ML19241B627
| ML19241B627 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/30/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19224D609 | List: |
| References | |
| SER-790530, NUDOCS 7907190641 | |
| Download: ML19241B627 (3) | |
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N SAFETY EVAldATION St THE OFFICE OF tiUCLEAR REACTOR REGULATICN SUPPORTING A" ENC"ENT NO. 36 TO LICENSE NO. DPR-16 JERSEY CENTRAL PO'WER 3 LIGHT COMPANY OYSTEP CREEK NUCLEAR GENERATING STATION DOCKET NO. 50-219 Introcuctior.
By letter dated May 19, 1979, Jersey Central Power & Light Comcany (the licensee) requestec an amendment to the Technical $;ecifications of License No. DPR-16 for the Oyster Creek Nuclear Generating Station. The amencment would revise the Tecnnical Specifications to extand the apolicacility of tne minimum water level safety limit to all modes of oceration, and to acd a new safety limit to require tnat two recirculation laces remain open curing all modes of operation except with the reactor vessel head removed.
Discussion On May 2,1979, during the cerformance of the isolation condenser automatic actuation surveillance test a false reactor high cressure scram occurred at the Oyster Creek Nuclear Generatiag Station. Subsequenctly, a turbine trio occurred on low load. This initiates an automatic transfer of acwer to the startuo transformers.
Startuo transformer SA provides the cower for the A feedwater train, and startue transformar (R orovides ocwer for the 3 & C feedwater comos.
Hcwever, 53 was out cf service for maintenance so when tne main turoine generator tric;ed the cwer succly to the B & C feedwater pumos was lost.
'he A feedwater ;uma tri:ced because Of the hydraulic transient caused by the loss of the 3 L C feedwater ;umas. Therefore all tnree feedwater cumos were tr': ed.
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- . I:na ge /31<es.ve e :' sec anc a'1 :f : e wc ' ch :/ ass ' es nere a:en. These five cycass lines afc nc: til:- 1 large encugn fl:w of vater from tne utsice of tne core region, the annulus, to the core regian. As 3 result, :ne aater nas '-iling away in :ne core region f acter tnan it was aeing returned nrougn :ne cycass lines anc :ne a er level toeve :ne core tecreasec telcw the Icw-Icw-icw level al a rm.
inen one of :ne recirenlation 1000 pumo ciscnarge valves nat reccenec
- ne na ar ficw fr:m :ne annulus to One care recicn was larce encu:n ::
- cce 51 e f r One aate- :cii+n; Of' in - e : re tgi:.c sa':re naier 9ve! i Crelsed 1:0Ve Fe ' w-l:n-l; level 111rm.
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. Exi sti g Technical Specification 2.1.0 defines tne low-law-low level as
- lafe
- j iimit wnen the aact:r is in the shutd:wn concition.
Den
- ncugn the reactor mode switen was not in tne snutdewn mode, the licensee and One NRC have been treating tne May 2,1979 event as if a safety limit had ceen violated. Therefore, the reactor was olaced in the cold shutdown condition and the licensee and tne NRC perfor ed a thorough evaluation of the minimum water level that occurred during tne event to determine if any fuel damage had occurred.
In addition, the NRC conducted an evaluation of the follow uo actions crocosed by One licensee to prevent recurrence.
This license amendment, is one cart of that follow uo action.
The NRC evaluation of the e.ent, the condition of the core, and all of tne corrective measures taken to prevent recurrence is being cescribed in a secarate Safety Evaluation Recort. This evaluation is related only to the oroposed license amendmen*.
Evaluation As a result of tne analysis of the May 2,1979 event it was recognized ty the licensee and the NRC staff that the low-10w low water level safety limit should te accliec and clearly cefined to incluce all modes of operation when the reactor vessel contains fuel. The basis for One current technical specifications limit was to assure aceouate margi1 for removing cecay neat frcm the fuel curing :eriods anen
- ne reactor is shutdown and corresoonas to tne lowee -ecctor vesse' aater level that can be monitored. TP
- basis for the safety 1.mit during coeration is to assure adecuate margin of water above the tor of the care to prevent core uncovery during antici;:ated transients.
It is considered prudent to have a measurable water level limit for the safety limit for all modes of coeration. Therefore, the licensee has recuested tnat the technical s:ecifications be modified to make the low-low-low water level (1 feet 3 inches above the too of tne active fuel zone) a safety limit 3:alica::le to all modes of oceration including transient conditions.
