ML20082M808
| ML20082M808 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 12/05/1983 |
| From: | Sherwood G GENERAL ELECTRIC CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-44, REF-GTECI-EL, TASK-A-44, TASK-OR 831202, JNF-086-83, JNF-86-83, MFN-222-83, NUDOCS 8312060286 | |
| Download: ML20082M808 (40) | |
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[**i GENER AL $ ELECTRIC NUCLEAR POWER SYSTEMS DMslON
$/N82N08N2i-5040 NE2 h JN F-086-83 December 2, 1983 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D.C. 20555 Attention: Mr. D.G. Eisenhut Division of Licensing
SUBJECT:
IN THE MATTER OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT (GESSAR II)
DOCKET N0. STN 50-447 APPENDIX 15E - STATION BLACK 0UT CAPABILITY Attached please fir.d a draft of new GESSAR II Appendix 15E pertaining to station blackout capability. This appendi>. concludes that the GESSAP, II station blackout capability exceeds ten (10) hours. The assessed capability assumes credit for operator actions that are straightforward and where means exists to enable the operator to execute the action. Where features and/or equipment are not present, potential design improvements are recommended.
It is anticipated that upon completion of NRC review a formal amendment on j
the GESSAR II docket will be submitted. This is anticipated to occur in j early 1984.
l If there are any questions on the:information provided herein please contact
! J.F. Quirk at (408) 925-2606 or J.N. Fox of my staff at (408) 925-5039.
l Very truly yours, i
l l
enn .S rw a er l Nuclear Safety & Licensin Operation I
- Attachment cc
- F.J. Miraglia (NRC) l D.C. Scaletti (NRC) i C.0. Thomas (NRC) l R.M. Ketchel (GE-Washington Liaison Office) 0 d L.S. Gifford (GE- Bethesda Liaison Office)
R. Villa (GE) [0 3 t
I l
8312060286 832202 DR ADOCK 05000447 PDR
', s GESSAR II
+ 238 NUCLEAR ISLAND mm g
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ht II APPENDIX 15E STATION BLACK 0UT CAPABILITY
e APPENDIX 15E CONTENTS SECTION TITLE PAGE 15E APPENDIX 15E - STATION BLACK 0UT CAPABILITY 15E.1 INTRODUCTION AND CONCLUSIONS 15E.1.1 Introduction 15E.1.2 Conclusions 15E.2 DEFINITION OF STATION BLACK 0UT 15E.3 INDICATION OF STATION BLACK 0UT 15E.4 INSTRUMENTATION REQUIREMENTS 15E.5 PLANT RESPONSE FOLLOWING A STATION BLACK 0UT 15E.5.1 Areas 15E.5.1.1 RCIC Room 15E.5.1.2 Remote Shutdown Panel Area 15E.5.1.3 Suppression Pool 15E.5.1.4 Drywell 15E.5.1.5 Control Room 15E.5.1.6 Fuel Pool 15E.5.2 Energy Supplies 15E.5.2.1 Pneumatic Supply 15E.5.2.2 125 VDC - Bus E 15E.5.2.3 125 VDC - Bus F 15E.5.2.4 125 VDC - Bus G 15E.5.2.5 125 VDC - Bus H 15EA ATTACHMENT A TJ APPENDIX 15E - ACRS QUESTIONS l PERTAINING TO AC/DC POWER SYSTEM RELI ABILITY 15EA.1 DC RELIABILITY 15EA.2 GRID RELIABILITY 15EA.3 DIESEL GENERATORS 15EA.4 LOW POWER TESTING / SIMULATED LOSS OF 0FFSITE POWER ISEB ATTACHMENT B TO APPENDIX 15E - RCIC ROOM HEATUP DURING A STATION ELACK0UT l
15EB.1 PURPOSE 15EB.2 INTRODUCTION 15EB.3 MODELING AND ASSUMPTIONS 15EB.4 INPUT PARAMETERS l 15EB.5 RESULTS AND DISCUSSION l
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e APPENDIX 15E TABLES TABLE TITLE PAGE 15E-1 Variables Assessed For Station Blackout Assessment 15E-2 Power Supplies to Instruments Needed For a Blackout APPENDIX 15E FIGURES FIGURE TITLE PAGE 15EB-1 RCIC Room Temperature Response Following a Station Blackout 4 15EB-2 RCIC Room Temperature Response Following a Station Blackout - Sensitivity to High Water Temperature 15EB-3 RCIC Room Temperature Response Following a Station Blackout - Sensitivity to Low Steam Leakage Rate 1
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e 15E.1 INTRODUCTION AND CONCLUSIONS 15E.1.1 Introduction This appendix
- vovided to demonstrate that the GESSAR II design has substantial capability to prevent a core damaging event well beyond the two-hour value recommended by NUREG-0626 and assumed in the Probabilistic Risk Assessment (Section 150.3).
Attachment A contains responses to pertinent questions on station blackout of interest to the staff. These are addressed in more detail in other parts of this appendix.
15E.1.2 Conclusions The GESSAR II station blackout capability exceeds ten (10) hours. The assessed capability assumes credit fcr operator actions that are straight-forward and where means exist to enable the operator to execute the action.
Where features and/or equipment are not present, potential design improvements are recommended. These operator actions and potential design improvements are summarized below:
- 1. Operator Actions
- b. Shift of RCIC pump suction to the condensate storage tank.
