ML20246K306
| ML20246K306 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 08/04/1989 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20246K304 | List: |
| References | |
| SAO-89-002, SAO-89-002-R01, SAO-89-2, SAO-89-2-R1, NUDOCS 8909050431 | |
| Download: ML20246K306 (24) | |
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SAFETY ANALYSIS FOR OPERATION (SAO)
SAO NO.89-002, Rev. 1 INITIATION DATE: 8/4/89
SUBJECT:
Control Room Habitability
.This SAO is being prepared to analyze the operability of the Fort Calhoun Station with the existing control room charcoal adsorber filter system until a new, upgraded filter system is installed and to allow the existing system to be inoperaale during the final tie-in period of the new system.
REFERENCES:
a.
- 1. NRC to OPPD letter dated May 28, 1987
- 2. LIC-87-496 dated August 31, 1987
- 3. LIC-88-110 dated March 8, 1988 '
- 4. Calculation No, FC05092, "LOCA Control Room Habitability Analysis JC0 Revised Containment Leakage Contribution", Rev. A. '
!. EXISTING CONDITIONS Radiological dose calculations were performed for control room personnel
(- during a postulated loss of coolant accident. The calculations determined that the whole body gamma and beta skin doses were below the values specified in Standard Review Plar. 6.4. However, the thyroid dose due to radioiodine was determined to exceed the SRP 6.4 value of 30 REM.
An NRC Fort Calhoun control room survey was conducted on October 27 through .
October 30, 1987, which determined that excessive amounts of unfiltered air were being pulled into the control room envelope. A Justification for Continued Operation was prepared as calculations indicated that these amounts of unfiltered air would result in operators receiving more than the 30 REM thyroid dose limit in a thirty day period. The NRC survey observations are contained in Reference 1. OPPD's response to this survey letter and the District's commitment to changes are contained in Reference
- 2. Several questions resulted which were addressed in a letter to the NRC (Reference 3). Major items in this letter were that RM-065 would be activated by VIAS, the toxic gas monitoring system was adequate even with 1650 CFM of control room inleakage, and the interin analysis for exceeding the 30 REN thyroid dose limit was sent to the NRC.
e909050431 090004 PDR ADDCK 050002G5 P PNV L__-__--------_------_------
'~ With the modifications made during the 1988 outage, RM-065 has been moved and is now automatically activated by VIAS, and the unfiltered air inleakage has been greatly reduced due to relocation of the control room air conditioning units such that inleaka e is no longer a concern. These control room changes did not alter the c arcoal filter train (VA-63 and VA-64)capabilitiesbutdidgreatlyreducetheunfilteredairinleakage which correspondingly reduced the thyroid radiciodine dose. However, it still exceeds the 30 REM limit established by SRP 6.4. SAO 89-002 was presared following the 1988 outage to address operability of the Fort Calaoun Station under these conditions. This revision to that SA0 incor> orates justification for an extended outage of the existing filter unit >eyond the seven day limit allowed by the Technical Specification LCO. The extended outage will allow the completion of a modification so that requirements established by SRP 6.4 will be met.
Installation of the new filter units (modification Mt-FC-87-20) requires the existing filter unit to be taken out of service for a period of time while final tie-ins of the new filter units are made. This time period is estimated to be approximately 14-15 days. Every effort has been made to shorten this duration, i.e., prefabrication, around-the-clock work, detailed scheduling, maximizig mapower, etc., but there is a large amount of work scheduled to be completed during the final tie-in period.
If the existing filter unit were found to be inoperable, Technical Specification 2.12 dictates that the plant could operate safely for seven
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days prior to starting cold shutdown procedures. The Technical Specifications address an inadvertent situation and allow a reasonable corrective action period before shutting the plant down. This SAO is to justify _ intentionally rendering the existing filter unit inoperable while the plant is on line to replace this unit with a superior system. The construction activities necessary for the system switchover cannot be -
accomplished within the Technical Specification LCO seven day period.
Therefore, a 17 day, carefully planned, monitored, and executed LCO time limit is proposed. This is a one time only request for a waiver of compliance from the Technical Specification LCO seven day limit.
II. SAFETY ANALYSIS FOR OPERABILITY WITH EXISTING CONDITIONS There are seven accident scenarios that could expose the control room operators to excessive radioiodine. These were all analyzed and it was
' found that the loss of coolant accident generated the greatest radioiodine threat to the control room operators.