Thi s cn. more clearly cefines tne safety limit for reactor vessel
.ater 1.ve' for all moces of oceration, tne 10a-10w-1:w ua er :evel l imi t i s no t, cnangea, therefore, tne pr0;osed cnange is accectacle.
"e iicensee has also crocosed to add a safety limi t as section 2.1
.2nicn recui es that during all moces of coeration excect anen :ne reacter head is removed and the reactor is #icoded 50 a level above
- re mainsteam ro::!ss at least two (2) recirculation 1000 cumo sucticn valves anc neir associated discnarge valves will be in the #ull ocen
- osition. This will assure that at all accr:criate times the water in tre care anc in the annulus aill be in hydraulic cormunication to v y 2,1979 event at Cyster : reek resulting
- reclude recurrence of One a
- - m di'#eren-level s in rese egions.
485 132
. ~ne effects of discnarge valve position on steady-state water level in tne core region nave seen evaluated from a hycraulic analysis of tne recirculation lines. This analysis mocelec tne recirculation line gecmetry with stancard fluid mechanic methocs. This modeled
- ne geometric pressure loss coefficients wnicn incluces a factor fo r
- ne fi tec rotor recirculation pump.
Tne cressure loss coefficient for the recirculation pumo has been estaciished from in situ tests. The other pressure loss coefficients are frcm standarc methods and are acequate. The analysis assumed differential criving heacs between annulus and core regions which are witnin the range of values assumed for overall analyses. These metnoas were utili:.0c to calculate the natural circulation ficw througn one recircul ation loop.
The districution of coolant inventary (between annulus and core) has been accounted for based on no forced recirculation flow (cue to a reactor coolant pump trip on low-Icw water level) and a maximum of one unisolated recirculation loop. The above conditions will result in the most adverse distribution of coolant inventory within the reactor vessel.
Basec on our evaluation we have concluded tna; the proposed Technical Specification will assure continual hycraulic communication between the annulus anc the core curing all moces of operation including transients, and therefore, is acceptable.
Envi cr. mental "ensiderations_
We nave deteralined that tne amendment does not authorize a change in effluent ty;es or total amounts nor an increase in power level and will not result in any significant environmer.tal imcact. Having made this tetermina tion,
,,a nave further concluded that the amendment involves an action wnich is insignificant from tne stancooint of environmental imcact inc :ursuant ta 10 CR 551.5(c)(4) tnat an envi-;ccental imcact state'"ent or necative decicqtion and environmental inca:: 3 craisal
- ee: c; ce precarea in connec Nn ~1:n tne 'ssuance s tris amencmert.
.; : ;si:n ae ave c:ncluced, based On the : nsider3tions discussed above, that:
(l' because tnis amencrent does not invcive a significant increase in the rc:atil;;y or consecueaces of acciients creviously consicerec and does not involve a significan: :ecrease in 3 safety argin, the a"'encment dces not involve a signi#icant nazarcs consideration. (2') nere is reasonacle assurance na :ne 9eal:N anc sa'ety O' tre :uciic mil rot be encangered by Ocerat':n in the :roccsed mance, and /3,' sucn act v i i ies ail' be conductec in ccmcliance t
wit' :ne ~: missi n's regula:icns ar: :ne issuance of :nis amenc en: will not te i--M cai :: tre :: On :e'ense anc secur ty cr :: ne "ealtn and sa#ety i
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NUCLEAR REGULATORY COMMISSION
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"ay 20, 1979 Occket No. 50-219 Mr. I. R. Finfreck, Jr.
Vice Presicent - Generation Jersey Central Power & Light Company Madison Avenue at Punch Bowl Road Morristown, New Jersey 07960
Dear Mr. Finfrock:
The Ccamission has issued the enclosed Amendment No. 36 to Provisional Operating License No. JPR-16 for the Oyster Creek Nuclear Generating S ta ti on. This amencment consists of changes to the Technical Specifications in response to your application dated May 19, 1979.
The amendment modifies Section 2.1.D to extend the applicaoility of the minimum water level safety limit to all modes of operation, ano add a new safety limit in Section 2.1.F to require that two recirculation loops remain open during all modes of cperation except with the reactor vessel head removed.
Copies of our related Safety Evaluation and the Notice of Issuance are also enciesed.