- c. Vessel depressurization with SRVs to about 200 psig. Maintain vessel pressure above 150 psig with manual SRV control.
- 2. Potential Design Improvements
- a. Provide manual logic override of the RCIC suction transfer signal and test line closure signal from the control room.
e i
- b. Provide Enhanced Water Level Instrumentation (currently under review for Appendix 1D).
- c. Provide alternate power supply to RCIC gland compressor.
An ongoing evaluation of the 125 VDC battery capability is in progress.
However, if necessary to ensure 10-hour capability, emergency DC bus cross ties, or larger battery capacity, or other methods will be identified.
In addition to the above actions, the following contingency actions could be taken to provide even longer duration capability are:
- 1. Provide override capability for the RCIC room high temperature isolation logic to be used if room temperature exceeds about 150 F.
- 2. Extend SRV pheumatic supply by replacing air bottles if depleted.
A connection outside the fuel building would be more convenient.
15E.2 DEFINITION OF STATION BLACK 0UT Station blackout refers to the total loss of both off-site and on-site a.c.
electrical power. In draft information pertaining to proposed Regulatory Guides, the NRC consultants refer to " Emergency AC" loss in addition to offsite power loss. This could be interpreted as the Division 1 and 2 Standby Emergency Diesel Generators. Both HPCS and RCIC operate et high pressure and can be considered redundant water sources available for maintaining core cooling during design basis assumptions that assume a single failure (i.e.,
such as a D-G). This configuration is believed to be adequate to comply with the proposed regulatory requirements. For purposes of this assessment,
however, a failure of the HPCS diesel generator has been assumed in addition to loss of offsite power and the division 1 and 2 diesel generators thus providing a more severe impact on plant systems and the station battery.
A one-line diagram of the GESSAR II design is shown in Figure 8.3-1. Three divisions of 6.9 kv on-site power are provided; two by standby emergency diesel generators (in addition to preferred and alternate off-site power sources); the third by an off-site power source and a separate and diverse diesel generctor dedicated to division 3 eletrical power. Division 3 supports the High Pressure Core Spray (HPCS) system and all of its supporting auxiliaries.
The GESSAR II design also includes a steam turbine driven Reactor Core Isolation Cooling System (RCIC) which operates in an emergency independently of a.c.
electrical power. This system is designed to provide high pressure makeup to the RPV during isolation events and would thus be initiated automatically during a postulated blackout event. The plant response with RCIC alone has been reviewed, and the duration capability of the GESSAR II plant in excess of ten hours has been verified. This configuration is consistent with the station blackout definition in the Probabilistic Risk Assessment (Section 150.3).
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l In the evaluation certain assumptions have been made:
o No Loss of Coolant Accident (LOCA), stuck open relief valve (SRV) or failure to scram concurrent with the station blackout is considered.
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o o In evaluation of equipment, some capability beyond environmental qualification limits has been assumed. In assessing the ultimate failure capability of equipment the judgement of senior General Electric engineering personnel has been relied upon to provide guidance. Such judgements are explicitly call out in the following sections.
o Operator actions are identified where adequate time and skills would be expected to be available to a typical operating plant staff. No extra-ordinary actions on the part of the operator are onlu assumed; rather, ggstiaightforward, simple actions are allowed.
o No credit for off-site assistance from a utility maintenance crew uc *ng portable electric generators or batteries has been assumed for this assessment even though this possibility may exist within the time frame of interest. Such capability might be considered by an applicant to improve the restoration time for on-site emergency a.c. power if the situation warranted.
I 15 G. 3 INo\ C/rTrog OF s TAWc SLA CKOVT j l
'Ibe station blackout event is daracterized by a loss of all off-site l
power (preferred and alternate feeders) and a loss of divisions 1, 2 and 3 of on-site a.c. power. As noted in Section 1D.2.3.33 of the !
l asmsanent against Regulatory Guide 1.97, the class 1E power distribu-tion system monitors voltage on the three 6.9 kv a.c. buses and the four 125 V d.c. buses. This indication is displayed on panel P800 in the main control rom. A potential station blackout event would be first noticed by the plant operatccs by a change in the control rocn lighting whi& would alert him to evaluate both the plant and the electrical distribution systen status, By observation of the loss of bus voltage en the 6.9 kv tuses "E", "F" and 'C" and the breaker position for incoming voltage to these bums, the operator would be alerted to the tresence of a potential blackout event. Voltage indication on the d.c. buses E, F, G & B would assure the operator that power is available to control the event.
Prior to conducting the various operator actions needed to mitigate a blackout event, the operatcr must distinguish between a short duration event and a prolonged blackout. A short duration event would be one in whi& restcration of an off-site or on-site a.c. power source would occur prior to develognent of conditions requiring the operator actions
defined later in this agplanent. Minimizing the time to recognize this event is insortant so that the potential drain on the batteries is controlled.
Upon recognition of the c.c. power source failure, an auxiliary operator would be sent to ea& of the diesel generatz rooms to attenpt a manual start. Sinultaneously, the control roan operator should attengt to start ead diesel from the main control roan. In addition, the systen dispatcher would be contacted by the shift supervisor to determine the status and likelibcod of off-site power restoration. Accomplishment of these activities in addition to those related to controlling vessel water level and pressure is expected to take about 30 minutes.