L The operator 30 day thyroid radioiodine dose analysis is based on teveral i
' key parameters. The only parameters addressed by this SA0 are those that are non-standard for the loss of coolant accident analysis.
Actual measured parameters were used rather than conservative bounding values in an attempt to show that the 30 REM Ifmit could be achieved.
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The containment leak rate used for the LOCA event is 0.05% by weight l per day. This is based on the past test data: January 1983 results
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of 0.052% and October 1985 results of 0.043%, both with a 95%
confidence. No Type A ILRT test was conducted during the 1988 outage and the next scheduled test is in 1991. Regulatory Guide 1.4 requires the maximum Tech. Spec. value (0.1% per day for Fort Calhoun) to be used for the first day and 1/2 that value for the remainder of the accident. Based on actual test data, the containment leak rate is assumed to be 0.05% per day for this SAO.
The ESF leakage rate determined by Combustion Engineering for Fort Calhoun was 1243 cc/hr. This design basis number is then doubled when used for control room habitability analysis. Surveillance tests have been recently conducted (January 3, 1989) to determine the actual leakage. No leakage was observed, see ST-RHRS-1 Section F.1 and F.2 test results and set point procedure change No. 25963. Based on this data, the ESF contribution to operator dose is assumed to be negligible for this SAO.
The control room filtered air flow rate through the two inch charcoal beds (nohumiditycontrol)wasassumedtobe1100CFMwith90%ofall iodine species being removed. This is conservative as the Tech. Spec.
flo.a value is 1000 1 100 CFMandtherecently(January 2,1989) measured values'were 907 to 930 CFM.
Unfiltered air inleakage is considered from two sources; the control room observation door which is not equipped with an air lock (10 CFM')
and the VA-63 filtered air makeup fan (50 CFM).
The VA-63 filtered makeup air fan is located immediately downstream of the -
VA-64 makeup air charcoal filter. Any inleakage introduced via the fan seals would be unfiltered. The fan and the short section of duct between it and the filter are not considered to be leak-tight, so an unfiltered inleakege value of 50 CFM has been assumed from this source. This is considered to be conservative due to the very small leakage openings which would be involved (actual inleakage from this source could not be measured).
No other sources of unfiltered air inleakage need to be considered because:
- 1. The control room has been demonstrated to be at greater than 1/8 inch water gauge positive pressure relative to all adjacent areas, with the ventilation system operating in the filtered air makeup mode. The system is automatically brought to this mode during a DBA.
- 2. The control room observation rooe door is the only control room envelope bocndary door not equipped with an air lock. Inleakage does not have to be assumed for the other two envelope boundary doors which l
are equipped with air locks.
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- 3. The control room air conditioning units are located inside the control room envelope. Their related ductwork is also located in the control room envelope. Thus, any air inleakage on these units or their associated ductwerk will only involve envelope air.
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I 4. The 12" diameter makeup air duct downstream of the VA-63 fan is welded duct which has been leak tested prior to installation. The normal unfiltered fresh air makeup duct (also welded) is isolated using a bubble-tight (zero leakage) butterfly damper when the system is-operating in the filtered makeup mode.
i The containment air recirculation / filtration system flow rate is assun.ed to be 100,000 CFM, a Tech. Spec. valhe, with a removal efficienc i
elemental and particulate iodine and 30% for methyl iodides. y The of 90%
design for basis assumptions made by C-E established methyl iodide removal no lower than 50% of the total released in containment. This is considered extremely conservative as it was initially thought the charcoal beds would become saturated, which is unlikely. The 30% removal rate for methyl iodides is more in line with industry standards, given the relatively low concentrations of radioiodine expected to be in the containment atmosphere.
Based on the above analytical considerations, it was determined that the 30 day operator dose would be 103 REM which exceeds the SRP 6.4 acceptable level of 30 REM.
If the operators would take protective action by donning respiratory (O) protection gear, i.e., SCBA's, for the initial two hours of the accident, the 30 day operator radiofodine thyroid dose would be 29.6 REM (based on a protectionfactorof10,000forSCBA's)whichiswithintheestablished30 REM limit. If the operators would don the SCBA's for the first four hours of the accident, the dose would be reduced to 7.5 REM. Plant equipment and -
procedures are already in place for this operator action to ifmit radiciodine intake.