Sincerely,
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Dennis L. Ziemann,L.hief v
Operating Reactors 3 ranch =2 Division of Operating Reactors
Enclosures:
1.
Amencaent No. 26 to DPR-16 2.
Safety Evaluation 2.
Notice of Issuance cc w/encicsures:
See next pa9e 8
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Mr. I. R. Finfrock, Jr. May 30,1979 CC G. F. Trowbridge, Esquire Gene Fisher Shaw, Pittman, Potts and Trowbridge Bureau Chief 1800 M Street, N. W.
Bureau of Radiation Protection Washington, D. C.
20036 380 Scotts Road Trenton, New Jersey 08623 GPU Service Corporation ATTN:
Mr. E. G. Wallace Mark L. First Licensing Manager Deputy Attorney General 260 Cherry Hill Road State of New Jersey Parsippany, New Jersey 07054 Department of Law and Public Safety Environmental Protection Section Anthony Z. Ro t sman 36 West State Street Natural Resources Defense Council Trenton, New Jersey 08625 917 15tn Street, N. W.
Washington, D. C.
20005 Joseph T. Carroll, Jr.
Plant Superintendent Steven P. Russo, Esquire Oyster Creek Nuclear Generating 248 Washington Street Station P. O. Box 1060 P. O. Box 388 Toms River, New Jersey 08753 Forked River, New Jersey 08731 Joseph W. Ferrar0, Jr., Esquire Director, Technical Assessment Deputy Attorney General Division State of New Jersey Of fica of Radiation Programs Department of Law and Public Safety (AW-459) 1100 Raymond Boulevard U. S. Environmental Protection Newark, New Jersey 07012 Agency Crystal Mall 72 Ocean County Liorary Arlington, Virginia 20460 3 rick Township Branch 401 Chamcers Sricge Road U. S. Environmental Protection Srick Town, New Jersay C8723 Agency Region II Office Mayor ATIN:
EIS CCCROINATCR Lacey Townshio 26 Federal Pla:a P. C. Box 475 New York, New York 10007 rorked River, New Jersey 08731
- Ccnmi ssioner Je:artment of Public Utilities State Of New Jersey 101 : rmerce Street Newark, New Jersey 071C2
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JERSEY CENTRAL POWER & LIGHT CCMPANY C0CKET NO. 50-219 OYSTER CREEK NUCLEAR GENERATING STATION, UNIT NO. 1 AMENCMENT TO PROVISICNAL OPERATING LICENSE Amendment No. 36 License No. OPR-16 1.
The Nuclear Regulatory Ccmission (the Ccmission) has found that:
A.
The acclication for amendment by Jersey Central Power & Light Ccmcany (the licensee) dated May 19, 1979, complies with the standards and recuirements of the Atomic Energy Act of 19E4, as arended (the Act), and the Comission's rules and reculations set forth in 10 CFR Chaoter I 3.
The facility will coera*e in conformity with the acclication, One crovisions of the Act, and the rules and reculations of the Ccmission; C.
There is reasonable assurance fi) that the activities authorized by nis omensment can be concucted without endangerinn ne health and safety of the cublic, and (ii) that sucn activites will be conducted in con.cliance with the Ccmission's reculations; D.
The issuance of t.1is amencment will not be inimical to the carron defense and security or to the heal th and safety of' :ne cublic; and E.
~he issuance af this amencment is in accordance with 10 CFR Par
- 51 of the Commission's reculations anc all acclicable recuirrents have been satisfied.
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Accordingly, the license is amended by changes to the Technical Scecifications as indicated in the attachment to this license amendment and :aragraph 3.3 of Provisional Ocerating License flo.
CPR-16 is hereby amended to read as follows:
3.
Technical Scecifications The Technical Scecifications contained in Accendix A, as revised through Amendment tio. 36, are hereby incor orated in the license. The licensee shall ocerate the facility in accordance with the Technical Scecifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE fwCLEAR REGULATCRY COMMISSICN J
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j Dennis L. Ziemann, Chief Operating Reactors Branch #2 Division of Operating Reactors
Attachment:
Changes to the Technical S:ecifications Date of Issuance: May 30, 1979
- 485 125
ATTACHMEilT TO LICENSE AMENCMENT NO. 36 PROV!SIONAL OPERATING LICENSE NO. OPR-16 DOCKET NO. 50-219 Revise Ao::endix A Technical Suecifications by removinc the cages identified below and insertina the enclosed pages.
