ML17059A353
| ML17059A353 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/10/1994 |
| From: | AFFILIATION NOT ASSIGNED |
| To: | |
| Shared Package | |
| ML17059A354 | List: |
| References | |
| NUDOCS 9407150229 | |
| Download: ML17059A353 (878) | |
Text
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XO Plant S specific EAL
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Nine Mile Point Unit 1 6/10/94 Operations Support Services, Inc.
233 Mfater Street 2nd Floor Plymouth, MA 02360
Plant Specific EA( bideline (A,H,S)
Nine Mile Point Unlt1
¹: AV1 Any unplanned release of gaseous or llquld radloactlvlty to the envlroninent that exceeds two times the radiological Technical Speclflcatlons for 60 minutes or longer.
ip. Mode ppllcablllty Q1(pwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Reftjei) Q6(Defuel) ~ All U1.1 AU1.2 valid reading on one or more of the following 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 229ElK IQKI'ote:
If the monitor readings Ja~ sustained for longer than 60 minutes and the required tssessments cannot be completed within this period, then the declarathn must be made
>ased on the valid reading.
Plant Specific EALsideline (A,H,S)
Nine Mile l'oint United l
uses ie term "Unplanned, as used in this context, includes any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions {e. g., minimum dilution iw, maximum discharge fhw, alarm setpoints, etc.) on the applicable permit. Val means that a radiation monitor reading has been confirmed by the operators to be correct.
iplanned 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 degradation in the vel 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 e 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 is 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 ill likely exceed 60 minutes.
I lonitor indications ~ should'alculated on the basis of the methodohgy of the site Offsite Dose Calculation Manual {ODCM to demonstrate ompliance with 10CFR20 andloi'10CFR50 Appendix I requirements. Annual average meteorology sboukkha h used uihoraaliowecL
Plant Specific EAluideline (A,H,S)
Nine Mlle volnt Unit1
- ¹: AU2 Unexpected Increase In plant radlatlon or airborne concentration.
)p. Mode (ppllcablllty Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) ~ All U2.1 AU2%
~ ~@~ direct area radiation monitor readinge hiahuuultinafrummammntzlhdurmu 2355~ Ihft ~ MQKQItK 4+eie.
Plant Specific EALsideline (A,H,S)
Nine Mile Point United ases slid means that a radiathn monitor reading has been confirmed by the operators to be correct.
Il of the above events tend to have hng lead times relative to potential for radiological release outside the site boundary, thus impact to public health and safety is very low.
light of Reactor Cavity Seal failure, incidents at two dNerent PWRs and hss of watei in the Spent Fuel Pij/Fuel Transfer Canal at a BWR all occurring since 1984, explicit coverage of these pes 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 vent.
mzda~ull JrmdiatmlhQJgQQQjnhllQQaunainiag ggjjQIQjihjjljjjQIQI"nQIQshlQbdQQii JQIjnQgQQrm. hhKZ
- AL40 ~ applies to plants with licensed dry storage of older irradiated spent fuel to address degradation of this spent fuel.
- .AL¹4~addresses unplanned increases in inylant radiation levels that represent a degradation in the control of radioactive material, and represent a potential degradation in the level of afety of the plant. JntligQJigngf QIQQLQfjQJign JQjjQh JggIQQmng Jg2~ JijnQQ JhQQIQonQQJggljjJQ JIQQhtQnQQJQQJQQKQjjQQ JhQQQItQJIIQQQIQ mgIQ IQZgy JIJQIIJIJJQhly lbgn a mgJligiQ gf 3KQEL JQYQI2 2ER hHd QQJgghlk QIQggmIQQIIX klan gnQ IJQQQlJQ gjtQr GgnnQI JQjjQIQ. M JimQQ JJIQ Qlgnn QQIggljjJ gIgYjjJQQ Qn QgjjljtQIQjnJbIQQJIQIIJ, This EAL escalates to an Alert Per IG A3, if the increases imPalr JhQ JQZgJgJ safe oPeration. QGJZgrgigngQC IBADIQQfjngQKQgj'InQIIJQ[QIJln JJIJQ M. fkRYQICIjnnQQQQRRE QmQIgQQK58KIQIQIign dLE E mgjIEIIJIKLQQIJ JQmggIKL Qjfi.jilgnJQIjQIQfhaihIJQJ!yQjjgQQjilKJljnQQ J!IQQhunQQJgglgJ.
Plant Specific EALjideline (A,H,S)
Nine Mlle Pajnt Unit 0 AA1 Any unplanned release of gaseous or llquld radioactivity to the environment that exceeds 200 times radiological Technical Specifications for 15 minutes or longer.
- p. Mode ppllcablllty Q1(PwrOps) Q2(HSB) 93(HSD) Q4(CSD) Q5(Refuel) Q6(Dehiel) QAII h1.1 AA1.2 valid reading on one or more of the following ggttt+ift monitors that exceeds the Value Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates in excess of 2%9. 11am. hatch @Usus 2QJN~Lt ELtnrllz lote: If the monitor readings4s ~ sustained for longer than 15 minutes and the required issessments cannot be completed within this period, then the declaration must be made
>ased on the valid reading.
iA1.3 AA1.4
Plant Specific EAL44ideline (A,H,S)
Nine Mile f oint Unit 0 tses did means that a radiation monitor reading has been confirmed by the operators to be correct.
iis 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 her and the 200 multiplier, the
.sociated site boundary dose rate would be 10 mR/hr. The required release duration was reduced to 15 minutes ln recognition of the increased severity.
onitor indications shoulQe~ calculated on the basis of the methodology of thagPJQJ~
Annual average meteorology ehout4ha ~ used whera4iowod.
Plant Specific EALuideline (A,H,S)
Nine Mlle t.oint Unlt1 4: AA2 MaJor damage to Irradiated fuel or loss of water level that has or will tesult In the uncovering of Irradiated fuel outside the reactor vessel.
)p. Mode hppllcabllity Q1(Pwrops) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII A2.1 AA2.2 Report of visual observation of irradiated fuel uncovered.
Building Ventifathn Monitor limEhr fULEhf:
fULE!lhr 8.0~
8.0339K iA2.3
Plant Specific EALuideline (A,H,S)
Nine Mlle I-oint Unit 3 aSeS
>Is IC applies to spent fuel requiring water coverage and Is not intended to address spent fue'I which Is licensed for dry storage, which is discussed in NUMARC IC AU2, 'Unexpected Increase Plant Radiation or Airborne Concentration. NUREG4818, "Emergency Action Leveh for Light Water Reactors," forms the basis for these EALs. Ihaaham ALs ~jjgftg by the specific area where irradiated fuel is heated such as reactor cavity, reactor vessel, or spent fuel pool.
"Severe Accident in Spent Fuel Pools in Support of sere is time available to take corrective actions, and there is little potential for substantial fuel damage. In addition, NUREG/CR4982, eneric 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
- o. 9048, "KR-85 Hazards from Decayed Fuel" presents the following it its discussidn:
an exclusion area radius of one mile n the event of a serious accident involving decayed spent fuel, protective actions would be needed for personnel on site, while offsite doses (assuming orn 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 vent of an accident with decayed spent fuel.
activities appropriately focus on concern icensees may wish to reevaluate whether Emergency Action Levels specified In the emergency plan and procedures governing decayed fuel handling accidents could occur, for example, the spent fuel pool working floor. Furthermore, licensees may wish to determine if ir onsite workers and Kr-85 releases in areas where decayed spent fuel procedures address the means for limiting radiohgical exposures of onsite personnel who are in other areas of the plant. Among other things, mergency plans and corresponding implementing iovlng onsite personnel away from the plume and shutting of building air intakes downwind from the source may be appropriate."
hus, an Alert Chssification for this event h appropriate. Escalation, if appropriate, would occur via Abnormal Rad level/Radiological Effluent or Emergency Director judgement.
Plant Specific EAluideline (A,H,S)
Nine Mlle 1-oint Unit 1 AA3 Release of radloactlve material or Increases In radiation levels wlthln the faclllty that lmpedes operation of systems required to malntaln safe operations or to establish or malntaln cold shutdown.
)p. Mode Lppllcablllty Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII A3.1 AA3.2 Ialid radiation monitor reading greater than 15 mR/hr in Ibh (sf~ctffe) Q QQtf:~ in areas requiring infrequent access to maintain piant safety functions.
Plant Specific EAluideline (A,H,S)
Nine Mlle I-oint United ases alid means that a radiation monitor reading has been confirmed by the operators to be correct. gales adaraliaadutt iumamaniary.mttihmauratxradhliuahmhfhathrhfkREhr.
yQgogftli ~ Lfttttijggatilt amjthrmf jn lhh EALLft$2hi firmmlssary KaeggagL his IG addresses increased radiation leveh that impede necessary access to operating stations, or other areas containing equipment that must be operated manually, in order to maintain safe oeration 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 nd/or magnitude of the increase in radiation leveh is not a concern of this IG. The Emergency Director must consider the source or cause of the increased radiation levels and determine if any ther IC may be involved. For example, a dose rate of 15 mR/hr ln the control room may be a problem in itself. However, the increase may also be indicative of high dose rates in the ontalnment 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 oaauaL NMP-1 due to a radioactivity release or radiation shine resulting from a major accident at thaotber.
. This is appropriate if the increase impairs operations at the operating unit.
his IG is not meant to apply to increases in the containment domo radiation monitors as these are events which are addressed in the fission product barrier matrix ICs. Nor is it intended to pply to anticipated temporary increases due to planned events (e. g., lncore detector movement, radwaste container movement, deplete resin transfers, etc.)
NMP-1 abnormal operating procedures, emergency operating procedures, the 10CFR50 Appendix R analysis, Ientifying areas containing safe shutdown equipment.
areas requiring continuous occupancy ~~~ include the control room an eomom central ~ ~gggti~ security alarm station. The value of 15 mR/hr Is derived from the GDG 19 value of 5 rem in 30 days with adjustment for expected occupancy times. Although lection 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, is a 30 day duration implies an event potentially more significant than an Alert.
ased on gbttgrmaj radiation levels which result in exposure control measures intended to naintain doses within normal occupational exposure guidelines and limits (i. e., 10CFR20), and in doing so, will impede necessary access.
10
Plant Specific EAL bideline (A,H,S)
Nine Mlle I-oint Unlt1 AS1 Boundary dose resulting from an actual or imminent release of gaseotts radioactivity exceeds 100 mR Whole Body or 500 mR Child Thyroid for the actual or proJected duration of the release.
- p. Mode ppilcabillty Q1(PvhtrOps) Q2(HSB) Q3(HSD) Q4(GSD) Q5(Refuel) Q6(Defuei) ~ AII 31.1 AAW4 valid reading on one or more of the following monitors that exceeds or is expected to (ceed the above criterion lote: If the monitor reading(s) is sustained for longer than 15 minutes and the required
.ssessments cannot be completed within this period, then the declaration must be made iased on the valid reading.
S1.3
'alid dose assessment capability indicates dose consequences greater than 100 mR delakody or 500 mRcMd QDE thyroid.
~ AS1.4 Field sutvey results indicate site boundary dose rates exceeding 100 mR/hr XEQE expected to continue for more than one hour; or analyses of field sunray samples indicate cbMiQQP thyroid dose commitment of 500 mR for one hour of inhalation.
11
Plant Specific EALuideline (A,H,S)
Nine Mile t-ufnt Unit1 ases that a radiation monitor reading has been confirmed by the operators to be correct.
~
~lid means integrated 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 se 100 mR agnitude) between
~scription. The 500 mR Integrated 4NcLthyroM g}p Qgzhf dose was established lri consideration of the 1$ ratio of the EPA Protective Action Guidelines for iyroid.
~
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 who4&ody and Qgf based on site boundary dose of 100 mR/hour arhotahody TEDE or 500 mR/hour 0
'hichever is more limiting (depends on source term assumptions).
he 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 EALuidelins (A,H,S)
Nine Mlle t~lnt Unlt1 4I: 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.
)p. Mode tppllcablllty Q1 (pvvf ops) Q 2 (HSB) Q 3 (HSD) Q 4 (CSD) Q 5 (Refuel) Q 6 (Defvel) Q AII G1.1 s valid reading on one or more of the following monitors that exceeds or is expected to xceed the above criterion Vote: If the monitor reading(s) is sustained for hnger than 15 minutes and the required assessments cannot be completed within this period, then the declaration must be made iased on the valid reading.
EG1.3 lalid dose assessment capability indicates dose consequences greater than 1000 mR JQ?f IvhoJahody or 5000 mR QDPcMd thyroid.
AG1.4 Field survey results indicate site boundary dose rates exceeding 1000 mR/hr continue for more than one hour; or analyses of field survey samples indicate doss commitment of 5000 mR for one hour of inhalation.
~ expected to chÃQP thyroid
Plant Specific EALsideline (A,H,S)
Nine Mile t-~fnt Unit 0 ases
~lid means that a radiation monitor reading has been confirmed by the operators to be correcL ie 1000 the dose exceeds > rem 3+Kahohkody or 5 rem ~~
mR IQEmholahody and the 5000 mR @}pchi8 thyroid integrated dose are based on the EPA protective action guidance which indicates that public protective actions are indicated thyrokL This ls consistent with the emergency class description for a General Emergency. This level constitutes the upper level I the desirable gradient for the Site Area Emergency. Actual meteorobgy Is specifbally Identified ln the initiating condition since it gives the most accurate dose assessment. Actual ieteorobgy (including forecasts) should be used whenever possible.
ased on site boundaiy doses for either ehotatedy TEDE or~MthycoidQQEfhgaid, whichever is more limiting (depends on source term assumptions(s).
- imfhamhulahufmlmhrihhjl'miamifhaaartmjladimtharaaatt a(fhamhailurajaQmaudral mam.aalathtahmmfltttlhrihhlQ
'he FSAR source terms applicable to each monitored pathway should be used in conjunctbn with annual average meteorobgy in determining indications for the monitors on that pathway.
14
Plant Specific EALuideline (A,H,S)
Nine Mlle t~fnt Unlt1 HU1 Natural and destructive phenomena affectin the protected area.
>p. Mode ppllcablllty Q1 (Pwr Ops) Q 2 (HSB) Q 3 (HSD) Q 4 (CSD) Q 5 (Refuel) Q 6 (Devel) ~ All "J1.1 HU1.2 Report by plant personnel of tornado striking within protected area.
U1.3 HU1.4 ssessment by the control room that an event Vehicle crash into plant structures or systems within protected area
>~LulaQMtht has occurred. boundary.
lU1.5 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.
esuiting in visible damage to permanent structure or equipment.
fU1.7 15
Plant Specific EA!uideline (A,H,S)
Nine Mile t-~lnt Unit 0 ases he protected area boundary Is4+caNy that part within the security Isolation zone and h defined in the site sec'urity plan.
gi: EAL ¹giQ 1 , NMP-1 seismic instrumentation actuates at 0.01 g. Damage may be caused to some portions of the site, but should not affect bility of safety functions to operate. Method of detection caa4o abased on Instrumentation, validated by a reliable sourc As defined i 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 mothn h felt at the nuclear plant site and recognized as an arthquake 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
>ost plants with seismic instrumentathn, the seismic switches are set at an acceleration of about 0.01 g."
AL ¹Jggi.2 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
~quired for safe shutdown of the plant. If such damage h confirmed visually or by other Inhalant indications, the event may be escalated to Alert.
AL ¹~3 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
~It but does not register on any plant-specific instrumentation, etc.).
.* kuL "" " '<<"
tructures containing functions and systems required for safe shutdown of the plant. If the crash ls confirmed to affect a plant vital area, the event may be escalated to Alert.
or EAL ¹~5, only those exploshns of sufficient force to damage permanent structures or equipment within the protected area should be considered. As used here, an explosion is a rapid, iolent, 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
- .AL 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
.:mergency Director also needs to consider any security aspects of the explosion, if applicable.
AL ¹~6 is intended to address main turbine rotating component failures of sufficien magnitude to cause observable damage to the turbine casing or to the seals of the turbine generator.
)f 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 lassified 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 quipment. Escalation of the emergency classification is based on potential damage done by missiles generated by the failure or by the radiological releases These latter events would be classified by the radiological ICs or fission product barrier ICs.
- AL ¹Qgf.7 covers e precursors of more serious
'vents.
16
Plant Specific EALuideline (A,H,S}
Nine Mlle ~oint Unit 3
¹: HU2 Fire wlthln protected area boundary not extinguished wlthln 15 minutes of detection.
tp. Mode
.ppllcablllty Q1(PwrOps) Q2(HSB) 03(HSD) Q4(CSD) Q5(Refuel) Q6(Deftlel) %All U2.1 gttQmZf fire in buildings or areas contiguous to any of the followin+sit~ci~areas not xtinguished within 15 minutes of control room notification latm:
17
Plant Specific EALuideline (A,H,S)
Nine Mlle I-~tnt Unlt1 ases within ie purpose of this IC h 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 Iministration buildings, waste-basket fires, and other small fires of no safety consequence. This IC applies to buildings and areas that are not contiguous or immediately adjacent to plant vital
.eas.
Mode".
scalation to a higher emergency class is by IC HA2, 'Fire Affecting the Operability of Plant Safety Systems Required for the Current Operating 18
Plant Specific EAIuideline (A,H,S)
Nine Mlle t. ~tnt Unit 0 HU3 Release of toxic or flammable gases deemed detrimental to safe operation of the plant.
)p. Mode applicability Q1 (PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5 (Refvel) Q6(Defvel) QAII U3.1 HU3.2 teport or detection of g~gf toxic or flammable gases that could enter II "
Report by local, county or state officials personnel based on offsite event.
for potential evacuation of site 19
Plant Specific EALuideline (A,H,S)
Nine Mile ~ ~lnt Unit )
ases his IG 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 Iacuation 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 laterials, in the DOT Emergency Response Guide for Hazardous Materials.
20
Plant Specific EA! uideline (A,H,S)
Nine Mile ~ ~lnt United 4f: HU4 Confirmed security event which Indicates a potential degradatlon In the level of safety of the plant.
)p. Mode
<ppllcablllty Q1(Pwrops) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII U4.1 HU4.2 emb device discovered within plant protected area and ftltfoutside the fgfltt~lt pant vital Other security events as determined from~~ecitic) rea~ Safeguards Contingency Plan.
21
Plant Specific EALuideline (A,H,S)
Nine Mlle a rnt Unlt1 ases his EAL is based o . Security events which do not represent at least a potential
>gradation in the level of safety of the plant, are reported under 10CFR73.71 or ln some cases under 10CFR50.72. The plant protected area boundary @typical/ that part within the security olation zone and is defined in the (site-specific) security plan. Bomb devices discovered within the plant vital area would result in EAL escalation.
22
Plant Specific EAluideline (A,H,S)
Nine Mlie v~lnt Unit 1 HU5 Other condltlons existing which In the Judgement of the Emergency Dliector warrant declaration of an Unusual Event.
)p. Mode tppilcablllty Q1 (Pwr Ops) 9 2 (HSB) 9 3 (HSD) 0 4 (CSD) 0 5 (Refuel) 0 6 (Defuel) g All U5.1
)ther conditions exist which in the judgement of the Emergency Director indicate a potential
'egradation of the level of safety of the plant. g~l 23
Plant Specific EAluideline (A,H,S)
Nine Mite I-~nt Unlit ases iis 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 fail under the Unusual Event emergency class.
rom a broad perspective, one area that may warrant Emergency Director judgement ls related to likely or actual breakdown of site specific event mitigating actions. Examples to consider elude inadequate emergency response procedures, transient response either unexpected or not understood, failure or unavailability of emergency systems during an accident in excess of iat assumed in accident analysis,.or insufficient availability of equipment and/or suppoit personnel.
is also intended that the Emergency Directors Judgement not be limited by any list of events as defined heie or as augmented by the site. This list is provided solely as examples for onsideration and it is recognized that actual events may not always follow a pre~nceived description.
24
Plant Specific EAIuideline (A,H,S)
Nine Mile I- tnt Uril) 1 A: HA1 Natural and destructive phenomena affecting the plant vital area.
)p. Mode hpplfcablflty 01(PwrOps) 02(HSB) 03(HSD) 04(CSD) 05(Refuel) Q6(Defuel) +All
'A1.1 HA1.2
~~Gal winds greater
~ Tornado or4igh mph strike within the protected area boundary.
than lA1.3 HA1.4
~f the following plant structures:
hll h~
lA1.5 HA1.6
/chicle crash Noctis plant vital areas.
lA1.7 25
- Plant Specific EAfuideline (A,H,S)
Nine Mjle t-uInt Untt1 ases ach 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 axsrred 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 dual magnitude of the damage. Escalation to a higher emergency class, if appropriate, will be based on System Malfunction, Fission Product Barrier Degradation, Abnormal Rad eieases/Radiohgical Effluent, or Emergency Director Judgement ICs.
AL ¹~1 should be based o+cltmpoc~ FSAR design basis of ~. Seismic events of this magnitude can cause damage to safety functions.
¹~2 should be based ogsit~poci~ FSAR design basis of 125 mph. Wind hads of this magnitude can cause damage to safety functions.
'hese
.AL
- .AL ¹~3 should specifygahacpocITQ structures containing systems and functions required for ~ ~gpgrgfjgn AL ¹~4
- .AL ¹55f 5 is intended to address such items as plane or helicopter crash, or on some sites, train crash, oc barge crash into a plant vital area.
'tlal" '
nciude 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
>r penetrated areas containing safety-related equipment the potential exists for substantial degradation of the level of safety of the p'lant
- -AL ~~covers EALs can also be g irecursore of more serious events. In particular, sites subject to severe weather as defined in the NUMARG station blackout initiatives, should Include an EAL based on activation of the severe weather mitigation procedures (e. g., precautionary shutdowns, diesel testing, staff callwuts, etc.).
26
Plant Specific EALuideline (A,H,S)
Nine Mlle taint Unlt1
¹: HA2 Fire or explosion affecting the operability of plant safety systems required to establish or maintain safe shutdown.
tp. Mode
<<ppllcablllty Q $ (pwr Qps) Q 2 (HSB) Q 3 (HSD) Q 4 (CSD) Q 5 (Refuel) Q 6 (Def0el) ~ All A2.1 he following conditions exist:
. Fire or exphsion in any of the followin+sf~pocNc) areas:
- i. Affected system parameter indications show degraded performance or plant personnel eport visible damage to permanent structures or equipment within the structures or
~quipment within the specified area.
27
uses IIM"hl W Plant Specific EAL Nine Mlle Pt
>> ~
sideline (A,H,S) ntUnlt)
W OIMSIRl insulted for equipment and plant areas required for the applicable mode. This will make it easier to determine if the fire or expbsion is potentially affecting one or more redundant trains of ifety systems. Escalation to a higher emergency class, if appropriate, will be based on System Malfunction, Rssion Product Barrier Degradation, Abnormal Rad fluent, or Emergency Director Judgement ICs.
Releases/Radiobgical lith regard to explosbns, only those expbsions of sufficient force to damage permanent structures or equipment required for safe operation within the klentified plant areas should be to nearby insidered. As used here, an explosion is a rapid, violent, unconfined combustbn, or a catastrophb failure of pressurized equipment, that potentially imparts significant energy to classification. No attempt is made in this ructures and materials. The inclusbn of a "report of visible damage should not be interpreted as mandating a lengthy damage assessment prior declaration of an Alert and the activation of the TSC will provide the Emergency AL to assess the actual magnitude of the damage. The occurrence of the expbsbn with reports of evidence of irector with the resources needed to perform these damage assessments. The Emergency Director also needs to consider any security aspects of the explosions, if applicable.
Mttrftitgm; 28
Plant Specific EAliouideline (A,H,S)
Nine Mlle ~oint Unit 3 4f: HA3 Release of toxic or flammable gases wlthln a facility structure which Jeopardizes operation of systems required to malntaln safe operations or to establish or malntaln cold shutdown.
)p. INode ippllcablllty Q1(PwrOps) Q2(HSB) Q3(HSD) 04(CSD) Q5(Refuel) 06(Defuel) ~ AII A3.1 HA3.2 ieport or detection of toxic gases within a f~gilgtttittgfacility structure~ in concentrations Report or detection of flammable gases within a ~ttllgtfdttg structure in concentrations that hat will be life threatening to plant personnel: will~ the safe operation of the plant:
29
0 Plant Specific EAluideline (A,H,S)
Nine Mlle f.ulnt Unit 1
'ases his IG h based on gases that have entered a plant structure&fectiap the safe operation of the p'lant. This IG applies to buildings and areas ontiguous to plant vital areas or other significant buildings or areas (L e., Service Water Pump house). The intent of this IG is not to include buildings (i. e., warehouses) or other areas that are ot contiguous or immediately adjacent to plant vital areas. It ls appropriate that increased monitoring be done to ascertain whether consequential damage has occurred. Escalation to a higher mergency class, if appropriate, will be based on System Malfunction, Fission Product Barrier Degradation, Abnormal Rad Releases/Radiological Effluent, or Emergency Director Judgement
~ 0 II 30
Plant Specific EAuideline United (A,H,S)
Nine Mlle rOlnt A'A4 Security event ln a plant protected area.
3p. Mode Applicability Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) AII IA4.1 HA4.2 ntrusion into plant protected area by a4e<144eree aasu!YQlSREl. Other security events as determined from~~peeitie)
Qgglg~ Safeguards Contmgency Plan.
31
ases
"~is h ~
iis event to a Site Area Emergency.
Plant Specific EALuideline (A,H,S)
Nine Mlle volnt Unlt1 class of security events represents an escalated threat to plant safety above that contained ln the UnusUal Event. For the purposes of this IC, b
32
Plant Specific EALOuideline (A,H,S)
Nine Mile eolnt Unlt1
¹: HA5 Control room evacuation has been Inltlated.
>p. Mode ipplfcablllty Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII A5.1 for ontrol room evacuation.
33
0 Plant Specific EAl uideline (A,H,S)
Nine Mile i-oint Unit 1 ases lith the control room evacuated, additional support, monitoring and direction through the Technical Support Center and/or other Emergency Operations Center is necessary. Inability to stabiish plant control from outside the control room will escalate this everrt to a Site Area Emergency.
~hrmmLI; 34
Plant Specific EAluideline (A,H,S)
Nine Mlle volnt Unlt1
¹: HA6 Other conditions existing which In the Judgement of the Emergency Director warrant declaration of an Alert.
ip. Mode ppllcablllty Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII.
W6.1 ther conditions existing which in the judgement of the Emergency Director indicate that lant safety systems may be degraded and that increased monitoring of plant functions Is arranted.
35
Plant Specific EAluideline (A,H,S)
Nine Mile saint Unit 3 aseS iis 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 Alert emergency class.
36
Plant Specific EALuideline (A,H,S)
Nine Mlle I-ufnt Unit1
¹: HS1 Security event In a plant vital area.
tp. Mode ippllcablllty Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) ~ All 81.1 HS1.2 itrusion into I~ttjjgtttjtgplant vital arear'y ~ostilo4orce ~ +ttftmggg Other security events as determined from~~peNio)
Qggttfj~nt( Safeguards Contingency Plan.
37
Plant Specific EAluideline (A,H,S)
Nine Mile I-elnt Unit 0 ases his class of security events represents an escalated threat to plant safety above that contained In the Alert tG In thatakostitaforcomadltftrzuy has progressed from the protected area to ie vital area 38
Plant Specific EAluideline (A,H,S)
Nine Mlle tolnt Unit 0 4f: HS2 Control room evacuation has been initiated and cannot be established.
)p. Mode tppllcablllty Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) ~ AII S2.1 he following conditions exist:
. Control room evacuation has been initiated.
AND
- i. Control of thaptant cannot be established per+~pecific)-
exodus withi+sitoapocific) jg ninutes.
39
Plant Specific EAIuideline (A,H,S)
Nine Mile rutnt Uril) 1 ases xpeditious transfer of safety systems has not occurred but fission product barrier damage may not yet be Indicated.+Ne~ecifg time for transfer hbased on analysis or assessments as to
~w quickly control must be reestablished without core uncovering and/or core damage. This time should not exceed 15 minutes. In cold shutdown and refueling modes, operator concern is
'rected 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 bairfer integrity. Escalation of this event, if appropriate, would be by Fission roduct Barrier Degradation, Abnormal Rad Releases/Radiological Effluent, or Emergency Director Judgement ICs.
6thraauMtnitmtmfimtrtfthanilblinanrfada nfant rtuaratiana.hdant mntruLmhatfmaurlmariltr unthaabilittr tamaintain thaamabzinaauhuf mnrfltinn. Iharafara. it hannzrtriata ta
'mumthalQantihhLanthatitamnttaaizaathanaatifnrazaamlina tkhanmntrrtlling thaufantfuunutlihidathaQaahul Bzun.
40
ll Plant Specific EA'uideline (A,H,S)
Nine Mlle t-vjnt United HS3 Other conditions which In the judgement of the Emergency Director warrant declaration of Site Area Emergency.
)p. Mode
<pplicablllty Q 1 (Pwr Ops) 0 2 (HSB) 0 3 (HSD) 9 4 (CSD) 0 5 (Refuel) 0 6 (Devel) ~ All S3.1
)ther conditions which in the judgement of the Emergency Director warrant declaration of tite Area Emergency. 1~i 41
Plant Specific EALuideline (A,H,S)
Nine Mile ~oint Unlt1 ases his EAL h intended to address unanticipated conditions not addressed explicitly elsewhere but that warrant dec1aratbn 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 EAuideline (A,H,S)
Nine Mlle Point Urilt1 HG1 Security event resulting In loss of ablllty to reach and rnalntaln cold ihutdown.
)p. Mode applicability Q 1 (PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) %All G1.1 HG1.2
.oss of ~+ysicat control c4~ the control room due to security event.
usa af uhmiml mntrulaf fha ramuiaaltttldannaauabTiitxdualammurily.mant. - .
43
Plant Specific EAluideline (A,H,S}
Nine Mlle t-oint Unit 1 ases his IG encompasses conditions under which a hostile force has taken physical control of vital areaa required to reach and maintain safe shutdown. Iha gunuarn
~uttm tharaauturanfirnaintalnuuraamling. Iharafura tbla M.haahaanmutiliiatituralaut aiuaaut ~uuntrul ftumhafhtha uuntrut ruumanti ramuta ahtttdunnuanala.
~ la tha tuaa ul ahTiittr 44
Plant Specific EAluideline (A,H,Sj Nine Mile I-ulnt Unit 3 HG2 Other conditions existing which ln the Judgement of the Emergency Director warrant declaration of General Emergency.
)p. Mode hppllcablllty Q1 (pvvr Ops) Q 2 (HSB) Q 3 (HSD) Q 4 (CSD) Q 5 (Refuel) Q 6 (Defuel) R All G2.1 7ther conditions existing which in the Judgement of the Emergency Director indicate: (1) actual or imminent substantial core degradation with potential for loss of containment, or (2) etential for uncontrolled radio nuclide releases. These releases can reasonably be
.xpected to exceed EPA PAG plume exposure levels outside the site boundary.
45
Plant Specific EAIjideline (A,H,S)
Nine Mlle f-ufnt Unit 0 uses iis EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but that warrant declaratbn of an emergency because conditions exist which are believed by the nergency Director to fall under the General Emergency class.
46
Plant Specific EAluideline (A,H,S)
Nine Mile Point Unit 3 4f: SU1 Loss of all offslte power to essential busses for greater than 15 minutes.
)p. Mode ippllcabllity Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Def0el) %All U1.1 he following conditions exist:
. Loss of power t+sit~pecie)
~asfocmere for greater than 15 minutes.
AND
>. At least~~peciT>e) lttttt emergency generators are supplying power to emergency
>uses; 47
Plant Specific EAlouideline (A,H,S)
Nine Mlle vo1nt Unlt1 ases robnged 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 comp'iete hss of AC power (station iackout). Fifteen minutes was selected as a threshold to exdude transient or momentary power losses.
48
Plant Specific EAIuideline (A,H,S)
Nine Mlle r'oint Unit 1
'¹: SU2 Inablllty to reach required shutdown wlthln Technical Speclllcatlon Llmlts.
)p. Mode ippllcablllty 51(PwrOps) ~ 2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Deluel) QAll U2.1
'lant is not brought to required operating mode withiQeit~ecitic) Technical Specifications
.CO Action Statement Time.
49
Plant Specific EAI iideline (A,H,S)
Nine Mlle Point Unlt1 aSeS imiting Conditions of Operation (LCOs) require the plant to be brought to a required shutdown mode when the Technical SpeciTicatlon required configuration cannot be restored. Depending on ie 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 iquires a one hour report under 10CFR50.72 (b) non-emergency events. The plant fs within its safety envelope when being shut down within the alhwable action statement time in the Technical pecifications. 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 pecificatlons. 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
> how long a condition may have existed. Other required Technical Specification shutdowns that fnvolve precursors to more serious events are addressed by other System malfunction iazards, or Fission Product Barrier Degradation ICs.
50
Plant Specific EAluideline (A,H,S)
Nine Mlle F oint United SU3 Unplanned loss of most or all safety system annunciation or indication In the control room for greater than 15 minutes.
)p. Mode ippllcablllty g1(PwrOps) ~ 2(HSB) 83(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII U3.1 he following conditions exist:
.. Loss of mestocal~~eciflc) annunciators for greater than 15 minutes.
AND AND
- . In the opinion of the Shift Supervisor, the loss of the annunciators or indicatois iequtres ncreased surveillance to safely operate the unit(e),
AND L Annunciator or indicator loss does not result from planned action.
51
Plant Specific EALuideline (A,H,S)
Nine Mlle Point Unlt1 ases iis IC and its associated EAL are Intended to recognize the difficult associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication
]uipment.
ecognition of the availability of computer based indication equipment ls considered (SPDS, plant computer, etc.).
jnplanned" loss of annunciators or indicator excludes scheduled maintenance and testing activities.
~eciTic plant design and subsequent retrofits.
Namihnafmhhrthaad uafhhrlm.
~~
- ompensatory non-alarming indications: in this context includes computer based information such as SPDS . This should Include all computer systems available for this use depending on h 0M' Ib Ihkt threshold It h not intended that plant personnel perform a detailed count of instrumentation lost but ~ use gf thou~ judgement + fjtftgruff ~~
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 rovide increased monitoring of system operation to safely operate the unit~.
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 nnunciators is more Iltcely 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 pecific, 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.
he 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
'pecification action, the Unusual Event is based on SU2, inability to Reach Required Shutdown Within Technical Specification Limits."
Mt~Tie) Annunciators MacQcahe for this EAL must include those identified In the Abnormal Operating procedures, in the Emergency Operating Procedures, and in other EALs (e. g.,
rea, process, and/or effluent rad monitors, etc.).
SIto-specific) Annunciators ce4adicaM for this EAL must Include those identified in the Abnormal Operating procedures, in the Emergency Operating Procedures, and in other EALs (e.g.,
korea, process, and/or effluent rad monitors, etc.).
'ifteen minutes was selected as a threshold to exclude transient or momentary power hsses.
)ue 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.
I'his Unusual Event will be escalated to an Alert if a transient is in progress during the loss of annunciation or indication.
52
Plant Specific EALuideline (A,H,S)
Nine Mile I'oint Unit 1
- ¹: SU4 Fuel clad degradatlon.
)p. Mode applicability Q1(PwrOps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) ~ All U4.1 SU4.2
~~Q~~~citic) coolant sample activity 53
Plant Specific EALideline (A,H,S}
Nine Mlle I-oint Unit 1 tS8$
iis IC is fncjuded as an Unusual Event because it ls conskfered to be a potential degradation ln the level of safety of the plant and a potential precursor of more serious problems.
Kl 1
¹~1 addresses +rupee~~ radiation monitor reading , that provide indication of fuel clad integrity. ~
KL ¹~2ICs.addresses coolant samples exceeding coolant technical specifications for hdine spike. Escalation of this IC to the Alert level is via the fission product barrier degradation onitoring 54
Plant Specific EAiuideline (A,H,S)
Nine Mlle volnt Unlt1 4f: SU5 RCS leakage.