- 1. Warning & Procedures The calculations done to date show that the bulk of the iodine dose to an unprotected operator occurs in the first hours after radiation releases from containment. If the operators are aware of excessive radiciodine levels in the control room, (via alarm of RM-065), they will be able to implement respiratory protection measures for a sufficient time so that operator thyroid doses will be kept below the 30 REM SRP 6.4 value.
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U. A_A Control Room Iodine Monitor RM-065 is automatically started upon receipt of a Ventilation Isolation Actuation Signal:(VIAS) during a-08A. (VIASisthesamesignalthatautomaticallyali T -room ventilation system to the filtered makesp mode.)gns It can,the control-
?% -r therefore, be assumed that RM-065 will be in service before iodine !
> levels become excessive in the control' room. ~If RM-065 alarmsnin one hourorless(indicating 25%of1MPCforradiciodine), procedure 1 01-PAP-11: instructs the operators to don respiratory >rotection (i.e., .
SCBA's). 01-PAP-11 also states that DI-PAP-9.should >e. implemented to determine' actual radiciodine concentration. 'OI-PAP-9 contains its own;
.prctective action guidelines based on radiciodine concentrations. The automatic actuation of RM-065 and existing instructions in 01-PAP-11 and 01-pas-g provide assurance that excessive radiciodine levels will be detected-and respiratory protective measures implemented,by the operators. '
1" If a~LOCA were to occur.during a standard seven day LC0 period, A VIAS signal would automatically place the control room iodine monitor u (RM-065) into service. -Protective features and procedures identical 3
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~to those described in the preceding paragraph are placed into effect.-
These provide assurance that excessive radiciodine levels will be ,
detected and respiratory protective measures implemented by the-
. operators.
If a waiver of com>11ance is granted from this seven day LCO time: ..
O limit-there would se adequate compensatory actions in place to. protect the control room operators from radiciodine following a LOCA. The conditions or scenario'do not change whether the control room charcoal -
filter is disabled for seven days'or 17 days.
Obviously, the extended period exposes the plant to 10 more days of- -
power operation, durinft which an accident could~ occur. However, control room habitabil'ty would be closely. scrutinized during this.
time. 'In addition, the probability of the loss of control room filter capability is reduced due to the installation of redundant 99%
efficient units in place of the existing single ~90% efficient unit.
If the final tie- Ns of the w o filter units were not made until the 1990 refueling outage, the.sta: ion would operate for approximately Lfour months w<th'the existing single filter train.
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, 2. Plant Equipment ThedosecalculationsfartheoriginalSAO89-002(Rev.0) determined that SCBA's need to be I,rovided for control room personnel for T ..
approximately two to three hours after the radiciodine reaches the control room with the existing filter unit operating. There is a sufficient number of SCBA's readily available to equip-seven aperatt,rs (there are not assumed to be more than seven operatcrs in the control ;
room during the 6ccident . There is sufficient refill air on the turbine deck for five ope)rators for six hours (30 operator hours which corresponds to 4.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of air for seven operators. There are also spare SC8A bottles in other locations around the plant that could be made available to the operators. It is, therefore, concluded that there is more than sufficient bottled air present in the plant to keep control room operators in SC8A's for the time required by the dose calculations for.the original SAO 89-002.
The dose calculations (Ref. 4) performed for the original SAO 89-002 (Rev. 0) were also used for this revision for source term release information. The radiciodine release rates were extracted from calculation FC05092 (Ref. 4) to determine the time when releases decreased to nog igible levels. This time was determined to be approximately ei ht hours following laitiation of a LOCA. The release rates decline na nly due to the containment recirculation / filtration system operating as previously explained in this SA0.
Since the control room will be operating in the 100% internal recirculation mode during the period'when the existing filter is disabled no credit is taken for or control room air filtration. positive pressure, envalope integrity However, the extremely high protectionfactorofferedbySCBA's(10,000)willpreventthe .
operators from receiving inhalation doses in excess of the SRP 6.4 30 REM limit.
There are currently 30 operatoe hours of air available on the turbine dock as stated above. With the filter unit disabled, a total of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of air for 7 operators requires 56 operator hours of air to be available. Therefore, thirty additional operator hours of air will be added to the turbine dock storage area during the time the extended LC0 is entered.
