ML19224D613
| ML19224D613 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/30/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19224D609 | List: |
| References | |
| SER-790530-2, NUDOCS 7907130146 | |
| Download: ML19224D613 (3) | |
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION 1
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SAFETY EVALJATICN SY THE OFFICE OF adCLEAR REACTOR REGULATICN SUPPORTING Af'ENDMENT h0. 26 TO LICENSE NO. OPR-16 JERSEY CENTRAL POWER 3 LIGHT CCMPANY OYSTER CREEK NUCLEAR GENERATING STATICN DOC.KET NO. 50-219 Introduction
- iy letter dated May 19, 1979, Jersey Central Dower & Light Comoany (the licensee) requestea an amencment to the Technical Specifications of License No. OPR-16 for the Oyster Creek Nuclear Generating Station.
The amencment would revise the Tecnnical Specifications to extend tne applicacility of the minimum water level safety limit to all modes of operation, and to aad a new safety limit to require tnat two recirculation locos remain ccen curing all modes of operation except with the reactor vessel neac removec.
- iscussion On May 2,1979, during the cerfor~ance of the isolation condenser automatic actuation surveillance test a false reactor hign pressure scran occurred at the Dyster Creek Nuclear Generating Station.
Subsecuenctly, a turbine trio occurred on low load.
This initiates an autcratic transfer of acuer to the startuo transfor ers.
Startuo transfor er SA ;rovides the cower for the A feedwater train, and startuo transfor-ar CR orovides
- c.er for the 3 & C feedwater cumos.
However, 53 was cut of service f:r maintenance so when the main turoine generator tricped ne scher sucoly to the 3 & C feedwater ;umos was lost.
~he 2 feecwater pumo tricped because of the nydraulic transient c'.used by the loss of Pe 3 i C feedwater ;um:s.
'herefore all Pree feedwater :u os were P': Cad.
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e 'Os s
-ecwa e
- rsier. 1' H /e.- fe rec, c 'a-m 1000
- _' :^3'ge t al.es -ve'a
' asec anc 3 ? ' Of
.~e ac ' ach oj:3ss ines aera en.
Tnese fi ve Oftass lines did not 31'
- 3 larce encu:n fl:w Of 4ater feca tne outsice of tne core region, tne annulus, to the core
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region. as a resul t, tre sater nas boiling 3way in the core region faster nan it was ceing returned througn the cycass lines ana :ne uater level above tne core decreased belcw the icw-Icw-Icw level al a rm.
,nen one cf the recircu13 tion 1 cop sumo ciscnarge valves was rectenea
- ne oater ficw from the annulus to the core region was 'arge enougn to
- m;e s3te for the aate-Ociling off in t e
- re regic i so ne uater
'evei iac reasec 30 ve tre ':w ' w-i aw 'evel 3' arm.
ms Ed 7907130/46 Exi sti,g Tecnnical Specification 2.1.D defines tne low-low-low level as afe:j 1"it wnen the eac*:r is in the shutccan concition.
Den a
though the reactor mode switcn was not in :ne shutdcwn mode, tre liceriee and tne i RC have been tmting tne May 2,1979 event as if a safety limit had ceen violated. Therefore, the reactor aas placed in the colc snutdown condition and the licensee and the NRC performed a trorougn evaluation of the minimum water level that occurred during the event to determine if any fuel damage had occurred.
In addition, the NRC conducted an evaluation of the follow up actions crocosed by tne licensee to crevent recurrence.
This license amendment, is one cart of that follow uo action.
~he NRC evaluation of tre event, the condition of the core, and all of tne corrective measures taken to prevent recurrence is being cescriDed in a secarate Safety Evaluation Recort. This evaluation is related only to tne croposed license a:renement.
Evaluation As a result of the analysis of the May 2,1979 event it wac recognized by the licensee anc tne NRC staff that the low-low inw water level safety limit snoula :e acclied and clearly cefinec
- .o include all modes of ooeration ahen the reactor vessel contains fuel.