)p. Mode ippllcablllty ~ 1(PwrOps) ~ 2(HSB) 83(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII U5.1
- ither of the following conditions exist:
~. Unidentified 0 gpm gggtgzazgaaf ~ ftt 4gyg5 leakage greater than OR
- i. Identified ~ gogh' g4ryygg leakage greater than 25 gpm.
55
Plant Specific EAiuideline (A,H,S)
Nine Mlle l-oint Unit 3 ases iis IC is included as an Unusual Event because it may be a precursor of more serious conditions and, as a result, is conskfered to be a potential degradation of the level of safety of the plant.
ae 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 ne~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
. pressure boundary leakage. In either case, escalation of this IC to the Alert level ls via Fission Product Barrier Degradatfon ICs or IC SAS, "inability to Maintain Plant in Cold Shutdown."
nly operating modes ln which there is fuel ln the reactor coolant system and the system Is pressurized are specified.
Plant Specific EALsideline (A,H,S)
Nine Mile l-oint Unit 3
¹: SU6 Unplanned loss of all onslte or offslte communlcatlons capabllltles.
>p. Mode pplfcablllty Q1(PwrQps) Q2(HSB) Q3(HSD) Q4(CSD) Q5(Refvel) Q6(Defuel) ~ All J6.1 ither of the following conditions exist:
onsite communications capability affecting re ability to perform routine operations:
. Loss of all offsite communications capability:
57
uses dios/walkie talkies).
>d Zh 'I Plant Specific EAL+iideline (A,H,S)
Nine Mile vvtnt United ie purpose of this IG and its associated EALs is to recognize a loss of communications capability that either defeats the plant operations staff ability to perform routine tasks necessary for ant 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 ldressed by 10CFR50.72.
I., .'" '" """ ', 0 dedicated EPP phone systems. This EAL is intended to be used only when extraordinary means are being utilized to make communications possib'le (relaying of information from radio
~nsmissions, individuals being sent to offsite locations, etc.).
58
Plant Specific EAluideline (A,H,S)
Nine Mlle >oint Unit 1 A: SU7 Unplanned loss of required OC power during cold shutdown or refueling mode for greater than 15 minutes.
)p. Mode kpplfcablllty 01(PwrOps) 02(HSB) Q3(HSD) 54(CSD) S5(Refuel) Q6(Defuel) QAII U7.1 Ahoc ~ of the following conditions exist:
>us voltage Indications AND
>. Failure to restore power to at least onecoqutred DG bus within 15 minutes from the time of oss.
59
Plant Specific EALiideline (A,H,S)
Nine Mile Porrit Unit1 3SSS ie purpose of this IC and its associated EALs 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
>eratlons. This EAL is intended to be anticipatory in as much as the operating crew inay not have necessaly Indication and control of equipment needed to respond to the hss.
iplanned 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 tsis during shutdown periods. It is intended that the hss of the operating (operable) train is to be considered. If this loss results in the inability to maintain cold shutdown, the escalation to an ert will be per SA3 "Inabilityto Maintain Plant in Cold Shutdown.
Qaepocif+~ bus voltage should be based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value should incorporate a margin of at least i minutes of operationbefore the onset of inability to operate those hads. This voltage is usually near the minimum voltage selected whenbattery sizing is performed. Typically the value for e entire battery set is approximately 1 05 volts per cell. For a 56 string battery set the minimum voltage is typically 1.8f volts per cell.
60
Plant Specific EA'uideline (A,H,S)
Nine Mile t'-oint Unit 0
- ¹: SA1 Loss of all offslte power and loss of all onslte AC power to essential busses during cold shutdown or refueling mode.
)p. Mode
<pplfcablllty Q1(PwrOps) Q2(HSB) Q3(HSD) 84(CSD) 55(Refuel) ~ 6(Defuel) QAII A1.1 hll of the following conditions exist:.
- i. Loss of power t AND
>. Failure of~~peciffo)
AND within 15 minutes rom the time of loss of both offsite and onsite AG power.
61
Plant Specific EAL4iideline (A,H,S)
Nine I@lie Punt United BSBS Ns of all AC power compromises all plant safety systems requiring electric power I
'hen in cold shutdown, refueling, or defueled mode the event can be classified as an Alert, because of the signNcantly reduced decay heat, lower temperature and pressure, increasing the ne 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 bevels/Radhlogical Effluent, or Emergency Director Judgement ICs. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
ckfQDUERK 62
Plant Specific EALuideline (A,H,S)
Nine Mlle I-oint Unit 1
¹: w h SA2 W P "" h p
dd SIR
)p. Mode ippllcablllty ~ 1(PwrOps) ~ 2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAll A2.1 63
Plant Specific EAl uideline (A,H,S)
Nine Mlle t.oint Unlt1 ases Yis condition indicates failure of the automatic antf manttat protection system to scram the reactor tuiha atttant ifrhtuhuiauitttfaa tharaautur haiOumada aii~iuat. This condition is more an a potential degradation of a safety system in that a front line automatic protection system dkl not function in response to a plant transient and thus the plant safety has been compromised, id design limits of the fuel may have been exceeded. An Alert is indicated because conditions exist that lead to potential hss of fuel clad or RCS.
A manual scram is any set of actions
~huituoa.mudaa)Mituhur hB9.
hialQaotframiltiou GALJtattahaanauauifiuai!tr mudifiafitumuraaauratabt daIioaihauuodittundamihatfhxthauaoartuhaaaaasauuliatituhuilioumatar mautura. Ihafaitiira uf aiituroatiu iitiatiunufaraautur mzamfulitofratfbyatiuuaaafttl~ JoitiatiunautiuoamhhhuanharauiCxtaifanat Jharaauturuuottufuuoauiarfuaanutuuaaauuiaotiat Juaauf aithar ftiaf uhtfur BQR uttotfaiiaa. JJauattaafailuratuauiamuuotfitiuoainEMhara falloff nuthx JhaazLdaouaufuotiuaiityituthr thamaiiattlamaruinuf mihuritiuaiitx.thauanariuuuiCzuamulC iauuira iaaaifiuatiunufanhhrt furuuotfitiuoainmhhhtharaauturh Jnfautahtdtfui~tnaaaiaaultuf JhaauramaiuoaL Jththauuotimtatfuritiualjttr untfaruuntfitIuoaiautiiriou aiaauturauiamnhiuh umtathauutaotial thraat tuBRQurhafulatfiotauritx. Jtiaahuimuuitant tuautathat thafaituiauf thaauztur urutautiunattatamtu Jnitiataanatitumatiuauram tfuaanui Jofar autttalur uutaotiaf thraa uf uthar a)tatamanur h Jt. In and.uf itmtlf.aurauttraur iu fhmunuruduut haular rfauratfatiuo. Iha BERaaootanuuthar aafaihhoutiun htit iu initaitaraautur zuraina. Iharafura.uouaiha tautur Jiaahaanatiuaaauilxaurammad.faihtraainiha BER axatam uanha~raouulaot aafattr Jmuaut. Jf Jmmadiataautnual autiunatu auram tha raautur arazzaaafut fultuninu rauuuoitiun uf an titumatiuauramfaihtrhiharahnuthraat tuaitharuiaot aafaixur fimunurutfuut Jntauoivralatatfh.thaatdumattuautamfaiturh Ihhtfayiatiunhuuoahtaotmahthauhiiumuhxufmahou
~aoxfaiturauf anautumatlumunmuml tuioitiataaatiuufmfulauramMmultfht JmmatJJatttlxfulilototdhxanuuaratur Jnitiatatfatanualauram. Iharafuiatha Ehl.murdiouinihaIBahaii olxmakarafftraouatuthafailurauf Jmmafiata manual mama)'4
Plant Specific EAluideline (A,H,S)
Nine Mlle ~oint Unlt1
¹: SA3 lnablllty to maintain plant In cold shutdown.
>p. Mode ppllcablllty D1{PwrOps) Q2(HSB) Q3(HSD) %4{CSD) Q5(Refuel) Q6{Defuel) QAII W3.1 he following conditions exist:
Toepocaucohcroaco-Whau4hov
~ E OR 65
iE Plant Specific EALsideline (A,H,S)
Nine Mlle Pr int Unit1 uses would be via iis EAL addresses complete loss of functions required for core cooling during refueling and cold shutdown modes. Escalation to Site Area Emergency or General Emergency
>normal Rad Leveh/Radiological ENuent or Emergency Director Judgement ICs.
jncontrolled" means that system temperature increase is not the result of planned adions by the plant staff.
he EAL guidance related to uncontrolled temperature rhe ls necessary to preserve the anticipatory philosophy of NUREG4654 for events starting from temperatures much lower than the cold hutdown temperature limit.
scalation to the Site Area Emergency is by IC SS5, "Loss of Water Level in the Reactor Vessel that has or will Uricover Fuel in the Reactor Vessel," or by Abnormal Rad Levels/Radiological
.ffluent ICs.
66
Plant Specific EAuideline (A,H,S)
United Nine Mlle I-oint A: 'A4 Unplanned toss of most or all safety system annunclatlon or Indication In control room with either (1) a slgnlfleant transient In progress, or (2) compensatory non-alarming Indicators are unavailable.
3p. Mode hpplicabllity ~ 1(pwrops) g2(HSB) ~ 3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII lA4.1 rhe folhwing conditions exist:
- s. Loss of amstocall+~pecifh) annundators or greater than 15 minutes.
AND
- i. In the opinion of the Shift Supervisor, the loss of the annunciators or Indicators requires increased surveillance to safely operate the unit(e).
AND
- . Annunciator or indicator hss does not result from planned acthn.
AND
- d. Either of the following:
~ A significant plant transient is in progress OR are unavailable.
67
Plant Specific EA sideline (A,H,S)
Nine Mlle Po<<>t Unit) ases sis IC and its associated EAL are intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication Iuipment during a transient. Recognition of the availability of computer based Indication equipment Is considered (SPDS, plant computer, etc.).
'lanned" loss of annunciators orindicators includes scheduled maintenance and testing activities.
lt Is not intended that plant personnel perform a detailed count of the instrumehtation lost but the specific opinion o f the Shift
~ use gf thcucal~~ Judgement Supervisor that additional operating personnel will be required to ireshold for determining the severity of the plant conditions. This judgement is supported by rovide increased monitoring of system operation to safely operate the unit(s).
a portion of is further recognized that most plant designs provide redundant safety system Indication powered from separate uninterruptable power supplies. While failure of large associated with assessment of plant conditions. The loss of nnunciators is more likely than a failure of a large portion of indications, the concern Is included In this EAL due to difficulty status. This will be addressed by the specific Technical Specification.
pecific, or several, safety system indicators should remain a function of that specific system or component operability related to the instrument hss will be reported vfa 10CFR50.72. If the shutdown is not in compliance with the Technical he initiation of a Technical Specification imposed plant shutdown
'pecification action, the Unusual Event is based on SU2 "Inability to Reach Required Shutdown Within Technical Speclcation Limits."
Qt~ecific) Snnunciators ocJadicatoas for this EAL must include those identified in the Abnormal Operating Procedures, ln the Emergency Operating Procedures, and in other EALs (e. g.,
rea, process, and/or effluent rad monitors, etc.).
ECCS injections, or thermal Significant Transient includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, ewer oscillations of 10% or greater.
Compensatory non-alarming Indications in this context Includes computer based Information such as SPDS. This should include all computer systems available for this use depending on of the annunciation system and all computer monitoring are unavailable to the extent that the additional operating
- pecific plant design and subsequent retrofits. If both a major portion iersonnel are required to monitor indications, the Alert is required.
)ue 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.
%is Alert will be escalated to a Site Area Emergency if the operating crew cannot monitor the transient in progress.
68
Plant Specific EAIuideline (A,H,S)
Nine Mlle t ~tnt Unit1 4: SA5 AC power capablllty to essential busses reduced to a single power source for greater than 15 minutes such that any addltlonal single failure would result In station blackout.
)p. Mode ippllcablllty ~ 1(pwrOps) ~ 2(HSB) ~ 3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAll A5.1 he following conditions exist (a and b):
- t. Loss of power tQsit~pecific) for greater than 15 minutes.
AND
>. Onsite power capability has been degraded to gabt one g~gilglttlmt)tcaiaof) 69
Plant Specific EAl ~ideline (A,H,S)
Nine Mile Petttt Unlt1 aSOS his IG and the associated EALs are intended to provide an escalation from IG SU1 'Loss of AII Offsite Power to Essential Busses for Greater than 15 Minutes. The condition indicated by this
- 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 hss of all offsite
~war and loss of onsite emergency dlesels with only one train of emergency busses being backfed from the unit main generator, or the loss of onsite emergency diesels with only one train of mergency busses being backfed from offsite power. The subsequent loss of this single power source would escalate the event to a Site Area Emergency in accordance with IG SS f "Loss of li Offsite and Loss of All Onsite AG Power to Essential Busses."
xample EAL ¹~1b 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 allure vulnerability.
70
Plant Specific EA'uideline (A,H,Sj Nine Mlle i ~lnt Unl)1 N: SS1 Loss of all offslte power and loss of all onslte AC power to essential biisses.
)p. Mode Lppffcabflity 1(PwrOps) g2(HSB) k3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII S1.1
.oss of all offsite and onsite AC power as indicated by:
- i. Loss of power to ~~pecifie) 4caasfarmers.
AND
~. Failure oQsit~ecifie) 1
@specific) M minutes from the time of hss of both offsite and onsite AC power..
71
Plant Specific EAIuideline (A,H,S)
Nine Mlle l-vfnt Unit 3 ases
~ss of all AC power compromhes all plant safety systems requiring electric power I Prohnged loss of all AC power ill cause core uncovering and loss of containment integrity, thus this event can escalate to a General Emergency. Th+elt~ pec~ time duration should be selected to exclude transient or iomentaty power hsses, but shoukl not exceed 15 minutes.
scalation to General Emergency is via Fisson Product Barrier Degradatlon or IC SGh; "Prolonged Loss of AII Offsite Power and Prohnged Loss of All Onsite AC Power."
IftfQQUEQK 72
Plant Specific EAlo'uideline (A,H,S)
Nine Mile i ~lnt Unit 1 A: SS2 Failure of Reactor Protection system Instrumentatlon to complete or Inltlate an automatic reactor scram once a Reactor Protection system setpolnt has been exceeded and manual scram was not successful l
)p. flrfode hppllcablllty 51(PwrOps) ~ 2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII S2.1 Mmanualmunarmhmafhmunaianal hlhnmfhxamaaual mun~hihhahui hmihaamatur.
73
Plant Specific EALuideline (A,H,S)
Nine Mlle w~int Unlt1 ases nder 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
~nditions 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 ecessary to better assure timely recognition and emergency response. Escalation of this event to a General Emergency wou'Id be via Fisshn Product Barrier Degradation or Emergency
'irector Judgement ICs.
74
Plant Specific EAuideline (A;H,S)
Nine Mlle i ~fnt Unlt1 A: SS3 Loss of all vital DC power.
3p. Mode hp pflcablllty ~ 1(PwrOps) ~ 2(HSB) ~ 3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuei) QAII S3.1
.oss of all vital DG power based o+ct~pociTio) ~tlat bus voltage indications for greater than 15 minutes.
~
75
Plant Specific EA'uideline (A,H,S)
Nine Mlle I-~tnt Unlt1 ases ass of all DC power compromises ability to monitor and control plant safety functions. Prolonged loss of all DC power will cause core uncovering and hss of containment integrity when there is gnificant decay heat and sensib'Ie heat In the reactor system. Escalation to a General Emergency would occur by Abnormal Rad Levels/Radiological Effluent, Fission Product Barrier egradatlon, or Emergency Director Judgement ICs. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
76
Plant Specific EAIuideline (A,H,S)
Nine Mlle >~tnt Unit 3 O'S4 Complete loss of function needed to achieve or hialntaln hot shutdovrn.
)p. Mode Lppllcablllty ~ 1(PwrOps) 82(HSB) 53(HSD) 04(CSD) p5(Refuel) Q6(Defuel) QAII S4.1 77
Plant Specific EAL4uideline (A,H,S)
Nine Mlle f-~tnt Unjt 0 uses iis EAL addresses complete loss of functions, including ultimate heat sink and reactivity control, required foi hot shutdown with the reactor at pressure and temperature. Under these inditions, 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 Emergericy woukf cur by Abnormal Rad Leveh/Radiohglcal Effluent, Fission Product Barrier Degradation, or Emergency Directoi Judgement ICs.
78
Plant Specific EAluideline (A,H,S)
Nine Mile ~ ~int Unit 1 A: SS5 Loss of RPV water level that has or will uncover @el In the RPV.
3p. Mode Applicability ~ 1(PwrOps) S2(HSB) S3(HSD) ~ 4(CSD) 55(Reftjel) Q6(DeNel) QAII
'S5.1 79
Plant Specific EA'~sideline (A,H,S)
Nine Mlle P~,nt United tases Inder the conditions specified by this IC, severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured.
0 I'lhl I I hb h5'U Idl t lib mmtmtlgnhrtftnftmllxamidaratfihahtttfar huttntfanr gfihararminmhhhlgf:aliuff&lagmthoarnmr @mr. Mrmtrataattghammantmtiunrtf mnhtaiihift gaa.haaafhgththa fttal
'htI htf antf BGRharrhramuathmuzurrad. 3hmfarLdadamthnaf a Kta hraa Etnttrganehzarmntad.
fhufrztanmnmtmtignaiagmm inmolurztiunniththauramtnmrtf mmmnhghhal daflagratignffmiaLLaRFlabxdrrtmtnantfaPI muganhumtiauaf thamataintnttntirmartat tea ref hauffaitaraffimtbfaraiaaaaratamuhiharauuimtf KRZamCrfagiaratignrtfa Qmral Emurganmrftaulrad
%us, declaration of a Site Area Emergency is warranted under the conditions specfffed by the IC. Escalation fo a General Emergency is via radiological effluence lC AG1.
80
Plant Specific EAuideline (A,H,S)
Nine Mite ~ ~lnt Unlt1 A: SS6 Inablllty to monitor a slgnlfleant transient In progress.
3p. Mode Applicability S1(PwrOps) 52(HSB) ~ 3(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII iS6.1 All of the following conditions exist:
- t. Loss of~~ooifio) annunciators 0
). are unavailable.
AND
- . Indications needed to rnonitor~taepocifio)
AND
$. Transient in progress.
81
Plant Specific EAIuideline (A,H,S)
Nine Mlle r ~tnt Unit 1 Iases his IG 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 ontrol room staff cannot monitor safety functions needed for protection of the public.
Sitaepecif+ 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.,
ad monitors, etc.).
Compensatory non-alarming Indications" in this context Includes computer based information such as SPDS. This should include all computer systems available for this use depending on pecific 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 ewer osciliations of 10% or greater.
Q~ocif+ Indications needed to monitor safety functions necessary for protection of the public must include control room indications, computer generated indications and dedicated innunciation 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,
~ 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 instrumentathn of this magnitude Is of such significance during a transient that the cause of the toss is not an ameliorating actor.
82
0 Plant Specific EAieluideline (A,H,S)
Nine Mile i ~lnt Unlt1 A: SG1 Prolonged loss of all offslte power and prolonged loss of all onslte AC power.
3p. Mode Appllcablllty ~ ~(PwrOps) 82(HSB) 83(HSD) Q4(CSD) Q5(Refuel) Q6(Defuel) QAII G1.4
'rohnged loss of all offsite and onsite AG power as indicated by:
- t. Loss of power tQci~pecific)
AND
- i. Faffure o~~pecific)
AND
- . At least one of the following conditions exist:
~ Restoration of gttz~ at least one emergency bus within~~peciTic) 4 hoiirs ls not likely OR
~
(
83
Plant Specific EAIuideline United (A,H,S)
Nine Mlle I-vfnt
'ses ms of all AC power compromises all plant safety systems requiring electric power i . Prohnged hss of all AC power ill lead to bss of fuel clad, RCS, and containment. Th+sitazpocific) hours to restore AC power can be based on site blackout coping analysis performed in conformance with 10CFR50.63 and egulatory 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 arrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response.
his 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
~rs as early as h appropriate, based on a reasonable assessment of the event trajectory.
he 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 vent could result in a loss of valuable time in preparing and implementing public protective actions.
i addition, under these condithns, fission product barrier monitoring capability may be degraded. Although Ij may be difficult to predict when power can be restored, it is necessary to give the mergency Director a reasonable Idea of how quickly (s)he may need to declare a General Emergency based on two major considerations:
. Are there any present Indications that core cxeiing 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 iformation.)
. If there are no present Indications of such core cooling degradation, how likely h it thaj power can be restored in time to assure that a loss of two barriers with a potential loss of the third arrier can be prevented?
hus, indication of continuing core cooling degradation must be based on fission product barrier monitoring with particular emphasis on Emergency Director judgement as it relates to imminent ass or potential loss of fission product barriers and degraded ability to monitor fisshn product barriers.
84
Plant Specific EAguideline (A,H,S)
Nine Mil~ r ~lnt Unlt1 A: SG2 Failure of the Reactor Protection system to complete an automatic scram and manual scram was not successful and there ls lndlcatlon of an extreme challenge to the ablllty to cool the core.
3p. Mode Applicability 1(PwrOps) g2(HSB) Q3(HSD) Q4(CSD) Q5(Refuel) 06(Defuel) DAll lG2.1 AND
~. Either of the folhwing:
a.
(S'.(s'R 85
Plant Specific EAtluideline (A,H,S)
Nine Mila a ~lnt United tases wtomatic and manual scram are not considered successful if action away from the reactor control console h required to scram the reactor.
lnder the conditions of this IC and its associated EALs, the efforts to bring the reactor subcritlcal have been tlnsuccessful and, as a result, the reactor is producing more heat than the aaximum decay heat load for which the safety systems were designed. Although there are capabilities away from the reactor control console, such as standby quid control In BWRs, the continuing temperate rise indicates that these capabilities are not effective. This situation could be precursor for a core melt sequence.
For BWRs, the extreme challenge to the ability to cool the core is intended to mean that the eactor vessel water level h below 2/3 coverage of active fuel another consideration ls the inability to initially remove heat during the early stages of this sequence.
For BWRs, ci~peclflc) consklerations include inability to remove heat via the main condenser, or via the suppression pool or torus (e. g., due to high pool water temperature).
n the event either of these challenges exist at a time that the reactor has not been brought behw the power associated with the safety system design (typically 3 to 5% power) a core melt equence exists. In this situation, core degradation can occur rapidly For this reasori, the General Emergency declaration Is intended to be anticipatory of the fission product barrier matrix leclaration to permit maximum offsite intervention time.
86.
Plant Specific EAL sideline (FPB)
Nine Mlle Point Unlt1 BWR FPB fC¹: FC1 Barrier: Fuel Claddin Type: Loss Descrfptlotl: Primary Coolant Activity Level FC1.1 Coolant activity greater than Bases:
Assessment by the NUMARC EAL Task Force indicates that this amount of coolant activity h well above that expected for iodine spikes and corresponds to about 2% to 5'/o fuel clad damage. This amount of clad damage indicates significant clad heating and thus the fuel clad barrier h considered hst.
There is no equivalent "Potential Loss" EAL for this item.
2.HhhBLB.annuzhtz&
2.555BLELznuamltz&
87
Plant Specific EAL sideline (FPB)
Nine Mlle Point Unit 3 BWR FPB IC¹: FC2 Barrier: Fuel Claddin Type: Loss/Po .Loss Descrfptfoo: Reactor Vessel Water Level FC2.1 Level less than Bases:
The "Loss" EA+si~peci~ value corresponds to the level which is used in EOPs to indicate challenge ol core cooling.
oWctiua4uot. Thisisthemlnimumvaluetoassurecorecoolingwithoutfurtherdegradationoftheclad. The "PotentialLoss" EAListhesameastehRGSbarrier "Loss EAL4belowand corresponds to the+~pec~ water level at the top of the active fuel. Thus, this EAL indicates a "Loss of RGS barrier and a "Potential Loss" of the Fuel Glad Barrier. This EAL appropriately escalates the emergency class to a Site Area Emergency.
88
Plant Specific EAt'>ideline (FPB)
Nine Mile Point Unit) C BWR FPB IC¹: FC3 Barrier: Fuel Claddin Type: Lo s
==
Description:==
Drywell Radiation Monitoring FC3.1 Drywell radiation monitor reading greater tha+sit~pocITQ Q99g. R/hr.
Bases:
Qggg R/hr is a value which Indicates the release of reador coolant, with elevated activity indicative of fuel damage, into the drywell. The readingWouldbe gag calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and hdine Inventory associated with a concentration of 300 pCI/gm dose equivalent 1-131 into the d~eli atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are there fore indicative of fuel damage (approximately 2/o 5'/o clad failure depending on core inventory and RCS volume). This value is higher than that specified for RCS barrier hss EAL ¹3. Thus, this EAL Indicates a hss of both fuel clad barrier and RCS barrier.
Caution: it is important to recognize that ln the event the radiation monitor ls 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 thh item.
89
e Plant Specific EAl iideline (FPB)
Nine Mlle Point United BWR FPB ICN: FC4 Barrier: Fuel Claddin Type: Loss Oescrlptlon: Other (Site-Specific) indications FC4.1 Bases:
This EAL is to cover other~4specifg indications that may fndicate loss or potential loss of the fuel clad barrier, including indications from containment air monitors or any other (sit~~~ instrumentation.
90
Plant Specific EA' >ideline (FPB)
Nine Mlle Point Unit 3 BWR FPB ICO: FC5 Barrier: Fuel laddin Type: Loss/Po . Los
==
Description:==
Emergency Director Judgement FC5.1 Any condition in the judgement of the Emergency Director that tndbates loss or potential bss of the fuel cladding barrier.
Bases:
This EAL addresses any other factors that are to be used by the Emergency Director ln 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 SGt, "Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power", for additional Informatbn.)
91
Plant Specific EAl'sideline (FPB)
Nine Mile Point Unit 0 BWR FPB IC¹: RCS1 Barrfer: RCS Tg)8: Loss a-i.a ~
Descrlptlon: RCS Leak Rate RCS1.1 f04~~ Indfcationg 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 1 0 miilirem. Thus, this EAL ls included for consistency with the Alert emergency classmcatfon.
92
Plant Specific EAI iideline (FPB)
Nine Mlle PoInt United BWR FPB IC¹: RCS1 Barrier: RCS Type: Pote Ial Los
==
Description:==
RCS Leak Rate RCS1.2 RCS leakage greater than 50 gpm inside the drywell RCS1.3 Bases:
The potential hss of RCS based on leakage is set at a level Indicative of a small breach of the RCS but which ls 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 hss of inventory is possible. Many BWRs may be unable to measure an RCS leak of this size because the leak would likely increase dtywell pressure above the drywell isolation setpoint. The system normally used to monitor leakage is typically isolated as part of the drywell holation 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 drywell is determined from site-specific hhttitttttttt 5afa Qpttrallag LRYgh4arars in the areas of the main steam line tunnel, rnala4vcbiao-goaorator, RCIC, QPCI, etc., which indicate a direct path from the RCS to areas outside primary containment.
93
Plant Specific EAL iideline (FPB)
Nine Mile Point United BWR FPB ICN: RCS2 Barrier: RCS Tg)eo Loss DesCrlptloo: Drywell Pressure RCS2.1 PI!ERE KQialpzUzi gressure ~ psig Bases:
The~~pec~cbywoQgggtitp~fgjgtfttftttfpressure is based on the diywell high pressure alarm setpolnt A higher value may be used if supporting documentation is provided which indicates the chosen value h 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
0 Plant Specific EAI. 'deline (FPB)
Nine Mile Point Unit 3 BWR FPB IC¹: RCS3 Barrier: RCS Type: Loss Descrlpttott: Drywell Radiation Monitoring RCS3.1 Drywall radiation monitor reading greater tha+sft~pecifg+ Rki Bases:
WJM dispersal of the reactor coolant noble gas and hdine inventory associated with normal operating concentrations 0. e., within T/S) into the drywell atmosphere. This reading will be less than that specified for fuel clad barrier EAL ¹3. Thus, this EAL would be indicative of a RGS leak only. If the radiation monitor reading increased to that value specified by the fuel clad barrier EAL
¹3, fuel damage would also be indicated.
However, if the site-specific physical location of the drywell radiation monitor is such that radiation from a chud 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 iideline (FPB)
Nine Mlle Potnt United BWR FPB IC¹: RCS4 Barrier: RCS Type: Loss
==
Description:==
Reactor Vessel Water Level RCS4.1 Level less than Bases:
This 'Loss" EAL is the same as "Potential Loss'fuel clad barrier EAL ¹2. Thgsitaepec~ 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 213 coverage of active fuel. This EAL appropriately escalates the emergency class to a Site Area Emergency. Thus, this EAL indicates a loss of the RCS barrier and a potential loss of the fuel clad barrier.
96
0 Plant Specific EAI sideline (FPB)
Nine Mlle Point Unit )
BWR FPB ItA: RCS5 Barrier: RCS Type: Loss
==
Description:==
Other (site-specific) indications RCS5.1 Bases:
This EAL is to cover othergsit~pecifg indications that may indicate toss or potential toss of the RGS barrier.
97
Plant Specific EAuideline (FPB)
Nine Mile Point Unit 5 BWR FPB IC¹: RCS5 Barrier: RCS Type: Po ential Loss Descrlptlotl: Other (site-specific) indications RGS5.2 Bases:
This EAL is to cover other~~pecifg indications that may indicate hss or potential loss of the RCS barrier.
98
Plant Specific EAL'Qeline (FPB)
Nine Mlle Point Unit 3 BWR FPB IC¹: RCS6 Barrier: RCS Type: LosslPot. Loss Desorlptloll: Emergency Director Judgment RCS6.1 Any condition ln 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 ln determining,whether the RCS barrier is lost or potentially lost. 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 lost or potentially lost.
(See also IC SG1, "Prolonged Loss of Offsite Power and Prolonged Loss of All Onsite AC Power,", for additional information.)
99
Plant Specific EAL 'beeline (FPB)
Nine Mile Point Unit 3 BWR FPB l(A'C1 Barrier: Prima Containment Type: Loss Oescrlptlon: Drywell Pressure PC1.1 PC1.2 Bases:
100
Plant Specific EA'.4uideline (FPB)
Nine Mile Point Unit 3 BWR FPB IC¹'C1 Barrfer: Prima Containment Type: Poten ial Loss Description t Drywell Pressure PC1.3 +~elf+
PC1.4 Exphsive mixture of exists Bases:
l "t'"""
of an explosive mixture means a hydrogen and oxygen concentrathn of at least the hwer deflagrathn limitcucvo exists.
Existence 101
Plant Specific EALsideline (FPB)
Nine Mlle Point Unit1 BWR FPB ICO: PC2 Barrier: Prima Containme . Type: Loss Desorlptloll: Containment IsoIatlon Valve Status after Containment Isolation Signal PC2.1 Jdainmmnihm Emargane Qandanaarataatnlinaa PC2.2 Intentional venting per EOPs:
PC2.3 anti altttar; Baal:tltrlhiidinahmIamparatzaaahntta thairmazimumaahrtnaratina hntahQB BmtzQuMintthamBadiathnLmhatmtlathairmartinifmaahuuarating latch Bas ceo in tE? Qr mttra araaa This EAL is intended to cover containment isolation failures alhwing a direct flow path to the environment such as failure of both MSIVs to close with open valves downstream to the turbine or to the condenser.
EKlllmtftmhaahtana4bd QEQ2J.aaamuuTmantBQRhaitagamumrtndar hilttra tn gggfjtlnn,~ 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 ls no Potential Loss'AL associated with this item.
Bahrftnnft:4 102
Plant Specific EAuideline (FPB)
Nine Mlle Paint Unit 3 BWR FPB ICA: PC3 Barrier: Prima Containment Type: Pote ial Loss Descrlptloo: Significant Radioactivity Inventory in Containment PC3.1 Bases:
asER8d 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 bss of containment, such that a General Emergency declaration is warranted.. HUREG-1228, "Source Estimatbns 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%%d, a radiation monitor reading corresponding to 20%%d fuel clad damage bag specified here.
There is no "Loss EAL associated with thh item.
103
0 Plant Specific EAL tideline (FPB)
Nine Mlle Point Unit 0 BWR FPB IC¹: PC4 - Barrier: Prirrta Containme Type: Pot ial L ss
==
Description:==
Reactor Vessel Water Level PC4.1 Bases:
I The conditions In this potential bss EAL represent imminent melt sequences whbh, if not corrected, could lead to vessel failure and increased potential for containment failure. In conjunction with the level EALs in the fuel and RGS barrier columns, this EAL will result in the declaration of a General Emergency loss of two barriers and the potential loss of a third. If the emergency operating procedures have been Ineffective in restoring reactor vessel level 'huta tht fgggf ~heal there is not a "success" path. ~
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 scenarbs, 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 will be effective should be apparent within the time provided. The Emergency Director should make the declaration as soon as it ls determined that the procedures have been, or will be ineffective.
There is no loss EAL associated with this item.
104
Plant Specific EAL iideline (FPB)
Nine Mlle Point Unit )
BWR FPB IC¹: PC5 Barrier: Prima Containme Type: Los Descrfptlon: Other (site-specNc) indications PC5.1 Bases:
This EAL is to cover other4sitmpecN+ indications that may indicate hss or potentiai toss of the containment barrier.
105
Plant Specific EA'sideline (FPB)
Nine Mlle Point Unit 3 BWR FPB 1C¹: PCS Barrier: Prima Contalnme Type: Pote lal Loss
==
Description:==
Other (site-specific) indications PC5.2 Bases: p This EAL is to cover other+~pec~ indications that may indicate loss or potential loss of the contalnmen) barrier.
106
0 Plant Specific EAl jideline (FPB)
Nine Mlle Point Unit 3 BWR FPB ICff: PC6 Barrier: Prima Containme Type: Lo s/Po Lo
>. 'a Descriptloll: Emergency Director Judgment PC6.1 Any condition in the Judgment of the Emergency Director that Indic>a>tes hss or potential Ioss of the containment barrier Lmafaziulamftatlndlnaturaamx Jnnludft; jnaaashtftatarunmnftntmtLQGhmsnnaaft Bauhiunttzulaiamtdanntasft fa9nMfing jnltlaljacrfUtailnanatttlamaaturuaaurn Bases: > ~
This EAL addresses any other factors that are to be used by the Emergency bfrectot ln> determining whether the containment barrier ls lost or potentially hst. In addition, the inability to monitor the barrier should also be incorporated in this EAL as a factor ln Emergency Dfiector judgement that the barrier may be considered lost or potentially lost. (See also IC SG1, "Prohnged Loss of Ail Offsite Power and Prohnged Loss of Ail Onsite AG Power; for additional Information.)
107
OSSI 92-402A-2-NMP 1 NMP1 Fission Product Barrier EAI. Evaluation, Rev. 0 Fission Product Barrier Evaluation Revision 0 Nia ara Mohawk Power Co Nine Mile Point Unit 1 Operations Support Services, Inc.
233 Water Street 2nd Floor Plymouth, MA 02360
OSSI 92-402A-2-NMP l NMPl Fission Product Barrier EAI. Evaluation, Rev. 0 Evaluation of NMP-1 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 example, there are five FC barrier example FALs, six RCS barrier example EALs, and six PC example 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 following list under the column labeled "NESP-007", NUMARC EALs 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-1 are identified with a yes/no under the column labeled "NMP-1".
OSSI 92-402A-2-NMP1 NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 NUMARC NMP-1 Barrier EAL ¹ ~SS ~POt. LO ~PO LOSS 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(RCS 1. 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 classification key given at the beginning of Table 3, the number of exaznple EALs, and the number of loss and potential loss conditions, the set of conditions that can yield a given emergency classification 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.
OSSI 92-402A-2-NMP1 NMP1 Fission Product Barrier EAI. Evaluation, Rev. 0 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 and 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 modified 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
OSSI 92-402A-2-NMP I NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 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 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 fission product barrier evaluation is presented in Appendix 2.