In conclusion, the Fort Calhoun Station operators, through compensatory measures, can be protected from excessive radiciodine that would be released during a LOCA, without a more efficient control room charcoal filter or no filter at all. Therefore, these conditions will provide for continued safe operation of Fort Calhoun Station durir.g the extended LCO time period.
III.SA0 DURATION AND SPECIAL CONDITIONS This SA0 applies to conditions which could occur during a design basis loss of coolant accident. Although this SA0 could apply to other accidents
( involving release of radionuclides, those releases are bounded by the DBA LOCA which is the worst-case accident. The modification package will address providing 30 additional operator hours of air to be available while replacing the filters.
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~ RJG-17-1989 10:40 FROM O.P.P.D.-510 I0 FCRT C4.-CLt4 22
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This SA0 remains in effect until:
sAtosmodificationsareinstalled.(1)Aemainingcontrolro od The extend 4d outaos 17 da Post-and 3) Control rooe operator dose modificationtestingiscompleted;leted(whichreflectthemodifiedsystem.
calculations and analyses are comp
-System modifications will include installation of redundant filter trains which will have a higher filtration efficiency than the single filter train presently in place. The target date for installation completion, including all testing and analyses, is November 30, 1988.
Preoared Ey/Date:
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/ Otte Concurred with/Date: < -
supervisor Date Concurred With/Date: _
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05)erteent Manager / pate ConcurredWith/Date: 0--- ls.(*A m 1 __
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junager - a w oate Concurred With/Date: T
- 17 Manager - Safety Review group Date ,
, Approvedty/ Dates 6 18MV ._
/8 k Manager - Fort Calhoun Station Date CLOSURE APPROVAL 1
Manager - wani Manager - R alA PRC Chatruan
- PRC APPROVED c: Division Mana - Nuclear Operations Supervisor - ntenance AU6 I71900 Supervisor - Operations PRC MTG. MINUTES
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- l. .R8 05-24-89 C3 Preliminary Applicability Screening
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7.1 Purpose /De'scription' Reference Section 7.1 WL h% eau* bbe'ie is heino ho th moi +n
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7.2 Procedure Applicability Reference Sections Checks 7.1/7-.2 Check Acolicability Cateoory References
[] Procedure Change Standing Orders G-30/G-46
- b4 Modification mppc-814e Standing Order G-21/G-46
[] Incident Report Standing Order R-4
[] Technical Specification Violation Standing Order R-3
[] Maintenance Standing Order M-101 l.
[] Temporary Modification Standing Order 0-25 Control
[] Test / Experiment Standing Orders G-19,0-30/G-46
[] New Procedure Standing Orders G-3, 0-16, 0-17 0-18, 0-18A, M-26/G-46
[] Change to Technical
. Specifications NOD Policy / Procedure G-12 h Other: .5Ao c>oz_ R.i A-1 FC/ FORM /04 4
'. FC-154
., R8 05-24-89 ATTACHMENT A Preliminary Applicability Screening Page1ofI(
-~ Identify Structures, Systems and/or ._
7.3 ' Components which may be affected Reference Section 7.3 Those directly affected: CMko\ leem WO AC ,. (14 -6 A , UA-M/ ,
Those indirectly affected: Asy Au ,'//,'o j (ne/sa / Raam i,, 7.4' Summary of Procedural Requirements Reference Sections 7.1/7.2/7.3 ('
Is further evaluation of the activit equired?
YES
[K h& cs kau a4m Ad 4_v ce e d /b hC__
NO [] Because: A ha hm #Al doses 6 O can h,4 19 Lee M .
Th ak s Ka Han mu en b e_ h eon lm 4e d 4o de fauvro% t.
VRnen untedieWeA sahAm Auewim Qv Mt.
J AN AFFIRMATIVE (YES) ANSWER TO THE ABOVE QUESTION REQUIRES FURTHER EVALUATION TO DETERMINE IF THE ACTIVITY CONSTITUTES A 10 CFR 50.59 CHANGE, TEST AND/OR EXPERIMENT OR IS DEEMED TO INVOLVE AN UNREVIEWED SAFETY QUESTION AS SPECIFIED BY TECHNICAL SPECIFICATION 5.5.1.7.b.