'Jhus recognition of a station blackout event and th:e initiation of any blackout specific operator actions is expected to be delayed for about 30 minutes.
15 E . 4 'I N sTru; m e WA%Qtd GL5h IMEM 6NT J Instrtunentation required to monitor plant status chring a blackout event has been selected fran a review of the type A through E variables gpowdI%
discussed - ion ID which is the response to Reg Guide 1.97 require-ments. This list has been augmented slightly to account for spcific variables sud a roan tanperatures and certain valve and breaker position indications which are needed to determine plant conditions.
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ISE-I Table 4Egists the variables considered and whether x not they are needed for the blackout sequence. Se basis fx selection generally is based on the need for the operator to follow Emergency Procedure Guide-lines (or take other actions whidt may later be established) during the period of interest. As such, type A variables are identified as need-ing indication during the blackout event while variables whidi are more representative of monitacing core damage or breaks of the reactoc coolant boundary or effluent releam are excluded. m I SG-7 2 Table 4=2 pows the power stqplies in the GESSARgdesign for the instrtanents needed. All indications needed to follow the blackout event are oc will be powered frm 125V d.c. sources.
cold d bgekup, he applicang6 provide d.c.A :ch;d power to the condensate storage tank level indicata and to ensure local control rom tapera-ture indication as available.
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L5E.5 ptaw7 fasSPdMJE FOL W 4JG A STR OQ 8tAcupT 1he key plant areas whi& could potentially effect the ability of t:he plant desicp to acccanodate a station blackout are:
o RCIC ro m o Renote shutdown pmel area o Stpsression pool and containment o Drywell o Control roan o Fuel pool In addition non-electrical a.c. plant energy supplies will be consumed and need to be addresad to asmss the plant capability. These are:
o Pnetsnatic Air Stpply Syster. (ADS) o D.C. Power Distribution Systen
'Jhese areas and energy supplies will be discussed in subsequent su -
g revemodi sections. An estimate of limiting condition, design or operatcr actions needed are noted in ea&.
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. .'r=:4 ncrc par =
- 4. Raasen for Concern o Rom taperature increase without area cooling could cause a loss of RCIC control due to equipnent failure, o Isolation and turbing trip due to leak detection systs trip. (Trip aetpoint approx.170 F) could prevent RCC fra operating .
vollves loecow s o Ste m line drain:4may fail after air supplygexhausted causing syste dunage on restart.
- h. Plant Response l o Approx. 1 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (w/GT suction) o Approx. 133 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (w/SP suction) j See Attachment B o Approx.101 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (w/10 lb/hr ste m);
Critir.nl r, _- m-gs T. imitation EE Differmtial Coil Approx. 170 F water ta p.
Magnetic speed Sensor 225 P Capability Instrumentation 212 >12 hrs l
C. A===a coerator Actions o Manual switd> of RCC suction to GT at about 30 min, o Override RCC high tmp isolation if rom tap > apgox.150 F (not expected) o Manual RPV level control of RCC to avoid L8 trip and rest. art.
- d. Potential Wrlifications/ Actions o Ensure override capability exists for RCC rom isolation signal.
o Ensure override capability for RC C suction transfer.
o Provide logic changes to pennit low flow RCC injection. Requires override capability on test line to GT to obtain flow split between GT return and vesel.
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. &. $ up Asee:p Damr*o h*rhm Panel Area
- 4. Da==<m for conrwrn o RCIC electronics could fail if area taperature exceech 150 F.
o Access needed if control rom tacomfortable or electronics erratic.
- b. Plant Response o Not evaluated, tmt very little heat source ince R e ote Shutdown Station pane t1 @brgized tmtil control tran fer switch is thrown, o Expect area taperature to renain
<150 F for 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> C. Assumad Ooerator Actions None.
. Potential Mrvlifications/ Actions None.
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.- 7.r= :p arrw== ion pnni A.Daanan for h rn o Hi$ sippression pool taperature could cause NPSH limits (apgox.
1757) and reduced lube oil cooling j to ft.CJ C.
- o Hi$ suppression pool level causes suction transfer.
o Hi@ containment air taperature may cause erratic RW indication.
o High supgession pool taperature and level increases contairunent loads, b . Plant Dannnrune
. ting T L Notes:
@T - Jip Tcbased on Tsp + judcynent (hrs) ( F) (g)
F (ft) T based on Table 15D.2-2 Ecalculation 1 135 100 +2 5 190 175 +5 Capability >10 hours.
10 220 220 @ weir 15 225 225 0 weir 20 30 30 @wdr Instrtunents qualified to 185 Fy capability likely to 250 P.
C, Assuned Ooerator Actions o Manual switchover back to CST within 1 hr. eliminates potential NPSH proble.
o Maintain vessel pressure below heat capcity taperature limit per EEGs
- ensure written procedures contain heat capcity taperature limit curve 6 - may need to exceed heat capcity taperature limit slightly after approx. 6 hrs, but acceptable because no additional depressurization required. Consistent with EPGs.
d Potential Modifications / Actions o Ensure manual override capability for RCIC suction transfer 1
__ n _.
GG SS AC. 'E 2.3 6 No cuna. Tst. Aw 0 ATTACH Mister A
-To APPswoW 15 G AC R. S Guss T\oNS Ps9 TAW \NG TO A C /v c POWsit. S YSTE M e.tst \ A G t L s TY ,
\5E A.I O C R.s1.A A su a-T'1' Onaation The tRC Staff has issued a report (NURIn-0666) on the reliability of d.c. power system in which a 2-train d.c. systs found to meet minimm NRC requirments was evaluated. As a result, the d.c. pwer systs was identified as a ptentially high contritutor to core melt.