The basis for tne current tacnnical specifications limit was to assure aceouate
argin f0r removing ceCay heat frCm tne fuel :Uring Ceriods snen One reactor is snutdown and corresoonas to tre lowe<* recctor / esse' aater level that can be monitored.
The basis #0e the safety limit during coeration is to assure adecuate margin of water abova the too o# tre core to prevenc core uncovery during anticicatec transients.
It is considered crudent to have a measurable water level limit for the safety limit for all ode 3 of coeration. Therefore, tne licensee nas reouested tnat the tecnnical soecifications be modified to make the low-low-low water level (a feet 3 incnes above the too of the active fuel zone) a afety limit acclicaole to all modes of oceration including transient nnditions.
~ is cnange nore clearly cefines the safety limit for reactor vessel n
e ter level for all moces of operation, ne Ica-lcw-1:w aater ' eve' limit is not cnanged, tnerefore, tne proposed :nange is acceptacle.
The licensee nas also crocosed to acc a safety 1 mi* as section 2.7 -
anicn requires that during all modes of oceration exceot unen One reactor head is re oved and the reacter is flooded to a level above tne ainsteam nozzles at least *wo (2) recirculation Icoo cumo suction valves and their associated discharge valves will be in the #ull ocen oosition.
This ail! assure that at all accrocriate times the water in the core and in the annuius will be in hydraulic corrunication to creclude recurrence of the 'ay ?,1979 event at Oyster Creek resultirg t
frem cif#erent levels in trese regions.
039 6,4 Q CIJG @v >;s!I!T M N, 369 R0 h
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-ne effects of discnarge valve position nn steady-state water level in tne core region have been evaluated from a hydraulic analysis of tne recirculation lines. This analysis mcdelea tne recirculation line gecmetry with standard fluid mechanic methods. This modeled tne gecmetric pressure loss coefficients wr.ich incloces a f actor fo r
- ce fi xed rotor recirculation pump.
Tne cressure loss coefficient for tne recirculation ;ume has oeen estaclisned from ir. si tu tests.
The other pressure loss coefficients are from standard methods and are adequete. The analysis assumed differe'tial driving heacs cetween annulas and core regions which are witnin tne range of values assumed for overall analyses.
These metnocs were utilized to calculate the natural circulation flow througn one recircul ation Icop.
Tne districution of coolant inventory (Detween annulus and core) has Deen accounted for based on no forced recirculation flow (due to a reactor coolant pump trip on low-1:w water level) and a maximum of one unisolated recirculation loop. The above conaitions will result in the most aaverse distribution of coolant inventory witnin the reactor vessel.
Basec on our evaluation we have concluded tnat tne croposed Technical Specification will assure continual nycraulic communication between the annulus ana tne core during all mo:es of operation including transients, and therefore, is acceptaole.
Environmental Consicerations
.ie have determined that tce amendment does not autnorize a change in effluent tyces or total amounts nor an increase in power level and will not result in any significant environmental imcact. Having made this determination, we nave furtner conclaced tnat the amendment involves in action wnich is insignificant frcm the standcoint of environmental imct and oursuant to 10 CR 551.5(c)(4) tnat an envi-onmental imcact statement or necative declaration and environmental irca:t 3ccraisal ee: c; ce precarec in cannecti:n at :n :ne ssuance
- :r s imencme-:.
= c w:n 369 040 We ave conchced, based cn the considerations discussed above, that:
(1 ) because this 3mendment does not involve a sicnificant increase in the
- rctacility or consecuerces of accidents crevicusly considered and does not involve a significant decrease in a safety -'arcin, the arend-ent does not involve a signi#icant hazards consi
- eration, (I) tnere is reasonable assurance
- nat the realth and sa#ety c tre 0;clic will rot be endangered by ocera*ien e
in the crocosed manner, and (3) sucn activities will be conducted in comoliance wi*n the Co-mission's regulations and :ne issuance of this amencnent will not be inimical to the :c cn defense and 56curi*y or to the health and safety o# the cualic.
3III
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