AN UNEQUIVOCAL NEGATIVE (NO) ANSWER TO THE ABOVE QUESTION INDICATES THAT FURTHER EVALUATION OF THE ACTIVITY IS NOT REQUIRED.
Prepared by k ./Y . h,i M Date LIP dlB 7 Reviewed S Date 7/2o[M A-2 W FC/ FORM /04
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FC-154 R8 05-24-89.
ATTACHMENT A Page 1 of h Preliminary Applicability Screening Identify Applicable Sections Reference Sections
-7.5.1 of the USAR 7.3/7.5.1 Section Paraoraoh Pace /Fioure lN. 2.3 l4 2 ~b. I> 2.,*A M, f, k
_ / 4.2'b -I FN ~lO 9.10 9.10.2..H e 9. /o.4.4 */,10- 7, 9.10- G
/[.Z- //. 2. 3.17 /l . L - 2.~5 . 2 <s'
}4./f }4.1S~.~1.2- /4'o l5~ "L4 7.5.2 !aentify Applicable Sections Reference Sections of the Technical Specifications 7.3/7.5.2 Section Paracraoh Pana/Finure
- 2. I 2. GhkA h 2.- 6 al e
O 7.6 Summary & Preliminary Reference Sections
, Safety Evaluation Sc:wning 7.1/7.2/7.3/7.4/7.5 Does the proposed change / resultant event require / involve changes in the facility as descri>ed in the USAR7 YES D4 N0 f Because: TAs k>ttI be fenla.c]ed e % %s noennb1 u favk hmwr41 ch eeA%%t
+ Ms. k eALxMtm ste m Ct\W Does the sub4Lm utu be he_We A uta h t tw Me_ M Ve thA . '
)roposed change / resultant event regrte/ involve chsnges in the procedures as descri>ed in the USAR7 YES N0 Ahma[ ] un.bQ Because: Edsahat 'musaduth rua <dhnAo 4e e nh h i swr mmhanda euuhrw.
Does the proposed change / resultant event require / involve changes in the conduct of tests or experiments no.1 described in the USAR7 te. h W[ e] l Srs[t &Because: Lk 4*Ms nALbelbel A Mm bo-YES NO K1 i b 40s Ao neuw CMh oW ( lent om t 4-o7 4k =D wM FAsh cd A N N M L Could the activityiffect nuclear safety 1A USAR7 in a h Nsilha4tCwuodh way not previously'eva'uatedCNR-R-in the B1 to) '
YES Mnm[$]e u[ee]Bepause:NO Y thAR andA GERLmot % cov\b(
E Woi h mhotn h h , NW hAc. * . t M AQL O Mtn lA% WG*#A'ilr e_ h. h kh h
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FC-154 R8 05-24-89 ATTACHMENT A Page 9 of h Preliminary Applicability Screening Sunnary & Preliminary Reference Sections 7.6(con't)- Safety Evaluation Screening 7.1/7.2/7.3/7.4/7.5 Does the activit Definition 4.7) y being evaluated involve the preparation of a new procedure (See or a procedure change which affects nuclear safety?
YES[] NO[y] Because
/lo Am pcaL m ru,08 Mneckum ban M.A.
ch" e t 6.utt*adel% '
Does the activity being evaluated involve propsed tests and exper%ts that affect nuclear safety?
YES Bge:eMrs com4csdrs uwc & & +e Q sAholh mis, Ao M cAn h e l eu c uw ot u noe CMaLA. W .soGehaawhhcs % k h tvasL ww 6 (mt s K.-87-.heV CAL --~ E -
vudL.
Does the activity being evaluated involve proposed changes to the Technical Specifications?
NO A 4me h h \ hem ?eno.-4tedh ehrtn=4L 11 M YES miu [ (tcA .DQ th Because:
& AD , denno(4 tan'A w c %tu-(Ardlo eA qel cN6huetAc.AL U a" mA Mile im 2& 4 Aan N M.
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Does the activity being evaluated involve proposed changes or modifications to the 4
plant systems or equipment that affect nuclear safety?
m YES N NO[] Because: %
ees h b r3 ne uw ord4Lcenkes\ h t atitMA cemhMh orJ4 udl\ be taktd eMsAtWe 41 N MA%thu A bio. rm6exc otro. 595, h %n A b e.ctn a o ehts h h V.M ut 04 tA dthA. %L dress'e.,s. M.
mVW Does I e activity being mM evaluated Paeaff involve a violation o ucJt-/t.-97-tc)f the Technical Specifications?