0 RECOGNITION CATEGORY F CO CO FISSION PRODUCT BARRIER DEGRADATION (0 INITIATINGCONDITION MATRIXTABLE S BWR O UNUSUAL EVENT SITE AREA EMERGENCY GENERAL EMERGENCY FU1 Any loss or any Any loss or any Loss of both fuel clad FGl Loss of any two potential loss of tential loss of either and RCS barriers containment e el clad or RCS. OR AND Potential loss of both Potential loss of third Op. Modes: Op. Modes: fuel clad and RCS barrier.
Power operauon Hot Power operation Hot OR Standby/Startup Standby/Startup Potential 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 barrier. (BWR)
Hot Shutdown Op. Modes:
Power operation Hot Standby/Startup (BWR)
Hot Shutdown O
NOTES:
- 1. Although the logic used for these initiating conditions appears overly complex, it fs necessary to reflect the following considerations:
O
~ The fuel clad barrier and the RCS barrier are weighted more heavily than the containment barrier (see Sections 3.4 and 3.8 for more information on this point). Unusual Event ICs associated with RCS and Fuel Clad barriers are addressed under System Malfunction ICs.
~ At the Site Area Emergency level, there must be some ability to dynamically assess how far present conditions are for General Emergency.
For example, ifFuel Clad barrier and RCS barrier "Loss" EALs existed, this would indicate to the Emergency Director that, in additional to offsite dose assessments, continual assessments of radfoacuve inventory and containment integrity must be focused on. If, on the other hand, both Fuel Clad barrier and RCS barrier "Potential Loss" EALs existed, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.
~ The 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. Q C
0
- 2. Fission Product Barrier ICs must be capable of addressing event dynamics. 'Ihus, the EAL Reference Tables 3 and 4 state that IMMINENT(i. p e., within 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) loss or potential loss should result in a classiflcation as ifthe affected threshold(s) axe already exceeded, particularly for the higher emergency classes.
RECOGNITION CATEGORY F CO (0
INITIATINGCONDITION M'ATRIXTABLE S BWR (0 Fuel Clad Barrier Example EALS' O Potential Loss
- 1. m la A tf e Coolant actfvfty greater than (site-spedQc) value Not Applicable
- 2. eact se Wat el Level less than (site-spedffc) value Level less than (sfte-specfffc) value dfatfon Monftorl Drywell radiation monitor readfng greater than (site-specfQc) Not Applicable R/hr
- 4. 0 c i dfcatf (site-spedfic) as applicable (site-spedQc) as applicable
- 5. e to d men Any condition in the judgment of the Emergency Dfrector that fndyfcates loss or potential loss of the fuel clad barrier Deterndne whfch combinatfon of the three barriers are lost or have a potential loss and use the following key to dassffy the event. Also. an event for multiple events could occur which result fn the condusfon that exceeding the loss or potential loss thresholds fs imminent (f. e.,
within 1 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />). In this fmmfnent loss situation, use judgment and dassffy as ffthe thresholds are exceeded.
RECOGNITION CATEGORY F CO INITIATINGCONDITION MATRIXTABLE 3 BWR RCS Barrier Example EALS' O Potential Loss
- 1. CS 1~k Rate (sfte-specific) fndfcatfon of main steam line break RCS leakage greater than 50 gpm inside the drywell OR unfsolatble primary system leakage outside drywell as indicated by area temperature or area radiation alarm Pressure greater than (site-spec!Qc) psig Not applicable
- 3. e dfatf o ito n Drywell radiation monitor reading greater than (sfte-specfQc) Not applicable R/hr 0R
- 4. t V W Level less than (site-specific) value Not applicable O (site-specific) as applfcable (site-specific) as applicable recto J d Any condition fn the]udgment of the Emergency Director that r indicates loss or potential loss of the RCS barrier Q C
0 p
y9 O
RECOGNITION CATEGORY F 0)
(6 INITIATINGCONDITION MATRIXTABLE S BWR CO Primazy Containment BarrIer Example EALs' O Rapfd unexplained decrease following fnlUal increase (site-speci Qc) psfg and fncreasfng OR OR Drywell pressure response not consistent with LOCA conditions explosive mixture exists
- 2. ta t solatfon Valve a t o t Failure of both values fn any one line to dose and downstream Not applicable pathway to the environment exfsts OR IntenUonal venting per EOPs Not applfcable OR Unfsolable primary system leakage outsIde drywell as fndfcated Not applicable by area temperature or area radfatlon alarm Not applicable Containment radiaUon monitor readfng greater than (site-specfffc)
R/hr
- 4. ea to V s 1Wate ve Not applicable Reactor vessel water level less than (site-specific) value and the maximum core uncovery time limit fs fn the unsafe region (site-specific) as applicable (site-specific) as applfcable
- 6. e e tor Jud e Any condition in the ]udgment of the Emergency Director that indicates loss or potenUal loss of the containment barrier 0
OSSI 92-402A-2-NMP1 NMPl Fission Product Barrier EAL Evaluation, Rev. 0 Table A BWR Fission Product Barrier Unusual Events NESP-007 Remarks Loss or pot. loss of PC FCXa-fess Condition not supported in PEG.
PC4b-less Condition not supported in PEG.
QCRMess 21 PQQb-less 1 PCRe-1ess 2 P-CS-loss Condition not supported in PEG.
BC6-less 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-NMP 1 NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 Table B BWR Fission Product Barrier Alerts NESP-007 Remarks s g e g g s s s u b
FCR-less 8 I~ass 7 i:::-: --'--:-':-ii'i PCS-less Subsumed in "Judgment" EAL.
8 Condition not supported in PEG.
Subsumed in "Judgment" EAL.
Loss or pot. loss of RCS RCS4-Iess 8 RCSS-Iass Condition not supported in PEG.
RCS6-1ass Subsumed in "Judgment" EAL.
15 23 Condition not supported in PEG.
Subsumed in "Judgment" EAL.
OSSI 92-402A-2-NMP1 NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 Table C BWR Fission Product Barrier Site Area Emergencies NESP-007 Remarks Loss of PC and RCS 16 18 8
Condition not supported in PEG.
Subsumed in "Judgment" EAL.
8 8
9 Condition not supported in PEG.
Subsumed in "Jud ent" EAL.
19 10 ll Condition not supported in PEG.
Subsumed in "Judgment" EAL.
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 FC and RCS FC2-yes-1 8 8
12
OSSI 92-402A-2-NMP1 NMPl Fission Product Barrier EAL Evaluation, Rev. 0 Table C BWR Fission Product Barrier Site Area Emergencies NESP-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 in "Judgment" EAL.
Subsumed in "Judgment" EAL.
Subsumed in "Judgment" EAL.
Subsumed in "Judgment" EAL.
Pot. loss of FC and loss of RCS 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.
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 in PEG.
8 13
OSSI 92-402A-2-NMP 1 NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 Table C BWR Fission Product Barrier Site Area Emergencies NESP-007 Remarks 8
8 Condition not supported in PEG.
Subsumed in "Judgment" EAL.
Condition not supported ln 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 in PEG.
Subsumed ln "Judgment" EAL.
Subsumed in "Judgment" EAL.
Subsumed in "Judgment" EAL.
Condition not supported in PEG.
Subsumed ln "Judgment" EAL.
Pot. loss of RCS and 1oss of PC 19 8
19 20 12 23 8
19 24 12 Condition not supported ln PEG.
Condition not supported in PEG.
Condition not supported ln PEG.
Condition not supported in PEG.
Condition not supported in PEG.
14
OSSI 92-402A-2-NMP 1 NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 Table C BWR Fission Product Barrier Site Area Emergencies NESP-007 Remarks Subsumed in "Judgment" EAL.
Subsumed in "Judgment" EAL.
Subsumed in "Judgment" EAL.
Subsumed in "Judgment" EAL.
Subsumed in Judgment" EAL.
A'2 Pot. loss of RCS and loss of PC
"!:PC:
Condition not supported in PEG.
Condition not supported in PEG.
-I "-iii i'"~m~t'"="':d::: '~: " ':it'i I':::: "lilies i" Condition not supported in PEG.
Subsumed in "Judgment" EAL.
Condition not supported in PEG.
Condition not supported in PEG.
21 22 13 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.
Condition not supported in PEG.
Condition not supported in PEG.
Condition not supported in PEG.
Subsumed in "Judgment" EAL.
ess Subsumed in "Judgment" EAL.
Subsumed in "Judgment" EAL.
Condition not supported in PEG.
Subsumed in "Judgment" EAL.
15
OSSI 92-402A-2-NMP1 NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 Table D BWR Piston Product Barrier General Emergencies NESP-007 Remarks Loss of FC + loss of RCS + pot. loss of PC 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.
Subsumed in "Judgment" EAL.
Subsumed in "Judgment" EAL.
16
OSSI 92-402A-2-NMP1 NMP1 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 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.
Subsumed in "Judgment" EAL 17
OSSI 92-402A-2-NMP1 NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 Table D BWR Fission Product Barrier General Emergencies NESP-007 Remarks Subsumed fn "Judgment" EAL.
Condition not supported in PEG.
Subsumed fn "Judgment" EAL 25 25 26 27 Condition not supported in PEG.
Subsumed fn "Judgment" EAL.
25 25
.26 27 Condition not supported fn PEG.
Subsumed fn "Judgment" EAL 25 25 26 27 Condftion not supported in PEG.
Subsumed in "Judgment" EAL.
25 25 26 27 Condition not supported fn PEG.
Subsumed in "Judgment" EAL.
Condition not supported in PEG.
Condition not supported in PEG.
Condition not supported fn PEG.
Condition not supported fn PEG.
Condition not supported in PEG.
Condition not supported in PEG.
Subsumed in "Judgment" EAL.
Subsumed in "Judgment" EAL.
18
OSSI 92-402A-2-NMPl NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 Table D HWR Fission Product Barrier General Emergencies NESP-007 Remarks Subsumed in "Judgnent" 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 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 19
OSSI 92-402A-2-NMP1 NMPl 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-NMP1 NMPl 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.
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 FC 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.
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.
21
OSSI 92-402A-2-NMP 1 NMPl 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 ln "Judgment" EAL.
Condition not supported in PEG.
Condition not supported ln PEG.
Condition not supported in PEG.
Subsumed in "Judgment" EAL.
Condition not supported ln PEG.
Subsumed in "Judgment" EAL.
Conditiori 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 ln PEG.
Subsumed in "Judgment" EAL.
28 Condition not supported ln PEG.
Subsumed in "Judgment" EAL.
25 Condition not supported in PEG.
Subsumed in "Judgment" EAL.
Condition not supported in PEG.
CondNon not supported in PEG.
CondNon not supported in PEG.
Subsumed in "Judgment" EAL.
Condition not supported in PEG.
22
OSSI 92-402A-2-NMP1 NMP I Fission Product Barrier EAL Evaluation, Rev. 0 Table D HWR Fission Product Barrier General Emergencies NESP-007 Remarks Subsumed in "Judgment" EAL.
CondlUon 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.
28 Condition not supported in PEG.
Subsumed in "Judgment" EAL.
28 Condition not supported in PEG.
Subsumed in "Judgment" EAL.
25 CondiUon not supported in PEG.
Subsumed in "Judgment" EAL Condition not supported ln PEG.
Condition not supported in PEG.
CondiUon not supported in PEG.
Subsumed in "Judgment" EAL.
CondiUon not supported in PEG.
Subsumed in "Judgment" EAL.
CondiUon 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 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 ln PEG.
OSSI 92-402A-2-NMP I NMP1 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.
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.
Condition not supported in PEG.
Subsumed in "Judgment" EAL 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. =
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.
Loss of PC + loss of FC + pot. loss of RCS Condition not supported in PEG.
Condition not supported in PEG.
Condition not supported in PEG.
Condition not supported in PEG.
24
QSSI 92-402A-2-NMPl NMPl Fission Product Barrier EAL Evaluation, Rev. 0 Table D BWR Fission Product Barrier General Emergencies NESP-007 Remarks 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.
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.
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.
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.
CondiUon not supported in PEG.
25
OSSI 92-402A-2-NMP1 NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 Table D HWR Fission Product Barrier General Emergencies NESP-007 Remarks Condition not supported in PEG.
'-::::"-'-'-'-.:::ii'i'i':::.::::::::::::::::-::ilia,':::::::::,:::::::::'::-::::-':": -"':--":-'-*"-":i!'~i~::i'-ii:':i:::;::i ""~i ""!'-~ii'ii'-'i"'i:::::l""i:::":8 29 Condition not supported in PEG.
io
- 30 Condition not supported in PEG.
Subsumed in "Judgment" EAL.
,::: "":-.i"-""i":-'ll"::.". I',i'l ~i" '""'"""-"'*"-'-':.i!i'i:':'iI 'l'A:-"'-'iiiii""""""-""""i'-""'i-"ii"'ii"*'i~iiii: """i 31 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 oss 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-NMP1 NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 Table D HWR Fission Product Barrier General Emergencies NESP-007 Remarks Condition'not supported in PEG.
Subsumed in "Judgment" EAL.
12 12 Condition not supported in PEG.
Subsumed in "Judgment" EAL.
PC2"..",~N::- @:,. r.;..'.:Fc<j', g<<'y. ':, <rg.;g.-,: ": ': '.:;:5 > ';.0 '".., '>,o<i N4 'x':.'.: '.q~<s'~v? ': i: dv<@c" ':y<<"..<<<5I'~i '"~M'j<<<;:yI:::,':;g .j >..>q(':
33 Condition not supported in PEG.
Subsumed in "Judgment" EAL.
""f"""""""""PCS'7 '""'-"": "' BR~' "'h'cih'WO'Wl WAF -<+~~
32 Condition not supported in PEG.
Subsumed in "Ju ent" EAL.
fi 'i': 'P'-::.": !<m Fi""-":::::ll:::::::::::::: ~~<"i:::::,:"-'-"-::::ei'>> <i<:::"'-ii::
34 Condition not supported in PEG.
Subsumed in "Judgment" EAL.
24 24 Condition not supported in PEG.
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.
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.
27
OSSI 92-402A-2-NMP1 NMPl 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, 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 in PEG.
Condition not supported in PEG.
Condition not supported fn PEG, Subsumed in "Judgment" EAL.
Subsumed fn "Judgment" EAL.
Condition not supported in PEG.
Subsumed in "Judgment" EAL Subsumed in "Judgment" EAL.
Subsumed in "Judgment" EAL.
Condition not supported fn PEG.
Subsumed fn "Judgment" EAL.
Subsumed in "Judgment" EAL.
Subsumed in "Judgment" EAL.
Condition not supported in PEG.
Subsumed in "Judgment" EAL.
Subsumed fn "Judgment" EAL.
Subsumed in "Judgment" EAL.
Condition not supported in PEG.
Subsumed in "Judgment" EAL.
Subsumed in "Judgment" EAL.
Subsumed fn "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.
Condftfon not supported in PEG.
28
OSSI 92-402A-2-NMP1 NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 Table D BWR Fission Product Barrier General Emergencies NESP-007 Remarks CondlUon 'not supported in PEG.
Condition not supported in PEG.
Condition not supported in PEG.
CondlUon 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 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.
Condition not supported ln PEG.
CondiUon 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.
CondlUon 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.
Condftion not supported ln PEG.
Condition not supported in PEG.
Condition not supported ln PEG.
Condition not supported in PEG.
Condition not supported in PEG.
29
0 OSSI 92-402A-2-NMP1 NMP1 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 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.
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 ln 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 ln PEG.
Condition not supported in PEG.
Condition not supported in PEG.
Condition not supported in PEG.
35 35 35 35 35 Subsumed in "Judgment" EAL 35 35 35 35 35 Subsumed in "Judgment" EAI,
OSSI 92-402A-2-NMP1 NMP1 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-NMP 1 NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 Table D BWR 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-NMP1 NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 Table D HWR Fission Product Barrier General Emergencies NESP-007 Remarks 24, 28 Subsumed in "Judgment" EAL.
35 35 35 35 35 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 ln PEG.
Condition not supported ln PEG.
Condition not supported ln 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 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 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-NMP 1 NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 Table D HWR Fission Product Barrier General Emergencies MES P-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" 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" EAL.
Subsumed in "Judgment" EAL.
Subsumed in "Judgment" EAL.
Subsumed in "Judgment" EAL.
OSSI 92-402A-2-NMP I NMPI Fission Product Barrier EAI Evaluation, Rev. 0 Table D BWR Fission Product Barrier General Emergencies NESP-007 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 EML¹ PC2.3 is a loss of the primary containment, EAL¹ RCS1.3 requires an Alert declaration at the maximum ~norm I 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, 5ML¹s 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 classification, 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 35
OSSI 92-402A-2-NMP l NMPl Fission Product Barrier EAI Evaluation, Rev. 0 Table D HWR Fission Product Barrier General Emergencies NESP-007 Remarks of active fuel. Since FC2.1 will always be reached before PC4.1, EAL¹ PC4.1 is unnecessary and can be deleted.
- 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 classiQcation. 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 TAF 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 classification. 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 classiQcation is unnecessary and can be deleted.
- 10. EAL¹ FC3.1 and EAL¹ RCS3.1 identify drywell radiation as a condition requiring an emergency classiQcation. 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 classiQcation 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.
OSSI 92-402A-2-NMP1 NMP1 Fission Product Barrier EAI Evaluation, Rev. 0 Table D BWR Fission Product Barrier General Emergencies NESP-007 Remarks
- 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.
- 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 EAL¹ 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
.FC1.l. 1'¹ FC1.1 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 FC1.1 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 failure 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 37
OSSI 92-402A-2-NMP I NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 Table D BWR Fission Product Barrier General Emergencies NESP-007 Remarks the loss of the RCS boundary. By definition, 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.
- 22. To intentionally vent the primary containment in accordance with the EOPs, two fission product barriers must have been lost and a third barrier is about to be lost due to venting. By definition, 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 definition, 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 contairunent 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, EAL¹ 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.
QSSI 92-402A-2-NMP1 NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 Table D BWR Fission Product Barrier General Emergencies NESP-007 Remarks
- 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.
- 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 deQnition, 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 F~ FCl.l, FC2.1 or FC3.1 has previously been evaluated as justification of General Emergency. Therefore this combination of conditions is redundant and can be deleted.
39
OSSI 92-402A-2-NMP I NMP1 Fission Product Barrier EAL Evaluation, Rev. 0 Aypendiz 2 Sumxnaxy of Fission Product Barrier Evaluation The following summarizes the EALs which resulted from the analysis performed of the fission product barrier methodology of NUIKARC-007 for NMP-1:
~ 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 EAL Evaluation, Rev. 0 Aypendix 2 Suxnmazy of Fission Product Barrier Evaluation PC1.3-pot. loss PC1.4-pot. loss PC3.1-pot. loss PC4.1-pot. loss PC2.l-loss + FC1.1-loss, FC2.1-loss or FC3.1-loss PC2.3-loss + FC1.l-loss, FC2.1-loss or FC3.1-loss Emergency Director Judgement
OSSI 92-402A-3-BWR BWR EAL Binning Document, Rev. 0 EAL Binain Document gg~i~n New York Power Authori J. A. Mzpatrfck Nuclear Power Plant Indian Point Nuclear Power Plant Unit 3 Nia ara Mohawk Power Co oration Nine MiLe Point Unit 1 Nine MQe 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 92-402A-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) 1.3 Containment Radiation RCS3.1 (Alert)
FC3.1 (SAE)
PC3.1 (General) 1.4 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 92-402A-3-BWR BWR EAL Binning Document, Rev. 0 8.0 Containment 8.1 Primary Containment Pressure RCS2.1 (Alert)
FC1.1 + RCS2.1 (SAE)
PC1.3 (General)
PC2.2 (General) 8.2 Suppression Pool Temperature SS4.1 (SAE) 8.8 Combustible Gas Concentration SS5.2 (SAE)
PC1.4/PC2.2 (General) 8.4 Contaixunent Isolation Status PC2.1 (SAE)
PC2.1+ FC1.1 (General)
PC2.1 + FC2.1 (General)
PC2.1 + FC3.1 (General) 4.0 Secondary Containment 4.1 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 EGluent Monitors AU1.1 AA1.1 AS1.1 AG1.1 5.2 Dose Projections/ Environmental Measurements AU1.2 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 7.8 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 8.3 Man-Made Events HU1.4 HU1.5 HU3.1 HU3.2 HA1.5 HA3.1 HA3.2 8.4 Natural 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 PLAN MAINTENANCE PROCEDURE EPHP-EPP-0101 REVISION 00 UNIT 1 EMERGENCY CLASSIFICATION TECHNICAL BASES TECHNICAL SPECIFICATION REQUIRED Approved by:
R. B. Abbott Plant Manager Unit 1 Date Effective Date:
NOT TO BE USED AFTER SUBJECT TO PERIODIC REVIEW
LIST OF EFFECTIVE PAGES
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Hay 1994 Page i EPHP-EPP-0101 Rev 00
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Hay 1994 Page ii EPHP-EPP-0101 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 4.0 DEFINITIONS . ~ t ~ ~ 1
5.0 REFERENCES
AND COMMITMENTS 2 6.0 RECORD REVIEW AND DISPOSITION . ~ ~ ~ ~ ~ ~ ~ 2 I RODUCTIO PURPOSE . ~ ~ ~ ~ ~ ~ 3 DISCUSSION ~ ~ ~ ~ ~ ~ ~ ~ 3 TECHNICAL BASES 1.0 REACTOR FUEL ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 7 2.0 REACTOR PRESSURE VESSEL (RPV) 20 3.0 PRIMARY 'ONTAINMENT (PC) 27 4.0 SECONDARY CONTAINMENT (SC) 36 5.0 RADIOACTIVITY RELEASE . 42 6.0 ELECTRICAL FAILURES . 51 7.0 EQUIPMENT FAILURES 59 8.0 HAZARDS . 69 9.0 OTHER . 87 ATTACHMENT 1: UNIT 1 EMERGENCY ACTION LEVEL TECHNICAL BASES . 3 ATTACHMENT 2: WORD LIST/DEFINITIONS . 94 May 1994 Page iii EPMP-EPP-0101 Rev 00
1.0 PURPOSE To describe the technical bases for the emergency action levels at Unit l.
2.0 PRIMARY RESPONSIBILITY 2.1 Emer enc Pre aredness Grou Monitor/solicit any changes to the technical bases of each emergency action level.
Assess these changes for potential impact on the emergency action level.
Maintain the emergency action level technical bases, EPIP-EPP-01, and the Emergency Action Level Matrix/Unit l.
3.0 PROCEDURE 3.1 Emer enc Pre aredness Grou 3.1.1 Maintain a matrix of technical bases references for each emergency action level.
3.1.2 Evaluate each technical bases reference change for impact on the affected emergency action level.
3.1.3 Modify EPIP-EPP-Ol, Emergency Action Level (EAL) Matrix/Unit 1 and Attachment 1 of this procedure, as needed.
- 4. 0 DEFINITIONS See Attachment 2.
May 1994 Page 1 EPMP-EPP-0101 Rev 00
5.0 REFERENCES
AND COMMITMENTS 5.1 Licensee Documentation None 5.2 Standards Re ulations and Codes
~ NUMARC NESP-007, Methodology for Development of Emergency Action Levels.
5.3 Policies Pro rams and Procedures
~ EPIP-EPP-01, Classification of Emergency Conditions at Unit.
5.4 Su lemental References
~ Nine Mile Point Unit 1, Plant-Specific EAL Guideline 5.5 Commitments None 6.0 RECORD REVIEW AND DISPOSITION None Hay 1994 Page 2 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 UNIT 1 EMERGENCY ACTION LEVEL TECHNICAL BASES PURPOSE The purpose of this document is to provide an explanation and rationale for each of the emergency action levels (EALs) included in the EAL Upgrade Program for Nine Mile Point 1 (NMP-1). It is also intended 'to facilitate the review process of the NMP-1 EALs and provide historical documentation for future
.reference. This document is also intended to be utilized by those individuals responsible for implementation of EPIP-EPP-Ol "Classification of Emergency Conditions Unit 1" as a technical reference and aid in EAL interpretation.
DISCUSSION EALs are the plant-specific indications, conditions or instrument readings which are utilized to classify emergency conditions defined in the NMP-1 Emergency Plan.
The revised EALs were derived from the Initiating Conditions and example EALs given in the NMP-1 Plant-Specific EAL Guideline (PEG). The PEG is'he NMP-1 plant interpretation of the NUMARC methodology for developing EALs.
Many of the EALs derived from the NUMARC methodology are fission product barrier based. That is, the conditions which define the EALs are based upon loss or potential loss of one or more of the three fission product barriers.
The primary fission product barriers are:
A. Reactor Fuel Claddin FC : The fuel cladding is comprised of the zirconium tubes which house the ceramic uranium oxide pellets along with the end plugs which are welded into each end of the fuel rods.
B. Reactor Coolant S stem RCS : The RCS is comprised of the reactor vessel shell, vessel head, CRD housings, vessel nozzles and penetrations and all primary systems directly connected to the RPV up to the outermost primary containment isolation valve.
C. Primar Containment PC : The primary containment is comprised of the drywell, suppression chamber (torus), the interconnections between the two, and all isolation valves required to maintain primary containment integrity under accident conditions.
Although the secondary containment (reactor building) serves as an effective fission product barrier by minimizing ground level releases, it is not considered as a fission product barrier for the purpose of emergency classification.
The following criteria serves as the bases for event classification related to fission product barrier loss:
May 1994 Page 3 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont)
Unusual Event:
Any loss or potential loss of containment Alert:
Any loss or any potential loss of either fuel clad or RCS Site Area Emer enc :
Any loss of both fuel clad and RCS or Any potential loss of both fuel clad and RCS or Any potential loss of either fuel clad or RCS with a loss of any additional barrier General Emer enc :
Loss of any two barriers with loss or potential loss of a third Those EALs which reference one or more of the fission product barrier Initiating Condition designators (FC, RCS and PC) in the PEG Reference section of the technical bases are derived from the Fission Product Barrier Analysis.
The analysis entailed an evaluation of every combination of the plant specific barrier loss/potential loss indicators applied to the above criteria.
Where possible, the EALs have been made consistent with and utilize the conditions defined in the NHP-1 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.
May 1994 Page 4 EPHP-EPP-0101 00 'ev
ATTACHMENT 1 (Cont)
DISCUSSION (Cont)
The EALs are grouped into nine categories to simplify their presentation and to promote a rapid understanding by their users. These categories are:
- 1. Reactor Fuel
- 5. Radioactivity Release
- 6. . Electrical Failures
- 7. Equipment Failures
- 8. Hazards
- 9. Other Categories 1 through 5 are primarily symptom based. The symptoms are indicative of actual or potential degradation of either fission product barriers or personnel safety.
Categories 6, 7 and 8 are event based. Electrical Failures are those events associated with losses of either AC or vital DC electrical power. Equipment Failures are abnormal and emergency events associated with vital plant system failures, while Hazards are those non-plant system related events which have affected or may affect plant safety.
Category 9 provides the Emergency Director (Shift Supervisor) the latitude to classify and declare emergencies based on plant symptoms or events which in his judgment warrant classification. This judgment includes evaluation of loss or potential of one or more fission product barriers warranting emergency classification consistent with the NUMARC barrier loss criteria.
Categories are further divided into one or more subcategories depending on the types and number of pl'ant 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).
May 1994 Page 5 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont).
DISCUSSION (Cont)
For each EAL, the following information is provided:
Classification: Unusual Event, Alert, Site'Area Emergency, or General Emergency Operating Mode Applicability: One or more of the following plant operating conditions are listed: Power Operation, Startup/Hot Standby, Hot Shutdown, Cold Shutdown, Refuel and Defueled EAL: Description of the condition or set of conditions which comprise the EAL Basis: Description of the rationale for the EAL PEG Reference(s): PEG IC(s) and example EAL(s) from which the EAL is derived Basis Reference(s): ,Source documentation from which the EAL is derived The identified operating modes are defined as follows:
Power 0 erations Reactor is critical and the mode switch is in RUN.
St rtu ot Standb This mode is subsumed in the Power Operations mode.
Hot Shutdown Mode switch is in SHUTDOWN and reactor coolant temperature is )212 F.
Cold Shutdown Mode switch in SHUTDOWN and reactor coolant temperature is g212 'F.
Refuel Mode switch in REFUEL.
Defueled RPV contains no irradiated fuel.
May 1994 Page 6 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 1.0 EACTOR 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 offs'ite radiological consequences.
The following parameters/indicators are indicative of possible fuel failures:
Coolant Activ't : 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.
~ff will 11-: 1 tt 1 t tttty, yf 1ftt release fission products to the reactor coolant. Those products which are gaseous or volatile in nature will be carried over with the steam and eventually be detected by the air ejector off-gas radiation monitors.
Containment Radiation Monitors: Although not a direct indication or measurement of fuel damage, exceeding predetermined limits on containment high range radiation monitors under LOCA conditions is indicative of possible fuel failures. In addition, this indicator is utilized as an indicator of RCS loss and potential containment loss.
Other Radiation Monitors: Other process and area radiation monitoring systems are specifically designed to provide indication of possible fuel damage such as Area Radiation Monitoring Systems.
Refuelin Accidents: Both area and process radiation monitoring systems designed to detect fission products during refueling conditions as well as visual observation can be utilized to indicate loss or potential loss of spent fuel cladding integrity.
May 1994 Page 7 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 1.0 REACTOR FUEL Coolant Activit 1.1.1 Unusual Event Coolant activity > 25 pCi/gm I-131 equivalent 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 Refer ence(s):
SU4.2 Bases Reference(s):
- 1. -
Radiological Technical Specifications, Appendix A to Facility Operating License No. DPR-63, Article 3.2.4.a 1.1.2 Alert Coolant activity > 300 pCi/gm I-131 equivalent Node Applicability:
Power Operation, 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.
May 1994 Page 8 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont).
1.1. 2 (Cont)
PEG Reference(s):
FC1.1 Basis Reference(s):
- 1. General Electric NED0-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions 1.2 Off- as Activit 1.2.1 Unusual Event Valid offgas radiation ~ hi-hi alarm 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. This offgas radiation level corresponds to the Technical Specification allowable limit of 500,000 pCi/sec (recombiner discharge gross noble gases beta and/or gamma). The hi-hi alarm setpoint has been conservatively selected because it is operationally significant and is readily recognizable by Control Room operating staff. The system isolates when both RN-12A and 12B alarm.
The hi-hi offgas radiation alarm is nominally set in accordance with the Offsite Dose Calculation Manual.
PEG Reference(s):
SU4.1 Basis Reference(s):
- 1. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Amendment 66, Article 3.6.15.c
- 2. Nl-ARP-H1, annunciator Hl-2-7 May 1994 Page 9 EPMP-EPP-0101 Rev 00
ATTACHHENT 1 (Cont) 1.3 Containment Radiation 1.3.1 Alert Drywell radiation > 20 R/hr Node Applicability:
Power Operation, Hot Shutdown Basis:
The drywell radiation reading is a value which indicates the release of reactor coolant to. the drywell. The reading is calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal 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.
The calculation referenced resulted in an EAL value of 24 R/hr.
However, a value of 20 R/h was selected as it is observable on existing instrumentation.
It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. Drywell High Range Radiation Monitors have a range of 0 to ES R/hr on recorder RR 201.7-36C pen 1 and '2. They are installed in the following drywell locations:
RAm 201.7-36 Az 340 , El 263'"
RAm 201.7-37 Az 310', EL 301'"
PEG Reference(s):
RCS3.1 Basis Reference(s):
- 1. Nl-RG197-EIL1, Important Design Features of Regulatory Guide 1.97 Instruments
- 2. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Amendment 72, 76, Table 3.6.11-1
- 3. Calculation 1H21C003, Rev. 0 Hay 1994 Page 10 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 1.3.2
~ ~ Site Area Emer enc Drywell radiation > 3000 R/hr Mode Applicability:
Power Operation, Hot Shutdown Basis:
The drywell radiation reading is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell. The reading is calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 pCi/gm dose equivalent I-131 into the drywell atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations allowed within Technical Specifications (including iodine spiking) and are therefore indicative of fuel damage (approximately 2X - 5X clad failure depending on core inventory and RCS volume). The reading is higher than that specified for EAL 1.3.1 and, thus, this EAL indicates a loss of both the fuel clad barrier and the RCS -barrier.
The calculation referenced resulted in an EAL value of 3090 R/hr.
However, a value of 3000 R/hr was selected as it is observable on existing instrumentation.
It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. Drywell High Range Radiation Monitors have,a range of 0 to E8 R/hr on recorder RR 201.7-36C pen 1 and 2. They are installed in the following drywell locations:
RAm 201.7-36 Az 340', El 263'"
RAm 201.7-37 Az 310', EL 301'"
PEG Reference(s):
FC3.1 Basis Reference(s):
- 1. Nl-RG197-EILl, Important Design Features of Regulatory Guide 1.97 Instruments
- 2. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Amendment 72, 76, Table 3.6.11-1
- 3. Calculation 1H21C003, Rev. 0 Hay 1994 Page ll EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 1.3.3 General Emer enc Drywell radiation ~ 4.0E6 R/hr Node Applicability:
Power Operation, Hot Shutdown Basis:
The drywell radiation reading is a value which indicates significant fuel damage well in excess of that required for loss of the RCS barrier and the fuel clad 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, this amount of activity in containment could have severe consequences if released.
It is, therefore, prudent to treat this as a potential loss of the containment barrier and upgrade the emergency classification to a General Emergency.
The calculation referenced resulted in an EAL value of 3.9E6 R/hr.
However, a value of 4.0E6 R/hr was selected as it is observable on existing instrumentation.
It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. Drywell High Range Radiation Monitors have a range of 0 to E8 R/hr on recorder RR 201.7-36C pen 1 and 2. They are installed in the following drywell locations:
RAm 201.7-36 Az 340', El 263'"
RAm 201.7-37 Az 310', EL 301'"
PEG Reference(s):
PC3.1 Basis Reference(s):
- 1. Nl-RG197-EIL1, Important Design Features of Regulatory Guide 1.97 Instruments
- 2. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Amendment 72, 76, Table 3.6. 11-1
- 3. Calculation 1H21C003, Rev. 0 Hay 1994 Page 12 EPMP-EPP-0101 Rev 00
ATTACHHENT 1 (Cont) 1.4 Other Radiation Nonitors 1.4.1 Unusual Event Any sustained ARH reading > 100 x alarm (OP-50A) or offscale hi resulting from an uncontrolled process Node Applicability:
All Basis:
Valid elevated area radiation levels usually have long lead times relative to the potential for radiological release beyond the site boundary, thus impact to public health and safety is very low.
This EAL addresses unplanned increases in radiation levels inside the plant. These radiation levels represent a degradation in the control of radioactive material and a potential degradation in the level of safety of the plant. Area radiation levels above'100 times the alarm setpoint have been selected because they are readily identifiable on ARH instrumentation. The 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 ARHS whose upper range limits are less than 100 times the alarm setpoint, a value of offscale high is used. This EAL escalates to an Alert, level of safe plant operation.
if the'ncreases impair the PEG Reference(s):
AU2.4 Basis Reference(s):
- 1. Nl-EOP-5/6, Secondary Containment Control / Radioactivity Release Control
- 2. EP I P-EPP-13
- 3. OP-50A, Area Radiation Honitoring System, Attachments 2 and 3 Hay 1994 Page 13 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 1.4.2 Alert Sustained RB Vent Monitor RN07A5 or B5 > 5 mR/hr OR any sustained refuel floor rad monitor > 8.0 R/hr or offscale hi, Table 1.1 Table 1.1 Refuel Floor Rad Monitors West End of Shield Wall, RB 340 (¹18)
Rx Bldg. East Wall El 340'¹25)
Refuel Bridge (high range) (Process Mon.)
Refuel Bridge (low range) (¹29)
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,9.0.would occur via Emergency Director judgment in EAL Category The basis for the reactor building ventilation monitor setpoint (5 mR/hr) is a spent fuel handling accident and is, therefore, appropriate for this EAL.
May 1994 Page 14 EPMP-EPP-0101 Rev 00
ATTACHHENT 1 (Cont).
