ML19220C072
| ML19220C072 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/06/1976 |
| From: | Ross D Office of Nuclear Reactor Regulation |
| To: | Deyoung R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904280174 | |
| Download: ML19220C072 (24) | |
Text
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0 Docket No. 50 "^-
Richard C. DeYoung, Jr., Acting Assistant Director for LWR's, DPM TEREZ MILZ ISLAND 2 SER INPUT UPDATE Plant Nane: Three Mile Island 2 Docket No.: 50-320 Milestone No.:
24-21 Licensing htage: OL Responsible Branch & Project Leader:
LWR-2, H. Silver Technical Review Branch Involved: Reactor Systens Ersuch Description of Review: SER Input Update Requested Cc=pletion Date: July 22, 1976 Review Status: Conplete The enclosed report contains en update of the evaluation perfor=ed by the Reactor Systens Branch on Three Mile Island 2.
This review included sections 1.5, 4.1, 4.4, 5.1, 5.2.2, 5.3, 5.5, 6.3.1, 6.3.2, 6.3.3, 6.3.4 and Chapter 15 from the Standard Forut Revision 1.
The previous version of the Safety Evaluation Report contained too =any open ite=s for issuance. In the intervening period several issues have been resolved and others introduced. The enclosed report represe'- ts the current status of Reactor Syste=s Branch sections. A su==sry of open issues is as follcws:
1.
Sump test, sunp screen development leading to verification of available NPSH for injection pumps.
2.
Stema line break analysis.
3.
Feedwater line break analysis review by staff.
4.
Additional break analysis for ECCS evaluation.
5.
Preoperational tests of long tern cooling modes controlling boren precipitation. Provision for flow neasurement.
h 790428009
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86 233 6
AUS 3 *575 R. C. DeYoung, Jr. 6.
Overpressure control during startup and shutdown.
7.
Applicatien of Standard Technical Specifications.
This docu:sent replaces the previous Reactor Systens Branch input to this Safety Evaluation Report.
.gInal big lI d 6
.ne D. F. Ross, Jr., Assistant Director for Reactor Safety Division of Systens Safety Office of Nuclear Reactor Regulation
Enclosure:
SER Input Updzite ec:
S. Hanauer R. Hein ean D. Ross K. Kniel H. Silver T. Novak S. Israel J. Watt Central File #
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THREE MILE ISLAND 2 3AFETY EVALUATION REPORT 1.5 Recuire=ents for Further Technieni Infor ation The applicant has identified in Section 1.5 of the Three Mile Island 2 FSAR development programs applicrble to the design.
These programs were initiated to establish the final design.
A "Once Through Steas Generator Test" vas co=pleted and reported in B&'4 Topical Report 3rJ-10027. NRC approval was documented in conjunction with a previous 347 related Safety Evaluation.
Thermal and Hydraulics programs were verified through a 1/6-scale model flow test.
Test data analysis and docu=entation were submitted as 3&W Tooical Report 3NJ-10037 (Rev. 2). NRC approval was documented in conjunction with a previous 35W related Safety Evaluation.
C.
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. 4.0 REAC*CR 4.1 Simrv Descriotion.
The design of the Babcock and Wilcox pressurized water reactor for Three Mile Island 2 is s4 diar to others (Davis-Besse 1 and Rancho Seco) recently approved for operation. The core consists of 17 fuel assenblies having 208 fuel reds each. The design heat output is 2772 MWt.
Full and part length control rods, dissolved boren, and burnable poison rod asse=blies are utilized for reactivity control.
4.4 Thermal and Hvdraulie Des 12n The Thermal-Eydraulic design of the core for the Three Mile Island 2 plant was reviewed. The scope of the review included the design criteria and design basis as i=ple=ented in the fina* core design.
We have evaluated Three Mile Island 2 en the basis of a design pcwer of 2772 MWt vi k 15x15 fuel assenblies. As shown in Table 4.4-1, the ther=al and hydraulic design parameters for Three Mile Island 2 are identical to Rancho Seco and s' diar to Davis-Besse 1.
The principal criterion for the ther al-hydraulic design of a reactor is to prevent fuel rod damage b" providing adequate heat transfer to the various core heat generation patterne occurring during nor=al operations and anticipated transie.ts. The c.pplicant has denenstrated through tr a use of the Westinghouse W-3 correla tion f or departure from nuc1 sate boiling heat flux ratio (DNBR) that a DN3R greater than 1.3 is saintained for steady-state design over power (112")'
and anticipated transient conditions.