Because: N um)Qun NO[]
YES[ 'TA. UN cesM mb
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AN AFFIRMATIVE (YES) ANSWER TO ANY OF THE ABOVE QUESTIORS REQUIRES FUR EVALUATION TO DETERMINE IF THE ACTIVITY IS DEEMED TO INVOLVE AN UNREVI QUESTION AS DEFINED IN 10 CFR 50.59 AND AS SPECIFIED BY TECHNICAL SPECI 5.5.1.7.b. PROCEED TO SECTION 7.7, BACKGROUND INFORMATION DEVELOPMENT AND ANALYSIS IF AH UNEQUIVOCAL NEGATIVE (NO) ANSWER TO OF THE ABOVE QUESTIONS CAN BE MADE, FURTHER EVALUATION OF THE ACTIVITY IS NOT RE IRED.
P repared by b.,ug vpate M2L/B9 Reviewed b{ t$k 0 _
f A-4
- Date 7/2 ole 9 > >
FC/ FORM /04
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, FC-154 R8 05-24-89 C
ATTACHMENT B Page E,,of S l
Background Information Development and Analysis Identify plant s acific design, Reference Sections 7.7.1 -- operating and tecinical documents 7.3/7.7.1=
Docusant Title ID "" W r Revision n<4e Mh iM/C TL t t. O Ts\L 7.7.2 Identify Applicable Reference Sections NRC/INP0/ Standards Documents 7.3/7.7.2 Title ID M" W r Revision G,hL A sRD L L4 EA h5L 7.7.3 Identify related Reference Sections drawings 7.3/7.7.3 l Title ID Number Revision MA I
7.8 Identify safety functions Reference Sections accidents and impact on 7.5/7.7/7.8 safety analyses List safety functions: 7 n w k.A. A te n d ftn m m sa k h n WK ~n tvulM ,LML embmwe<hs ami comM t M %u te? A Aasat .
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list applicable accidents: LotA , AN A , F#4. s&T2., &s dem hk 7mWii . C t. A Giorum 'ard LM
\ J W
' fur A ' Ss, / rali\TJAR T Terre e.n s, Are T hF U'Y Describe impact.pn previous safety analysis: Y ctwu d GACt avmbit (LE w am %.0 ewm 4 tm con h \ derfs CMniema\ %%L A/Sn .
M. t' A O ddh-BY-oL te\vs. on UR-6A / A c4 h & wtoDo2ah(q W A chamo# o w' t l ca u hrt mnput4th ace a t9AAs -F<n 4A I Ve> b fvs. Moxhnum. 9 d B-1 FC/ FORM /04
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FC-154 R8 05-24-89 ATTACHMENT B Page_f:L. of Background Information Development and Analysis 7.9 --
Determine affect on safety Reference Section functions 7.8/7.9 Operating / Design Not Evaluation Criteria Aeolicable Aeolicable Supporting once / attached Containment Integrity []
[N /
Seismic Analysis [] [$ /
System / component [] [1 /
performance Single failure criteria [] [1 /
High Energy Line Break [] [4] /
Each Accident Identified in 7.8 [$ [] ANdelmeM Separation Criteria [] [$ /
{
Control Room Habitability @ [] Md*Nm&d C Fire Protection or Fire Loads [] [}] /
Release of Radioactivity [] [$] /
Probability of Operator Error [] [$3 /
Design bases, assumptions, or [)Q [] $bh O values used in the USAR Materials Compatibility [] [j] /
Potential Consequences of [] [1] /
Procedure Errors Missile Protection [] [$ /
Heavy Loads [] [y] /
Natural Phenomena, such as [] [t] /
flood, wind, lightning Environmental Qualification [] [Y] /
B-2
. FC/ FORM /04
FC-154 R8 05-24-89
{ ATTACHMENT B Page 1 of 1 Background Information Development and Analysis 7.9(con't) Determine affect on safety References Section functions 7.8/7.9 Operating / Design Not Evaluation Criteria Aeolicable Aeolicable Supporting once / attached Electrical Failure [] [V /
Battery / Electrical Bus Loading [] [y] /
Diesel Loading [] $] /
Mechanical Failure [] [$ /
Control Signal Failure [] E4] /
Potential for internal plant [] [$3 /
flooding Security [] [$] /
Applicable Technical Specifications / Basis
[1] [] b L-^110 Installation [] [p /
Other: [] [p] /
7.10.1 Technical Specification Change Reference Sections Determination 7.8/7.9/7.10.1 Does the proposed change / resultant event require / involve YES []
a change in Fort Calhoun Station Technical Specifications?
