Scenario 15, question 10: EAL 8.1.3 needs to state that the areas of concern are those defined by Security. Seems that these areas would all require keycard to access. Need to make it clear at SAE level that it is Security's call. ls Security personnel knowledgeable enough to know that it pertains to an intrusion. NMP2 will contact Security to ensure their words are same meaning as EALs.

Resolution 11/9 EAL clearly states that the concern is with a plant security vital area. Awaiting NMP confirmation that security words mean same as EAL intent.

status 0 Open 0 Resolved/Awaiting Disposition 0 Closed

~ Record No. 50 Date 9/20/93 Name M. C. Daus Orlglnatlng Site Site Applicability NMP-2 C3 JAF C3 IP-3 C3 NMP-2 C3 Generic BWR H General C3 IP-2 C3 NMP-1 C] Ginna C3 Generic PWR mpact C3 NUMARC-007 C3 Procedure C3 Verification C3 Training C3 Hardware C3 EAL C3 Technical Bases C3 Validation C3 Deviation C3 None Cat. N/A I C¹

• • No. ** Emer. Class.

Comment Scenario 15, question 12: EAL 8.1.4: Should be an AND no OR. Change to "Loss of plant control from the control room" because an RSP takeover could transfer the control room capability to control the plant.

Resolution Changed as suggested.

status 0 Open 0 Resolved/Awaiting Disposition 0I Closed Record No. 5I Date 9/20/93 Name M. C. DauS rlglnatlng Site Site Appllcabllity NMP-2 C3 JAF C3 IP-3 Q NMP-2 CI Generic BWR g General Impact C3 IP-2 C3 NMP-1 Q Ginna 0 Generic PWR C3 NUMARC-007 C3 Procedure C3 Verification C3 Training C3 Hardware C3 EAL C3 Technical Bases C3 Validation C3 Deviation C3 None Cat. N/A I C¹

• • No. ** Emer. Class.

Comment Scenario 16, question 14: To fit desk, side by side would be best. Make them portable. Should be one board.

Resolution Agree.

status 0 Open 0 Resolved/Awaiting Disposition 0 Closed

YE

.,Record No. 52 Date 9/20/93 Name M. C. DauS Originating Site Site Applicability NMP-2 CI JAF 0IP-3 8NMP-2 C3Generic BWR QGeneral Impact DIP-2 PNMP-1 PGinna QGenericPWR 0 NUMARC-007 CI Procedure Cl Verification Q Training 0 Hardware Q EAL Q Technical Bases CI Validation 0 Deviation 0 None C¹ ** No. **

cat. N/A I Emer. Class.

Comment Scenario 19, question 19: Low lake level is 233.1 ft for service water pump operability based on USAR level for intake. USAR minimum lake level prevents ever reaching minimum intake level.

Therefore, could go to lake level instead of low intake level? Can it be measured? Check in simulator. It is on instrument LI-502 on panel 601.

Resolution No. Could have an intake problem causing low level when lake levels are ok.

Can be measured in control room.

status OOpen OResolved/Awaiting Disposition 0+Closed

l~

A V

ff I

C

OSSI 92-402A-7 EAL Valldatlon Procedure. Rev. 0 Attachment 3 VaHdation Exercise ChecMist Date: 10 7 93 Checklist No.: 10 Yes No ~NA

14. Did the EAL identiQcation scheme adequately support location of the EAL condition within the classification procedure' S Cl 0 Comments: None.

15. Additional Comments: None.

3-4

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