The applicant could be asked what his assessment of his d.c. systs is and what consideration he has given to the recommendations of !bRID-0666.
Resconser We do not favor the use of such a minimum systs as considered in IUREG-0666, For example, it has a single bus tie breaker with too much ptential for The GWSSActII common cause failure. Our originalgoesign allowed d.c. cross-connection capability with . dual cross-tie breakers and double key interlocks. GE '
ape.d do kde E d . C. Cro3.s.c.o wccdson c.cio n d k w c, \ 4 i d C ck w N Shosw 4-( d 4h s c.etpabsigh do4 J noi cowk<*k kC b d. c_ sy4~ e kak ht1-The following is govided in respnse to the remmmendations in NURID-0666:
(1) Prohibits certain design and operational fectures of the d.c.
power systs such as use of a tie breaker which could compromise divisional independence. As noted above, GESSAR[ecomplies although we believe cross-connection cambility is appropriate for specific conditions during shutdown and occurrences which require last resort flexibility (such as station blackout). GESSAR s four 28- !
saf ty-related batteries, cech of which has two chargers so that charger I maintenance does not require use of cross-connections nor cause draw-down on the battery.
(2) Addresses testing and maintenance activities. Tnese are acocm-plished by the applicant. We agree with these recmmenda-tions, and the GESSgdesign allows their implenentation.
(3) Requires staggered test and maintenance activities to minimize the potential for human error related common cause failure. This is controlled in the field, but we agree that these actions are appropriate.
(4) Requires design and operational features to be adequate to maintain reactor core cooling in the hot standby condition follosiing the loss of any one d.c. power bus and a single independent failure of any other systen required for shutdown cooling. Although we cannot disagree with the intent of this recommendation, a judgment as to what features are needed should be tenpered with an assess-ment of the reliability of the d.c. power loads and sources. We have concentrated on maintaining full separation and independence between division 1 and division 2 d.c. systens to provide this reliability. ' .
6With four independent d.c. systens and with three inde ndent a.c. sfs M Ne $ N $t$sIan pability m Nu k N U Rsh- 0 6 66 a cow a on.
For example, a potentially adverse capability loss would follo.i from the loss of both RHR systens, but the suppression pool can
-t9-
storo decay hest for sevrral hours, during which it may be
.' possible to recover active decay heat renoval.
(SEA. 2. Grid Reliability Oi== tion What is the applicant's assessment of grid reliability and what procedures exist for restoring offsite power to the plant in the event of this loss.
Resoonser 'Ihe grid is the responsibility of the applicant, and we assume he will meet the NRC requirenents in this area. On less of normal preferred offsite power, there is autoratic transfer to the alternate offsite power source and, if necessary, to the onsite diesel generators.
Restoring preferred power is acwmplished tranually by the control roan operator. The specific procedures for restoration of power in the switchyard or transmission systens would be developed by the applicant.
Station Blackout Analysis Ouestioni What are the results of the applicant's station blackout analyses? Has the applicant made a best-estimate analysis of the accident sequence and evaluated what might be done to improve the plant, cr has a conservative analysis been made with a core Iaelt assumed after scme specified degradation of the battery?
% i .: evokuaktow HspedJ +o b ek #3 bo" 8 '
Resocnser A Our best-estimate analys s to the extent that it is complete is the primary subject of this supplement. We have identified potential systen design and procedural improvements, and we will implemed 4kow up on concurrowce Qo Oe. N RC +b a+ +Ln3 s aki s % Oors aso\vs 4-k t3sw ,
-3o-
Our probabilistic risk assessment mnsidered station blackout capability in a mnservative manner (core cooling lost in two hours due to battery depletion and loss of RCIC control). We believe the more realistic treatr.ent considering automatic and manual d.c. load shedding shows a substantially longer capability.
15GA.3 Diesel Generators Ouestion: What is the applicant's assessment of his diesel generator syst s? To what extent has LER and operating experiences been used to improve the design?
Resronse Our HPG diesel generator has undergone extensive testing (including 300 tests without failure) which has been dacumented for the NRC. From this testing and from field experience we have high confidence in the design. Extensive review of the design spcifiation, the installation design and the auxilian systs design for the larger diesel c evwows4rakeJ l
generators (division 1 and 2) enmuummusummpuum high availability from l
these units.
l l5E A e 4 Low Power Testina/ Simulated Loss of Offsite Power Ouestion: Has the applicant performed low power testing and a simulated loss of offsite pwer test? If so, what are the results and what has the applicant learned?