NO [p Explain: A TK e6L k Aof hannI/L . 61t an fea n%4b CL ho - Ltma_ im M.h a MWw h _t ~
im t.rhJ U I
, NOTE: AN AFFIRMATIVE (YES) ANSWER REQUIRES NUCLEAR REGULATORY C0tHISSION
/. AUTHORIZATION OF THE ACTIVITY PRIOR TO IMPLEMENTATION. SUBMIT REQUEST
("
FOR LICENSE AMENDMENT AS REQUIRED BY 10 CFR 50.90/50.92 AND IN ACCORDANCE WITH NOD POLICY / PROCEDURE G-12.
B-3 FC/ FORM /04
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.,. FC-154 R8 05-24-89 ATTACHMENT B Page 1 of Background Information Development and Analysis
- ~.
7.10.2 ~ Determination of a 10 CFR 50.59 References Section Change, Test and/or Experiment 7.8/7.9/7.10.2 Does the proposed change / resultant event require / involve changes / modifications in facility as described in the USAR7 YES [%
Explain: N celAn\ rom h e ~A 4.1%e t O- (d -bE(oC h MM\ M e4's AeA wA Y m ie,he f,h / /,4 ,%s M t
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U llai em ,NA , / 1 Does the presosed change / resultant event require / involve YES []
changes in t1e procedures as described in the USAR7 NO [W Explair : % OMsh % 1 dad- meeabraL nu n ueauala in cAtoc C W ' rinn td thn e 4d h. come % ers as 4..bA / u_ Adn h\ hh etia=A.d -w R h wt
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O Does the proposed change / resultant event require / involve YES []
(
the conduct of tests or experiments E2.t. described in the USAR7 Explain: 44D O n 4askt er enudead aA9 alamed- a +k M Q.Attaaen.VeA w W 3 M M 'u'e n f a % J w m alo "
Ohte-c C- QU.? rn M/ .<A e m_' ale #tih ' A' bt-neu 4.N . t% h A
Does the activity affect nuclear safety in a way not YES [)Q previously evaluated in the USAR?
Explain: h ctiund O O AIL Md o udue&L *1 CA9 L'L k + Le r avniseu l hoe m cAnuM CM & s,ue w t
NOTE: AN AFFIRMATIVE (YES) ANSWER TO ANY OF THE ABOVE QUESTIONS REQUIRES FURTHER EVALUATION (ATTACHMENT C, UNREVIEWED SAFETY QUESTION DETERMINATION) TO DETERMINE IF THE ACTIVITY IS DEEMED TO INVOLVE AN UNREVIEWED SAFETY QUESTION AS SPECIFIED BY 10 CFR 50.59. PROCEED TO SECTION 7.11, UNREVIEWED SAFETY QUESTION DETERMINATION.
Prepared by 2)I b W Date Mad @1 Reviewed b d Date'i/tof$4 n i i %
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UnreviewedSafetyQuestion(USQ) Determination I7.11 ~~ List Safety Issues Reference Sections 7.8/7.9/7.11 1 Safety Issue 1:s O An S4_ bb CcAhmim RNol'tm h SOMu ahor etis i t h tLArw w to hw ha e m e id m e. 1 3. on 4
-l
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I Safety Issue 2: b f A-Safety Issue 3: u h.
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Safety Issue 4:
Lif 7.12.1 USQ Determination Summary References Section 7.11/7.12.1 Has the probability of occurrence or consequences of an accident or malfunction of safety-related equipment previously evaluated in the USAR been increased?
Determine for Safety Issue 1: See hWekel b E YES DQ NO >l.