1.4.2 (Cont)
Area radiation levels on the refuel floor at or above the Haximum Safe Operating value (8.0 R/hr) are indicative of radiation fields which may limit personnel access. Access to the refuel floor is required in order to visually observe water level in the spent fuel pool. Without access to the refuel floor, it would not be possible to determine the applicability of EAL 1.5.2. For those radiation monitors whose upper range limits are less than 8.0 R/hr, a value of offscale high is used.
PEG Reference(s):
AA2. 1 Bases Reference(s):
- 1. NUREG-0818, Emergency Action Levels for Light Water Reactors
.2. NUREG/CR-4982, Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82, July 1987
- 3. NRC Information Notice No. 90-08, KR-85 Hazards from Decayed Fuel
- 4. Nl-ARP-L1, annunciator Ll-4-3
- 5. Niagara Hohawk Power Corporation Hemo File Code NHP31027, Exposure Guidelines for Unusual/Accident Conditions 1.4.3 1llert Sustained area radiation levels > 15 mR/hr in either:
Control Room OR Central Alarm Station (CAS) and Secondary Alarm Station (SAS)
Node Applicability:
All Hay 1994 Page 15 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 1.4.3 (Cont)
Basis:
This EAL addresses increased radiation levels that impede necessary access to operating stations requiring continuous occupancy to maintain safe plant operation or perform a safe plant shutdown. Areas requiring continuous occupancy include the Control Room, the central alarm station (CAS) and the secondary security alarm station (SAS).
The security alarm stations are included in this EAL because of their importance to permitting access to areas required to assure safe plant operations.
The value of 15 mR/hr is derived from the GDC 19 value of 5 rem in 30 days with adjustment for expected occupancy times. Although Section III.D.3 of NUREG-0737, "Clarification of TMI Action Plan Requirements", provides that the 15 mR/hr value can be averaged over the 30 days, the value is used here without averaging. A 30 day implies an event potentially more significant than an Alert. 'uration
,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 NMP-1 due to a radioactivity release or radiation shine resulting from a major accident at the NMP-2 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.).
PEG Reference(s):
AA3.1 Basis Reference(s):
- 1. GDC 19
- 2. NUREG-0737, Clarification of THI Action Plan Requirements",
Section III.D.3 May 1994 Page 16 EPMP-EPP-0101 Rev 00
STIA II ENT lC t) 1.4. 4
~ ~ ~lett 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 Safet Function Areas Reactor Building
. Turbine Building Screen and Pump House Off Gas 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. This bases of the value is described in NHPC memo File Code NNP31027 "Exposure Guidelines For Unusual/Accident Conditions". The areas selected are consistent with those listed in other EALs and represent those structures which house systems and equipment necessary for the safe operation and shutdown of the plant.
It is the impaired ability to operate the plant that results in the actual or potential degradation of the level of safety of the plant.
The cause or magnitude of the increase in radiation levels is not a concern of this EAL. The Emergency Director must consider the source or cause of the increased radiation levels and determine 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 NNP-1 due to a radioactivity release or radiation shine resulting from a major accident at the NNP-2 or JAFNPP. Such a declaration would be appropriate if the increase impairs safe plant operation.
May 1994 Page 17 EPHP-EPP-0101 Rev 00
STTACHIIENT 1 lC t) 1.4.4 (Cont)
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. NUREG-0818, Emergency Action Levels for Light Water Reactors
- 2. NUREG/CR-4982, Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82, July 1987
- 3. NRC Information Notice No. 90-08, KR-85 Hazards from Decayed Fuel
- 4. Nl-SOP-20, Loss of SFP/Rx Cavity Level/Decay Heat Removal
'.5. Niagara Mohawk Power Corporation Memo File Code NHP 31027, Exposure Guidelines for Unusual/Accident Conditions 1.5 efue 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 Basis:
The above event has a long lead time relative to the potential for radiological release outside the site boundary, thus impact to public health and safety is very low. However, in light of recent industry events, classification as an Unusual Event is warranted as a precursor to a more serious event.
The spent fuel pool low water level alarm setpoint is actuated by LS-26C which alarms at El 338'". The definition of "... cannot be restored and maintained above ..." allows the operator to visually observe the low water level condition, if possible, and to attempt water level restoration instructions as long as water level remains above the top of irradiated fuel. Water level restoration instructions are performed in accordance with procedure Nl-SOP-20, Loss of SFP/Rx Cavity Level/Decay Heat Removal.
May 1994 Page 18 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont).
1.5.1 (Cont)
When the fuel transfer canal is directly connected to the spent fuel pool and reactor cavity, there could exist the possibility of uncovering irradiated fuel in the fuel transfer canal. Therefore, this EAL is applicable for conditions in which irradiated fuel is being transferred to and from the RPV and spent fuel pool.
PEG Reference(s):
AU2.1 Basis Reference(s):
None 1.5.2 alert
~
Report of visual observation of irradiated fuel uncovered 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 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."
Hay 1994 Page 19 EPHP-EPP-0101 Rev 00
1.5.2 (Cont)
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 irr adiated 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. Nl-SOP-20, Loss of SFP/Rx Cavity Level/Decay Heat Removal, provides appropriate instructions to report a visual observation of irradiated fuel uncovery.
This EAL applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage.
PEG Reference(s):
AA2.2 Basis Reference(s):
- 1. NUREG-0818, Emergency Action Levels for Light Water Reactors
- 2. NUREG/CR-4982, Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82, July 1987
- 3. NRC Information Notice No. 90-08, KR-85 Hazards from Decayed Fuel
- 4. Nl-SOP-20, Loss of SFP/Rx Cavity Level/Decay Heat Removal 2.0 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.
May 1994 Page 20 EPMP-EPP-0101 Rev 00
ATTACHMENT I (Cont) 2.0 (Cont)
There are two RPV parameters which are indicative of conditions which may pose a threat to RPV or fuel cladding integrity:
~ RPV Water Level: RPV water level is directly related to the status of adequate core cooling, and therefore fuel cladding integrity. Excessive ( > Tech. Spec.) reactor coolant to drywe11 leakage indications are utilized to indicate potential pipe cracks which may propagate to an extent threatening fuel clad, RPV and primary containment integrity. Conditions under which all attempts at establishing adequate core cooling have failed require primary containment flooding.
~ Reactor Power Reactivit Control: The inability to control reactor power below certain levels can pose a direct threat to reactor fuel, RPV and primary containment integrity.
2.1 PV Water Level
.E.!..! ~E1 E Unidentified drywell leakage ~ 10 gpm OR Reactor coolant to drywell identified leakage > 25 gpm Node Applicability:
Power Operation, Hot Shutdown Basis:
The conditions of this EAL may be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for 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.
Only operating modes in which there is fuel in the reactor coolant system and the system is p} essurized are specified.
May 1994 Page 21 EPMP-EPP-0101 Rev 00
BTTACIIIIENT I (C t) 2.1.1 (Cont)
PEG Reference(s):
SU5.1 Basis Reference(s):
None 2.1.2 Site Area Emer enc RPV water level cannot be restored and maintained > -84 in. (TAF)
Node Applicability:
Power Operation, Hot Shutdown, Cold Shutdown, Refuel Basis:
The RPV water level used in this EAL is the top of active fuel (TAF).
This value corresponds to the level which is used in EOPs to indicate challenge to core cooling and loss of the fuel clad barrier. This is the minimum water level to assure core cooling without further degradation of the clad. Severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured water level is not maintained above TAF.
if RPV Uncovery of the fuel irrespective of the event that causes fuel uncovery is justification alone for declaring a Site Area Emergency.
This includes events that could lead to fuel uncovery in any plant operating mode including cold shutdown and refuel. Escalation to a General Emergency occurs through radiological effluence addressed in EAL 1.3.3 for drywell radiation and in the EALs defined for Category 5.0, Radioactivity Release.
PEG Reference(s):
SS5.1 FC2.1 Bases Reference(s):
- 1. Nl-ODP-PR0-0302, EOP Technical Bases May 1994 Page 22 EPMP-EPP-0101 Rev 00
ATTACHHENT 1 (Cont) 2.1.3 General Emer enc Drywell Flooding required Node Applicability:
Power Operation, Hot Shutdown Basis:
The condition in this EAL represents a potential for imminent melt sequences which, if not corrected, could lead to failure RPV and increased potential for primary containment failure. If the EOPs have been ineffective in restoring RPV water level above the top of active fuel, loss of the fuel clad barrier may be imminent. Therefore, declaration of a General Emergency is appropriate when entry to the Drywell Flooding EOP is required.
PEG Reference(s):
PC4.1 Basis Reference(s):
- 1. Nl-ODP-PR0-0302, EOP Technical Bases 2.2 Reactor Power Reactivit Control 2.2.1 ert All immediate manual scrams fail to shut down the reactor Node Applicability:
Power Operation Nay 1994 Page 23 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 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 inta 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. Nl-ODP-PR0-0302, EOP Technical Bases
.'2.2.2 S te rea Eme e c All immediate manual scrams fail to shut down the reactor AND Boron injection is required Node Applicability:
Power Operation 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.
May 1994 Page 24 EPMP-EPP-0101 Rev 00
ATTACHHENT 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, Liquid Poison system 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. NI-ODP-PR0-0302, EOP Technical Bases 2.2.3 Gene al Emer enc All immediate manual scrams fail to shut down the reactor AND RPV water level cannot be restored and maintained > -108 in.
Mode Applicability:
Power Operation Basis:
Under the conditions of this EAL, the efforts to bring the reactor subcritical have, been unsuccessful and, as a result, the reactor is producing more heat than the. maximum decay heat load for which the safety systems were designed.
Hay 1994 Page 25 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (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 (-108 in.). This RPV water level is used in the EOPs to define the lowest RPV water level in a failure-to-scram event above which adequate core cooling can be maintained. 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. Nl-ODP-PR0-0302, EOP Technical Bases 2.2.4 General Emer enc All immediate manual scrams fail to shut down the reactor AND Torus temperature and RPV pressure cannot be maintained < HCTL Node Applicability:
Power Operation Basis:
Under the conditions of this EAL, the efforts to bring the reactor subcritical have been unsuccessful and, as a result, the reactor is producing more heat than the maximum decay heat load for which the safety systems were designed.
May 1994 Page 26 EPMP-EPP-0101 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 containment failure.
In this situation, core degradation can occur rapidly For this reason, the General Emergency declaration is intended to be anticipatory of the loss of two fission product barriers and a potential loss of a third thus permitting the maximum offsite intervention time.
An immediate manual scram is any set of actions by the reactor operator at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor subcritical including manual scram push buttons, ARI and mode switch.
PEG Reference(s):
SG2.1 Basis Reference(s):
- 1. NI-ODP-PR0-0302, EOP Technical Bases 3.0 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 27 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont)
(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.
~ Primar Containment Pressure: Excessive primary containment pressure is also indicative of either primary system leaks into containment or loss of containment cooling function. Primary containment pressures at or above specified limits pose a direct threat to primary containment integrity and the pressure suppression function.
~ Torus Tem erature: Excessive torus water temperatures can result in a loss of the pressure suppression capability of containment and thus be indicative of severely degraded RPV and containment conditions.
~ Combustib e G s Concentrations: The existence of combustible gas concentrations in containment pose a severe threat to containment integrity and are indicative of severely degraded reactor core and/or RPV conditions.
~ Containment Isolation Status: The existence of an unisolable steam line break outside containment constitutes a loss of containment integrity as well as a loss of RCS boundary. Should a loss of fuel cladding integrity occur, the potential for release of large amounts of radioactive materials to the environment exists.
3.1 Containment Pressure
- 3. 1. 1 alert Drywell pressure cannot be maintained < 3e5 psig due to coolant leakage Mode Applicability:
Power Operation, 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.
Hay 1994 Page 28 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 3.1.1 (Cont)
PEG Reference(s):
RCS2.1 Basis Reference(s):
- 1. Nl-ARP-F1, annunciator 1-5
- 2. Nl-ARP-F4, annunciator 1-4
- 3. Nl-EOP-4, Primary Containment Control 3.1.2 Site Area Emer enc Drywell pressure cannot be maintained < 3.5 psig AND Coolant activity > 300 pCi/gm Node Applicability:
Power Operation, 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 29 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 3.1.2 (Cont)
Bases Reference(s):
- 1. Nl-ARP-Fl, annunciator 1-5
- 2. Nl-ARP-F4, annunciator 1-4
- 3. General Electric NED0-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions
- 4. Nl-EOP-4, Primary Containment Control 3.1.3 General Emer enc Primary containment venting is required due to PCPL Node Applicability:
Power Operation, 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
-bases 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 Bases Reference(s):
- 1. Nl-ODP-PR0-0302, EOP Technical Bases May 1994 Page 30 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 3.2 Torus Tem erature 3.2.1 Site Area Emer enc Torus temperature and RPV pressure cannot be maintained < HCTL (non-ATWS)
Node Applicability:
Power Operation, Hot Shutdown Basis:
This EAL addresses complete loss of functions, including ultimate heat sink, required for hot shutdown with the reactor at pressure and temperature. Under these conditions, there is an actual major failure of a system intended for protection of the public. Thus, declaration of a Site Area Emergency is warranted.
Functions required for hot shutdown consist of the ability to achieve reactor shutdown and to discharge decay heat energy from the reactor to the ultimate heat sink. Inability to remove decay heat energy is reflected in an increase in torus temperature. Elevated torus temperature is addressed by the Heat Capacity Temperature Limit (HCTL). The HCTL is a function of RPV pressure and torus water temperature. If RPV pressure and torus temperature cannot be maintained below the HCTL, primary containment integrity is challenged and declaration of a Site Area Emergency is warranted.
PEG Reference(s):
SS4.1 Basis Reference(s):
- 1. Nine Mile Point Nuclear Station Unit 1 Appendix 'R'eview Safe Shutdown Analysis, Figure V-1 Addresses: "Hot Shutdown Systems" "Functional Perf. Criteria Req. for Station Shutdown" May 1994 Page 31 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 3.3 Combustible Gas Concentration 3.3.1 Site Area Emer enc
~ 4A Hz exists in DW or torus Node Applicability:
Power Operation, Hot Shutdown Basis:
4X hydrogen concentration is the lowest hydrogen concentration which, in the presence of sufficient oxygen, can support upward flame propagation. This hydrogen concentration is generally considered the lower boundary of the range in which localized deflagrations may occur. To generate such a concentration of combustible gas, loss of both the fuel clad and RCS barriers must have occurred. Therefore, declaration, of,a Site Area Emergency is warranted.
If hydrogen concentrations increase in conjunction with the presence of oxygen to global deflagration levels (i.e. ~ 6X hydrogen and 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. Nl-ODP-PR0-0302, EOP Technical Bases 3.3.2 General Emer enc Primary containment venting is required due to combustible gas concentrations Node Applicability:
All May 1994 Page 32 EPMP-EPP-0101 Rev 00
ATTACHNENT 1 (Cont).
3.3.2 (Cont)
Basis:
6X hydrogen concentration in the presence of 5X oxygen concentration is the lowest concentration at which a deflagration inside of the primary containment could occur. When hydrogen and oxygen concentrations reach or exceed combustible limits, imminent loss of the containment barrier exists. To generate such levels of combustible gas, loss of the fuel clad and RCS barriers must have occurred. Venting of the containment irrespective of the offsite radioactive release rate is required by EOPs for this condition.
PEG Reference(s):
PC1.4 PC2.2 Basis Reference(s):
- 1. Nl-ODP-PR0-0302, EOP Technical Bases Containment Isolation Status 3.4.1 Site rea Emer enc NSL, EC steam line or Reactor Water Clean-up Isolation failure resulting in a release pathway outside primary containment Node Appl icabil ity:
Power Operation, Hot Shutdown Basis:
This EAL covers 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 to the condenser. 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.
May 1994 Page 33 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 3.4.1 (Cont)
PEG Reference(s):
PC2.1 Basis Reference(s):
None 3.4.2 General Emer enc MSL, EC steam line isolation failure or Reactor Water 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 < -84 in. (TAF)
.DW radiation > 3000 R/hr Node Applicability:
Power Operation, Hot Shutdown Basis:
The conditions of 'this EAL include the containment isolation failures allowing a direct flow path to the environment. A release pathway outside primary containment exists when steam flow is not prevented by downstream 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 bases for declaration of Site Area Emergency in EAL 3.5.1.
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 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 34 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 3.4.2 (Cont)
The RPV water level used in this EAL is the top of active fuel (TAF).
This value corresponds to the level which is used in EOPs to indicate challenge to core cooling and loss of the fuel clad barrier. This is the minimum water level to assure core cooling without further degradation of the clad. Severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured water level is not maintained above TAF.
if RPV The drywell radiation reading is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell. The reading is calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 pCi/gm dose equivalent I-131 into the drywell atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations allowed within Technical Specifications (including iodine spiking) and are therefore indicative of fuel damage (approximately 2X 5X clad failure depending on core inventory and RCS volume).
It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. Drywell High Range Radiation Monitors have a range of 0 to E8 R/hr on recorder RR 201.7-36C pen 1 and 2. They are installed in the following drywell locations:
RAm 201.7-36 Az 340', El 263'"
RAm 201.7-37 Az 310', EL 301'"
PEG Reference(s):
PC2.1 and FCl.l PC2.1 and FC2.1 PC2.1 and FC3.1 May 1994 Page 35 EPMP-EPP-0101 Rev 00
0 ATTACHMENT 1 (Cont) 3.4.2 (Cont)
Basis Reference(s):
- 1. General Electric NED0-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions
- 2. Nl-ODP-PR0-0302, EOP Technical Bases
- 3. Nl-RG197-EILl, Important Design Features of Regulatory Guide 1.97 Instruments
- 4. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Amendment 72, 76,, Table 3.6.11-1
- 5. Calculation 1H21C003, Rev 0 4.0 SECONDARY CONTAINMENT 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 a direct release into secondary containment:
Secondar Containment Tem eratures: Abnormally high secondary containment area temperatures can also pose a threat to the operability of vital equipment located inside secondary containment including RPV water level instrumentation. High area temperatures may limit personnel accessibility to vital areas.
High area temperatures may also be indicative of either primary system discharges into secondary containment or fires.
Secondar Containment Area Radiation Levels: Abnormally high area radiation levels in secondary containment, although not necessarily posing a threat to equipment operability, may pose a threat to personnel safety and the ability to operate vital equipment due to a lack of accessibility. Abnormally high area radiation levels may also be the result of a primary system discharging into the secondary containment and be indicative of precursors to significant radioactivity release to the environment.
May 1994 Page 36 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 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, Nl-EOP-5 Node Applicability:
Power Operation, Hot Shutdown Basis:
The presence of elevated area temperatures in the secondary containment may be indicative of an unisolable primary system leakage outside the primary containment. These conditions represent a loss of the containment barrier and a potential loss of the RCS barrier.
PEG Reference(s):
PC2.3 RCS1.3 Basis Reference(s):
- 1. 'Nl-ODP-PR0-0302, EOP Technical Bases
- 2. Nl-EOP-5 4.1.2 General Emer enc Primary system is discharging outside PC AND RB area temperatures are
> maximum safe operating levels in two or, more areas, Nl-EOP-5 AND any:
~ Coolant activity > 300 pCi/gm 1-131 equivalent
~ RPV water level < -84 in. (TAF)
~ DM radiation > 3000 R/hr Node Applicability:
Power Operation, Hot Shutdown Hay 1994 Page 37 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 4.1.2 (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.
When secondary containment area temperatures are accompanied by elevated coolant activity, RPV water level below TAF, or high drywell radiation, declaration of a General Emergency is appropriate due to loss of the primary containment barrier, RCS barrier, and loss or potential loss of the fuel clad barrier.
Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This amount of coolant activity is well above that expected for iodine spikes and corresponds to about 2X to 5X fuel clad damage. When reactor coolant activity reaches this level, significant clad heating has occurred and thus the fuel clad barrier is considered lost.
The RPV water level used in this EAL is the top of active fuel (TAF).
This value corresponds to the level which is used in EOPs to indicate challenge to core cooling and loss of the fuel clad barrier. This is the minimum water level to assure core cooling without further degradation of the clad. Severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured water level is not maintained above TAF.
if RPV The drywell radiation reading is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell. The reading is calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 pCi/gm dose equivalent I-131 into the drywell atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations allowed within Technical Specifications (including iodine spiking) and are therefore indicative of fuel damage (approximately 2X 5X clad failure depending on core inventory and RCS volume).
It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. Drywell High Range Radiation Monitors have a range of 0 to E8 R/hr on recorder RR 201.7-36C pen 1 and 2. They are installed in the following drywell locations:
RAm 201.7-36 Az 340', El 263'"
RAm 201.7-37 Az 310', EL 301'"
May 1994 Page 38 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 4.1.2 (Cont)
PEG Reference(s):
PC2.3 and FC1.1 PC2.3 and FC2. 1 PC2.3 and FC3.1 Basis Reference(s):
- 1. Nl-ODP-PR0-0302, EOP Technical Bases
- 2. General Electric NED0-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions
- 3. Nl-RG197-EIL1, Important Design Features of Regulatory Guide 1.97 Instruments
- 4. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Amendment 72, 76, Table 3.6.11-1
- 5. Calculation 1H21C003, Rev 0
- 6. N1-EOP-5 4.2 Reactor Bui 1 din Radi ati on Level 4.2. 1 Site Area Emer enc Primary system is discharging outside PC AND RB area radiation levels are > levels specified in Nl-EOP-5)'Io Node Applicability:
Power Operation, Hot Shutdown Basis:
The presence of elevated area radiation levels in the secondary containment may be indicative of an unisolable primary system leakage outside the primary containment. These conditions represent a loss of the containment barrier and a potential loss of the RCS barrier.
PEG Reference(s):
PC2.3 RCS1.3 Nay 1994 Page 39 EPHP-EPP-0101 Rev 00
ATTACHNENT 1 (Cont) 4.2.1 (Cont)
Basis Reference(s):
- 1. Nl-ODP-PR0-0302, EOP Technical Bases
- 2. Nl-EOP-5 4.2.2 General Emer enc Primary system is discharging outside PC AND RB area radiation levels are > levels specified in Nl-EOP-5, in two or more areas gag
~ Coolant activity > 300 pCi/gm I-131 equivalent
~ RPV water level < -84 in. (TAF)
~ DW radiation > 3000 R/hr Node Applicability:
Power Operation, Hot Shutdown Basis:
The presence of elevated area radiation levels in the secondary containment may be indicative of an unisolable primary system leakage outside the primary containment. These conditions represent a loss of the containment barrier and a potential loss of the RCS barrier.
When secondary containment radiation levels are accompanied by elevated coolant activity, RPV water level below TAF, or high drywell radiation, declaration of a General Emergency is appropriate due to loss of the primary containment barrier, RCS barrier, and loss or potential loss of the fuel clad barrier.
Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This amount of coolant activity is well above that expected for iodine spikes and corresponds to about 2X to 5X fuel clad damage. When reactor coolant activity reaches this level, significant clad heating has occurred and thus the fuel clad barrier is considered lost.
The RPV water level used in this EAL is the top of active fuel (TAF).
This value corresponds to the level which is used in EOPs to indicate challenge to core cooling and loss of the fuel clad barrier. This is the minimum water level to assure core cooling without further degradation of the clad. Severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured if RPV water level is not maintained above TAF.
Hay 1994 Page 40 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 4.2.2 (Cont)
The drywell radiation reading is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell. The reading is calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and 'iodine inventory associated with a concentration of 300 pCi/gm dose equivalent I-131 into the drywell atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations allowed within Technical Specifications (including iodine spiking) and are therefore indicative of fuel damage (approximately 2X 5X clad failure depending on core inventory and RCS volume).
It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. Drywell High Range Radiation Monitors have a range of 0 to E8 R/hr on recorder RR 201.7-36C pen 1 and 2. They are installed in the following drywell locations:
RAm 201.7-36 Az 340', El 263'"
RAm 201.7-37 Az 310', EL 301'"
PEG Reference(s):
PC2.3 and FCl.l PC2.3 and FC2.1 PC2.3 and FC3.1 Basis Reference(s):
- 1. Nl-ODP-PR0-0302, EOP Technical Bases
- 2. General Electric NED0-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions
- 3. Nl-RG197-EILl, Important Design Features of Regulatory Guide 1.97 Instruments
- 4. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Amendment 72, 76, Table 3.6. 11-1
- 5. Calculation 1H21C003, Rev 0
- 6. Nl-EOP-5 May 1994 Page 41 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 5.0 RADIOACTIVITY RELEASE Many EALs are based on actual or potential degradation of fission product barriers because of the increased potential for offsite radioactivity release. Degradation of fission product barriers though, is not always apparent via non-radiological symptoms.
Therefore, direct indication of increased radiological effluents are appropriate symptoms for emergency classification.
At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions.
There are two basic indications of radioactivity release rates which warrant emergency classifications.
~ Effluent Monitors: Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits.
Dose Pro 'ection and or Environmental Measurements: Projected offsite doses (based on'ffluent monitor readings) or actual offsite field measurements indicating doses or dose rates above classifiable limits.
5.1 Effluent Monitors 5.1.1 Unusual Event A valid reading on any monitors from Table 5.1 "UE" column for > 60 min.
Table 5.1 Effluent Monitor Classification Thresholds Monitor UE Al ert SAE GE Stack (RNlOA/8) >300 cps >3.0E4 cps >5.0 E6 cps N/A EC Vent >10 mR/hr Z30 mR/hr h310 mR/hr N/A SW Effluent ~900 cpm 290,000 cpm N/A N/A RW Discharge >2 x batch ~200 x batch N/A N/A Mode Applicability:
All Hay 1994 Page 42 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 5.1.1 (Cont)
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.
Two times the monitors alarm setpoints have been selected for use in this EAL. The alarm setpoints for the listed monitors are conservatively set to ensure Technical Specification radioactivity release limits are not exceeded. The value shown for the UE level is two times the high alarm setpoint for the Emergency Condenser vent-monitor and the Service Water effluent monitor, and two times the high-high alarm setpoint for the main stack (OGESM) monitor.
The following radiation monitors are not included in this EAL:
Reactor Building Vent Monitors: Reactor -building ventilation discharges to the main stack. Radioactivity release from the reactor building would, therefore, be assessed by the main stack monitor.
Containment Spray Raw Water Monitors: These monitors detect radiation in the discharge from their respective processes. The monitors are located upstream of the Service Water monitor. Therefore, the Service Water radiation monitor adequately detects offsite radioactivity releases from these systems.
PEG Reference(s):
AU1.1 May 1994 Page 43 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont).
5.1.1 (Cont)
Basis Reference(s):
- 1. Nl-OP-50B Process Radiation Monitoring System
- 2. Nl-ARP-Hl Annunciator Hl-1-8
- 3. Nl-CSP-f308, Attachment 2: current settings per Jim Mosher
- 4. Nl-CSP-(215, Service Water Alarm Setpoint Determination, Attachment 2: current settings per Jim Mosher
- 5. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications
- 6. Calculation 1H21C003, Rev 0 5.1. 2 alert
-A valid reading on any monitors from Table 5.1 "Alert" column for > 15 min.
Table 5.1 Effluent Monitor Classification Thresholds Monitor 'UE Alert SAE GE Stack (RN10A/B) h300 cps h3.0E4 cps >5.0 E6 cps N/A EC Vent h10 mR/hr 230 mR/hr ~310 mR/hr N/A SW Effluent ~900 cpm Z90,000 cpm N/A N/A RW Discharge h2 x batch ~200 x batch N/A N/A Node Applicability:
All Basis:
Valid means that a radiation monitor reading has been confirmed by the operators to be correct. This event escalates from the Unusual Event by increasing the magnitude of the release by a factor of 100 over the Unusual Event level (i. e., 200 times Technical Specifications).
Prorating the 500 mR/yr bases of the 10CFR20 non-occupational DAC limits for both time (8766 hr/yr) and the 200 multiplier, the associated site boundary dose rate would be 10 mR/hr. The required release duration was reduced to 15 minutes in recognition of the increased severity.
May 1994 Page 44 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 5.1.2 (Cont)
The following radiation monitors are not included in this EAL:
Reactor Building Vent Monitors: Reactor building ventilation discharges to the main stack. Radioactivity release from the reactor building would, therefore, be assessed by the main stack monitor.
Containment Spray Raw Water Monitors: These monitors detect radiation in the discharge from their respective processes. The monitors are located upstream of the Service Water monitor. Therefore, the Service Water radiation monitor adequately detects offsite radioactivity releases from these systems.
PEG Reference(s):
AA1.1 Basis Reference(s):
- 1. Nl-OP-SOB, Process Radiation Monitoring System
- 2. Nl-ARP-H1, Annunciator Hl-1-8
- 3. Nl-CSP-f308, Attachment 2: current settings per Jim Mosher
- 4. Nl-CSP-(215, Service Water Alarm Setpoint Determination, Attachment 2: current settings per Jim Hosher
-5. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications
- 6. Calculation 1H21C003, Rev 0 5.1.3 Site Area Emer enc A valid reading on any monitors from Table 5. 1 "SAE" column for > 15 min.
Table 5.1 Effluent Monitor Classification Thresholds Monitor UE Alert SAE GE Stack (RN10A/B) >300 cps >3.0E4 cps >5.0 E6 cps N/A EC Vent ~10 mR/hr h30 mR/hr h310 mR/hr N/A SW Effluent ~900 cpm ~90,000 cpm N/A N/A RW Discharge >2 x batch ~200 x batch N/A N/A Hay 1994 Page 45 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 5.1.2 (Cont)
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 IH21C003, Rev. 0.
PEG Reference(s):
AS1.1 May 1994 Page 46 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont).
- 5. 1. 3 (Cont)
Basis Reference(s):
- 1. Nl-OP-50B, Process Radiation Monitoring System
- 2. Nl-ARP-H1, Annunciator Hl-1-8
- 3. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications
- 4. Calculation 1H21C003, Rev. 0 5.2 Dose Pro ections Environmental Neasurements 5.2.1 Unusual Event Confirmed sample analyses for gaseous or liquid release rates > 2 x technical specifications limits for > 60 min.
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.
PEG Reference(s):
AU1.2 Basis Reference(s):
- 1. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Article 3.6.15.a(1) and Article 3.6.15.b(l) (a) and (b)
May 1994 Page 47 EPMP-EPP-0101 Rev 00
0 0
ATTACHMENT 1 (Cont) 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 bases of the 10CFR20 non-occupational DAC limits for both time (8766 hr/yr) and the 200 multiplier, the associated site boundary dose rate would be 10 mR/hr. The required release duration was reduced to 15 minutes in recognition of the increased severity.
PEG Reference(s):
AAl.2 Basis Reference(s):
- 1. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Article 3.6.15.a(l) and Article 3.6.15.b(l)(a) and (b) 5.2.3 Al ert 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 May 1994 Page 48 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 5.2.3 (Cont)
Node Applicability:
All Basis:
Offsite integrated doses in excess of 10 mR TEDE or dose rates in excess of 10 mR/hr TEDE represent an uncontrolled situation and hence, a potential degradation in the level of safety. This event escalates.
from the Unusual Event by increasing the magnitude of the release by a factor of 100 over the Unusual Event level (i. e., 200 times Technical Specifications). Prorating the 500 mR/yr bases of 10CFR20 for both time (8766 hr/yr) and the 200 multiplier, the associated site boundary dose rate would be 10 mR/hr.
As previously stated, the 10 mR/hr value is based on a proration of 200 times the 500 mR/yr bases of 10CFR20, rounded down to 10 mR/hr.
PEG Reference(s):
AA1.2 Basis Reference(s):
- 1. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Article 3.6.15.a(l) and Article 3.6.15.b(1)(a) and (b)
.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 Al ert 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 49 EPMP-EPP-0101 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 dur ation 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. Facility Operating License No. DPR-63, Appendix A, Radiological Technical. Specifications 5.2.5 General Emer enc Dose projections or field surveys which indicate doses / dose rates >
Table 5.2 column "GE" at the site boundary or beyond Table 5.2 Dose Pro 'ection Env. Measurement Classification Thresholds Al ert 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 50 EPHP-EPP-0101 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 CDE thyroid. This is consistent with the emergency class description for a General Emergency. This level constitutes the upper level of the desirable gradient for the Site Area Emergency. Actual meteorology is specifically identified since it gives the most accurate dose assessment. Actual meteorology (including forecasts) should be used whenever possible. In establishing the dose rate emergency action levels, a duration of one hour is assumed. Therefore, the dose rate EALs are based on a site boundary dose rate of 1000 mR/hr TEDE or 5000 mR/hr CDE thyroid, whichever is more limiting.
PEG Reference(s):
AG1.3 AG1. 4 Basis Reference(s):
- 1. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications 6.0 ELECTRICAL FAILURES Loss of vital plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.
The events of this category have been grouped into the following two loss of electrical power types:
II 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.
May 1994 Page 51 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 6.1 Loss of AC Power Sources 6.1.1 Unusual Event Loss of power for > 15 min. to all:
~ T-101N
~ T-101 S
~ T-10 backfed through T-1 or T-2 Node Appl icabi1 ity:
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.
Backfeeding of the Station Transformer T10 has been included to allow for those conditions in which maintenance is being performed on the Station Reserve Transformers or 115 kv system. It is recognized that this is not a readily available source of emergency power under emergency conditions and should only be taken credit for those conditions under which backfeeding has already been established.
PEG Reference(s):
SU1.1 Basis Reference(s):
- 1. Nl-OP-45, Emergency Diesel Generators
- 2. NI-OP-30, 4.16 Kv, 600V, and 480V House Service May 1994 Page 52 EPMP-EPP-0101 Rev 00
0 ATTACHMENT i tC t) 6.1.2 Alert 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 to all:
T-101N T-101S T-10 backfed through T-1 AND failure of both DGs to power emergency buses AND failure to restore power to PB102 or PB103 in g 15 min.
When in cold shutdown, refueling, or defueled mode this event is classified as an Alert. This is because of the significantly reduced decay heat, lower temperature and pressure, thus increasing the time to restore one of the emergency busses, relative to that specified for the Site Area Emergency EAL. Escalating to the Site Area Emergency, if appropriate, is by Abnormal Rad Levels/Radiological Effluent, or Emergency Director Judgment ICs. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
Backfeeding of the Normal Station Transformer has been included to allow for those conditions in which maintenance is being performed on the Station Reserve Transformers or 115 kv system. It is recognized that this is not a readily available source of emergency power under emergency conditions and should only be taken credit for those conditions under which backfeeding has already been established.
PEG Reference(s):
SA1. 1 Basis Reference(s):
- 1. Nl-OP-30, 4.16 Kv, 600V, and 480V House Service
- 2. Nl-OP-45, Emergency Diesel Generators May 1994 Page 53 EPNP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 6.1.3
~ ~ Al ert Available emergency bus AC power reduced to only one of the following sources for >15 min.:
~ DG102 (PB102)
~ DG103 (PB103)
~ T-101N
~ T-101 S Node Applicability:
Power Operation, 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. The subsequent loss of this single power source would escalate the event to a Site Area Emergency.
PEG Reference(s):
SA5.1 Basis Reference(s):
- 1. Nl-OP-45, Emergency Diesel Generators
- 2. Nl-OP-30, 4.16 Kv, 600V, and 480V House Service
- 6. 1.4 Site Area Emer enc Loss of all emergency bus AC power for >15 min.
Node Applicability:
Power Operation, Hot Shutdown May 1994 Page 54 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 6.1.4 (Cont)
Basis:
Loss of all AC power compromises all plant safety systems requiring electric power. This EAL is indicated by:
Loss of power to T-101N and T-101S, and T-10 backfed through T-1 AND failure of both DGs to power any emergency buses AND failure to restore power to PB102 or PB103 in g 15 min.
Prolonged loss of all AC power will cause core uncovery and loss of containment integrity, thus this event can escalate to a General Emergency. The time duration selected, 15 minutes, excludes transient or momentary power losses.