Resmnse:
'TE s s s O e. V4.S f obJihId of M Apph c.a d ,
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15G .5. h f 3 Lg &vw11 OL Reason for Cancern o Hi@ drywell tanperature could cause RN level instrtment reference leg bolloff.
o Hi@ drywell tenperature mi@t exceed qualification levels for drywell equipnent.
o Hi@ drywell tenperature could cause SRV solenoid failure.
b Plant Jesponse Approx.135 F during plant operation ,
<270 F price to depressurization at 30 min. Capability: l unlimited
<200 F after depressurization to 200 psi' Drywell equignent qualified for >300 F C- Ammunad Ocerator Actions o Depressurization to approx. 200 psi to limit drywell heatup, o Maintain gessure >118 psi to avoid reference leg flooding, o Maintain RN water level apgox. + 20" on Enhanced Level Instrtment.
g, Recnmanded Modifications / Actions o Enhanced water level instrtment (ELI) compensates for drywell and contairment tenperature effects. (Previously rectrumended. See SWsMt App tN 1D.)
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, g1a (hntrol parun 4.Raasan for Concern o HiW1 control rom taperature could cause caputer/microgocessor controls to fail.
-- o High taperature could make the control rom uninhabitable.
b- M Response o PGCC floor sectign heat sinks expected to prevent Capbility:
heatup above 105 F. miimited.
o Beidity could beame mcomfortable but not tminhabitable.
MicroprocesEors (ELI, ERIS, etc.) mreliable above apgox.105 F but backup information is available at Rmote Shutdown Station (RSS).
- c. Assmed Operator Actions o Transfer control to rmote shutdown station (RSS) if control rom beames tminhabitable. (not expected)
- d. Potential Mndifications/Actiorm Kone.
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- 1 V~ -
~ . _ . _ _ . _ _ _ - _ , __ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ . _ _ _ _ . _
IS G. 5,( . 6
. h zuel A e l _ _ . _
O. Emmaco for enevwrn o Loss of fuel Ir,ol cooling could cause fuel pool to boil away.
le, M ant nesponse o Approx.14 hrs to boiling Basis: Judgnent probably lcoger with o Approx. 77 hrs to fuel tmcovery less hot fuel Capability >75 hrs.
C. Anstmed Ocerator Actiens None, but SRV air bottle replacement (see ineumatic supply) could be hanpered by fuel building environnent after approx.14 hrs.
d,PotentialModifications/ Actions Consideration of moving extra air bottles to corridor outside fuel building. Not required for station blackout.
-: 6 -
15 E. 5.2. E n ev% S w p p b __
"'8
- T1y 97= =dH c nmply 7
- a. g wire == of censi==+1nn o ADS /SRV o Drywell and contairinent vactaan breakers
- b. Fatim tad Duration (5000 & available)
SRV Depressurization apgrox. 50 actuations @ 8 T/ actuation = 400 &
Ongoing SRV use approx.
2 x x8 = 240 Cal Leakage @ l TB/ valve x 8 valves = 8 CMI Di Vacutan Breakers approx. h 9 x 2 VB = 4 GB total approx. 250 CHI 0
= 18 hrs. Capacity >18 hrs C, Ooeratnr Actions to Extend Duration o Air bottle replacernent after depletion possible if necessary (not expected).
o Rotate use of ADS /SRV valves to permit time for acctanulators to redlarge and give preference to Division 2 ADS /SRV values, o Monitor SRV position indimtion to indicate need for switch to other values (valves clom when air supply lost).
, Potential Modifications None i
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. L S s . s. 2 . t
. . . , _ _ p _,J?s vm - == E
- 4. jlajor sources of con ==-+1on See Table 8.3-6
- b. Estimated Duration (1950 mp hours (AH), 2 hr)
RCIC Gland Cagessor Modification (see belm) delete 58A Capability Shed load appox 35A (see below) >
- hrs.
Steady state load apgox. 251-58-35 = 158A C. Operator Actions 1.0 Extend Duration o Shed the follwing loads (at apgox. 30 min.)
- !#E panel H13-P669 (NSPS) - 25A fra EPS inverter Bnergency lighting (fuel building) - 10A
- d. Potentid Modifications o Power RCIC gland comgessor fra an alternate source, o Delete 125 VDC emergency lighting systen except for control building or move to Bus J.
o Provide Emergency crosstie capability with dual crosstie breakers and double key interlocks if needed for longer duration.*
o Provide larger capacity battery if needed for longer diration*.
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- 2he capability of this battery with load shedding is being evaluated. If the estinated &lration is less than about 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, the addition of crossties or expanded battery size will be reviewed to determine the q+i== configuration for achieving a 10-hour capability.
l 15 E . 5. 2 . 3 1 M vrr - == r
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a . Major hwr== of con ==+1on i
See Table 8.3-7 b . Estimated Duration (1500AH, 2hr)
Shed Loads approx. 40A (see below) Capability Steady State Load = 175 -40 = 135A >
- hrs.
C. Operator Actions D Extend Duration Shed the following leads at approx. 30 min.
- lets panel H13-P670 (NSPS) -25A
- Dnergency lighting -15A
<d.potentialunaifications o Delete 125 VDC emergency lighting in auxiliay tuilding o Provide larger capacity battery if needed for longer duration."
d 1edding is being evaluated. If
- The the estinated chration is less than about hours, the addition of capability of this battery with load p)X crossties or expanded battery size will be reviewed to determine the optinun configuration for ac211eving a 10-hour capability.
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15 E .5. 2. 4 in vre mise a
, 2r- ggy g Mgot mire == of Nai-+4m See Table 8.3 Estimated nicatim (400 AH, 6 hr)
Shed Loa h = 25A (see below) Capability SS lead = 78 -25 = 53A >
- hrs.
Operator Ac+ims to Ertand Duration Shed the following load at apgox. 30 min.