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., FC-154 R8 05-24-89 ATTACHMENT C Page ICL of UnreviewedSafetyQuestion(USQ) Determination Reference Sections 7.12.1 USQ Determination Summary 7.11/7.12.1 Determine for Safety Issue 2: MA YES []
NO [)G Determine for Safety Issue 3: Li % YES []
NO [)G Determine for Safety Issue 4: ga YES []
NO [)d NOTE: AN AFFIRMATIVE (YES) ANSWER TO ANY OF THE ABOVE QUESTIONS ESTABLISHES THAT THE PROPOSED ACTIVITY INVOLVES AN UNREVIEWED SAFETY QUESTION AND AS SUCH REQUIRES NUCLEAR REGULATORY COMMISSION AUTHORIZATION OF THE ACTIVITY PRIOR TO IMPLEMENTATION.
7.12.2 USQ Determination Summary References Section 7.11/7.12.2 Has the probability of an accident or malfunction of a different type than any previously evaluated in the USAR been created?
Determine for Safety Issue 1: AppA&rb3 h E YES []
NO @
Determine for Safety Issue 2: ulA YES []
NO
[/J Determine for Safety Issue 3: AMA- YES []
NO [)G C-2 FC/ FORM /04
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( ATTACHMENT C UnreviewedSafetyQuestion(USQ) Determination Page,1L.of.]_[a Reference Sections 7.12.2 USQ Determination Summary 7.11/7.12.2 2 Determine for Safety Issue 4: sJ,1A YES []
NO [)t]
NOTE: AN AFFIRMATIVE (YES) ANSWER TO ANY OF THE AB0VE QUESTIONS ESTABLISHES THAT THE PROPOSED ACTIVITY INVOLVES AN UNREVIEWED SAFETY QUESTION AND AS SUCH REQUIRES NUCLEAR REGULATORY COMMISSION AUTHORIZATION OF THE ACTIVITY PRIOR TO IMPLEMENTATION.
Reference Sections 7.12.3 USQ Determination Summary 7.11/7.12.3 Has the margin of safety, as defined in the basis for any Technical Specification or in the USAR, been reduced? y Determine for Safety Issue 1: h e S//atAnd b m E. YES [$]
NO [1 Determine for Safety Issue 2: OlM- YES []
NO [f Determine for Safety Issue 3:
u/a YES []
NO [4]
Determine for Safety Issue 4: $ YES []
^
NO [)9 NOTE: AN AFFIRMATIVE (YES) ANSWER TO ANY OF THE ABOVE QUESTIONS ESTABLISHES THAT THE PROPOSED ACTIVITY INVOLVES AN UNREVIEWED SAFETY QUESTION AND AS SUCH REQUIRES NUCLEAR REGULATORY COMMISSION AUTHORIZATION OF THE ACTIVITY PRIOR TO IMPLEMENTATION.
Prepared by2 8f d _ T C te A/El.189 I Date 7 f2o[g4 L; s J ' ~ h 'im Reviewedb(in11 f_ d 1 a'in tan n . ',,,. - num C-3 y-- --
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ATTACHMENT D Page 7 of Conclusions, Review and Approval 1 .n _ - _ _
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7.13 Summary Reference Section 7.13
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7.15 NRC Approval Reference Section 7.15 Is NRC authorization required prior to implementation? YES QQ NO [] Exp, lain: % Tat 6 te m t s ec h 4. 6 u n Tzan cauh TNL We W he amm'A+ he e ce id tA u+elhwn ' wM sk 2V ho as effe> _ _Sa #w daaa !% //.m,% e div n ohu h .4 M . (dytk el e m &th McAM Awd h A . % bem WG. i M tJe ,k w J k wmt /w deh a L ht M fr YM s er k ErjL a mu ,M Am ! I'14 v4 d>> > , k e M Eumi+ov-k no11a eL ksi-
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r NOTE: IF THE RESULTS OF THIS SAFETY EVALUATION CONCLUDE THAT THE ACTIVITY BEING EVALUATED INVOLVES A CHANGE TO THE TECHNICAL SPECIFICATION OR CONSTITUTES AN UNREVIEWED SAFETY QUESTION, NUCLEAR REGULATORY COMMISSION AUTHORIZATION OF THE ACTIVITY IS REQUIRED PRIOR TO IMPLEMENTATION. 7.16 Certify Performance Reference Section 7.16 "I hereby certify that this Safety Evaluation is complete and accurate to the best of my knowledge." Performed By:
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I Date: )4lEll81 Reviewed and Approved By: 1d Date: 'l[2.cftg 0-2 FC/ FORM /04 I - - - - - - - - - - - - _ - _ _ .}}