PEG Reference(s):
SS1.1 Basis Reference(s):
- 1. Nl-OP-45, Emergency Diesel Generators
- 2. Nl-OP-30 4.16 Kv, 600V, and 480V House Service
- 3. Nl-SOP-18, Station Blackout 6.'1. 5 General Emer enc Loss of all emergency bus AC power AND either:
Power cannot be restored to any emergency bus in < 4 hrs OR RPV water level cannot be restored and maintained > -84 in. (TAF)
Node Applicability:
Power Operation, Hot Shutdown Basis:
Loss of all AC power compromises all plant safety systems requiring electric power. Prolonged loss of all AC power will lead to loss of fuel clad, RCS, and containment. Although this EAL may be viewed as redundant to the RPV Water Level EALs, its inclusion is necessary to better assure timely recognition and emergency response.
May 1994 Page 55 EPMP-EPP-0101 Rev 00
ATTACHHENT 1 (Cont) 6.1. 5 (Cont)
This EAL is specified to assure that in the unlikely event of prolonged station blackout, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency occurs as early as is appropriate, based on a reasonable assessment of the event trajectory.
The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of the situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing public protective actions.
In addition, under these conditions, fission product barrier monitoring capability may be degraded. Although it may be difficult to predict when power can be restored, the Emergency Director should declare a General Emergency based on two major considerations:
- 1. Are there any present indications that core cooling is already degraded to the point that Loss or Potential Loss of fission product barriers is imminent?
- 2. If there are no present indications of such core cooling degradation, how likely is it that power can be restored in time to assure that a loss of two barriers with a potential loss of the third barrier can be prevented?
Thus, indication of continuing core cooling degradation must be based on fission product barrier monitoring with particular emphasis on Emergency Director judgment as it relates to imminent loss or potential loss of fission product barriers and degr aded 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 Hay 1994 Page 56 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 6.1.5 (Cont)
Basis Reference(s):
- 1. Nl-OP-45, Emergency Diesel Generators
- 2. Nl-OP-30 4.16 Kv, 600V, and 480V House Service
- 3. Nl-SOP-18, Station Blackout, pg. 1
< 106 vdc on battery board ll and 12 for >15 min.
Mode Applicability:
Cold Shutdown, Refuel Basis:
The purpose of this EAL is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations. This EAL is intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss.
The bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value incorporates a margin of at least 15 minutes of operation before the onset of inability to operate loads.
PEG Reference(s):
SU7.1 Basis Reference(s):
- 1. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Basis for articles 3.6.3 and 4.6.3
- 2. Nl-OP-47A, 125 vdc Power System Hay 1994 Page 57 EPMP-EPP-0101 Rev 00
0 ATTACHMENT 1 (Cont).
6.2.2 Site Area Emer enc
< 106 vdc on battery board 11 and 12 for > 15 min.
Node Applicability:
Power Operation, Hot Shutdown Basis:
Loss of all DC power compromises ability to monitor and control plant safety functions. Prolonged loss of all DC power will cause core uncovering and loss of containment integrity when there is significant decay heat and sensible heat in the reactor system. Escalation to a General Emergency would occur by other EAL categories. Fifteen minutes was selected as a .threshold to exclude transient or momentary power losses.
The bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value incorporates a margin of at least 15 minutes of operation before the onset of inability to operate loads.
PEG Reference(s):
SS3.1 Basis Reference(s):
- 1. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Basis for articles 3.6.3 and 4.6.3
- 2. Nl-OP-47A, 125 vdc Power System May 1994 Page 58 EPMP-EPP-0101 Rev 00
0 ATTACHHENT 1 (Cont) 7.0 E UIPHENT FAILURES Numerous plant system related equipment failure events which warrant emergency classification, based upon their potential to pose actual or potential threats to plant safety, have been identified in this category.
The events of this category have been grouped into the following event types:
~ Technical S ecifications: Only one EAL falls under this event type related to the failure of the plant to be brought to the required plant operating condition required by technical specifications.
~ S ste 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 ab'lit : Certain events which degrade the plant operators ability to effectively assess plant conditions or communicate with essential personnel within or external to the plant warrant emergency classification.
Under this event type are losses of annunciators and/or communication equipment.
7.1 Technical S ecifications 7.1.1 Unusual Event Plant is not brought to required operating mode within Technical Specifications LCO Action Statement Time Node Applicability:
Power Operation, Hot Shutdown Hay 1994 Page 59 EPHP-EPP-0101 Rev 00
ATTACHNENT 1 (Cont)
- 7. 1. 1 (Cont)
Basis:
Limiting Conditions of Operation (LCOs) require the plant to be brought to a required shutdown mode when the Technical Specification required configuration cannot be restored. Depending on the circumstances, this may or may not be an emergency or precursor to a more severe condition. In any case, the initiation of plant shutdown required by the site Technical Specification requires a one hour report under 10CFR50.72 (b) non-emergency events. The plant is within its safety envelope when being shut down within the allowable action statement time in the Technical Specifications. An immediate Notification of an Unusual Event is required when the plant is not brought to the required operating mode within the allowable action statement time in the Technical Specifications. Declaration of an Unusual Event is based on the time at which the LCO-specified action statement time period elapses under the site Technical Specifications and is not related to how long a condition may have existed. Other required Technical Specification shutdowns that involve precursors to more serious events are addressed by other EALs.
PEG Reference(s):
SU2.1 Basis Reference(s):
- 1. Radiological Technical Specifications, Appendix A to Facility Operating License No. DPR-63, article 3.0.1 7.2 S stem 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, Hot Shutdown May 1994 Page 60 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 7.2.1 (Cont)
Basis:
This EAL is intended to address main turbine rotating component failures of sufficient magnitude to cause observable damage to the turbine casing or to the seals of the turbine generator. Of major concern is the potential for leakage of combustible fluids (lubricating oils) and gases (hydrogen cooling) to the plant environs.
Actual fires and flammable gas build up are appropriately classified through other EALs. This EAL is consistent with the definition of an Unusual Event while maintaining the anticipatory nature desired and recognizing the risk to non-safety related equipment.
PEG Reference(s):
HU1. 6 Basis Reference(s):
None 7.2.2 Aler t Control Room evacuation Node Applicability:
All Basis:
With the Control Room evacuated, additional support, monitoring and direction through the Technical Support Center and/or other Emergency Operations Facility is necessary. Inability to establish plant control from outside the Control Room will escalate this event to a Site Area Emergency.
PEG Reference(s):
HA5.1 Basis Reference(s):
- 1. Nl-SOP-9.1, Control Room Evacuation May 1994 Page 61 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 7.2.3 Alert Reactor coolant temperature cannot be maintained < 212 'F Mode Applicability:
Cold Shutdown, Refuel Basis:
This EAL addresses complete loss of functions required for core cooling during refueling and cold shutdown modes. Escalation to Site Area Emergency or General Emergency would be through other EALs.
A reactor coolant temperature increase that approaches or exceeds the cold shutdown technical specification limit warrants declaration of an Alert irrespective of the availability of technical specification required functions to maintain cold shutdown. The concern of this EAL is the loss of ability to maintain the plant in cold shutdown which is
.defined by reactor coolant temperature and not the operability of equipment which supports removal of heat from the reactor.
PEG Reference(s):
SA3.1 Basis Reference(s):
- 1. Facility Operating License No. DPR-63, Appendix A, Radiological Technical Specifications, Amendment 99, Article l.l.a 7.2.4 Site Area Emer enc Control Room evacuation AND Control of core cooling systems cannot be established in g 15 min.
Mode Applicability:
All Hay 1994 Page 62 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 7.2.4 (Cont)
Basis:
This EAL indicates that expeditious transfer of safety systems has not occurred but fission product barrier damage may not yet be indicated.
The time interval for transfer is based on analysis or assessments as to how quickly control must be reestablished without core uncovering and/or core damage. In cold shutdown and refueling modes, operator concern is directed toward maintaining core cooling such as is discussed in Generic Letter 88-17, "Loss of Decay Heat Removal." In=
power operation, hot standby, and hot shutdown modes, operator concern is primarily directed toward monitoring and controlling plant parameters dictated by the EOPs and thereby assuring fission product barrier integrity.
PEG Reference(s):
HS2.1 Basis Reference(s):
- 1. Generic Letter 88-17, "Loss of Decay Heat Removal"
- 2. Nl-SOP-18, Station Blackout 7.3 Loss of Indicatio s Alarm Communication Ca abilit 7.3.1 Unusual Event Unplanned loss of all annunciators or indicators on all panels L, K, H, F, G for > 15 min. AND increased surveillance is required for safe plant operation Node Applicability:
Power Operation, Hot Shutdown Basis:
This EAL recognizes the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment. 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.
May 1994 Page 63 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 7.3.1 (Cont)
It is not intended that plant personnel perform a detailed count of instrumentation lost but the use of judgment by the Shift Supervisor as the threshold for determining the severity of the plant conditions. This judgment is supported by the specific opinion of the Shift Supervisor that additional operating personnel will be required to provide increased monitoring of system operation to safely operate the plant.
It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptable power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by 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 Refers ence(s):
SU3.1 Basis Reference(s):
- 1. Nl-OP-22, Process Computer
- 2. Nl-OP-55, Safety Parameter Display System May 1994 Page 64 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 7.3.2
~ ~ Unusual Event Loss of all communications capability affecting the ability to either:
Perform routine onsite operations OR notify offsite agencies or personnel Node Applicability:
All Basis:
The purpose of this EAL is to recognize a loss of communications capability that either defeats the plant operations staff ability to perform routine tasks necessary for plant operations or the ability to communicate problems with offsite authorities. The loss of offsite communications ability is expected to be significantly more comprehensive than the condition addressed by 10CFR50.72.
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 PBX Gaitronics Portable headsets Station radios ENS RECS UHF radios The offsite communications loss must encompass the loss of all means of communications with offsite authorities, Table 7. 1. This EAL is intended to be used only when extraordinary means are being utilized to make communications possible (relaying of information from radio transmissions, individuals being sent to offsite locations, etc.).
May 1994 Page 65 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 7.3.2 (Cont)
PEG Refer ence(s):
SU6.1 Basis Reference(s):
- 1. Nl-OP-51, Communications System 7.3.3 Alert Unplanned loss of all annunciators or indicators on all panels L, K, H, F, G for > 15 min.
AND Increased surveillance is required for safe plant operation AND either:
Plant transient in progress OR plant computer and SPDS are unavailable Node Applicability:
Power Operation, Hot Shutdown
.Basis:
This EAL recognizes the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment during a transient. Recognition of the availability of computer based indication equipment is considered (SPDS, plant computer, etc.).
Unplanned" loss of annunciators or indicators does not include scheduled maintenance and testing activities.
It is not intended that plant personnel perform a detailed count of the instrumentation lost but the use of the value as a judgment by the shift supervisor as the threshold for determining the severity of the plant conditions. This judgment is supported by the specific opinion of the Shift Supervisor that additional operating personnel will be required to provide increased monitoring of system operation to safely operate the plant.
May 1994 Page 66 EPMP-EPP-0101 Rev 00
ATTACHHENT 1 (Cont)
V.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.).
"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 crew cannot monitor the a Site Area Emergency transient in progress.
if the operating PEG Reference(s):
SA4.1 Basis Reference(s):
- 1. Nl-OP-22, Process Computer.
- 2. Nl-OP-55, Safety Parameter Display System Hay 1994 Page 67 EPHP-EPP-0101 Rev 00
ATTACHMENT t (C t) 7.3.4 Site Area Emer enc Loss of all annunciators or indicators on all panels L, K, H, F, G AND plant computer and SPDS are unavailable AND Indications to monitor all RPV and primary containment EOP parameters are lost AND Plant transient is in progress Node Applicability:
Power Operation, Hot Shutdown Basis:
This EAL recognizes the inability of the Control Room staff to monitor the plant response to a transient. A Site Area Emergency is considered to exist if the Control Room staff cannot monitor safety functions needed for protection of the public.
Annunciators for this EAL should be limited to include those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures,'nd in other EALs (e. g., rad monitors, etc.).
'"Transient" includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25K thermal power change, ECCS injections, or thermal power oscillations of lOX or greater.
Indications needed to monitor safety functions necessary for protection of the public must include Control Room indications, computer generated indications and dedicated annunciation capability.
The specific indications should be those used to determine such functions as the ability to shut down the reactor, maintain the core cooled and in a eoolable geometry, to remove heat from the core, to maintain the reactor coolant system intact, and to maintain containment intact.
"Planned" actions are excluded from this EAL since the loss of instrumentation of this magnitude is of such significance during a transient that the cause of the loss is not an ameliorating factor.
May 1994 Page 68 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont).
7.3.4 (Cont)
PEG Reference(s):
SS6.1 Basis Reference(s):
- 1. Nl-OP-22, Process Computer
- 2. Nl-OP-55, Safety Parameter Display System
- 3. Nl-ODP-PR0-0302, EOP Technical Bases, 8.0 HAZARDS Hazards are those non-plant system related events which can directly or indirectly impact plant operation or reactor plant and personnel safety.
The events of this category have been grouped into the following types:
Securit Threats: Thi's 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: Man-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.
May 1994 Page 69 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 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. Bomb devices discovered within the plant vital area would result in EAL escalation.
PEG Reference(s):
HU4.1 HU4.2 Basis Reference(s):
- 1. Nine Mile Point Nuclear Station Physical Security and Safeguards Contingency Plans 8.1.2 Alert Intrusion into plant Protected Area by an adversary Mode Applicability:
All May 1994 Page 70 EPMP-EPP-0101 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 unauthorized personnel inside the Protected Area boundary can be considered a significant security threat.
Intrusion into a vital area by unauthorized personnel will escalate this event to a Site Area Emergency.
NMP-1 and NMP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also see S&W Drawing No. 12187-SK-032483-25, Issue No.
1, Site facilities Layout Status as of 8/1/89.
PEG Reference(s):
HA4.1 HA4.2 Basis Reference(s):
- 1. Nine Mile Point Nuclear Station Physical Security and Safeguards Contingency Plans
- 2. S&W 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 Mode Applicability:
All Basis:
This class of security events represents an escalated threat to plant safety above that contained in the Alert in that unauthorized personnel have progressed from the Protected Area to the vital area.
May 1994 Page 71 EPMP-EPP-0101 Rev 00
ttitttt)NT t tt t) 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 remote shutdown capability Mode 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):
HG1.1 MG1.2 Basis Reference(s):
None Nay 1994 Page 72 EPHP-EPP-0101 Rev 00
ATTACHMENT 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 M 15 min. of Control Room notification Table 8.2 Plant Areas RadWaste Solidification and Storage Bldg.
Security West Bldg.
Table 8.3 Plant Vital Areas Control Room Building Auxiliary Control Room Cable Spreading Room Reactor Bldg.
Turbine .Bldg.
Diesel Generator Area Screen and Pump House Off Gas Bldg.
Node Applicability:
All Basis:
The purpose of this EAL is to address the magnitude and extent of fires that may be potentially .significant precursors to damage to safety systems. This excludes such items as fires within administration buildings, waste-basket fires, and other small fires of no safety consequence.
PEG Reference(s):
HU2.1 May 1994 Page 73 EPMP-EPP-0101 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 RadWaste Solidification and Storage Bldg.
Security West Bldg.
Table 8.3 Plant Vital Areas Control Room Building Auxiliary Control Room Cable Spreading Room Reactor Bldg.
Turbine Bldg.
Diesel Generator Area Screen and Pump House Off Gas Bldg.
Node Applicability:
All Basis:
The listed areas contain functions and systems required for the safe shutdown of the plant. The NMP-1 safe shutdown analysis was consulted for equipment and plant areas required for the applicable mode.
May 1994 Page 74 EPMP-EPP-0101 Rev 00
ATTACHMENT 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. Nl-SOP-9, Fire In Plant
- 2. Nine Mile Point Nuclear Station Unit 1 Appendix 'R'eview Safe Shutdown Analysis
- 3. NUREG 0737, Section II.B.2-2 8.3 Nan-Nade Events 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. NMP-1 and NMP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also, refer to SKW Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89.
May 1994 Page 75 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont)
(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.
PEG Reference(s):
HU1.4 Basis Reference(s):
- l. USAR Figure 1.2-1
- 2. S8W Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89 8.3.2 Unusual Event Report by plant personnel of an explosion within Protected Area boundary resulting in visible damage to permanent structures or equipment 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.
For this EAL, only those explosions of sufficient force to damage permanent structures or equipment within the Protected Area should be considered. As used here, an explosion is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts significant energy to near by structures and materials. No attempt is made in this EAL to assess the actual magnitude of the damage. The occurrence of the explosion with reports of evidence of damage (e. g., deformation, scorching) is sufficient for declaration. The Emergency Director also needs to consider any security aspects of the explosion.
May 1994 Page 76 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 8.3.2 (Cont)
PEG Reference(s):
HU1.5 Basis Reference(s):
- l. 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 8.3.3 Unusual Event Report or detection of a release of toxic or flammable gases that could enter or have entered within the Protected Area boundary in amounts that could affect the health of plant personnel or safe plant operation OR
-Report by local, county or state officials for potential evacuation of site personnel based on offsite event
'ode Applicability:
All Basis:
This EAL is based on releases in concentrations within the site boundary that will affect the health of plant personnel or affecting the safe operation of the plant with the plant being within the evacuation area of an offsite event (i. e., tanker truck accident releasing toxic gases, etc.). The evacuation area is as determined from the DOT Evacuation Tables for Selected Hazardous Materials, in the DOT Emergency Response Guide for Hazardous Materials.
NMP-1 and NMP-2 share no common safety systems, but their respective Protected Area boundaries share common borders in some places.
Therefore it is possible that a toxic or flammable gas incident happening on one site could affect the other site.
Should an explosion occur within a specified plant area, an Alert would be declared based on EAL 8.2.2 PEG Reference(s):
HU3.1 HU3. 2 May 1994 Page 77 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 8.3.3 (Cont)
Basis Reference(s):
None 8.3.4 lert 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 Control Room Building Auxiliary Control Room Cable Spreading Room Reactor Bldg.
Turbine Bldg.
Diesel Generator Area Screen and Pump House Off Gas Bldg.
Node Applicability:
All Basis:
This EAL addresses events that may have resulted in a plant vital area being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage.
NMP-1 and NMP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also see SLW Drawing No. 12187-SK-032483-25, Issue No.
1, Site Facilities Layout Status as of 8/1/89.
This EAL addresses such items as plane, helicopter, train, car, or truck crash, or impact of other projectiles into a plant vital area.
PEG Reference(s):
HA1. 5 May 1994 Page 78 EPMP-EPP-0101 Rev 00
ATTACHHENT 1 (Cont) 8.3.4 (Cont)
Basis Reference(s):
- l. 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 r
- 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 Control Room Building Auxiliary Control Room Cable Spreading Room Reactor Bldg.
Turbine Bldg.
Diesel Generator Area Screen and Pump House Off Gas Bldg.
Node Applicability:
All Basis:
This EAL is based on gases that have entered a plant structure precluding access to equipment necessary for the safe operation of the plant. This EAL applies to buildings and areas contiguous to plant vital areas or other significant buildings or areas. The intent of this EAL is not to include buildings (i. e., warehouses) or other areas that are not contiguous or immediately adjacent to plant vital areas. It is appropriate that increased monitoring be done to ascertain whether consequential damage has occurred.
Hay 1994 Page 79 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 8.3.5 (Cont)
PEG Reference(s):
HA3.1 HA3.2 Basis Reference(s):
- 1. USAR Figure 111-6, Station Floor Plan Elevation 281'-0" and 291'-0" r 8.4 Natural Events 8.4.1 ~1E Earthquake felt in plant by any operator AND either:
NHP-1 seismic instrumentation actuated OR Confirmation of earthquake received on NHP-2 or JAFNPP seismic instrumentation Node Applicability:
All Basis:
NHP-1 seismic instrumentation actuates at 0.01 g.
Damage to some portions of the site may occur but it should not affect ability of safety functions to operate. Methods of detection can be based on instrumentation validated by a reliable source, operator assessment, or indication received from NHP-2 or JAFNPP instrumentation. As defined in the EPRI-sponsored "Guidelines for Nuclear Plant Response to an Earthquake", dated October 1989, a "felt earthquake" is:
"An earthquake of sufficient intensity such that: (a) the inventory ground motion is felt at the nuclear plant site and recognized as an earthquake based on a consensus of Control Room operators on duty at the time, and (b) for plants with operable seismic instrumentation, the seismic switches of the plant are activated. For most plants with seismic instrumentation , the seismic switches are set at an acceleration of about 0.01 g."
Hay 1994 Page 80 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 8.4.1 (Cont)
PEG Reference(s):
HU1.1 Basis Reference(s):
- 1. Nl-ARP-H2 annunciator H2-1-7
- 2. Nl-SOP-11, Earthquake
- 3. Nl-ARP-H2, Annunciator H2-1-3
- 4. EPRI document, "Guidelines for Nuclear Plant Response to an Earthquake" 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 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.
NMP-1 and NMP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also see SKW 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. S8W Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89 May 1994 Page 81 EPMP-EPP-0101 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 lant Vital Areas Control Room Building Auxiliary Control Room Cable Spreading Room Reactor Bldg.
Turbine Bldg.
Diesel Generator Area Screen and Pump House Off Gas Bldg.
Node Applicability:
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 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
- 3. NUREG 0737, Section II.B.2-2 Hay 1994 Page 82 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 8.4.4 Unusual Event Lake water level > 248 ft OR forebay water level < 238 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. The low level is based on intake forebay level and corresponds to the minimum intake water level for operability of Emergency Service Water, Emergency Diesel Generator cooling water, Containment Spray Raw Water and Diesel and Electric Fire Pump.
PEG '-Reference(s):
HU1. 7 Basis Reference(s):
- 1. Nl-ARP-H2, Annunciator H2-1-3
- 2. EPIP-EPP-Ol, Classification of Emergency Conditions Unit 1
- 3. Nl-SOP-7, Service Water Failure/Low Intake Level
- 4. DER I-92-g-0489 8.4.5 Alert Earthquake felt in plant by any operator AND NMP-1 seismic instrumentation indicates > 0.11 g Node Applicability:
All May 1994 Page 83 EPMP-EPP-0101 Rev 00
STTACIIMENT 1 tC t) 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 bases earthquake of 0.11 g. Seismic events of this magnitude can cause damage to plant safety functions.
PEG Reference(s):
HAl.l
'Basis Reference(s):
- 1. Nl-ARP-H2, annunciator H2-1-7
- 2. Nl-SOP-11, Earthquake 8.4.6 Alert Sustained winds > 125 mph OR Tornado strikes a plant vital area, Table 8.4 Table 8.4 Plant Vital Areas Control Room Building Auxiliary Control Room Cable Spreading Room Reactor Bldg.
Turbine Bldg.
Diesel Generator Area Screen and Pump House Off Gas Bldg.
Node Applicability:
All Nay 1994 Page 84 EPHP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 8.4.6 (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 bases of 125 mph. Wind loads of this magnitude can cause damage to safety functions.
NMP-1 and NMP-2 Protected Area boundaries are illustrated in USAR Figure 1.2-1. Also see S&W Drawing No. 12187-SK-032483-25, Issue No.
1, Site Facilities Layout Status as of 8/1/89.
PEG Reference(s):
HA1.2 Basis Reference(s):
- 1. FSAR Section VI.C.l.l, Wind and Snow Loadings, 6/91
- 2. Nl-SOP-10, High Winds
- 3. USAR Figure 1.2-1
- 4. S&W Drawing No. 12187-SK-032483-25, Issue No. 1, Site Facilities Layout Status as of 8/1/89
- 5. NUREG 0737, Section II.B.2-2 8.4.7 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 Control Room Building Auxiliary Control Room Cable Spreading Room Reactor Bldg.
Turbine Bldg.
Diesel Generator Area Screen and Pump House Off Gas Bldg.
May 1994 Page 85 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 8.4.7 (Cont)
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.
PEG Reference(s):
HA1.3 Basis Reference(s):
- 1. USAR Figure III-6, Station Floor Plan Elevation 281'-0" and 291'-0"
- 2. NUREG 0737, Section II.B.2-2 8.4.8 8lert Lake water level > 254 ft OR forebay water level < 236 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.
May 1994 Page 86 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 8.4.8 (Cont)
This EAL covers high and low lake water level conditions that exceed levels which threaten vital equipment. The high lake level is based upon the maximum probable flood level. The low forebay water level corresponds to the minimum level before damage may occur to the service water pumps.
PEG Reference(s):
HA1.7 Basis Reference(s):
- 1. Nl-SOP-7, Service Water Failure/Low Intake Level
- 2. DER 1-92-0-0489 9;0 ~OT Ef The EALs defined in categories 1.0 through 8.0 specify the predetermined symptoms or events which are indicative of emergency or potential emergency conditions, and which warrant classification.
While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and.
subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Shift Supervisor or Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria, based upon their judgment.
9.1.1 Unusual Event Any event, as determined by the Shift Supervisor or Emergency Director, that could lead to or has led to a potential degradation of the level of safety of the plant.
Node Applicability:
All May 1994 Page 87 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 9.1.1 (Cont)
Basis:
This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the Unusual Event emergency class.
I From a broad perspective, one area that may warrant Emergency Director judgment is related to likely or actual breakdown of site specific event mitigating actions. Examples to consider include inadequate emergency response procedures, transient response either unexpected or not understood, failure or unavailability of emergency systems during an accident in excess of that assumed in accident analysis, or insufficient availability of equipment and/or support personnel.
Another example to consider would be exceeding a plant safety limit as defined in Technical Specifications.
PEG Reference(s):
HU5. 1 Basis Reference(s):
None 9.'1. 2 Unusua E e t Any event, as determined by the Shift Supervisor or Emergency Director, that could lead to or has led to a loss or potential loss of containment.
Node Applicability.
Power Operations, Hot Shutdown Basis:
This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the containment barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this EAL as a factor in Emergency Director judgment that the barrier may be considered lost or potentially lost.
May 1994 Page 88 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 9.1.2 (Cont)
PEG Reference(s):
PC6.1 Basis Reference(s):
None 9.1.3 Alert Any event, as determined by the Shift Supervisor or Emergency Director, that could cause or has caused actual substantial degradation of the level of safety of the plant.
Mode Applicability:
All Basis:
This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the Alert emergency class.
PEG Reference(s):
HA6. 1 Basis Reference(s):
None 9.1.4 Alert Any event, as determined by the Shift Supervisor or 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, Hot Shutdown May 1994 Page 89 EPMP-EPP-0101 Rev 00
ATTACHMENT 1 (Cont) 9.1.4 (Cont)
Basis:
This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the fuel clad or RCS barriers are lost or potentially lost. In addition, the inability to monitor the barriers should also be considered in this EAL as a factor in Emergency Director judgment that the barriers may be considered lost or potentially lost.
PEG Reference(s):
FC5.1 RCS6.1 Basis Reference(s):
None "9.1.'5 Site Area Emer enc I
As determined by the Shift Supervisor or Emergency Director, events are in progress which indicate actual or likely failures of plant systems needed to protect the public. Any releases are not expected to result in exposures which exceed EPA PAGs.
Node Applicability:
All Basis:
This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warr ant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency class description for Site Area Emergency.
PEG Reference(s):
HS3.1 Basis Reference(s):
None Hay 1994 Page 90 EPMP-EPP-0101 Rev 00
ATTACHNENT 1 (Cont) 9.1.6 Site Area Emet enc Any event, as determined by the Shift Supervisor or Emergency Director, that could lead or has led to either:
Loss or potential loss of both fuel clad and RCS barrier 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, Hot Shutdown Basis:
This EAL addresses unanticipated conditions affecting fission product barriers which are not addressed explicitly elsewhere. Declaration of an emergency is warranted because conditions exist which are believed by the Emergency Director to fall under the emergency class description for 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 Hay 1994 Page 91 EPMP-EPP-0101 Rev 00
ATTACHMENT I (Cont) 9.1.7 General Emer enc As determined by the Shift Supervisor or Emergency Director, events are in progress which indicate actual or imminent core damage and the potential for a large release of radioactive material in excess of EPA PAGs outside the site boundary.
Node Applicability:
All Basis:
This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to be consistent with the General Emergency classification description.
Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the site boundary.
PEG Reference(s):
HG2.1 Basis Reference(s):
None 9.1.8 General Emer enc Any event, as determined by the Shift Supervisor or Emergency Director, that could lead or has led to a loss of any two fission product barriers and loss or potential loss of the third.
Loss of containment indicators may include:
~ Inconsistent or unexpected LOCA response
~ Rapid unexplained decrease following initial increase in containment pressure Node Applicability:
Power Operations, Hot Shutdown May 1994 Page 92 EPMP-EPP-0101 Rev 00
NITACNIIENT I EC tt 9.1.8 (Cont)
Basis:
This EAL addresses unanticipated conditions affecting fission product barriers which are not addressed explicitly .elsewhere. Declaration of an emergency is warranted because conditions exist which are believed by the Emergency Director to fall under the emergency class description for the General Emergency class.
Rapid unexplained loss of pressure (i. e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of containment integrity. Drywell pressure should increase as a result of mass and energy release into containment from a LOCA. Thus, drywell pressure not increasing under these conditions indicates a loss of containment integrity.
PEG Reference(s):
FC5.'1 RCS6.1 PC6.1 PC1.1 PC1.2 Basis Reference(s):
None Hay 1994 Page 93 EPHP-EPP-0101 Rev 00
ATTACHNENT 2 WORD LIST DEFINITIONS Actuate 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.
Ade uate Co e Coolin Heat removal from the reactor sufficient to prevent rupturing the fuel clad.
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 able 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.
Hay 1994 Page 94 EPHP-EPP-0101 Rev 00
ATTACHMENT 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 making 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 Confirmation To validate, through visual observation or physical inspection, that an assumed condition is as expected or required, without taking action to alter the "as found" configuration.
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.
~D1echer e Removal of a fluid/gas from a volume or system.
~Dr e11 That component of the BWR primary containment which houses th'e RPV and associated piping.
May 1994 Page 95 EPMP-EPP-0101 Rev 00
ATTACHMENT 2 (Cont)
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.
Exi 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.
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.
May 1994 Page 96 EPMP-EPP-0101 Rev 00
ATTACHMENT 2 (Cont)
Increase To become progressively greater in size, amount, number or intensity.
Indicate To point out or point to; to display the value of a process variable; to be a sign or symbol.
Initiate The act of placing equipment or a system into service, either manually or automatically. Activation of an function or protective feature (i.e. initiate a manual scram).
In ectio The act of forcing a fluid into a volume or vessel.
~lbl Not able to perform it's intended function
~Itruai o The act of entering without authorization Loss Failure of operability or lack of access to.
~Maintain Take action, as necessary, to keep the value of the specified parameter within the applicable limits.
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.
May 1994 Page 97 EPMP-EPP-0101 Rev 00
ATTACHMENT 2 (Cont)
Nonitor Observe and evaluate at a frequency sufficient to remain apprised of the value, trend, and rate of change of the specified parameter.
~Met1 f 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.
~0erab1 e Able to perform it's intended function ger~fo tg To carry out an action; to accomplish; to affect; to reach an objective.
Primar Containment The airtight volume immediately adjacent to and surrounding the RPV, consisting of the drywell and wetwell in a BWR plant.
Primar S stem The pipes, valves, and other equipment which connect directly to the RPV or reactor coolant system such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.
Remove To change the location or position of.
May 1994 Page 98 EPMP-EPP-0101 Rev 00
ATTACHMENT 2 (Cont)
~Re ort To describe as being in a specific state.
~Re uvre To demand as necessary or essential.
Restore Take the appropriate action requires to retur n 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.
~Seta 1e 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."
~Shutdo As applied to the BWR reactor, subcritical with reactor power below the heating range.
May 1994 Page 99 EPMP-EPP-0101 Rev 00
ATTACHHENT 2 (Cont)
Site Area Emer enc Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels except near the site boundary.
Su ression ool The volume of water in a BWR plant intended to condense steam discharged from a primary system break inside the drywell.
Sustained Prolonged. Not intermittent or of transitory nature 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.
Uncontro11ed An evolution lacking control but is not the result of operator action.
~Un 1 armed 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.
Unusual Event Events are in process or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
Hay 1994 Page 100 EPHP-EPP-0101 Rev 00
ATTACHMENT 2 (Cont)
Valid Supported or cor roborated on a sound bases.
Vent 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."
Vital Area Any plant area which contains vital equipment.
May 1994 Page 101 EPMP-EPP-0101 Rev 00
NVIP-1 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 ) 2$ pCi/gm I-181 equivalent Valid offgas radiation > hi-hi alarm 1.1.2 Alert 1.2.2 Alert Coolant activity > 300 pCi/gm I-131 equivalent Valid offgas radiation > 10 x hi-hi alarm Power Operation, Hot Shutdown Power Operation, Hot Shutdown 1-2
Category 1.0 Reactor Fuel 1.0 Reactor Fuel 1.0 Reactor Fuel 18 Containment Radiation 1.4 Other Radiation Monitors 1.8.1 Alert 1.4.1 Unusual Event Drywell radiation 2 20 R/hr Any sustained ARMreading > 100 x alarm (OP-50A) or ofFscale hi resulting from an uncontrolled process Power Operation, Hot Shutdown 1.8.2 Site Area Emergency 1.4.2 Alert Drywell radiation > 3000 R/hr Sustained RB Vent Monitor RN07A5 or B5 > 5 mR/hr Power Operation, Hot Shutdown OR Any sustained refuel Qoor rad monitor ) 8 R/hr or ofFscale hi, Table 1.1 1.3.3 General Emergency Drywell radiation 2 4.0E6 I/hr Power Operation, Hot Shutdown 1.4.8 Alert Sustained area radiation levels > 15 mRQxr 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 1.5 Refueling Accidents 1.4.4 Alert 1.5.1 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 1.5.2 Alert Table 1.1 Refuel Floor Rad Monitors Report of visual observation of irradiated fuel uncovered West End of Shield Wall, RB 340 (N18)
Rx Bldg. - East Wall El 340'N25)
Refuel Bridge (high range) (Process Mon.)
Refuel Bridge (low range) (029)
Table 1.2 Plant Safet Function Areas Reactor Building Turbine Building Screen and Pump House 08'as 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 2.2.1 Alert Unidentified drywell leakage 2 10 gpm Allimmediate manual scrams fail to shut down the OR reactor Reactor coolant to drywell identified leakage > 25 gpm Power operation, startup Ihot standby Power operation, 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
-84 in. (TAF) AND Boron injection is required Power Operation, Hot Shutdown, Cold Shutdown, Refuel Power Operation, Startup/Hot Standby 2.1.8 General Emergency 2.2.8 General Emergency Drywell Flooding required Allimmediate manual scrams fail to shut down the Power Operation, Hot Shutdown reactor AND RPV water level cannot be restored and maintained >
-108 in.