The revised caces are identified by the cactioned amendment number and contain vertical lines indicating the areas of change.
REMOVE INSERT 2.1 -2 2.1-2 2.1-23 2.1 -J 2.1-4 2.1 Ja 485 126
, 1.,
D.
During all modes of reactor operation with irradiated l
fuel in the reactor vessel, the water level shall not be less than 4 '-S" above the top of the normal active fuel :ene.
E.
The existence of a minimum critical power ratio (MCPR) less than 1.32 for 7 x 7 fuel and 1.34 for 3 x S fuel shall constitute violation of the fuel claddi:. integrity safety limit.
F.
During all medes of operation except when the reactor head is off and the reactor is flooded to a level abcve the main stes: noccles, at least two (2) recirculation loop suction valves and their associated discharge valves will be in the full open position.
a Bases:
The fuel cladding represents one of the primary physical barriers which separate radioactive material frc: the environs. The integrity of this cladding barrier is related to its relative freede frca perforations or cracking. Although sc=e corrosion or use-related cracking =ay occur during the life of the cladding,
- ssicn product migration. :ro :gis source is incremental 3.y cumulative, continucusly measurable and tolerable. Fuel cladding perfora icns, hcwever, could result from thermal effects if reactor operation is significantly above design conditions and the associated protection system se: point. While fission product migration frca cladding perforation is just as measurable as that from use-related cracking, the thermally-caused cladding perferations signal a threshold, beyond which still greater thermal conditions may cause gross rather than incremental cladding deterioration.
Therefore, the fuel cladding safety limit is defined in terms of the reactor operating conditions which may result in cladding p e-forat io n.
A critical heat flux occurrence results in a decrease in hea:
transferred frc the clac and, therefore, high clad :e:peratures and the possibility of clad failure. However, the existence of a critical heat flux occurrence is not a directly cbservable parameter in an operating reac:cr.
Further:cre, the critical heat flux correlation data which relates observable carameters c the critical heat flux :sgnitude is statistical in 'ature.
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2.1-2a The cargin to boiling transition is calculated fre: plant operating parameters such as core pressure, core flow, feedwater temperature, core power, and core power distribution. The cargin for each fuel assenbly is characteri:ed by the critical power ratio (CPR) which is the ratio of the bundle power which would produce onset of transition boiling divided by the actual bundle power. The minimus value of this ratio for any bundle in the core is the minimum critica' pcwer ratio (3'CPR) (10).
The safety lini: curves shewn in Figure 2.1.1 represent conditions which assure with better than 95 percent confidence a 95 percent probabili y of avoid:ng a critical heat flux occurrence. The cri:ical pcwer value was determined using the design basis critical pcwer correlation given in Reference 1.
The operating range with
- CPR >1.32 for 7 x ' fuel and 1.34 for S x 3 fuel is below and to the right of these curves.
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f 2.1-4 The range in pressure used for Specification 2.1.A in the esicu-lation of the fuel cladding integrity safety limit is frc: 600 to 1250 psia. Specification 2.1.3 provides a requirement on power level when operating below 600 psia or 104 flew.
In general, Specification 2.1.3 will only be applicable during startup or shutdown of the plant. A review of all the applicable icw pressure and low ficw data (6,7) has shown the lowes: data point for transition boiling to have a heat flux of 144,000 STU/hr-f 2 To insure applicability to the BWR fuel rod geccetry, and provide a =argin, a fac:c; of one-half was used, giving a critical heat flux of 72,000 STU/hr-ft2 This is ecuivalent to a core average power of 354 Mit (18.3% si rat ed). This value is applicable to ambient pressure and no flow conditions.
For any greater pressure or flew conditions, there is increased margin.
During transient operation, the heat flux (thermal power-to-water) would lag behind the neutron flux due to the inheren hes: transfer time constant of the fuel of S-9 seconds. Also, the limiting safety system scra settings are at values which will not allow the reactor to be operated above the safety lini during normal operation or during other plant operating situatiers which have been analyzed in detail (2,3,4,S,9,10).
If the scra: occurs such tha: the neutron flux dwell time above the limiting safety system setting is less than 1.75 seconds, the safety liti will not be exceeded for normal trubine er generator trips, which are the most severe normal operating tran-sients expect ed.