1995 panel H13-P671 (NSPS) -25A Potential Mndificatims Larger capcity battery if needed for longer duration.
- Ihe capability of this battery with load shedding is being evaluated. If the estinated & ration is less than about'#A hours, the addition of crossties or expanded battery size will be reviewed to determine the l optimum configuration for achieving a 10-hour capability.
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I 5 E.5.2 5
, 191 vrr nim n Egigt sourans of cmai=rtim See Table 8.3-9 Ritimated Duration (425 AH, 2 hr)
Load Shed = 25A Qtpability SS Load = 100 -25 = 75A > * . hrs.
Operator Actions 1,Q Extend Duration Shed the following load at approx. 30 min.
Shed 18tS Panel H13-P672 (NSPS) -25A Potential Modifications None AO
- The capability of this battery with load edding is being evaluated. If the estinated duration is less than about hours, the addition of crossties or expanded battery size will be reviewed *.o determine the optimum configuration for adiieving a 10-hour capability.
- 2. l -
nBLE oe. is s-I waIABus assessed nu srurrn Bucsarr assessnour RG 1.97 RG 1.97 Discussion Needed in Black-Variable TVDe Category Subsection out Secuence?
naartivity control Neutron Flux A,B 1 1D.2.3.1 No*
(value, rate, trend)
Control Rod Position B 3 1D.2.3.2 No*
Boron Concentration B 3 1D.2.3.3 No (sample) enre coolina Coolant Level in the A,B,C 1 1D.2.3.4 Yes Reactor (value, trend)
Maintaining Reactor Coolant System Intecrity RCS Pressure A,B,C 1 1D,2.3.5 Yes (value + alarm)
Drywell Sump Level B,C 3 1D.2.3.6 No (value + alam)
Drywell Pressure B,C,D 1,2 1D.2.3.7 No Primary Containment E 1 1D.2.3.8 No Area Radiation C 3 Suppression Pool A,C,D 1.2 1D.2.3.9 Yes Water Level Maintaining Contalment Intearity _
Primary Contairunent B 1 1D.2.3.10 Yes**
Isole.cion Valve Position (Excluding 01eck Valves)
Prinary Contairunent A 1 1D.2.3.ll Yes Temperature
- A'IWS plus blackout is not considered in this study. Failure to scre ctn be inferred from ainormal watet level and pressure response.
C*Plus RCIC minimLIR flCW.
- 1. 2 -
DELE *sF t 5 E-l VARIABLE ASSESSED POR STEPIM BUCKOUT ASSESSMTr (Ceb )
RG 1.97 RG 1.97 Discussion Needed in Black-Variabla Tvne Category s h ion out Seauence?
4 Maintaining Containnent Intearitv fcantiruwi Primary Containnent A,B,C 1 1D.2.3.12 Yes Pressure (value, rate, tr ed,
+ alaan)
Drywell/Containnent A,C 1 ID.2.3.13 No Hydrogen Concentration (value)
Secondary Containnent C,E 2 1D.2.3.14 No Area Radiation (value)
Secondary Contairunent C,E 2 1D.2.3.15 No Noble Gas Effluent Primary Cmtainnent C 3 ID.2.3.16 No Noble Gas Effluent Suppression Pool A,D 1,2 1D.2.3.17 Yes Tenperature Drywell Air Tenperature A, D 1,2 1D.2.3.18 Yes Fuel Cladding Barrier Monitoring l Coolant Radiation WA WA 1D.2.3.19 -
l (value + alaan)
Coolant Gama C 3 1D.2.3.20 No (1 sample /6 hours) rcults within 72 hr Systern (Veration Main Stean Line Isolation D 2 1D.2.3.21 No Valve Leakage Control Systen Pressure Contairunent Spray Flow D 2 1D.2.3.22 No
- Z.'s -
l
DBLE 4s3 I S E -l VARIABLES ASSESSED PCR STATION BMT ASSESSMENT ( CedwM)
RG 1.97 m 1.97 Discussion Needed in Black-Var 4mh1a Tvoe Qttegory 33tgiection otit Samarice?
1 System Operation fcentim e Re idual Beat Remwal D 2 1D.2.3.22 No (RHR) Systen F1w ;
RHR Service Water F1w D 2 1D.2.3.23 No LN Pressure Coolant D 2 1D.2.3.22 No Injection Systen F1w -
Reactor Core Isolation D 2 1D.2.3.24 Yes Cooling Systen Fiw RCIC Rocan '14rnp. - - -
Yes Control Rocan 'Dunp. - - -
Yes High Pressure Coolant D 2 1D.2.3.24 No Spray Systen F1w Core SIray Systen F1w D 2 1D.2.3.24 No Standby Liquid Control D 2 1D.2.3.25 No Systen (5,CS) F1w SLCS Storage Tank Level D 3 ID.2.3.26 No SRV Position D 2 1D.2.3.27 Yes FeecMter Flw D 3 1D.2.3.28 No CET Level D 3 1D.2.3.29 Yes ESF Cooling Water F1w D 2 1D.2.3.3 0 No
! ESF Cooling Water Tenperature D 2 1D.2.3.30 No f Bich Radioactivity Tank Level D 3 1D.2.3.31 No Bnergency Vent Damper Position D 2 1D.2.3.32 Yes Stancby Energ Status D 2 1D.2.3.33 Yes*
CIncluding breaker position.
l
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. N # IS E-l
.varmus msmszD xm ernum aucuxrr msmsme Ccod-,\)
RG 1.97 RG 1.97 Discussion Needed in Black-Variable ,Tvoe g m hnact irm g Sernm_nce?