Power Operation, 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 Torus temperature and RPV pressure cannot be maintained < HCTL Power Operation, Startup IHot Standby
Category 8.0 Primary Containment 3.0 Primary Contahunent 8.0 Primary Containment 3.1 Containment Pressure 8.2 Torus Temperature 3.1.1 Alert 8.2.1 Site Area Emergency Primary containment pressure cannot be maintained < Torus temperature and RPV pressure cannot be 8.5 psig due to coolant leakage maintained < HCTL (non-ATWS)
Power Operation, Hot Shutdown Power Operation, Hot Shutdown 8.1.2 Site Area Emergency Primary containment pressure cannot be maintained
< 8.6 psig AND Coolant activity > 300 p.Ci/gm Power Operation, Hot Shutdown 3.1.3 General Emergency Primary containment venting is required due to PCPL Power Operation, Hot Shutdown 3-1
Category 8.0 Primary Contaixnnent 8.0 Primary Containment 8.0 E'rimary Containment 8.8 Combustible Gas Concentration 8.4 Contaimnent Isolation Status 8.8.1 Site Area Emergency 8.4.1 Site Area Emergency
> 4% H2 exists in DW or torus MSL, EC steam line or RWCU isolation failure resulting in a release pathway outside primary Power Operation, Hot Shutdown containment Power Operation, Hot Shutdown 8.8.2 General Emergency Primary containment venting is required due to 8.4.2 General Emergency combustible gas concentrations MSL, EC 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 < -84 in. (TAF)
~ DW radiation > 8000 R/hr Power Operation, Hot Shutdown 8-2
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, Nl-EOP-5 operating levels in two or more areas, Nl-EOP-5 Power Operation, Hot Shutdown Power Operation, Hot Shutdown 4.1.2 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, N1-EOP-5 operating levels in two or more areas, N1-EOP-5 AND any: AND any:
~ Coolant activity > 800 p,Ci/gm I-131 ~ Coolant activity > 800 p,Ci/gm I-131 equivalent equivalent
~ RPV water level < -84 in. (TAF) ~ RPV water level < -84 in. (TAF)
~ DW radiation > 3000 R/hr ~ DW radiation > 8000 R/hr Power Operation, Hot Shutdown Power Operation, Hot Shutdown
Category 6.0 Radioactivity Release /Area Radiation 5.0 Radioactivity Release /Area Radiation 5.0 Radioactivity Release /Area Radiation 6.1 Effluent Monitors 6.2 Dose Projections/ Environmental Measurements/ Release Rates 5.1.1 Unusual Event 6.2.1 Unusual Event A valid reading on any monitors Table 5.1 column "NUE" for > 60 min. Confirmed sample analyses for gaseous or liquid release rates > 2 x technical specifications limits for >
60 min.
5.1.2 Alert A valid reading on any monitors Table 5.1 column 6.2.2 Alert "Alert"for > 15 min, Confirmed sample analyses for gaseous or liquid release rates > 200 x technical specifications limits for
> 15 min.
6.1.8 Site Area Emergency A valid reading on any monitors Table 5.1 column "SAE" for > 15 min. 6.2.8 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 6.0 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.
5-2
Category 6.0 Radioactivity Release / Area Radiation Monitor GE Stack (RN10A/B) 2300 cps 23.0E4 cps 25.0E6 cps N/A EC Vent >10 mR/hr a30 mR/hr 2310 mR/hr N/A SW EQluent 2900 cpm >90,000 cpm N/A N/A RW Discharge a2 x batch &00 x batch 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 mEVhr 1000 mR/hr CDE Thyroid rate N/A 500 mR/hr 5000 mR/hr 5-3
Category 6.0 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 B Loss of power for > 15 min. to all: Loss of all emergency bus AC power for >15 min.
~ T-101N
~ T-101S Power Operation, Hot Shutdown
~ T-10 backfed through T-1 or T-2 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. 54hrs OR Cold Shutdown, Refuel, Defuel RPV water level cannot be restored and maintained > -84 in. (TAF) 6.1.3 Alert Power Operation, Hot Shutdown Available emergency bus AC power reduced to only one of the following sources for >15 min.:
~ DG102 (PB102)
~ DG103 (PB103)
~ T-101N
~ T-101S Power Operation, Hot Shutdown
Category 6.0 Electrical Failures 6.0 Electrical Failures 6.2 Loss of DC Power Sources 6.2.1 Unusual Event
< 106 vdc on battery board 11 and 12 for >15 min.
Cold shutdown, Refuel 6.2.2 Site Area Emergency
< 106 vdc onbatteryboard 11 and 12 for > 15 min.
Power Operation, Hot Shutdown
Category V.O Equipment Failures 7.0 Equipment Failures V.O Equipment Failures 7.1 Technical Speci6cationXRequirements V.2 System Failures or Control Room Evacuation V.1.1 Unusual Event 7.2.1 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, Hot Shutdown seals.
Power Operation, Hot Shutdown 7.2.2 Alert Control Room evacuation V.2.3 Alert Reactor coolant temperature cannot be maintained
< 212 'F Cold Shutdown, Refuel 7-1
Category 7.0 Equipment Failures V.O Equipment Failures 7.0 Equipment Failures 7.2 System Failures or Control Room 7.8 Loss of Indications/Alarm/Communication Evacuation Capability V.2.4 Site Area Emergency 7.8.1 Unusual Event Control Room evacuation Unplanned loss of all annunciators or indicators on all AND panels L, K, H, F, 0 for > 15 min.
Control of core cooling systems cannot be established AND in 5 15 min. Increased surveillance is required for safe plant operation Power Operation, Hot Shutdown V.3.2 Unusual Event Loss of all communications capability aFecting the ability to either:
Perform routine onsite operations OR Notify o6'site agencies or personnel .
7-2
Category 7.0 Equipment Failures V.O Equipment Failures V.3 Loss of Indications/Alarm/Communication Capability V.3.3 Alert Unplanned loss of all annunciators or indicators on all panels L, K, H, F, 0 for > 15 min.
AND Increased surveillance is required for safe plant operation AND either:
Plant transient in progress OR Plant computer and SPDS are unavailable Power Operation, Hot Shutdown 7.3.4 Site Area Emergency Loss of all annunciators or indicators on all panels L, K,H,F,G AND Plant computer and SPDS are unavailable AND Indications to monitor all RPV and primary containment EOP parameters are lost AND Plant transient is in progress Power Operation, Hot Shutdown 7-3
Category 8.0 Hazards 8.0 Hazards 8.0 Hazards 8.1 Security Threats 8.2 Fire or Explosion 8.1.1 Unusual Event 8.2.1 Unusual Event Bomb device or other indication of attempted sabotage Con6rmed flre in any plant area, Table 8.2 or Table discovered within plant Protected Area 8.3, not extinguished in ~ 15 min. of Control Room noti6cation 8.1.2 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 &om the Control Room AND Loss of remote shutdown capability 8-1
Categ 8.0 Hazards 8.0 Hazards 8.0 Hazards 8.3 Man-Made Events 8.3 Man-Made Events 8.3.1 Unusual Event 8.3.4 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 toxic or flammable gases that could enter or have entered within the Protected Area boundary in amounts that could affect the health of plant personnel or safe plant operation OR Report by local, county or state officials for potential evacuation of site personnel based on offsite event 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-1 seismic instrumentation actuated Forebay water level < 238 ft OR Confirmation of earthquake received on NMP-2 or JAFNPP seismic instrumentation 8.4.5 Alert Earthquake felt in plant by any operator 8.4.2 Unusual Event AND NMP-1 seismic instrumentation indicates > 0.11 g Report by plant personnel of tornado striking within plant Protected Area boundary 8.4.6 Alert 8.4.8 Unusual Event Sustained winds > 125 mph OR Assessment by Control Room personnel that a natural Tornado strikes a plant vital area, Table 8.3 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 > 254 ft OR Forebay water level < 236 ft
Category 8.0 Hazards Table 8.2 Plant Areas
~ RadWaste Solidification and Storage Bldg.
~ Security West Bldg.
Table 8.3 Plant Vital Areas
~ Control Room Building
~ Auxiliary Control Room
~ Cable Spreading Room
~ Reactor Bldg.
~ Turbine Bldg.
~ Diesel Generator Area
~ Screen and Pump House
~ OffGas Bldg.
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 Shik Supervisor or Any event, as determined by the ShiS Supervisor or Emergency Director, that could lead to or has led to a Emergency Director, that could lead or has led to a loss potential degradation of the level of safety of the plant. or potential loss of either fuel clad or RCS barrier.
Power Operation, Hot Shutdown 9.1.2 Unusual Event 9.1.5 Site Area Emergency Any event, as determined by the Shik Supervisor or As determined by the Shik Supervisor or Emergency Emergency Director, that could lead to or has led to a Director, events are in progress which indicate actual loss or potential loss of containment. or likely failures of plant systems needed to protect the public. Any releases are not expected to result in Power Operation, Hot Shutdown exposures which exceed EPA PAGs.
9.1.3 Alert Any event, as determined by the Shift Supervisor or Emergency Director, that could cause or has caused actual substantial degradation of the level of safety of the plant.
Category 9.0 Other 9.0 Other 9.0 Other 9.1.6 Site Area Emergency 9.1.8 General Emergency Any event, as determined by the Shift Supervisor or Any event, as determined by the Shik Supervisor or Emergency Director, that could lead or has led to Emergency Director, that could lead or has led to a loss either: of any two fission product barriers and loss or potential Loss or potential loss of both fuel clad and RCS loss of the third.
barrier OR Loss of containment 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, Hot Shutdown
~ Rapid unexplained decrease following initial increase in containment pressure Power Operation, Hot Shutdown 9.1.7 General Emergency As determined by the Shift Supervisor or Emergency Director, events are in progress which indicate actual or imminent core damage and the potential for a large release of radioactive material in excess of EPA PAGs outside the site boundary.
OSSI 93-402A-10-NMP 1 NMP-1 EAL Veriftcatton & Valfdatfon Report, Rev. 0 Emer enc Action Level Verification Br. Validation Re ori Revision 0 New York Power Authori J. A. Fitzpatrick Nuclear Power Plant Indian Point Nuclear Power Plant Unit 3 1
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 93-402A-10-NMP1 NMP-1 EAL Verification Bc Validation Report, Rev. 0 Table of Contents Section ~Pa e Introduction ..........................~..............,.............~............. ~~ ~e ~~ eooo ~ ettoo ~ oteo ~~~~ 1 2, Preparations. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~~~~~~ 1
- 3. P i OCeSS ~o~ ~~~o~~ oto ~~~~o~~~~~ ~~~~~~~~ t~ ~ ~ ~~ ~~ ~ o~o ~~ ~o~ ~ o~o ~~ o~ ~~~ ~ ~~ ~ .4 3 .1 Verification Activities .................,......,.................. .4 3.2 Validation Activities ...~.........~..... ~;...,..... ~ o ~ o ~ ~ ~ o ~ eeoooo ~ ~ too 5 4 Comment Resolution ~ ~ t ~ ~ oooo ~ ~ o ~ o ~ ~ o ~ oeto ~ ~ ~ o ~ ~ o ~ ~ ~ etoooooooottoeto ~o~ ~ ~ t ~ ~ eeo 9 5 References ...................................., ~....... ~ ~ otto ~ o ~ t ~ ~ ~ o ~ o ~ ~ ~ eeoo ~ ~ ~ o ~~ ot ~ ~ ~ ot ~ ~ oo ~ ~ otto ~ ~ ~ o 9 EAL Verification Checklists ......,. ~o~~~ 1 1 EAL Verification Comment Database ...,............~......,..........................................,...2-1 EAL Validation Scenarios. .3-1 EALValidation Summary Sheets. ..........,............. ...................
~ ~ ~ ~~~ ~o~o~~~~~o~o~ tt 04
~
EALValidation Exercise Checklists.. 5-1 EAL Validation Comment Database .. ~ .... ............
~ ~ ~..... ~ ..........~....... .6-1
OSSI 93-402A-10-NMP 1 NMP-1 EAL Verification Bc Validation Report, Rev. 0
- l. Introduction The verification process was performed to ensure the NMP-1 Emergency Action Levels (EALs) and classification procedures are written correctly and are technically correct. The NMP-1 EAL verification was conducted prior to the EAL validation exercises. Verification activities were completed according to Reference l.
The validation process was performed to ensure that the NMP-1 FALs and classification procedures are usable and operationally correct, and to ensure that responsible emergency response organization personnel are able to arrive at consistent interpretations of EALs under varying conditions. The NMP-1 EAL validation exercises were conducted on October 6, 1993 at the Nine Mile Point Training Center NMP-1 control room simulator. Validation activities were completed according to Reference 2.
The NMP-1 EAL verification/validation was one of six verification/validations conducted by OSSI at each of the six participating plants in the NYPA EAL Upgrade Project.
- 2. ~i Mr. C. K. Walker (OSSI) was assigned EAL verification and validation team leader. For EAL verification, he was responsible for:
~ Determining the extent to which the EAL documentation is verified.
~ Selecting team members to conduct EAL verification reviews.
0 OSSI 93-402A-10-NMP1 NMP-1 EAL Verification & Validation Report, Rev. 0
~ Providing appropriate source documents so team members can conduct verification reviews.
~ Coordinating resolutions to any verification review comments.
~ Coordinating update of EAL program documentation consistent with the resolution of verification review comments.
~ Determining the extent to which each selected EAL is validated.
For EAL validation, Mr. Walker was responsible for:
~ Selecting team members to participate as validation exercise observers and as emergency response organization personnel during EAL validation exercises,
~ Preparing a validation exercise test plan and schedule. (EALs selected for validation are documented on the Validation Summary Sheet which served as the validation test plan.)
~ Obtaining appropriate scenarios to test emergency response organization classification activities while using the EALs.
~ Coordinating resolutions to validation exercise comments.
~ Coordinating update of EAL program documentation consistent with the resolution of validation exercise comments.
Mr. J. P. Staley (OSSI) was assigned to the verification team and was responsible for:
~ Becoming familiar with appropriate verification source documents and the NMP-1 EALs to be verified.
~ Performing assigned EAL verification reviews.
OSSI 93-402A-10-NMP 1 NMP-I EAL Verification & Validation Report. Rev. 0
~ Completing verification checklists for technical accuracy and written correctness reviews.
~ Assisting in the preparation of resolutions to veriQcation review comments.
Mr. M. C. Daus (OSSI) and Mr. J. Jones (NMPC) were assigned EAL validation exercise observers. They were responsible for:
~ Becoming familiar with appropriate NMP-I EAL development documents and the EALs to be validated.
~ Observing emergency response organization participants using the EALs while responding to simulated emergency events.
~ Completing the validation exercise checklists
~ Assisting in the preparation of resolutions to validation exercise comments.
Several members of the NMP-1 plant staff were also assigned to the validation team to play the role of emergency response organization positions. Their names and titles are listed on the EAL Validation Summary sheets (Attachment 4). They were responsible for:
~ Becoming familiar with the EALs to be validated.
~ Using the EALs while responding to simulated emergency events.
~ Completing the validation exercise checklists.
OSSI 93-402A-10-NMP1 NMP-1 EAL VerlAcatlon & Valldatlon Report, Rev. 0
- 3. Process 3.1 Verincation Activities The technical accuracy and written correctness of the upgraded EALs were verified through table-top reviews which addressed the following EAL attributes:
Written Correctness Human engineering factors of the EAL Writer's Guide Format, appearance and terminology consistent, to the extent possible, among BWR and PWR plants involved in the NYPA EAL Upgrade Project EAL structure EAL terminology is clear and well defined Technical Accurac Technical completeness and appropriateness for each classification level Potential for classification upgrade only when there is an increased threat to public health and safety Logical progression in classification for combinations of multiple events Consistency of EALs, to the extent possible, among BWR and PWR plant designs The EALs were reviewed in terms of the evaluation criteria embodied in the checklists for technical accuracy and written correctness (Attachments 1 and 2 of Reference 1). EAL verification reviews for technical accuracy and written correctness were accomplished by a comparative Table-Top evaluation of the following:
OSSI 93-402A-10-NMP 1 NMP-1 EAL VeriAcation & Validation Report, Rev. 0
~ Written correctness of the EALs including human factors guidance of the EAL writer's guide.
~ Technical accuracy of the EALs compared to the EAL Technical Basis, EAL Fission Product Barrier Evaluation, Plant-Specific EAL Guideline, EAL Binning Document, and NUMARC NESP-007, Revision 2 (including NUMARC/NRC Questions and Answers).
~ Compatibility of the EALs with the plant.
~ Numerical values, quantitative and calculated information, The Walk-thru method of verification was performed during EAL validation where necessary EAL references to equipment, indications and instrumentation were checked against control room hardware as represented in the simulator control room.
Verification reviews were performed using the applicable sections of the EAL verification checklists (Attachment 1). All discrepancies were documented on EAL Comment Forms in the EAL Verification Comment Database. A printout of this database is provided in Attachment 2.
3.2 Validation Activities The usability and operational correctness of the upgraded EALs were validated through observation of emergency response organization personnel responding to simulated emergency events using the EALs. The group of EALs selected for validation were sufficiently representative to test that the EALs possess the following attributes:
~Usabili
~ User friendliness
~ Ease of place-finding
OSSI 93-402A-10-NMP1 NMP-1 EAL VerlQcation & Valtdatton Report, Rev. 0
~ Ease of place-keeping
~ Ease of upgrading and declassifying 0 erational Correctness Potential for classification upgrade only when there is an increased threat to public health and safety Technical completeness and appropriateness for each classification level A logical progression in classification for combinations of multiple events EALs not selected for validation were compared to the validation checklist criteria at the conclusion of the validation exercises.
The EALs were validated in terms of the evaluation criteria embodied in the checklist for EAL evaluation. EAL validation exercises were conducted using the Table-Top method and the Simulator method. Scenarios were developed and used in the performance of the Table-Top and Simulator methods of validation (Attachment 3). The scenarios provided the means for validation team observers to view emergency response organization personnel consulting the EALs for proper emergency classification.
In the classroom, members of the EAL validation team were introduced to the upgraded EALs by the team leader. Classification categories and subcategories were discussed as were the technical basis for individual EAL conditions. This served to familiarize all validation team participants with the content of the new EALs and-their relationship to the existing classification procedures. Members of the validation team were also briefed on the validation process described in Reference 2. Copies of the upgraded EALs were made available to team members during the validation exercises.
The EAL validation test plan is given on the EAL Validation Summary Sheets in Attachment 4. For each EAL validation scenario, the following activities were performed:
OSSI 93-402A-10-NMP1 NMP-1 EAL Verification 6 Validation Report. Rev. 0
- 1. The validation team members assumed the emergency response organization roles they were expected to fulfill during an actual emergency.
- 2. The team leader described the initial plant conditions.
- 3. When emergency response organization personnel were familiar with initial plant conditions, the team leader announced the start of the scenario exercise and described changes in key plant parameters (for the Table-Top method) or he instructed the simulator facility instructor to place the simulator in RUN (for the Simulator method).
- 4. The emergency response organization personnel described the actions they would perform (for the Table-Top method) or they manipulated appropriate plant controls in the simulator as needed to respond to changing plant conditions (for the Simulator method).
- 5. The emergency response organization personnel consulted the upgraded EALs according to Emergency Plan procedures and made appropriate classifications,
- 6. The team observers occasionaly asked questions of the emergency response organization personnel during the exercise.
- 7. When final plant conditions were reached, the validation team leader stopped the exercise and held a post scenario briefing during which:
~ Team members jointly discussed problems and comments noted during the exercise.
OSSI 93-402A-10-NMP1 NMP-1 EAL Verlf1catton & Valtdatton Report, Rev. 0
~ Team members jointly completed the EAL Validation Exercise Checklists (Attachment 5).
~ Possible reasons for noted problems and comments were discussed.
~ In some cases, portions or all of the exercise were reperformed to gain a better perspective of noted problems and comments.
The validation team leader ensured the following information was recorded on each Validation Exercise Checklist:
Date and plant name Validation team member names and titles EAL identification number of EALs validated Scenario description Validation method Following each post-scenario briefing, team members compared observations and determined if any problems and comments noted thus far required modification of the test plan to achieve validation objectives.
When all validation exercises were completed, the team leader, with the assistance of the team members, consolidated all exercise problems and comments by:
~ Reviewing every problem and comment recorded on the EAL Validation Exercise Checklists.
~ Recording problems and comments in the EAL Validation Comment Database. A printout of this database is provided in Attachment 6.
~ Recording EAL Comment numbers on the EAL Validation Exercise Checklists.
e OSSI 93-402A-10-NMP1 NMP-1 EAL VerIAcation R Valldatlon Report, Rev. 0
- 4. Comment Resolution 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 EAL development were used to identify the scope of the discrepancy and to prepare appropriate solutions.
They prepared resolutions to the discrepancies, determined the impact the final resolutions have on EAL Program documentation, determined the impact the final resolutions have on the plant, and identiQed any required follow-up actions.
Results of the verification and validation comment resolution process were documented in the EAL Verification Comment Database (Attachment 2) and the EAL Validation Comment Database (Attachment 6), respectively.
- 5. References
- l. OSSI 92-402A-6, Emergency Action Level Verification, Revision 0
- 2. OSSI 92-402A-7, Emergency Action Level Validation, Revision 0
OSSI 93-402A-10-NMP1 NMP-1 EAL Verification Bc Valldatfon Report, Rev. 0 Attachment 1 EAL Veri6cation Checklists 1-1
OSSI 92-402A-6-NMP 1 EAL 'Verification Procedure, Rev. 0 Attachment 1 Technical Accuracy Plant: Nine Mle Point 1 Date: 9 20 93 Verifier: J. P. Stale EALU rade Pro ect En ineer 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'? A 0 0 1.2 Does the operating mode applicability of each PEG initiating condition agree with the NESP-007 operating mode applicability? 0 A 0 1.3 Is each PEG EAL derived from a corresponding NESP-007 example EAL applicable to plant specific design'? A 0 0 1.4 Do PEG EALs reflect the intent of the NESP-007 example EALs? 0 ~ 0 1.5 Does the PEG EAL bases reflect the intent of the NESP-007 EAL bases which are applicable to plant specific design? 0 A 0 1-1
OSSI 92-402A-6-NMP 1 EAL Verification Procedure, Rev. 0 Attachment 1 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)? CI ~ 0 1.7 Is each applicable PEG fission product barrier EAL properly considered in the fission product barrier evaluation for this plant? ~ 0 Q 1-2
OSSI 92-402A-6-NMP1 EAL Verification Procedure, Rev. 0 Attachment 1 Technical Accuracy Yes ~N 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'? S Cl Cl 2.2 Is each TB EAL derived from one or more corresponding PEG EALs as defined by the FPBE and EAL Binning Document? Q ~ Q 2.3 Do TB EALs reflect the intent of the PEG EALs from which they are derived? Q ~ Q 2.4 Does the operating mode applicability of each TB EAL agree with the corresponding PEG EAL operating mode applicability? Q ~ Q 2.5 Does the TB EAL bases reflect the intent of the PEG EAL bases and FPBE? 5 Q Cl 2.6 Are the references listed for each TB EAL appropriate and consistent with the PEG:
~ PEG Reference(s)? Q ~ Q
~ Basis Reference(s)? S Q Q 1-3
OSSI 92-402A-6-NMP 1 EAL Verification 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)' 0 ~ 0 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'? S 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 classification for combinations of multiple events within a category'? ~ 0 0 1-4
OSSI 92-402A-6-NMP1 EAL VerIficatfon Procedure, Rev. 0 Attachment 1 Technical Accuracy Yes No NA
3.1 Does the set of EAL categories and subcategories agree with the TB categories and subcategories, respectively? 8 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 applicability'? ~ 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? 5 Q Q 4.2 Is each EAL adequately supported by plant instruments, approved instructions, or other appropriate sources of information'? S Q Q 1-5
OSSI 92-402A-6-NMP1 EAL VerifIcatfon 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
'nformation'
~ 0 0 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' 0 5 0 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-NMP1 EAL Verification Procedure. Rev. 0 Attachment 2 Written Correctness Plant: ¹ine Mile Point 1 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? ~ Q Q 1.2 Is each-subcategory clearly associated with its category? ~ Q Q
- 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 number? 5 Q Q 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? 5 Q Q 2-1
OSSI 92-402A-6-NMP 1 EAL 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:
I
.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'? ~ o o
'3.3 Are EALs consisting of multiple conditions formatted such that each condition and its relationship to other conditions are easily understood'? r o o 3.4 Is wording and abbreviations/ acronyms used in the EALs consistent with the definitions provided in Attachments 1 and 2 of the EAL Writer's Guide? ~ o Q k
3.5 Are EAL conditions expressed quantifiably where possible'? r o o 3.6 Where used, do limit modifiers (<, >, <, >)
simplify presentation of EAL conditions'? ~ Q o 2-2
OSSI 92-402A-6-NMP1 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.)? ~ a a
- 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'? ~ a o 4.2 When an EAL must express an alternate combination of two conditions, are the conditions joined by the logic term OR? ~ o a 4.3 Is the use of AND and OR within the same EAL avoided where possible'? a' a 4.4 Is each EAL condition clear and concise'? ~ a o 2-3
OSSI 92-402A-6-NMP I EAL Verification Procedure, Rev. 0 Attachment 2 Written Correctness Yes No 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' ~ 0 0 5,2 Are column headings with applicable engineering units provided for tables with multiple columns of information' r 0 0 5.3 Where vertical lines separate columns of information, is readability improved? ~ 0 0 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' ~ Q 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-NMP I EAL VeriAcatlon Procedure, Rev. 0 Attachment 2 Written Correctness Yes No NA 694 Is word usage consistent among the EALs'? ~ Q Q 695 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'? ~ 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'? E Q Q
- 7. EAL format:
791 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 bullet'? ~ Q Q 792 Are EAL limit values, value modifiers and value engineering units printed in bold print?
All discrepancies have been recorded on EAL Comment Forms and forwarded to the Verification Team Leader.
Signature: 9 t: ~929 93 2-5
OSSI 92-402A-6-NMP 1 EAL Verification Procedure, Rev. 0 Attachment 3 Inter-Plant EAL Comparison Plants: J. A. FitzPatrick Date: 9 20 93
¹ne Mile 1
¹ine Mile 2 Verifier: J. P. Stale EAL Pro ect En ine r 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'? ~ 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-NMP I EAL VeriAcatIon 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? Cl S 0
- 6. Within the constraints of BWR and PWR plant design, is EAL word usage the same for each plant? ~ o o All discrepancies have been recorded on EAL Comment Forms and forwarded to the Verification Team Leader.
Signature Date: ~920 93 3-2
OSSI 93-402A-10-NMP1 NMP-1 EAL VeriAcation & Validation Report, Rev. 0 Attachment 2 EAL Verification Comment Database 2-1
~ ~ . ~ ~
a ~ ~ a ~ ~
Record No. 5 Date 9/20/93 Name M. C. DauS Originating Site Site Applicability JAF Q JAF Q IP-3 O NMP-2 O Generic BWR 8 General NMP-1 Impact Q IP-2 Q NMP-1 Q Ginna O Generic PWR NMP-2 O NUMARC-007 O Procedure Q Verification Q Training Q Hardware g EAL 3 Technical Bases O Validation O Deviation Q 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!nncirlor ovnlanatinn in tho hacic that inrlinatoc that tho ovictinn ia(nrdinn onnnrnnaccoc uihon 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 0I Closed Record No. 7 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability JAF O JAF O IP-3 O NMP-2 Q Generic BWR 8 General NMP-1 O IP-2 Q NMP-1 O Ginna O Generic PWR Impact NMP-2 O NUMARC-007 O Procedure O Verification O Training O Hardware IHI EAL Q Technical Bases Q Validation O Deviation Q None ca t. System Malf. Ic¹ 3 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 Q JAF Q IP-3 Q NMP-2 Q Generic BWR g General NMP-1 Impact 0 IP-2 C3 NMP-1 0 Ginna p Generic PWR NMP-2 0 NUMARC-007 0 Procedure 0 Verification 0 Training 0 Hardware g EAL 8 Technical Bases 0 Validation 0 Deviation 0 None ca t. Hazards Ic¹ 3 No. 1 Emer. Glass. URUSual 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 Originating Site Site Applicability JAF CJ JAF 0 IP-3 C3 NMP-2 0 Generic BWR rel General NMP-1 Impact P IP-2 P NMP-1 P Ginna 0 Generic PWR NMP-2 0 NUMARC-007 Cl Procedure C3 Verification' Training 0 Hardware g EAL H Technical Bases CI Validation Q Deviation Q None ca t. Barrier IC¹ ** 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 Oi Closed
Record No. 11 ,Date 9/2Q/93 Name M. C. Daus Orlglnating Site Site Applicability JAF P JAF P IP-3 P NMP-2 D Generic BWR E General NMP-1 Impact P IP-2 P NMP-1 P Ginna P Generic PWR NMP-2 P NUMARC-007 p Procedure p Verification D Training D Hardware g EAL D Technical Bases p Validation p Deviation D None ca t. N/A I c¹ ** No. ** 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 Oe Closed Record No. 13 Date 9/2Q/93 Name M. C. DauS Orlglnatlng Site Site Applicability JAF D JAF D IP-3 P NMP-2 H Generic BWR P General NMP-1 Impact P IP-2 D NMP-1 D Ginna D Generic PWR NMP-2 D NUMARC-007 D Procedure p Verification D Training D Hardware E EAL p Technical Bases p Validation D Deviation D 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 "AII".
Resolution Checked "All"for HU1, HA3 operating mode applicability in the PEG.
status 0 Open 0 Resolved/Awaiting Disposition 0 Closed
Record No. 15 ,Date 9/2Q/93 Name M. C. DauS originating Site Site Applicability JAF Q JAF QIP-3 QNMP-2 EGeneric BWR QGeneral NMP-1 QIP-2 QNMP-1 QGinna QGenericPWR Impact NMP-2 Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware 8 EAL Q Technical Bases Q Validation Q Deviation Q None cat. System Malf. Ic¹ 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 olcn o RWR FAI volirfotinn nnmmont status OOpen OResolved/Awaiting Disposition OIClosed Record No. I6 Date, 9/2Q/93 Name M. C. Daus Originating Site Site Applicability JAF QJAF QIP-3 QNMP-2 QGeneric BWR g General NMP-1 QIP-2 QNMP-1 QGinna QGeneric PWR Impact NMP-2 Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware IIEAL QTechnical Bases QValidation QDeviation QNone cat. System Malf. Ic¹ 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 onnlinohlo tn thic Il . hono>>co tho IC'. ic cnonifinollM writton fnr rnlrl ch>>trlnwn onrl rofiiol s~a~us 0 Open O Resolved/Awaiting Disposition Oo Closed
Record No. I7 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability JAF QJAF QIP-3 QNMP-2 EGenericBWR QGeneral NMP-1 Impact Q IP-2 Q NMP-1 Q Ginna Q Generic PWR NMP-2 Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware g EAL QTechnical Bases QValidation Q Deviation Q None cat. System Malf. Ic¹ 2 No. *'mer. 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'.honriorl RWA PFA Il . RR9 tn inrli irlo ctortiin/hnt ctonrlhu mnrlo onrl orlriori ohnuo ornlontotinn tn status OOpen O Resolved/Awaiting Disposition Oi Closed Record No. 18 Date 9/2Q/93 Name M. C. DauS Originating Site Site Applicability JAF QJAF QIP-3 QNMP-2 IEIGeneric BWR QGeneral NMP-1 QIP-2 QNMP-1 QGinna QGenericPWR Impact NMP-2 Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware g EAL Q Technical Bases Q Validation Q Deviation Q None cat. System Malf. Ic¹ 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?
Re so lut ion 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.
r'.honriorl RWA PFC~ Ir. c l~9 tn innl>>rlo ctort>>n/hnt ctonrlhu mnrlo onrl orlriorl ohnuo ovnlontotinn tn s~a~us O Open O Resolved/Awaiting Disposition Oe Closed
Record No. 22 Date 9/2Q/93 Name M. C. DauS Orig lnatlng Site Site Applicability NMP-2 Q JAF Q IP-3 Q NMP-2 Q Generic BWR H General NMP-1 Impact Q IP-2 Q NMP-1 Q Ginna Q Generic PWR Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware g EAL 8 Technical Bases Q Validation Q Deviation Q None I C¹ ca t. Abnorm. Rad. No. Emer. Class.
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 Orlglnatlng Site Site Appllcabillty NMP-1 Q JAF Q IP-3 H NMP-2 Q Generic BWR Q General NMP-2 Impact Q IP-2 g NMP-1 Q Ginna Q Generic PWR Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware gg EAL El Technical Bases Q Validation Q Deviation Q None cat ~ Abnorm. Rad. Ic¹ 1 No. ** Emer. Class. Site Al'ea Comment (verification) AS1: Note "laters" here, both units.
Resolution Still waiting for numbers from NMP.
status Oe Open 0 Resolved/Awaiting Disposition O Closed
Record No. 24 Date 9/2Q/93 Name M. C. DauS Originating Site Site Applicability NMP-1 OJAF OIP-3 ENMP-2 PGenericBWR PGeneral Impact Q IP-2 IH NMP-1 P Ginna P Generic PWR 0 NUMARC-007 CI Procedure 0 Verification 0 Training CI Hardware HEAL 8 Technical Bases Cl Validation D Deviation D None cat. Hazards lc¹ 5 No. Emer. Class. UnuSual EVent Comment (verification) NMP-1: HU5 bases: Para. 4 references HU1.5, should reference HU1.4.
Resolution Changed reference to HU1.4.
~
s~a~us OOpen 0 Resolved/Awaiting Disposition Oi Closed Record No. 25 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-1 0 JAF OIP-3 ONMP-2 HGeneric BWR OGeneral OIP-2 ONMP-1 OGinna ClGeneric PWR Impact Q NUMARC-007 Q Procedure 0 Verification C3 Training CI Hardware g EAL H Technical'Bases D Validation Q Deviation 0 None Ic¹ ** No.
- Emer. Class. UnuSual EVent ca t. Hazards Comment (verification) NMP-1, HA-1: If the basis for this EAL is 0.11 g, I don't understand why you can' declare the event unless NMP-2 gets confirmation of a (lesser) seismic event of 0.075 g. This is not explained in the basis.
Resolution Added to the basis for HU1.1 and HA1.1: Confirmation of an earthquake from NMP-2 seismic instrumentation and recognition by plant operations personnel is included in this EAL to ensure that event declaration does not result from a spurious seismic alarm.
Alen morlo cimilor rhonnoc tn NhilPP onrI.IAFNPP whoro onnlirohlo status 0 Open 0 Resolved/Awaiting Disposition Oi Closed
Record No. 26 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-1 QJAF QIP-3 QNMP-2 HGenericBWR QGeneral QIP-2 QNMP-1 QGinna QGenericPWR Impact Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware g EAL !3I Technical Bases Q Validation Q Deviation Q None ca t. Hazards Ic¹ 1 No.
- Emer. Class. Alert Comment (verification) NMP-1, NMP-2 HA-1: Last two paragraphs of generic basis statements are missing.
(All after "EAL 7 covers...)
Resolution Added the missing words to NMP1, 2 and JAF: These EALs can also be a 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.). [water levels rnrrncnnnrl tn fl~torhl status Oopen OResolved/Awaiting Disposition OIClosed Record No. 27 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-1 Q JAF QIP-3 ENMP-2 QGeneric BWR QGeneral NMP-2 QIP-2 HNMP-1 QGinna QGeneric PWR impact 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 7 No. 1 Emer. Class. 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 power!
Resolution Agree, but that's what we have from their data sources.
s~a~us O Open O Resolved/Awaiting Disposition Oi Closed
Record No. 28 Date 9/2Q/93 Name M. C. DauS Originating Site Site Applicability NMP-1 O JAF O IP-3 8 NMP-2 O Generic BWR O General NMP-2 Impact O IP-2 8 NMP-1 O Ginna O Generic PWR O NUMARC-007 O Procedure O Verification OTraining O Hardware g EAL 8 Technical Bases O Validation O Deviation O None Ca t. SyStem Malf. I C¹ 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.
status 0 Open 0 Resolved/Awaiting Disposition Oo Closed Record No. 30 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-1 O JAF O IP-3 O NMP-2 8 Generic BWR O General O IP-2 O NMP-1 O Ginna O Generic PWR Impact O NUMARC-007 O Procedure O Verification O Training O Hardware g EAL H Technical Bases O Validation O Deviation O None cat. System Malf. Ic¹ 1 No. ** Emer. Glass. Site Area Comment (verification) NMP-1, NMP-2 SS1 bases: Last two paragraphs deleted from PEG.
Resolution Added missing two paragraphs to NMP1, 2 and JAF.
status 0 Open 0 Resolved/Awaiting Disposition Oe Closed
Record No. 31 .Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-1 0JAF DIP-3 HNMP-2 OGeneric BWR C3General NMP-2 PIP-2 HNMP-1 PGinna OGenericPWR Impact Cl NUMARC-007 C3 Procedure Cl Verification 0 Training Cl Hardware g EAL ILTechnical Bases OValidation ODeviation CINone Ic¹ 2 No.
- Emer. Class. Site AI'ea cat. System Malf.