Following a turbine or generator trip, if it is determined that the bypass systes calfunctioned, analysis of plant data will be ased to ascertain if the safety limit has been exceeded, according to Specification 2.1.A.
The dwell time of 1.75 seconds in Specification 2.1.C provides increased cargin for less severe power transients.
Should a pcwer transien: occur, the eves: recorder would show the ti=e interval the neutron flux is over 1:s scra: setting. When the even recorder is cut of service, a safety lici: viola icn will be assumed if the neutron flux exceeds the scra: setting and aa-si red scran does not occur. The event recorder shall be rete::ed to an operable condition as soon as practical.
If reac:c: water level shculd drop below the op of the setive fuel, the acility Oc c0cl the core is reduced. This redu::icn in core ecoling capability :culd lead to elevated cladding temperatures and clad perfora:icn. With a water level above :ne ::p of -he a:-ize fuel, adecu' a - cling is maintained and the decay heat can easily be ace = cdated.
'he 1:wes: poin: 2: which the water level can presently be cni: red is l'-3" Ibove -he
- p Of the 10:ive fuel.
Although the 1:wes-re2c::: w1:er level 1: 1: vat:n ensures acec.ua:e core :: clin; is the
- p of :9.e acti te fue'., -he safe y lini-has been established at d'-5'
- OTOtide a Ocin' which can be OCnitored.
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F 2.1-Ja Specification ? assures that an adequate flow path exists frera the annular space, between the pressure vessel wall and the core shroud, to the core region. This provides for good co=unication between these areas, thus assuring that rea :cr water level ins tra:nent readings are traly indicative of the water level in the core regicn.
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7590-01 UNITED STATES NUCLEAR REGULATCRY COM5'!SSION DCCKET NO. 50-219 JERSEY CENTRAL PohER & LIGHT C0b9ANY NOTICE CF ISSUANCE OF AMENDMENT TO PROVISIONAL CPERATING LICENSE The U. S. Nuclear Reculc. tory Commission (the Ccmmission) has issued Amendment No. 36 to Prnvisional Operating License No. DPR-16, issued to Jersey Cantral Pcwer & Light Company (the licensee), which revised the Techaical Specifications for operation of the Oyster Creek Nuclear Generating Station (the facility) located ia Ocean County, New Jersey.
The amendment is effective as of its date of issuance.
The amencment modifies Section 2.1.0 to extend the apolicability of the min 1 mum water level safety limit to all mcdes of coeration, and add a new safety limit in Section 2.1.F to recuire that two recirculation loops remain acen during all modes of operation except with the reactor vessel head removed.
In secarate actions relating to this facility, the Ccmmission is:
(1) Amending License No. OPR-16 to allow oceration witn the nore restrictive uaximum Average Planar Linear Heat Generaticn Rate limits authorized by Amencment No. 30, dated March 14, 1973, extended to encomcass higher ex::csure fuel in the reactor; and (2) Authorizing resumotion of cceration after a May 2,1979, transient event invciving li:ense safety limits.
485 134
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7590-01
. The application for amencment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules anc regulations. The Commission has made approcriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amencment. Prior puolic notice of this amendment was not required since the amencment does not involve a significant hazards consideration.
9 The Commission has determined that the issuance cf this amendment will not result in any significant environmental impact and that pursuant to 10 CFR ! Sl.5(d)(4) an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with issuance of this amencment.
For further details with respect to this action, see (1) the application for amencment dated May 19, 1979, (2) Amencment No. 36 to License No. DPR-16, and (3) the Ccmnission's related Safety Evaluation.
All af these items and the Commission's secarate actions described above are available for cublic inscection at the Comnission's Public Cocument Reca,1717 H Street, N. W., Washington, D. C., and at the Ocean Ccunty Library, Erick Townshio 3 ranch, 201 Cham ers Bridge Road, 3 rick Tcwn, New Jersey Ca/23. A cecy of items (2) and (3) in addition : the secarate acticns may be obtained uccn recdes: addressed to tne 'J.
S.
Nuclear Regulatory Cornission, Washineton, D. C.
2C555, 20:en:icn:
2f rec:cr, Division cf :eratino Reactors.
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7590-01 Dated at Bethesda, Maryland, this 30th day of May,1979.
FOR THE NUCLEAR REGULATCRY COMMISSION
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DennisL.Ziemann'/cnter Operating Reactors Branch #2 Division of Operating Reactors d
e 485 136
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