Effluent Monitoring SG25 Ventilation Flw Rate E 2 1D.2.3.34 No Other Ventilation Flw Rates E 3 1D.2.3.34 No Particulate / Halogen E 3 1D.2.3.35 No Release (sample)
Errtirons Radioactivity E 3 1D.2.3.36 No Monitcring Meteccologt E 3 ID.2.3.37 No Post-Accident Sampling E 3 ID.2.3.38 No (sample) l l
1
TABLE SG \ 5 E- 1 POWER SUPPLIES '10 INSMMNIS EEEED POR A BLACKOUT Control Rom Power Variahla IndicM or SLgply Available? Notes RW Invel B21 R623A 120 Inst. Bus A Yes 1 R623B 120 Inst. Bus B RW Pressure B21 R623A 120 Inst. Bus A Yes 1 R623B 120 Inst. Bus B Yes 1 Suppression Pool Water Level P50-R600A,B 125 VDC Yes 3 Pri. Contaiment Isol. Valve Indimtion RPS Yes Position Lichts Pri. Containnent Tenperature T41-RR613A,B 125 VDC Yes 3 Pri. Containnent Pressure N1-RR618A,B 125 VDC Yes 3 Suppression Pool Tenperature P50-R600A,B 125 VDC Yes 3 Drywell Air Tenperature N1-BR611A,B 125 VDC Yes 3 RCIC Flow E51-R606 RPS Yes RCIC Rom Tenperature E31-R608 RPS Yes Control Roan Tenperature - -
Yes 5 SRV Fosition Indimting 125 VDC Yes Lights CST Level By applicant By appli mnt Yes 2 Dnergency vent Denqmr Position Indim ting 125 VDC Yes 3 Lights Stancby Energy Status 619 kv AC Volt 2neters Source Yes DC Volt 2neters Source Yes Air F53-R606A,B 125 VDC Yes 3
-LG-
. Netan to Table w $ g . {
- 1. Enhanced Water Level Instrtment to be powered from d.c. power.
- 2. D.C. power to be provided by applicant.
- 3. Power Supply frcan 125V d.c. to Reactor Island Logic Panels P881 or N82.
- 4. Exhaust air measurenent may be tnreliable. Local thennczneter to be supplied by applicant.
-L7-
1 l
G5SS AG "If
,' 2'69 N u caA z. Ist AND ATT AC NM15 k>T 6 TO A P Psuoi V iSE p.ctc R.o o s HsATVP Dum wG A s r Am oW BLAcvooT I SE B . L PURPOSE a% ch~e A The purpose of thii meneeendesAis to document the resnits of analysis perf ormed by Containment and Radiological Engineering on Reactor Core Isolation Cooling System (RCIC) room temperature response during a station blackout for thus GESSAR - U. .
geant The resnits indicate that a station blackout imposes no threat to the speration of RCIC with the RCIC room temperature reaching 122'F 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> into the transient, well below the point above which RCIC perf ormance would be degraded.
S:nsitivity results for some of the most important parmeters are also given.
65G8.2 4 INTRODUCTION A station blackout resnits in loss of all A.C. porer (both offsite and onsite source s) , initiating reactor isolation and scre. For this analysis all three diesel generators of a BWR plant are assumed inoperative, i.e., no Emergency Core Cooling System (ECCS) pumps are available: this leaves the batteay operated RCIC as the only system available for core cooling. Thus, it is essential that the RCIC remains cpe rationa l . An important requirement for the proper functioning of the RCIC is that the RCIC room temperature be maintained below the equipment operational limit.
The loss of all A.C. power also means the loss of Isshting, auxiliary equipment speration, area HVAC and drywell fan coolers, resniting in a drywell heatup. At some point reactor depressurization will be initiated to reduce the heat input to the drywell, although the reactor is assumed to be depressurized only to the point sufficiently above the RCIC shutoff pressure so that operation of the 2CIC can be beintai ne d.
RCIC initially draws water from the Condensate Storage Tank (CST). However, an ottomatic switchover to the suppression pool as the water
-3t-
l e
1
\
l s;rrso would occur if the CST water level drops too low or the suppression pool water l 1evel rises above a eertain point. Since the suppression pool heats up as a result I
cf SRV discharges and subsequent reactor depressurization, and since the design taaperature for the kCIC pump is 140*F, a manual switch back to the CST from the s:pprcesion pool as the RCIC w6ter source is required when the pool temperature cypreaahes 140*F. Since the time period when the RCIC takes section from the l styprocsion pool is relatively short (about 30 minutes) compared to the transient period of interest (up to 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />), the impact on RCIC room temperature in assuming that RCIC draws all water from the CST is insignificant.
SEB.3 hM0DELINGANDASSUMPTIONS To cedel the RCIC room temperature response, thermodynamic prcperties of steam and cir la the room are evaluated based on mass and energy balances. Heat source a and host sinks were considered. In addition, some steam has leaksd into the room through ths RCIC turbine 31 sad seal. The room is conservatively assumed to be isolated from the cdj acent rooms.