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.
s~a~us OOpen OResolved/Awaiting Disposition OClosed Record No. 32 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-1 C3 JAF Q IP-3 CINMP-2 EGeneric BWR QGeneral NMP-2 HIP-2 QNMP-1 OGinna OGeneric PWR Impact JAF C3 NUMARC-007 0 Procedure Q Verification 0 Training 0 Hardware g EAL 8 Technical Bases 0 Validation 0 Deviation 0 None ca t. System Malf. Ic¹ 2 No. ** Emer. Class. Site Area 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 Oe Closed
Record No. 33 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability NMP-1 P JAF P IP-3 Q NMP-2 g Generic BWR CI General NMP-2 Impact CI IP-2 0 NMP-1 0 Ginna 0 Generic PWR JAF 0 NUMARC-007 Cl Procedure 0 Verification D Training 0 Hardware g EAL 8 Technical Bases D Validation C] Deviation Cl None ca t. System Malf. lc¹ 5 No. ** Emer. Class. Site Area 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 8 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 Originating Site Site Applicability NMP-1 Q JAF P IP-3 Q NMP-2 Q Generic BWR 8 General NMP-2 Impact P IP-2 P NMP-1 P Ginna P Generic PWR JAF Q NUMARC-007 C] Procedure 0 Verification 8 Training C3 Hardware g EAL H Technical Bases Q Validation 0 Deviation C3 None cat ~ System Malf. lc¹ 5 No. ** Emer. Class. Site Al'ea 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%. 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 hnttor Rofino this FAI status 0 Open 0 Resolved/Awaiting Disposition OI Closed
Record No. 36 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-1 P JAF P IP-3 P NMP-2 P Generic BWR E General NMP-2 Impact Cj IP-2 P NMP-1 P Ginna P Generic PWR O NUMARC-007 p Procedure p Verification E Training p Hardware 8 EAL 8 Technical Bases p Validation p Deviation p None cat. System Malf. ic¹ 5 No. *'mer. 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 Oi Closed Record No. 37 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-1 P JAF P IP-3 P NMP-2 8 Generic BWR P General NMP-2 Impact P IP-2 P NMP-1 P Ginna P Generic PWR JAF p NUMARC-007 p Procedure p Verification g Training p Hardware g EAL 8 Technical Bases p Validation p Deviation p None Cat. SyStem Malf. IC¹ 5 No. ** Emer. Class. Site Area 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 0 Open 0 Resolved/Awaiting Disposition Oe Closed
Record No. 38 .Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-1 Q JAF Q IP-3 8 NMP-2 Q Generic BWR Q General NMP-2 Q IP-2 IHI NMP-1 Q Ginna Q Generic PWR Impact Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware g EAL 8 Technical Bases Q Validation Q Deviation Q None ca t. Barrier IC¹ ** 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 Qo Open -0 Resolved/Awaiting Disposition 0 Closed Record No. 39 Date 9/2Q/93 Name M. C. DauS Originating Site Site Applicability NMP-1 Q JAF Q IP-3 Q NMP-2 H 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 g EAL 8 Technical Bases Q Validation Q Deviation Q None ca t. Barrier lc¹ ** No. ** 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 rnnrlitinnc>>nrtor tho iiidnornont FAI Pr'.R 1 status P Open 0 Resolved/Awaiting Disposition 0 Closed
Record No. 40 Date 9/2Q/93 Name M. C. Daus Orlglnatlng Site Site Applicability NMP-1 0 JAF 0 IP-3 C3 NMP.-2 IH Generic BWR Cl General NMP-2 Impact CI IP-2 CI NMP-1 C3 Ginna Q Generic PWR JAF Q NUMARC-007 CJ Procedure 0 Verification 0 Training Cl Hardware 8 EAL 8 Technical Bases C3 Validation 0 Deviation 0 None ca t. Barrier I c¹ ** 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.
Re so lut ion 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=(~ rnnditinn onnnmnoccoc tho Nl ihilARr'. rnnditinn rnnnorninn RPEI wotor loMol onrl tho status 0 Open 0 Resolved/Awaiting Disposition Oi Closed Record No. 41 Date 9/2Q/93 Name M. C. Daus Orlglnatlng Site Site Applicability NMP-1 0 JAF Q IP-3 P NMP-2 8 Generic BWR P General NMP-2 Impact Q IP-2 P NMP-1 P Ginna C3 Generic PWR JAF Q NUMARC-007 0 Procedure 0 Verification Q Training 0 Hardware g EAL g Technical Bases 0 Validation 0 Deviation C] None ca t. Barrier lc¹ ** No. ** 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 0 Open 0 Resolved/Awaiting Disposition OI Closed
Record No. 46 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability NMP-1 OJAF OIP-3 ONMP-2 HGenericBWR OGeneral NMP-2 Impact P IP-2 Q NMP-1 P Ginna C3 Generic PWR 0 NUMARC-007 0 Procedure 0 Verification CI Training C3 Hardware IjEAL H Technical Bases Q Validation 0 Deviation Cl None cat. Barrier Ic¹ ** No.
- Emer. Class.
Comment (verification) Remark ¹1 5: 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 drywell >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 irfontifv on FAI hocorl nn o uoru orhitrorM niimhor (RA nnmh whon onM ciihctontiol omniint status 00pen 0 Resolved/Awaiting Disposition 0I Closed Record No. 62 Date 9/20/g3 Name M. C. Daus Originating Site Site Applicability NMP-1 C] JAF OIP-3 ONMP-2 IIGeneric BWR C]General NMP-2 C3IP-2 QNMP-1 C3Ginna OGeneric PWR Impact Q NUMARC-007 Cj Procedure 0 Verification 0 Training C3 Hardware g EAL IITechnical Bases 0 Validation 0 Deviation 0 None cat. Barrier I C¹ ** No. Emer. Class.
- 'omment (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 00pen 0Resolved/Awaiting Disposition OiClosed
Record No. 66 .Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-1 CI JAF 0 IP-3 Q NMP-2 g Generic BWR Q General NMP-2 HIP-2 QNMP-1 PGinna PGenericPWR Impact JAF 0 NUMARC-007 0 Procedure 0 Verification 0 Training 0 Hardware g EAL rajTechnical Bases QValidation ODeviation ONone cat. Barrier I C¹
- 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.
status 0 Open 0 Resolved/Awaiting Disposition Closed Record No. 67 Date 9/2Q/93 Name M. C. Daus Orig inatlng Site Site Applicability NMP-1 QJAF HIP-3 C7NMP-2 EGeneric BWR PGeneral NMP-2 Impact Q IP-2 P NMP-1 Q Ginna P Generic PWR JAF p NUMARC-007 C3 Procedure Q Verification C] Training p Hardware g EAL H Technical Bases 0 Validation C] Deviation 0 None cat. Barrier Ic¹ ** N
- Emer. Class. **
Comment (verification) AA2.1 is duplicated, in 1.4 and 1.5.
Resolution Deleted AA2.1 from subcategory 1.5.
status OOpen 0 Resolved/Awaiting Disposition Oi Closed
~
Record No. 69 Date 9/2Q/93 Name M. C. Daus Originating 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. Barrier I C¹
- No. ** Emer. Class.
Comment (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 9/2Q/93 Name M. C. Daus rig lnating Site Site Applicability NMP-1 Q JAF Q IP-3 Q NMP-2 E 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 rgj EAL 8 Technical Bases Q Validation Q Deviation Q None ca t. Barrier I C¹ ** No. ** Emer. Class.
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. 7I Date 9/2Q/93 Nsme M. C. Daus Originating Site Site Applicability NMP-1 Q JAF Q IP-3 P NMP-2 E Generic BWR P General NMP-2 Impact 0 IP-2 C3 NMP-1 Cl Ginna 0 Generic PWR JAF 0 NUMARC-007 0 Procedure 0 Verification 0 Training 0 Hardware IHI EAL H Technical Bases 0 Validation 0 Deviation Q None ic¹ ** No. ** **
cat. 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 Oo Closed Record No. 72 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-1 Q JAF 0 IP-3 0 NMP-2 g Generic BWR Q General NMP-2 Impact C] IP-2 0 NMP-1 D Ginna Q Generic PWR 0 NUMARC-007 0 Procedure C3 Verification 0 Training C3 Hardware g EAL H Technical Bases 0 Validation Q Deviation C3 None Ic¹ ** No. ** **
cs t. Barrier Emer. Class.
Comment (verification) NMP-1, 2, NUE 1.1.1: The stated basis for this doesn't read much like the PEG, although it seems OK...
Resolution Agree.
status 0 Open 0 Resolved/Awaiting Disposition Oe Closed
Record No. 74 Date 9/20/93 Name M. C. DauS Originating Site Site Applicability NMP-1 Q JAF Q IP-3 8 NMP-2 Q Generic BWR Q General NMP-2 Impact Q IP-2 IHI NMP-1 Q Ginna Q Generic PWR Q NUMARC-007 Q Procedure Q Verification Q Training Q Hardware g EAL 8 Technical Bases Q Validation Q Deviation Q None ca t. Barrier lc¹ ** No. ** Emer. Class.
- 'omment (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 1-131.
Resolution Added discussion in EAL TB basis to PEG EAL basis for NMP1, 2. JAF is ok.
status 0 Open O.Resolved/Awaiting Disposition Oe Closed Record No. 75 Date 9/20/93 Name M. C. Daus riglnating Site Site Applicability NMP-1 Q JAF Q IP-3 Q NMP-2 8 Generic 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 Q Hardware g EAL H Technical Bases Q Validation Q Deviation Q None ca t. Barrier I C¹ ** No. ** 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 OI Closed
Record No. 76 Date 9/2Q/93 Name M. C. Daus Originating Site Site Ap plica bill ty NMP-1 I7 JAF P IP-3 8 NMP-2 P Generic BWR CJ General NMP-2 Impact Q IP-2 g NMP-1 C] Ginna Q Generic PWR Q NUMARC-007 Q Procedure 0 Verification 0 Training 0 Hardware g EAL ETechnical Bases CIValidation I7Deviation QNone cat. Barrier I c¹ 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 Qe Closed Record No. 77 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-1 P JAF P IP-3 Q NMP-2 H Generic BWR C3 General NMP-2 Impact P IP-2 Q NMP-1 CI Ginna Q Generic PWR C3 NUMARC-007 C3 Procedure 0 Verification C3 Training 0 Hardware g EAL 8 Technical Bases 0 Validation 0 Deviation 0 None c¹ ** No. ** **
cat. Barrier I 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 0 Open 0 Resolved/Awaiting Disposition Oe Closed
Record No. 78 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-1 0 JAF 0 IP-3 0 NMP-2 H Generic BWR 0 General NMP-2 Impact 0 IP-2 0 NMP-1 0 Ginna 0 Generic PWR 0 NUMARC-QQ7 0 Procedure 0 Verification 0Training 0 Hardware g EAL 8 Technical Bases 0 Validation 0 Deviation 0 None ca t. Barrier I c¹ ** ** Emer. Class. **
Comment (verificatiori) 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 0 Open 0 Resolved/Awaiting Disposition Oe Closed Record No. 79 Date 9/2Q/93 Name M. C. DauS Originating Site Site Applicability NMP-1 0 JAF 0 IP-3 0 NMP-2 El Generic BWR 0 General NMP-2 Impact 0 IP-2 0 NMP-1 0 Ginna 0 Generic PWR
- 0 NUMARC-QQ7 0 Procedure 0 Verification 0 Training 0 Hardware
.g EAL 8 Technical Bases 0 Validation 0 Deviation 0 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 Oe Closed
Record No. 80 Date 9/2Q/93 Name M. C. DauS Originating Site Site Applicability NMP-1 O JAF O IP-3 H NMP-2 O Generic BWR O General NMP-2 O IP-2 O NMP-1 O Ginna O Generic PWR Impact O NUMARC-007 O Procedure O Verification O Training O Hardware g EAL 8 Technical Bases O Validation O Deviation O None ca t. Barrier I c¹ ** ** 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 0 Resolved/Awaiting Disposition 0 Closed Record No. 82 Date 9/2Q/93 Name M. C. Daus
'Originating Site Site Applicability NMP-1 O JAF O IP-3 O NMP-2 O Generic BWR E General NMP-2 O IP-2 O NMP-1 O Ginna O Generic PWR Impact O NUMARC-007 O Procedure O Verification O Training O Hardware 8 EAL 8 Technical Bases O Validation O Deviation O None ca t. Barrier I C¹
- No. ** Emer. Class.
- 'omment (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.
status 0 Open 0 Resolved/Awaiting Disposition Oi Closed
Record No. 84 ,Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-1 OJAF HIP-3 CINMP-2 QGeneric BWR g General NMP-2 Impact P IP-2 P NMP-1 Cl Ginna P Generic PWR 0 NUMARC-007 OProcedure OVerification OTraining OHardware g EAL 8 Technical Bases Cl Validation Q Deviation Q None cat. Barrier I C¹
- ** Emer. Class.
Comment (verification) NMP1, 2: For EAL 7.3.4, see comment 013 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 thin FAI status 0 Open 0 Resolved/Awaiting Disposition 0+ Closed Record No. 85 Date 9/2Q/93 Name M. C. Daus Originating Site Site Applicability NMP-1 0 JAF 0 IP-3 0 NMP-2 0 Generic BWR mt General NMP-2 Impact OIP-2 DNMP-1 C3Ginna QGenericPWR C3 NUMARC-007 C3 Procedure Q Verification C3 Training Q Hardware g EAL IHITechnical Bases OValidation 0 Deviation 0 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.
s~a~us OOpen OResolved/Awaiting Disposition 0+Closed
Record No. 87 Originating Site
'Date 9/20/93 Name Site Applicability M. C. Daus NMP-1 0 JAF 0 IP-3 C3 NMP-2 0 Generic BWR 8 General NMP-2 0 IP-2 0 NMP-1 0 Ginna C] Generic PWR Impact 0 NUMARC-007 0 Procedure 0 Verification Cl Training 0 Hardware 8 EAL 8 Technical Bases 0 Validation 0 Deviation 0 None Class.
cat. Barrier I C¹ Emer.
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 0 Resolved/Awaiting Disposition Oe Closed Record No. 89 Date 9/20/93 Name M. C. Daus Originating Site Site Applicability NMP-1 P JAF Q IP-3 8 NMP-2 P Generic BWR 0 General Impact Q IP-2 8 NMP-1 P Ginna P Generic PWR P NUMARC-007 Q Procedure 0 Verification 0 Training 0 Hardware g EAL H Technical Bases Q Validation 0 Deviation C3 None ca t. Barrier IC¹ ** ** Emer. Class.
Comment (verification) NMP1: In EAL 5.1.1, the Effluent Monitor Classification Thresholds table states (later)
CPS for the Stack Monitor at the alert level. The PEG says 30,000 CPS. The EC vent for the Table states (later) and the PEG states 1000 mR/hr.
Resolution PEG needs to be revised when NMP provides proper data.
status 0 Open 0 Resolved/Awaiting Disposition 0 Closed
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a
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Da te 9/20/93 M. C. Daus Recor'd"~No 1 Department NuCleal'ame Ex t. 408-274-9587 o riginating site Q JAF Q IP-3 Q NMP-2 0 NYPA OOSSI " st Mo"itic": 11/9/93 HIP-2 HNMP-1 QGinna PNMPC comment General:
8.2.2: Need to make sure equipment pertains to that needed for safe plant shutdown.
RN10A/B: should be replaced for RAGEMS because it will be on line at all times. OGESMS will go on standby.
- E'A'L'¹ ,',".,,'.
- "::."-',srrii:;:AP'Pllcabiiiiy'"',O'JAP'!CI'IP.-::3.'::,,";:!CINMP,'-'2lICIG'e'nerIc!BwR!!iaiP'erieraI.:-:;:;
~BP P
- .:Rec'o'gnition:."RCS ~i'IEm'er'";:,",Cla's's ** ",,"'.;,;,",, I C¹~*:'-"':i,.-";:,. "E'A'L¹. '~*
Resolution Added "... needed for safe plant operation" to EAL for all plants.
Even though OGESMS will go on standby in the future, the EALs must be written to agree with the plant as it will exist when the EALs are implemented. RAGEMS will not replace OGESMS until later.
Continuation Sheet Attached
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Date 9/20/93 Name M. C. Daus Record No. 2 Department NuClear Ext. 408-274-9587 originating site Q JAF Q IP-3 Q NMP-2 Q NYPA Q OSSI Last Modified: 11/9/93 Q IP-2 8 NMP-1 P Ginna 0 NMPC Comment Draft: A Scenario 2, question: Need to add reference to OP-50A, Area Radiation Monitoring System, Revision 7 Attachments 2 and 3, in EAL 1.4.1 for alarm setpoint values.
EAL¹i~]4 ],
'NUMARC Classification:
':ella Appllcarrrrrry CI JAF,:0:IP-3 CI NMP-2 '::CI Geneiic'.BWR PIP;2 HNMP.-1 'Q:Ginna PGeneric'PWR-
';Ben'ei'al' Race'Enlllcn: N/A .. Ernar. Class. ** IC¹~* 'EAL¹,~*
. Cat.
- ';0 EAL"':-,"':."""."'::.:'"-:.".-.---'-:-':=";ClTecshneical Ba/s'e's.".'-'.Q:II/alidcaetioari'- -:C3'DeryIatIoii C3 N'ocn'e.':--'"-"
Resolution Added reference as suggested.
Continuation Sheet Attached
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Da te 9/20/93 Name M. C. Daus Record No. 3 Department NuClear Ext. 408-274-9587 o rlglnatlng site Q JAF Q IP-3 Q NMP-2 Q NYPA Q OSSI Last Modlfled: 11/9/93 QIP-2 HNMP-1 QGinna QNMPC Comment Draft: A Scenario 7, question 10: Does a large break loca which is handled per design and does not threaten primary containment a General Emergency even if rad in drywell is equivalent to 20%
fuel damage?
EAL¹:~N(A site -'Applicability Q JAF p:IP-3 ., Q NMP-2 Q Generic BWR "-Q General.
OIP-2 HNMP-1 Q Ginna QGeneric"PWR NUMARC Classlf Ication:
Recognltlon Cat.
'/A Emer. Class. *" lc¹ **: .EAL¹'",
'}
';Q NUMARC-",007::-.-,:"';-O'Procedure:...;;:--.':.', -::=";Q:Verification Q Training:Q Hardware EA'L'-, =':::'-.':::-::::::-;:.'-':~".";.;:;:IQ~Te'chnIcal,"::Ba'se's'=:-.-':;:.Q:,Validation -'Q;Deviation Cj None.'esolution No.
Continuation Sheet Attached
OSSI 93-402A-10-NMP1 NMP-1 EAL Verification & Validation Report. Rev. 0 Attachment 3 EAL Validation Scenarios 3-1
OSSI 92-402A-7A-NMP 1 EAL Validation Procedure, Rev. 0 Attachment 2 Validation Exercise Scenario Checklist No.:
Plant; NMP-1 Simulator: S Table-Top: 0 Scenario 8 1 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.
T101N and T101S fail to energize when the generator trips (UE 6,1.1), DGs start -and energize emergency busses.
EC steam line ruptures due to pressure spike and EC isolation valves fail to isolate.
Emergency RPV depressurization due to secondary containment maximum safe operating temperature 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, (106 vdc on all batteries (SAE 6.2.2),
and loss of offsite power (UE 6.1.1); loss of annunciators and indicators and increased surveillance with transient in progress (Alert 7.3.3). [If explosion were to have occured while in cold shutdown/refuel, battery loss per UE 6.2.1 and loss of emergency AC per Alert 6.1.2.]
RPV water decreases CZAF (SAE 2.1.1); with primary system discharging outside primary containment and RB temperatures above MSO levels in two or more areas (GE 4.1.2).
Some fuel damage occurs with core uncovery. RB ARMs increase above MSO values in more than two areas; with primary system discharging outside primary containment (SAE 4.2.1, SAE 7.3.4, GE 4.2.2).
2-1
OSSI 92-402A-7A-NMP I EAL Validation Procedure, Rev. 0 Attachment 2 Validation Exercise Scenario Checklist No.:
Plant: NMP-1 Simulator: ~ Table-Top: Cl Scenario ¹2 Scenario Description(s):
Initial Conditions: Reactor power 100% for past 3 months, small steam leak from one turbine throttle valve, plant to shutdown tomorrow With the plant at 100% power and a small steam leak on turbine throttle valve, a dropped control rod results in fuel clad failure.
Reactor scrams; ARMs near HCUs exceed 100 times alarm setpoint (UE 1.4, 1),
Offgas activity increases (UE 1.2.1).
One MSL fails to isolate (SAE 3.5.1).
Offsite radioactivity release increases to the General Emergency level (UE 5.1.1, Alert 1.2.2 and 5.1.2, SAE 5.1.3, GE 5.1.4).
Drywell radiation readings increase (Alert 1.3.1).
Coolant sample results (>300 pCi/gm) support high offsite radiactivity readings (UE 1.1.1, Alert 1.1.2, GE 3.5.2).
Emergency RPV depressurization is required.
2-2
OSSI 92-402A-7A-NMP 1 EAL Valfdatlon Procedure, Rev. 0 Attachment 2 VaIMation Exercise Scenario Checklist No.:
Plant: NMP-1 Simulator: 8 Table-Top: 0 Scenario ¹3 Scenario Description(s):
Initial Conditions: Reactor power 60%, return to power delayed with feedwater heater problems.
Earthquake causes seismic activity alarms at JAFNPP and NMP-1/2 (UE 8.4.1).
Small loca into drywell, unidentified leakage >10 gpm (UE 3.1.1).
Drywell pressure > scram setpoint (Alert 3.2.1).
Multiple failures of RPV injection systems RPV water level decreases < TAF (SAE 2.1.1) 2-3
OSSI 92-402A-7A-NMP I EAL Valtdatlon Procedure. Rev. 0 Attachment 2 Validation Exercise Scenario Checklist No.:
Plant: NMP-1 Simulator: 0 Table-Top: C3 Scenario 0 4 Scenario Description(s):
Initial Conditions: Reactor power 100%, no equipment OOS.
Condensate header ruptures resulting in a loss of feed.
When the reactor scrams on low RPV water level, several control rods fail to insert (Alert 2.2.1). Reactor power remains above 5%.
HPCI does not remain operable.
Various other failures leave one low pressure ECCS pump for RPV makeup.
Boron injection is requred (SAE 2.2.2).
SRV operation heats suppression pool and leak from suppression pool causes water level to decrease. RPV pressure and suppression pool temperature cannot be maintained below the HCTL (SAE 3.3.1, GE 2.2.3);
RPV water level cannot be maintained below MSCRWL (GE 2.2.4),
Hydrogen concentration in the suppression chamber reaches 4% (SAE 3.4,1).
2-4
OSSI 92-402A-7A-NMP 1 EAL Validation Procedure, Rev. 0 Attachment 2 VaHdation Exercise Scenario Checklist No.:
Plant: NMP-1 Simulator: ~ Table-Top: 0 Scenario ¹5 Scenario Description(s):
Initial Conditions: Reactor power 75%, a shutdown is in progress for a drywell entry to locate unidentified leakage, wetwell is deinerted, drywell deinertion in progress with drywell oxygen concentration 10%. No equipment OOS.
Loss of offsite power occurs (UE 6.1.1).
Reactor fails to scram, 30 rods out (Alert 2.2.1).
When RPS fuses pulled, all rods fully insert.
All but one DG fails to start (Alert 6.1.3).
Remaining DG trips (SAE 6.1.4).
Major LOCA occurs, RPV water level cannot be restored and maintained above TAF (GE 6.1.5).
After 20 minutes, one DG returned to operation; available injection cannot restore RPV water level above TAF.
Drywell hydrogen >4% (SAE 3.4.1).
Drywell flooding is required (GE 3.1.2).
Hydrogen in drywell >6% (GE 3.4,2).
Primary containment is vented due to DWPL (GE 3.2.2).
2-5
OSSI 92-402A-7A-NMP1 EAL Valtdatton Procedure, Rev. 0 Attachment 2 Validation Exercise Scenario Checklist No.:
Plant NMP-1 Simulator: 0 Table-Top: ~
Scenario 8 6 Scenario Description(s):
After elevated offgas levels are noted, reactor coolant samples indicate coolant activity > 25 pCi/gm I-131 eq. (UE 1.1.1)
Following reactor shutdown and depressurization, coolant samples are taken indicating 390 pCi/gm I-131 eq. (Alert 1.1.2) 2-6
OSSI 92-402A-7A-NMP I EAL Validation Procedure, Rev. 0 Attachment 2 Validation Exercise Scenario Checklist No.:
Plant: NMP-1 Simulator: Q Table- Top: E Scenario 8 7 Scenario Description(s):
Reactor scrams on high drywell pressure. Drywell radiation levels indicate
[Later] R/hr. (Alert 1.3.1).
Following emergency RPV depressurization, drywell radiation levels of
[Later] R/hr are indicated (SAE 1.3.2).
At what level would you declare a General Emergency based on drywell radiation levels'GE 1.3.3) 2-7
OSSI 92-402A-7A-NMP I EAL Validation Procedure, Rev. 0 Attachment 2 VaHdation Exercise Scenario Checklist No.:
Plant: NMP-1 Simulator: 0 Table-Top: ~
Scenario 8 8 Scenario Description(s):
A HP Technician performing routine surveys measures Control Room area radiation levels of 20 mR/hr (Alert 1.4.3).
It is reported that an unshielded radiography source is in the Relay Room.
General area radiation levels in the relay room are approximately 20 R/hr (Alert 1.4.4) 2-8
OSSI 92-402A-7A-NMP 1 EAL Valldatfon Procedure, Rev. 0 V
Attachment 2 VaHdation Exercise Scenario Checklist No.:
Plant NMP-1 Simulator: Cl Table-Top: ~
Scenario 8 9 Scenario Description(s):
Refueling operations are in progress and a main steam line plug begins to leak causing the refueling cavity and spent fuel pool level to drop. The SFP low level alarm is recieved (UE 1.5.1).
A fuel bundle is on the grapple and in the cattle shute when the refuel floor is evacuated (Alert 1.5,2).
The refuel floor radiation monitors go offscale high (Alert 1.4.2) 2-9
OSSI 92-402A-7A-NMP 1 EAL Validation Procedure, Rev. 0 Attachment 2 Validation Exercise Scenario Checklist No,:
Plant: NMP- 1 Simulator: Q Table-Top: ~
Scenario 8 10 Scenario Description(s):
Chemistry reports stack effluent analysis indicates that effluents have been approximately 3 times Tech. Spec. 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).
P 300 times Tech. Spec. for the last 20 minutes (Alert 5.2.2) 2-10
OSSI 92-402A-7A-NMP 1 EAL Validation Procedure, Rev. 0 Attachment 2 Validation Exercise Scenario Date: Checklist No.:
Plant: NMP-1 Simulator: C3 Table-Top: ~
Scenario 8 11 Scenario Description(s):
Field survey teams report whole body dose rates at the site boundary of 20 mR/hr (Alert 5.2.3).
200 mR/hr (SAE 5.2.4)
Dose projections indicate child thyroid doses of 7200 mR (GE 5.2.5) 2-11
OSSI 92-402A-7A-NMP 1 EAL Validation Procedure, Rev. 0 Attachment 2 Validation Exercise Scenario Checklist No.:
Plant: NMP-1 Simulator: C3 Table-Top: ~
Scenario 8 12 Scenario Description(s):
The plant has entered a 24 LCO action statement at 0700 due to DG operability. At 1800 a plant shutdown is initiated. At 0700 the following day, coolant temperature is still 220 'F while attempting to initiate shut down cooling (UE 7.1.1)
Shutdown cooling cannot be established due to a failure of SDC suction valve.
Reactor temperature cannot be reduced to 212 'F (Alert 7.2.3) 2-12
OSSI 92-402A-7A-NMP 1 EAL Validatfon Procedure, Rev. 0 Attachment 2 VaHdation Exercise Scenario Checklist No.:
Plant NMP-1 Simulator: 0 Table-Top: 8 Scenario ¹ 13 Scenario Description(s):
A tanker carrying ammonia gas overturns on the access road releasing ammonia gas. The plume crosses onto the site, incapacitating numerous site personnel (UE 8.3.3)
The gas then enters the control room requiring the control room to be evacuated (Alert 8.3.5 & 7.2.2).
Control of RPV injection is not acheived after 15 minutes (SAE 7.2.4).
2-13
OSSI 92-402A-7A-NMP 1 EAL Validation Procedure, Rev. 0 Attachment 2 VaHdation Exercise Scenario Checklist No.:
Plant: NMP-1 Simulator: 0 Table-Top:
Scenario 8 14 Scenario Description(s):
A severe storm causes a loss of all telephone systems offsite. No radios respond to attempts to call offsite (UE 7.3.2).
Meteorological tower stripchart indicate sustained wind speeds of 130 mph (Alert 8.4.6).
The roof is ripped off of the Off Gas Building (Alert 8.4.7).
2-14
OSSI 92-402A-7A-NMP 1 EAL Validation Procedure, Rev. 0 Attachment 2- VaHdation Exercise Scenario Checklist No.:
Plant: NMP-1 Simulator: Q Table-Top: 8 Scenario ¹ 15 Scenario Description(s):
A bomb threat is recieved. A search reveals a bomb in a Control Room back panel cabinet (UE 8,1,1).
An unauthorized individual is recognized to have scaled the Protected Area fence (Alert- 8.1.2).
The individual is tracked into the Control Building (SAE 8.1.3).
The bomb explodes destroying the electrical panel and the panels behind it (GE 8.1.4 or Alert 8.2.2).
Instead of the Control Room, a bomb explodes in the Administative Building (UE 8.3.2).
2-15
OSSI 92-402A-7A-NMP 1 EAL Valfdatton Procedure, Rev. 0 Attachment 2 Validation Exercise Scenario Checklist No.:
Plant: NMP-1 Simulator: 0 Table-Top: ~
Scenario ¹ 16 Scenario Description(s):
A security truck rams the diesel fuel storage tank (UE 8.3.1).
The collision tears a hole in the oil tank (Alert 8.3.4).
The spilled oil catches fire and burns out of control for 30 minutes (UE 8.2.1).
2-16
OSSI 92-402A-7A-NMP 1 EAL Validation Procedure, Rev. 0 Attachment 2 Validation Exercise Scenario Checklist No.:
Plant: NMP- 1 Simulator: Q Table-Top: 0 Scenario 8 17 Scenario Description(s):
The control room operators notice ground motion and that the seismic activity alarm is recieved. JAFNPP calls and confirms the earthquake (UE 8.4. 1).
JAFNPP later calls and says the earthquake was of magnitude O. ll g (Alert 8.4.5).
As a result of the earthquake the Screen and Pump House is destroyed (Alert 8.4.7).
2-17
OSSI 92-402A-7A-NMP 1 EAL VaIIdatfon Procedure, Rev. 0 Attachment 2 Validation Exercise Scenario Checklist No.:
Plant NMP-1 Simulator: Cl Table-Top; 8 Scenario ¹ 18 Scenario Description(s):
A report to the control room states that a tornado has been sighted inside the security fence (UE 8.4.2).
An operator reports that he cannot get to the Screen and Pump House because of wind and debris (UE 8.4.3).
2-18
OSSI 92-402A-7A-NMP 1 EAL Valtdatton Procedure, Rev. 0 Attachment 2 Validation Exercise Scenario Checklist No.:
Plant: NMP-1 Simulator: Cl Table-Top: E Scenario ¹ 19 C
Scenario Description(s):
Lake flooding has resulted in measured lake levels of 248 ft. (UE 8.4.4).
Frizzle ice formation has caused the forebay water level to drop to [Later] ft.
(Alert 8.4.8) ~
2-19
OSSI 93-402A-10-NMP1 NMP-1 EAL Verification & Validation Report, Rev. 0 Attachment 4 EAL Validation S Sheets 4-1
Itl (Ir
~
I OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 1 Vacation Samaxazy Sheet N~P ~ r ft 6 3 Validation Team Members: RA-QCc i
~PA rL II (-ulCS ez iC.
Checkhst No.:
IL =
EAL Rev. No.:
I
& Checldtst No.:
!L
~ EALRev. No.: ~
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Q c. g,y-L Q lH Q C9 Valid. performed and comments recorded: Vshda. performed snd comments recorded: '
Ce~ )(e2 khan Validation earn Leader Date Validation T am Leader Date Checklist No.:
la!JI V EAL Rev. No.: ~ Checklist No.: ~ EALRev. No.: 8 ., I 55". j Qpc lb Q Qp c. Q El I. 8, Q p C 3%%A Q G3
( ~ l W Q ( Q Q 13 g Q Q Q Valid. performed and comments recorded: Valida. performed and comments recorded:
tof~gaZ uJ r o/ofvg Validatlo Team Leader Date Validation Te Leader Date I
Continuation Sheets Attached: I I
)
I
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 1 VaHdation Suznmazy Sheet ol~ Ap riant: N~ t Validation Team Members: LL Ft 45% /154+
j Checklist No.: ~L EAL Rev. No.: ~ Checklist No.: EAL Rev. No.:
I il VL = I ~L~g, Jim I ~ I ~
Q Q Q lZI
,.2 Q Q QL.
.S.l Q 5 Zr5 Q g t
~ ~ 4 Q 23 Q Q Q Valid, performed and comments recorded: Valida. perf rmed and comments recorded. 'i
~
io(a(eS hl Jo>
Validation earn Leader Date Ualidatio Team Leader Date l Checklist No.: Q~ EAL Rev. No.: b Checklist No.: ~>a fI EAL Rev. No.:
i<5 II I <<).
Cl el Q Q. Q s i Lf C Q ti'
<<r
/, jZl Q Q 8 Q ~ IC Valid. perfo ed and comments recorded: Valida. pe ormed and comments recorded:
IE)4 t'03 /lg 'ops c'alidati cL Team Leader Date Validatio Team Leader Date I, g
Continuation Sheets Attach'd: ~
1-2
OSSI 92-402A-7 EAL Validation Procedure, Rev, 0 Attachment 1 VaHdation Sunumxzy Sheet 0 C P>attt. Af 4 /
Validation Team Members: ~)m F
~
I I
Checkhst No.: ~~<45 EAI Rev. No.:
Jl 8' e Checklist No.:
O'
~ EAI Rev.,No.:
I
'I "j)l j I'
-.r~
0 .91 0
. ll ~
4 2 Q 0
'I Q ', I 0 Zl ~
- 2. 0 III
~ r7 0 g /(o(ogden Q 0
,/e/ Q g 0 0
~ l Valid, performed and comments recorded: Valida. pe rmed and comments recorded: ~
d .g.(< 1 Validation earn Leader Date Validation earn Leader Date Checklist No.: ~~ EAL Rev. No.: Checklist No.: I~~4~ EAL Rev. No.:
EaLH Bl ex F~a. Rm. ZZ
~ ~ 2- 0 ~ 7 0 Zl 0 ~
IQ Q
~ v 0 ~
9 0
.Z.K Q Cl 8 3'L '/. ( 0 Valid. perfoxmed and comments recorded: Valida. perfo ed and comments recorded:
C--
Validatio Team Leader Date Validation earn Leader Date Continuation Sheets Attached:
1-3
OSSI 93-402A-10-NMP1 NMP-1 EAL Veriilcatfon & Validation Report, Rev. 0 Attachment 5 EAL Validation Exercise Checklists 5-1
OSSI 92-402A-7 EAL VaBdation Procedure. Rev. 0 Attachment 3 VaHdation Exercise ChecRHst Date 10 6 93 Checklist No.: 1 Yes No ~NA
- 1. When the need for classification was initially recognized, were the EALs easily accessible to the user? 0 0 Cl Comments: None.
- 2. Where plant conditions required emergency classification, did the format and layout of the EALs support easy and rapid classification recognition' ~ o u Comments: None.
- 3. Was classification of any conditions not requiring emergency classification avoided' ~ o o Comments: None.
- 4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts' 0 CI Cl Comments: None.
3-1
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecRHst Date: 10 6 93 Checklist No.: 1 Yes No ~NA
- 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 ~
Comments: None.
- 7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriateV ~ 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 Valldatlon Procedure, Rev. 0 Attachment 3 Validation Exercise ChechHst Date; 10 6 93 Checklist No.: 1 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 Q Comments: None.
- 12. Did the EALs have adequate detail to be effectively evaluated'P ~ Q Q Comments: None.
- 13. Are the EALs devoid of excessive detail'? r 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 VaHdation Exercise ChecRHst Date: 10 6 93 Checklist No.: 2 Yes No ~NA
- 1. When the need for classification was initially recognized, were the EALs easily accessible to the user' a a a Comments: None.
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 0 Comments: None.
4, Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts'? ~ Cl 0 Comments: None.
3-1
OSSI 92-402A-7 EAI. Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise Chechlist Date: 10 6 93 Checklist No.: 2 Yes No ~NA
- 5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized'? 5 Cl Cl Comments: None.
- 6. Did the EALs and required Emergency Plan procedures interface properly? Cl Cl ~
Comments: None.
- 7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriate'? 5 Cl Cl Comments: None.
- 8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate'? a Cl Cl 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 Cl J
Comments: None.
3-2
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise ChecMist Date: 10 6 93 Checklist No.: 2 Yes No ~NA
- 10. Are the EALs devoid of any misleading or incorrect information'? ~ Cl Cl Comments: None.
- 11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.
necessary to effectively evaluate the EALs? ~ Cl Cl Comments: None.
- 12. Did the EALs have adequate detail to be effectively evaluated'? Cl ~ Cl Comments: Need to a d reference to OP-50A Area Radiation Monitorin S stem Revision 7 Attachments 2 and 3 in EAL 1.4.1 for alarm set oint values.
- 13. Are the EALs devoid of excessive detail'? ~ Cl Cl Comments; None.
- 14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure'? ~ Cl Cl Comments: None.
3-3
OSSI 92-402A-7 EAI. Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 6 93 Checklist Nc.: 2 Yes No ~NA
- 15. Additional Comments: None.
3-4
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise Checldist Date: 10 6 93 Checklist No.: 3 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' S Q Q Comments: None.
- 3. Was classification of any conditions not requiring emergency classification avoided' Q r Q Comments: EAL 6.1.5 states "RPV water level cannot be restored and maintained <TAF". This is condition for d ell fioodin er RL-25. Is this consistent with other EALs for GE usin TAF.
- 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 6 93 Checklist No.: 3 Yes No ~NA
- 5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized? Q ~ Q Comments: Consider swa in 6.1.2 and 6.1.3 so EALs for same modes are on same line.
- 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: ~Non .
3-2
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exerci.se CheckHst Date 10 6 93 Checklist 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? ~ Cl 0 Comments: None.
- 10. Are the EALs devoid of any misleading or incorrect information? ~ o u Comments: None.
- 11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.
necessary to effectively evaluate the EALs'? r o o Comments: None.
- 12. Did the EALs have adequate detail to be effectively evaluated? a o o Comments: None.
- 13. Are the EALs devoid of excessive detail'? ~ 0 Q Comments: None.
3-3
OSSI 92-402A-7 EAL Valldatlon Procedure, Rev. 0 Attachment 3 Validation Exercise ChechHst Date: 10 6 93 Checldist No.: 3 Yes No ~NA
- 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 CheckHst Date: 10 6 93 Checklist No.: 4 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 recognitionV ~ Q Q Comments: None.
- 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 Q Comments: None.
3-1
OSSI 92-402A-7 EAL Valtdatton Procedure, Rev. 0 Attachment 3 Validation Exercise ChecMist Date: 10 6 93 Checldist No.: 4 Yes No ~NA
- 5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized' ~ 0 Q 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 appropriateV Comments: None.
- 8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate'? ~ 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'? ~ CI 0 Comments: None.
3-2
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdatlon Exercise Checklist Date: 10 6 93 Checklist No.: 4 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? R Q Q Comments: None.
- 12. Did the EALs have adequate detail to be effectively evaluated? Q ~ Q Comments: EAL 3.4.1 and others that refer to su ression chamber ressure and su ression ool water level should be chan ed to torus ressure and torus lev l.
- 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.
3-3
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise Checklist Date: 10 6 93 Checklist No.: 4 Yes No ~NA
- 15. Additional Comments: None.
3-4
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise ChechHst Date 10 6 93 Checklist No.: ~
Yes No ~NA
- l. When the need for classification was initially recognized, were the EALs 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 classification recognition? ~ Q Q Comments: None.
- 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'? S Q Q Comments: None.
3-1
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise ChechHst Date: 10 6 93 Checklist No.: 6 Yes No ~NA
- 5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized? ~ Q Q Comments: ~Non .
- 6. Did the EALs and required Emergency Plan procedures interface properly' Q Q ~
Comments: None.
- 7. After initial classification, did subsequent classiQcation 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' 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'?
Comments: None.
3-2
OSSI 92-402A-7 EAL Valfdatfon Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecRHst Date: 10 6 93- Checklist No.: 6 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 Q Comments: None.
- 12. Did the EALs have adequate detail to be effectively evaluated? 0 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 VaHdation Exercise Chechlist Date: 10 6 93 Checklist No.: 7 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 EALs support easy and rapid classification recognition' r Q Q Comments: None.
- 3. Was classiQcation 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' E Q Q Comments: None.
3-1
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 6 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 S Comments: ~Non .
- 7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriate? ~ Q Q Comments: None.
- 8. Did the F~ 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? S Q Q Comments: None.
3-2
OSSI 92-402A-7 EAL Valfdatlon Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecRHst Date: 10 6 93 Checklist No.: 7 Yes No ~NA
- 10. Are the EALs devoid of any misleading or incorrect information' o r o Comments: Does a lar e break loca which is handled er desi n and does not threaten rima containment a General Emer enc even if rad in d ell i e uivalent to 20% fuel
~dama e'P
- 11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.
necessary to effectively evaluate the EALs'? r o o Comments: None.
- 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 o Comments: None.
3-3
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise ChechHst 10 6 93 Checldist No.: 7 Yes No ~NA
- 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 Validatfon Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 6 93 Checklist No.: 8 Yes ~N ~NA
- 1. When the need for classification was initially recognized, were the EALs easily accessible to the user' ~ Q Q Comments: None.
Where plant conditions required emergency classification, did the format and layout of the EALs support easy and rapid classification recognitionV ~ Q Q Comments: None.
- 3. Was classiAcation of any conditions not requiring emergency classification avoided' Comments: None.
- 4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts' S Q Q Comments: None.
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Datee 10 6 93 Checklist No.: 8 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'? ~ Q Q Comments: None.
- 8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate? S 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? S Q Q Comments: None.
3-2
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 6 93 Checklist No.: ~
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 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 detailV 5 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 Checldist Date: 10 6 93 Checklist No.: 9 Yes No ~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 classification,. did the format and layout of the EALs support easy and rapid classification recognition'? 5 0 0 Comments: None.
- 3. Was classiQcation of any conditions not requiring emergency classification avoided?
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 Checklist Date: 10 6 93 Checklist No.: ~
Yes ~N ~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 5 Comments: None.
- 7. After initial classification, did subsequent classiQcation 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'? 5 Q Q Comments: None.
3-2
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise Chechlist Date: 10 6 93 Checklist No.: 9 Yes No ~NA
- 10. Are the EALs devoid of any misleading or incorrect information'? 8 Q Q Comments: None.
ll. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.
necessary to effectively evaluate the EALs'? S 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.
3-3
OSSI 92-402A-7 EAL Valldatlon Procedure, Rev. 0 Attachment 3 VaHdation Exercise Chechlist Date: 10 6 93 Checklist No.: ~
Yes No ~NA
- 15. Additional Comments: What do Central and Seconda Alarm Stations? Are the needed for Alert declaration' If bundle is believed to be uncovered but was not s ecificall seen uncovered is declaration re uired'?
Should use "sustained" uncove to av id urious uncovernr, What about NI re lacement in which a detector is uncovered' 3-4
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise Checklist Date: 10 6 93 Checklist No.: 10 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 EALs support easy and rapid classification recognitionV ~ Q Q Comments: None.
- 3. Was classification 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' ~ Q Q Comments: None.
- 5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized' ~ Q Q 3-1
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecMist Date: 10 6 93 Checklist No.: 10 Yes No ~NA 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'? ~ 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.
- 10. Are the EALs devoid of any misleading or incorrect information? ~ Q Q 3-2
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise Checklist Date: 10 6 93 Checklist No.: 10 Yes No ~NA 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' a 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: It seems that an alarm would have been received before 2x tech s ecs is reco nized. It's unlikel th t Chemis woul identi this level without Arst havin been aware b~th 3-3
OSSI 92-402A-7 EAL ValidaUon Procedure, Rev. 0 Attachment 3 ValMation Exercise Checldist Date: 10 6 93 Checklist No.: 11 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: None.
- 3. Was classiAcation of any conditions not requiring emergency classification avoided V ~ Q Q Comments: None.
- 4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts' 5 Q Q Comments: None.
- 5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized' 3-1
OSSI 92-402A-7 EAL ValIdatfon Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 6 93 Checklist No.: 11 Yes No ~NA 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' 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.
- 10. Are the EALs devoid of any misleading or incorrect information? ~ Q Q 3-2
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 6 93 Checklist No.: 11 Yes No ~NA 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: None.
- 13. Are the EALs devoid of excessive detail' Comments: None.
- 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.
3-3
OSSI 92-402A-7 EAL Validation Procedure. Rev. 0 Attachment 3 VaHdation Exercise ChecRHst Date: 10 6 93 Checklist No.: 12 Yes No ~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 classification, did the format and layout of the EALs support easy and rapid classiQcation recognition' r o a Comments: None.
- 3. Was classiQcation of any conditions not requiring emergency classification avoided'? ~ CI CI Comments: None.
- 4. Did type and arrangement of EAL categories and subcategories adequately support emergency classification efforts' a o o Comments: None.
3-1
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise Checklist Date: 10 6 93 Checklist No.: 12 Yes No ~NA
- 5. Where plant conditions required emergency classiQcation, 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? ~ 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 EALsV ~ Q Q Comments: None.
3-2
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecRHst Date: 10 6 93 Checklist No.: 12 Yes No ~NA
- 10. Are the EALs devoid of any misleading or incorrect information? a 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: 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'? r 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 6 93 Checklist No.: 13 Yes No ~NA
- 1. When the need for classification was initially recognized, were the EALs easily accessible to the user'? r 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: None.
- 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? 8 Q Q Comments: None.
3-1
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise Checklist Date: 10 6 93 Checklist No.: 13 Yes No ~NA
- 5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized?
Comments: None.
- 6. Did the EALs and required Emergency Plan procedures interface properly'? a o r Comments: None.
- 7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriate'? ~ 0 0 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'?
Comments: None.
3-2
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise ChecMist Date: 10 6 93 Checklist No.: 13 Yes No ~NA
- 10. Are the EALs devoid of any misleading or incorrect informationV ~ Q Q Comments: None.
- 11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.
necessary to effectively evaluate the EALs'P ~ Q Q Comments: None.
- 12. Did the EALs have adequate detail to be effectively evaluated? ~ Q Q Comments: If halon or CO2 accidentl dischar ed into DG room would not want to declare but EAL 8.3.5 as written would re uire declaration. Ma be should consider ualifier for safe o eration or shutdown of the lant and access is re uired.
- 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 ChechHst Date: 10 6 93 Checklist No.: 13 Yes No ~NA
- 14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure' ~ o a Comments: None.
- 15. Additional Comments: None.
3-4
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise ChecMist Date: 10 6 93 Checklist No.: 14 Yes No ~NA
- l. When the need for classification was initially recogniied, 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' Comments: None.
- 3. Was classification of any conditions not 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 ChechHst Date: 10 6 93 Checklist No.: 14 Yes No ~NA
- 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 5 Comments: ~Non .
- 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.
3-1
OSSI 92-402A-7 EAL Valtdatlon Procedure, Rev. 0 Attachment 3 VaHdation Exercise Checklist Date: 10 6 93 Checklist No.: 14 Yes No ~NA
- 10. Are the EALs devoid of any misleading or incorrect information? ~ 0 0 Comments: None.
- 11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.
necessary to effectively evaluate the EALs? o ~ o Comments: Need to check ran e of wind s eed meter to see if it can read velocities s ecified in EAL.
- 12. Did the EALs have adequate detail to be effectively evaluated? 0 ~ 0 Comments: EAL 8.4.7 should include safe o eration and shutdown criteria. structures containin s stems and functions re uired for safe shutdown of the lant.
- 13. Are the EALs devoid of excessive detail? ~ o o Comments: None.
3-2
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise Checklist Date: 10 6 93 Checklist No.: 14 Yes No ~NA
- 14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedures ~ a o Comments: None.
- 15. Additional Comments: None.
3-3
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise Chechlist Date 10 93 Checklist No.: 15 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 classiQcation, did the format and layout of the EALs support easy and rapid classification recognition'? Q ~ Q Comments: EAL 8.1.4 should need to loose both CR and RSP befor GE is d clared. Chan e OR to AND.
- 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 classiQcation efforts' Q Q Comments: None.
3-1
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise Chechlist Date: 10 6 93 Checklist No.: 15 Yes No ~NA
- 5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized'? S Q Q Comments: None.
- 6. Did the EALs and required Emergency Plan procedures interface properly? Q Q 5 Comments: None.
- 7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriate'? S 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? a Q Q Comments: None.
3-2
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecMist Date: 10 6 93 Checklist No.: 15 Yes No ~NA
- 10. Are the EALs devoid of any misleading or incorrect information? 5 Q 0 Comments: None.
- 11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.
necessary to effectively evaluate the EALs'? r u o Comments: None.
- 12. Did the EALs have adequate detail to be effectively evaluated'? 0 ~ 0 Comments: EAL ..1 should clari that structures are those im ortant to safe lant o eration and shutdown.
- 13. Are the EALs devoid of excessive detail? ~ Cl 0 Comments: None.
- 14. Did the EAL identiQcation scheme adequately support location of the EAL condition within the classification procedure'? ~ 0 0 Comments: None.
3-3
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecRHst Date: 10 6 93 Checklist No.: 15 Yes No ~NA 15, Additional Comments: None.
3-4
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date: 10 6 93 Checklist No.: 16 Y~e 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 EALs support easy and rapid classiAcation recognition? r Q Q Comments: None.
- 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 Q Comments: None.
3-1
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise ChechHst Date: 10 6 93 Checklist No.: 16 Yes No ~NA
- 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 r 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? ~ Q 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? ~ Q o Comments: None.
3-2
OSSI 92-402A-7 EAL Valfdation Procedure, Rev. 0 Attachment 3 Validation Exercise ChecMist Date: 10 6 93 Checklist No.: 16 Yes No ~NA
- 10. Are the EALs devoid of any misleading or incorrect information' r o o Comments: None.
- 11. Did the EALs adequately specify controls,
,instrumentation, operator aides, procedures, etc, necessary to effectively evaluate the EALs'? r o o Comments: None.
- 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 o Comments: None.
- 14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure' r o o Comments: None.
- 15. Additional Comments: None.
3-3
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise ChechHst Date: 10 6 93 Checklist No.: 17 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 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 Valtdatton Procedure, Rev. 0 Attachment 3 VaHdation Exercise CheckHst Date: 10 6 93 Checklist No.: 17 Yes No ~NA
- 5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized' ~ 0 0 Comments: None.
- 6. Did the EALs and required Emergency Plan procedures interface properly' 0 0 5 Comments: ~Nne.
- 7. After initial classification, did subsequent classification escalation follow a logical progression in the EALs when appropriate' a 0 0 Comments: None.
- 8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate' r 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 EALsV ~ 0 0 Comments: None.
3-2
OSSI 92-402A-7 EAL Validation Procedure. Rev. 0 Attachment 3 VaHdation Exercise Checklist Date: 10 6 93 Checklist No.: 17 Yes No ~NA
- 10. Are the EALs devoid of any misleading or incorrect information'? ~ Cl 0 Comments: None.
- 11. Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc.
-necessary to effectively evaluate the EALs'?
Comments: ~Non .
- 12. Did the EALs have adequate detail to be effectively evaluated'? a r a Comments: EAL 8.4.5 does not s eci whose seismic instrumentation is used to sense it. Need to state s ecific unit.
EAL 3.1.1 exclude durin leak rate testin EAL 6.1.1 should sa backfed throu h Tl or T2.
- 13. Are the EALs devoid of excessive detailV ~ a o Comments: None.
3-3
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 Validation Exercise Checklist Date: 10 6 93 Checklist No.: 17 Yes No ~NA
- 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 Validation Exercise ChecMist Date: 10 6 93 Checklist No.: 18 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 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: 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 ChechHst Date: 10 6 93 Checklist No.: ~l Y~s 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' ~ Q Q Comments: None.
- 8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate? S 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'? S Q Q Comments: None.
3-2
OSSI 92-402A-7 EAL Valtdation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChecMist Date: 10 6 93 Checklist No.: 18 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 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?
Comments: None.
- 15. Additional Comments: None.
3-3
OSSI 92-402A-7 EAL Validation Procedure, Rev. 0 Attachment 3 VaHdation Exercise ChechHst Date; 10 6 93 Checklist No.: 19 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 EALs support easy and rapid classification recognition? ~ Q Q 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? ~ Q Q Comments: None.
3-1
4 OSSI 92-402A-7 EAL Validation Procedure. Rev. 0 Attachment 3 Validation Exercise ChecRHst Date: 10 6 93 Checklist No.: 19 Yes No ~NA
- 5. Where plant conditions required emergency classification, was the operating mode applicability of the EALs clearly recognized' ~ 0 0 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' ~ 0 0 Comments: None.
- 8. Did the EALs support escalation of emergency classification when plant conditions indicated that escalation was appropriate? ~ 0 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'? ~ 0 0 Comments: None.
3-2
OSSI 92-402A-7 EAI. Validatfon Procedure, Rev. 0 Attachment 3 Validation Exercise ChecRHst Date: 10 6 93 Checklist No.: 19 Yes No ~NA
- 10. Are the EALs devoid of any misleading or incorrect information'? ~ 0 0 Comments: None.
11 ~ Did the EALs adequately specify controls, instrumentation, operator aides, procedures, etc, necessary to effectively evaluate the EALsV ~ u o Comments: None.
- 12. Did the EALs have adequate detail to be effectively evaluated' o ~ a Comments: Consider use of "Confirmed" fire in EAL 8.2.1.
- 13. Are the EALs devoid of excessive detailV a o a Comments: None.
- 14. Did the EAL identification scheme adequately support location of the EAL condition within the classification procedure'? 5 0 0 Comments: None.
- 15. Additional Comments: None.
3-3
OSSI 93-402A-10-NMP1 NMP-1 EAL Verification & Validation Report, Rev. 0 Attachment 6 EAL Validation Comment Database 6-1
~ ~ ~ ~ e a ~ ~ a ~ ~ ~ ~ ~ ~
Date 9/20/93 M. C. DauS Record No. 4 NUCIeal'ame Ex t. 408-274-9587 Department o rlglnatlng site Q JAF Q IP-3 Q NMP-2 p NYPA Q OSSI Last Modlfled: 1 1/9/93 CIIP-2 g NMP-1 PGinna C3NMPC Comment Scenario 3, question 5: Consider swapping 6.1.2 and 6.1.3 so EALs for same modes are on same line.
E'aL¹:~N/A site Applicability p JAF p:IP-::3 - -'.p NMP-2 'Q Generic'BWR".'II:G'en'eral DIP-2 QNMP-"1 P Ginna QGeneiic,PWR NUMARC Classification:
Recognltlon-- N/A Emer. Class.
- I C¹
- EAL¹ *
'Cat.
"',Q NUMA'RC-.".007.;.',Q Piocedure.::- .-Q Verification=:.'-:OTraining;'-0 Hardware.
".;.0 EAL: .", ". -
":C3Te'chnIcal':Bases,-C3 Validation '-Q Deviation:.Q N'orie
'5 A Resolution The benefit of this suggestion is arbitrary and will not be incorporated.
Continuation Sheet Attached
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Date 9/20/93 Name M. C. Daus Record No. 5 Department NuClear Ext. 408-274-9587 orlglnatlng site 0 JAF 0 IP-3 D NMP-2 Q NYPA p OSSI Last Modlfled: 11/9/93 C]IP-2 HNMP-1 OGinna QNMPC Comment Draft: A Scenario 3, question 3: EAL 6.1.5 states "RPV water level cannot be restored and maintained
<TAF". This is condition for drywell flooding per RL-25. Is this consistent with other EALs for GE using TAF.
EALi¹!~NrA
"- -":"st'te'ppiicaiirisy 'Cl JAP-':l3 iP'3: "'CI NMP2"'mi Generic BWR:::::::.QGeneral.
C3¹IP,-2 Q NMP-1 P- GInna .Q Generic PWR'NUMARC-.rclasslflcation:
iReco'gnltion"
'Cat.
N/A 'Emer..class. ** Ic¹~*::,i .
'ta'ALn¹r'i'~
h~ ', ",' , *
- Q NUIVlARC;-,,007"
- C3:Procedure'::-:-,: .,;:.p~Verification.-".OTraini'ng .,--p Hardware' ".: .
.;C3 EAL-; . '=QTechnicalBases -:Q Validatio~ -
'.-';C3 Deviation Q Norie Resolution Yes.
Continuation Sheet Attached
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Da te 9/20/93 M. C. DauS
'Record No. 6 Department NuCleai'ame Ex t. 408-274-9587 o riginating site 0 JAF 0 IP-3 p NMP-2 QNYPA QOSSI Last Modified: 11/9/93 OIP-2 g NMP-1 OGinna C]NMPC Comment Scenario 4, question 12: EAL 3.4.1 and others that refer to suppression chamber pressure and suppression pool water level should be changed to torus pressure and torus level.
Met data from computer goes to 140 mph but back panel recorder only goes to 100 mph.
RAGEMS and OGESMS should be put in table for category 5. When NMP1 decides which to use they can cross them out.
EAL¹:~34 ] ,site Applicability Q JAF Q IP-;3 Q NMP-2 -Q Generic BWR .QGeneral DIP-2 HNMP-1 'Q Ginna QGenericPWR NUMARC Classification:
Recognition" N/A .,
'mer." 'Ciass.'* IC¹ ** 'EAL¹', **
C,at:-. -,
' p'-".: ',-'.,;Q.NUMA'RC'0¹07<,"Q'Procedure",-',-- ';.',.Cl Verification: -'HTraining;..:,:0 Hardwai:e..-'.'..':-'
"-""-:.'DEAL"'-::-';:,",::"''Q~Tec¹hnical Bases-,";0;Validation '0 Deviation ':C3'None-,'-
'esolution Changed to "torus" pressure and water level as suggested.
Agree that back panel recorder only goes to 100 mph. Use of computer to recognize wind speeds approaching 125 mph will be discussed in training (EAL 8.4.6).
RN 10A/B will be used in category 5 Table 5.1 until it is determined that RAGEMS has in fact replaced OGESMS.
Continuation Sheet Attached
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~ ~ ~ ~ ~ 0 Date 9/20/93 M. C. Daus Record No.'7 Department Nuoleal'ame Ex t. 408-274-9587 orlglnatlng site Q JAF Q IP-3 Q NMP-2 Q NYPA Q OSSI t.ast Modlfled: 11/9/93 Q IP-2 E NMP-1 Q Ginna Q NMPC Draft:
Comment A Scenario 9, question 15: What do Central and Secondary Alarm Stations? Are they needed for Alert declaration?
If bundle is believed to be uncovered but was not specifically seen uncovered, is declaration required?
Should use "sustained" uncovery to avoid spurious uncovery.
What about Nl replacement in which a detector is uncovered?
NUMARC'-,;Cia'sslfication':-
'Rscognlllon-": N/A Cat.
'::--'-"-..Q IP-'2;Q NMP-,:1;Q Ginna Q Geneiic PWR Emsr. Class.
-. - .-.'lc->:L IC¹~:EAL¹ EA'L¹!r~N/A::::: i*elfe I'ApplicabIill'lr,"'::;CIJAP .'CI IP-3',: 'DNMP-'2 CI GeneiiCBWR!'-.'. 8General'::',
..;Q s
"Q NUMARC-007 CI Procedure - ...;:Q Verification .:.::8 Trainirig:=.".',:.Q Hal'dware,:...
.Q.EAL, Te'chnical:Bases Q Validation 'grDeyiation',::C3-No'rie ',
Resolution CKW "Central and Secondary Alarm Stations" are CAS and SAS. Abbreviations appear to be better understood and have been replaced in EAL.
Declaration is required if bundle was not specifically seen uncovered. This will be covered in training.
NUMARC does not provide the latitude for a "sustained" uncovery in this EAL.
Nl replacement is not specifically addressed by this EAL but would lead to a declaration based on refuel floor monitor indications.
Continuation Sheet Attached
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Date 9/20/93 M. C. DauS
~Record No:- 8 Department Nucleal'ame Ex t. 408-274-9587 o riginating site 0 JAF 0 IP-3 0 NMP-2 Q NYPA Q OSSI Last Modified: 11/9/93 IHIP-2 HNMP-1 OGinna ONMPC Comment Draft: A Scenario 10, question 15: It seems that an alarm would have been received before 2x tech specs is recognized. It's unlikely that Chemistry would identify this level without first having been aware by the alarm.
'EAL¹.:~gp i"':::i':."ens'":appncabiot'y.'::'.c3'JAF-!D Ip-:3,: ':.".:cl'NMF.=2';:Q Gene'ii'c',BwR! ig'General 1
';: .,-':::,Cl:IP,-.:2 '.;[3 NMP--1-':Cl,Ginna ':Q:Generic'PWR NUMARC.;Clas'elf lcatlon Recognltlon ": N/A ': Emer'.: 'Class'. I Cff
- 'EAL-ff:.'*
Cat..
Q NUMA'RC-'007 ';Q:Procedure' '.
'Q Verificaetion ",QTrainirig,..'Q Hardware EAL .CI Technical- Bases '0 Validation,,C3 Deviation- .'.Cl None' ',
',0 Resolution Under most circumstances the comment is probably correct.
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Date 9/20/93 M. C. DauS
',Reco'rd':! No;",;,9 NUCIeal'ame 408-274-9587 Department Ext.
orlglnatlng site P JAF Q IP-3 Q NMP-2 Q NYPA P OSSI Last Modlfled: 11/9/93 C3IP-2 g NMP-1 C]Ginna QNMPC Comment Draft: A Scenario 13, question 12: If halon or CO2 accidently discharged into DG room, would not want to declare but EAL 8.3.5 as written would require declaration. Maybe should consider qualifier for safe operation or shutdown of the plant and access is required.
- .:::eA'iel~83 5 ':,:,",:,::,:.,::!::;siipi:::A'pp'ircaeIiiiy,,CI'JA'F,'"'Cl:IP,'-,3;::::.,:;:O'NMP.-'2:::;CIGeiierIc~BWR! I E:6 UMARC.',/Cia'sslflc'atlon:.'-'.-:"-"'-;:.",': -
".Recogpnltlo'n;::":,.,': N/A .**
";:.:.-:.;Eii'i'er".-"'.-:!Cla's's" ~r ' '-."~l C¹~*::", "'-"',E'A'L¹fN~*
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c:~.',.'...,.'esolution Added the phrase "... needed for safe plant operation".
Q Continuation Sheet Attached
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Date 9/20/93 M. C. Daus
'eco'rd':;:No';! 10 NUCleal'ame 408-274-9587 Department Ex t.
originating site Q JAF Q IP-3 Q NMP-2 Q NYPA QOSSI "ast Modified: 11/9/93 P IP-2 g NMP-1 Q Ginna P NMPC Draft: A Comment Scenario 14, question 11: Need to check range of wind speed meter to see if it can read velocities specified in EAL.
- ,jAi.'¹.:~84 6 );;:,";;sixie,,;:A'p'plI'c'abjjiIy':.;.',
"Rec'o'gnltlon':::,.:j
- ,'c, t'::.':;
N/A ~<',:.','.,'Eme'r'.: :--:,Cla'ss'~>
';I C¹~*: .'>IDEAL'¹'.'"-;~*
Resolution Agree that back panel recorder only goes to 100 mph. Use of computer to recognize wind speeds approaching 125 mph will be discussed in training (EAL 8.4.6).
Continuation Sheet Attached
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Da te 9/20/93 Name M. C. Daus Record 'No. 11 Department NUCleal Ext. 408-274-9587 originating site p JAF 0 IP-3 Q NMP-2 Q NYPA QOSSI Cl(P-2 ENMP-1 OGinna QNMPC comment Draft: A Scenario 14, question 12: EAL 8.4.7 should include safe operation and shutdown criteria for structures containing systems and functions required for safe shutdown of the plant.
EAr.¹:~34 7:,:ena npposanmry::I7 JAP 'CI.IP-3 CI NMP.-2:CI Generic BWR:ISGenerai g:IP-.2 nQ NMP,-1 Q".Ginna=.-Q,Generic PWR UMARC:.Classification:
r; s.:"
Reaognltian N/A lgmsr; Class. ** "
I c¹~* EAL¹:~*
Cat.' .
a
- . "C3.EA'L""-' "" '- -';0 Techiiicaal:Banses"",Q Validaation,:.:.::QrDevia¹tiori':::Q Noii'e "
Resolution Added the phrase "... needed for safe plant operation".
Continuation Sheet Attached
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Date 9/20/93 Name M. C. Daus
- ,'Record,"':No".,"t 2 Department Nuclear Ext. 408-274-9587 or)g)natIng s)te p JAF p IP-3 p NMP-2 p NYPA p OSSI "ast Modt<~ed: 11/9/93 P IP-2 g NMP-1 P Ginna P NMPC Comment Scenario 15, question 2: EAL 8.1.4 should need to loose both CR and RSP before GE is declared. Change OR to AND.
.,'Eiii,'¹,:~g~ 4,:;:;:::-::.,",'::,.'.js)t'e,".,..jp'piI'c'aiÃIIiy;",clJAF Re¹'c'ognlttori,.<",.
C N/A "-';;!Emer.':,Clas's. **; f
'I.C¹~!i'; .-'-;:EA'L'¹:~~* ¹ Resolution Changed OR to AND.
Continuation Sheet Attached
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Date 9/20/93 Name M. C. Daus Record. No'. 13 Department Nuoleal'x t. 408-274-9587 o rlglnatlng Site 0 JAF p IP-3 0 NMP-2 0 NYPA p OSSI Last Modlfle": 11/9/93 Q IP-2 g NMP-1 P Ginna P NMPC Comment Scenario 15, question 12: EAL 8.3.1 should clarify that structures are those important to safe plant operation and shutdown.
- EaL¹:~83 ]:.".c:,:::ella 'Appllcanllny:;"CIJAF:CI IP-3 .:,::!0 NMP-'.2 ';CI Generic BWR;:.':::::.:E'6'enerai: '
IP-2:P NMP--r1::::P ';Ginna "'P.Genceric::PWR, RUMARC, ¹Classlflcatlon:
'Rsccgnlllcn::I N/A Cat.
Emar.:Classr ** .: IC¹~:E'A'L¹:.:~
a
-..-":,-:D NUMARC-:007:; O'Procedure.:;:,.= .:r,.::P Verification -,QTiaining cC3 Hardware
. ";C3.EA¹L".-';:-".-':.-:,;.
':.C3;TechsnIcal'Ban'ses ",'.p Validation ":Q Deviation -:g Norie Resolution NUMARC states that this EAL must be declared when a crash and projectile impact "...
potentially damage plant structures containing functions and systems required for safe shutdown of the plant". The basis for this EAL adequately addresses the comment concern.
Continuation Sheet Attached
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Date 9/20/93 Name M. C. DauS
"'Record':. No.'4 Department NuClear Ex t. 408-274-9587 orlglnatlng site C] JAF p IP-3 p NMP-2 Q NYPA Q OSSI Last Modified: 11/9/93 HIP-2 glNMP-1 PGinna C]NMPC Comment Draft: A Scenario 17, question 12: EAL 8.4.5 does not specify whose seismic instrumentation is used to sense it. Need to state specific unit.
'E'AL8:~84 5 '::':;:,':;,Site:;::,;".AppllcablllIy':!CIU MUMARC::-::;Clas'slftcatlon .:. ,:
"Rec'o'gnltlon"'.,.'-:
N/A Cat~",",':-..: "'",;-".,.
'-. 'Emer. ',Cla'ss; . ":,I C¹~*, ." -.:.:.E'AL'¹~~*>',';;:,;'~,:,'",*,
Resolution Added plant specific name to the second condition of this EAL.
Continuation Sheet Attached
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Date 9/20/93 Name M. C. Daus
'Rer'co'rd""'No'"'- 1 5
::.-".'-" ." '; ".;,'"',"-",."'"..""epartment Ext. 408-274-9587 0 OSSI Nucleal'rlglnatlng Site 0 JAF 0 IP-3 0 NMP-2 CI NYPA Last Modified: 11/9/93 HIP-2 HNMP-1 PGinna QNMPC ttraft:~A Comment Scenario 17, question 12: EAL 3.1.1 exclude during leak rate testing.
,.',eitt'L'¹;~3t t '::::,'!!a!etta .'Appitac'atrtltteyc'CI JAP'lip:IP,,-'3tii':-,:Q'NMP,'-:2irCI Generic!'BWR:;.:i'EGenei'al,;:
'Recognltlon'.-,: N/A ' -
Emter."Cla'ssi" ** $;.,',-'--.':,i Ci ~*;-.":;.':.'SEA'Lg~ ~*
- .
- ;.',:.':;g EA'L: .""-, r:;", -,-';::;0:Tech'riical.'."Bases;.;-',D:,ValidatIoyii';: '::O,'D y
Resolution Added the phrase "... due to coolant leakage" to ensure this EAL is only declared when pressure increase is due to a loss of coolant.
Continuation Sheet Attached
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Da te 9/20/93 M. C. DauS Record No. 16 Department NuCleal'ame 408-274-9587 Ext.
orlglnatlng site p JAF p IP-3 p NMP-2 p NYPA p OSSl "ast Modlfled: 11/9/93 P IP-2 g NMP-1 P Ginna P NMPC Draft: A Comment Scenario 17, question 12: EAL 6.1.1 should say backfed through T1 or T2.
EAL¹:~6i ] eire Appncabmry 0 JAF i7 IP-3 Q NMP-'2 G Generic BWR mi General QIP-2 QNMP-1 Q Ginna ',.Q.GenericPWR UMARC Classlflcatlon:
Recognitlon ": N/A - -
Emer. Class. I C¹ ** EAL¹
- Cat; " %).
- C3 E'AL": .
"'-':. ':0TechriIcal ..Bases:: IC3'ValidatIon:;:Q DeyIartIony:::.C3'.Norie
esolution Changed to include "backfed through T-1 or T-2".
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Date 9/20/93 Name M. C. Daus Record':. No.'17 Nuoleal'xt.
Department 408-274-9587 o rlglnating site 0 JAF P IP-3 0 NMP-2 P NYPA P OSSI Last Modlfled: 11/9/93 P IP-2 talNMP-1 PGinna QNMPC Comment Scenario 19, question 12: Consider use of "Confirmed" fire in EAL 8.2.1.
"E'/rL'¹:~gp q i,:,i::::-:,',:::.:¹lie.';,'/rppircanrirry- CI'/JAF.::,:0:IP:..-,'3;::.;CJNMP-2iiCI'Generic'BWR':i'iaiGeneral::
,O'.IP-:2::,C3 NMP-"1, .;Q:GInna".-:Cl Generic PWR lUMARC'::::Classlficatlon: '-
Reco'gnlrlon'i
'C'at N/A::Emar. Class. *':,c:,:IC¹~* -
"EAL¹: ~*
",c '.
- .
- -; '-",Q NUMARC:'.007"-"Q Procedure: p'Vre'rifIcation':pTrainIngr Q Hardware'",':.'::;.'.",
.0:EAL . ':-.'-:=C3Technicalr Bases -,0 Validation', '0-D'eviation,-'.'0 None-.
Resolution Added "Confirmed" as suggested.
Continuation Sheet Attached
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