Esst Scurce s - The following heat sources are modeled:
o Steam Pipes - there is a six inch steam pipe upstream of the RCIC tzrbine, 60 f t long, with three inches of insulation, with the pipe temperature assumed equal to the reactor steam temperature of 552'F under normal operating conditions, and 388'F af ter reactor d3 pressurization to 200 psig): and a sixteen inch exhaust steam pipe downstream of the RCIC turbine, 40 f t long, with two inches of insulation, with pipe smaperature at 250*F because steam pressure dswastream of the turbine is held at 25 psia.
I o Water Pipes - two mainsulated water pipes, one suction pipe and the other discharge pipe, with dimensions of 8"I 38 ft and 6"I 36 f t, I carry water from the water source and inject it into the reactor. As contioned previously, the water source may be either the CST or the s ppression pool, thus the water temperature may vary from the CST temperature of 90'F sp to the suppression pool temperature. Depending cn the RCIC room temperature at a particular time, these water pipes ccy be either heat sources or heat sinks.
o Turbine - the RCIC turbine is insulated. The turbine temperature is taken as the average upstream and downstream steam tem pe r atur es .
Small portions of turbine that are not insulated are not modeled.
o RCIC Pump - the RCIC pump weighs 6600 lba and is not insulated. As in the case of water pipes, the RCIC pump may become a heat sink dapending on the room temperature and the water temperature. ,
E st Einks - The following heat sinks are modeled:
l
~33-l l
g - --w- ---yv -w , - ,ywy ,we-w,,--- ,- w.--- - - - - , ----w,,-c w e -, . . , , . - ---e-- rrww.w.--m.-w--m- .m.-ir-= w-w-wve- =mm=v----e.-ww-- wi-e- rP-v- - - - ---w
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o Concrete Walls, Floor and Colling - the walls are 26 f t tall, with oldths varying from 18 ft to 31 ft. Thicknes se s vary from 1 f t to 3 ft. These structures were conservatively assumed to be insulated on the outer surface.
o Tarbine Base Plate - it weighs 900 lba and is aninsulated.
o Room Cooler - it weighs 2000 lba and is mainsulated.
O As mentioned previously, the water pipes and RCIC pump become heat sinks if the RCIC room temperature is higher than the RCIC water temper a ture.
Analytical Assumptions - The tollowing assumptions were made in the analysis, with Jcstifications for these assumptions given subsequently:
0 Air and steam are uniformly mixed at all times.
- Air behaves like an ideal gas.
- No condensation on structural surf aces.
8 The RCIC room is isolated from the surroundings.
- Esat conduction is one dimensional through structures and walls.
5i200 the period of interest is several hours,Jsteam leaked into the room has sufficient time to diffuse and mix with air, therefore, the uniform mixing assumption is a good approximation. Also, since only low pressures and temperatures are cacenatered, the ideal gas law holds true for air.
Assumptions of no condensation on structural surf aces is conservative because the froe-canvection heat transfer coef ficient used in the absesce of condensa tion is ses11or than the condensing heat transfer coef ficient. Isolating the RCIC room is czother conservatism, because mass and energy are preventd from leaving the room through conduction, convection and radiation. Finally, the one-dimensional heat conduction assnaption is correct except at the corners of the walls, but the impact is s;ogligible.
35B.4
) INPUT PARAME'11RS The following initial conditions and key parameters were used in the analysis:
- Initial room temperature was 90'F.
- Steam leakage rate was 70 lbe/hr.
- '3 4 -
, - , , - , - - - . - -.--...y,,...m.- ,,-,_ - __ ,_, -. ..-.m..,,e._-.__.- ,,,,c_, , _ , ,o ,-m , ., .._ , . _ . . - - - . ,._,._.,-_..m.,_ -
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- No reactor depressurisation for the first 30 instes (as the operator is trying to determine appropriate actions) and the reactor was cooled down at 100'F/hr.
o Temperature of RCIC water was 90'F, which is the technical specification CST temperature, because the RCIC can take section from the suppression pool for only a short period of time and the operator cill switch the sucton back to the CST as the pool approaches 140'F.
n5E8 5
,iRESULTS APO DISCUSSIONS A timeshare computer progran has been developed to carry out the calculations dassribed above.
N RCIC room temperature response following a station blackout is given in Figure IEES -t .
then Q N temperature increases rapidly during the first hour of the transient, the rate of increase levels off sesequently. The room temperature rises to 119'F at cight hours of transient and 122*F at twelve hours of transient.
tSE8-l and issB-3 Figarcsge=ese=9 show the sensitivity resnits at high water temperature and low steam leskas3 rate, re spe ctively. With the water temperature at 140'F, the RCIC room temperature rises to 133'F st twelve hours, while at the steam leakage rate of 10 lbe/hr (which corresponds to new turbir.c gland seal condition) the room temperature scochss only 101*F at twelve hours. The high sensitivity to the steam leakage rate is d:o to the large latent heat of steam which is released upon condensing in the RCIC room. The sensitivity study also indicates that there is no impact of reactor cooldown rate on the RCIC room temperature response.
N obove results indicate that the RCIC room temperature twelve hours following a station blackout to be substantially below the equipment quaification limits of 212'F fer the first six hours and 150'F between six and twelve hours following a station blockett. This shows that proper operation of the RCIC can be maintained for many hsars during a station blackout to provide adequate core cooling.
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