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' 3-Table 4.4-1 Ther=al-Hvdraulic Desirn S"-,rv Co=carison of Three Mile Island 2, Davis-Besse 1, and Rancho Seco Three Mile Davis Rancho Island 2 3 esse Seco Design Core Heat Output, MWt 2772 2772 2772 Nominal Systes Pressure, psia 2200 2200 2200 Vessel Coelant Inlet Te:perature, F 557 555.4
'557 Vessel Ccoluit Outlet Te=perature, F 607.7 608.6 607.7 Total Heat T ransfer Surface Area in Core, ft 49734 49734 49734 Average Heat Flux, 3tu/h-ft 185,090 185,090 135,090 Maxi =us Hea*. Flux, 3:u/h-ft 576,885 554,2C0 576,885 Average Ther=al Output, kW/f t
- 6. 105 6.105 6.105 Maximus Design Ster =al Cutput, kW/ft 19.03 18.23 19.0 3 Md m Cladding Surface Te=perature, F 654 654 654 Average Core Fuel Te=perature, ?
1200 1200 1200 Maxi =um Fuel Te perature at Hot Spot, F
4170 4060 4170 0
Total Reactor Coolant Flow, 10 lb/h 137.8 131.32 137.3 Core Average Coolant 7elocity, fps 16.52 15.74 16.52 DNB Ratio at Design Overpower 1.39 1.41 1.29 DNB Ratio at Design Power 1.75 1.79 1.75 e
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The analyses indica:e adequate =argin over.the 1.3 ON3R 11:10 This persi:s the conservative assessment with 95* pr:babili:7, at the 95*:
confidence level, tha: :he het red in the core vil'. act experience a departure f:cm nucleate b'iling condition during nor=al operatics o
or ::ansients that are anticipated to occur with nederate frequency.
The i= balance ::1p se: points are established for the reactor p c-tection system functica to ensure that ther=al design cri:eri:n en DNBR and fuel te=perature li=1:s will not be exceeded. The i: balance 11=1: envelope and the appropria:e design condi:1:ns vill be ;;cvided in the Technical Specifica:icus for Three Mile Island 2.
The analysis presented in the FSAR censidered =azd-"
design condi:icas and nos probable c=nditiens. The for=c: is the nos: severe condition and was used to identify DN3R and fuel temperatures during nor:al and an icipated transients. The latter provides a convenien: cecparisen for :he expec:ed condi:1cus in the core.
Durin8 the Davis-3 esse reviev the staff lad rec.uire.d further cen-sideration of the effect of s:uck cpen internal vent valres en the ther=al hydraulic design and core cooling characteristics. 35W has responded en a generic basis by submitting a report, "2&W Cperating Experience of Reactor Internal 7ent Valves".
~he staff has evalua:ed this report and concluded. hat a fl:v penal:7 due to internal ven:
valve leakage need not be applied.
~he applican: nust, h:vever, i=cle=en: a p cgram of inspecti:n and tast of the vrives a: each re-fueling.
Single-locp opara: ice while the reac:c: is cri:ical is p::hibi:ed until the satisfac : 7 c:=pietion ef a singie-10ep :e2-p gram.
- his l'
':ati:n is Oc be noted in the Techni:al 5:ecifica:i:ns.
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-5 ~.
On the basis of our rcview of :he ene=a;-hydraulic charac: eristics of Three Mile Island 2 and cc=periscas with % ris-Besse 1 and Rancho Seco, we conclude tha: the cher=al hydraulic design of 2 :ee Mile Island 2 is accep:-; ole.
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5. 0-REACTOR CCOLMIT SYSTE'.
5.1 Sumnrv Descriction The Three Mile Island reactor coolant system consists of the reactor vessel, two vertical once-through steam generators, four shaft-sealed reactor coolant pu=ps, an electrically heated pressuri:er, and inter ?onnecting piping.
In most i=portant aspects the systes is s1=11ar to the Davis-3 esse and Rancho Seco syste=s.
The steam generators on the Davis-3 esse plant are physically located higher relative to the rest of the system than in this plant and Rancho Seco. We conclude that the overall design of Three Mile Island 2 is acceptable.
5.2.2 overoressure Protection The pressure relief system prevents overpressuri:ation of the reactor coolant pressure boundary under the cost severe transients and li=its the reactor pressure during nor=al operational transients. Overpressure protection for the reactor coolant pressure boundary is acco:clished by utilizing the two safety valves. These valves discharge :o the pressurizer quench tank through a ec==on header from the pressurizer.
The reactor coolant system (RCS) safety valvas, in conjunction vi:h the 22 steam generator safety valve's, and the reactor protection system, will protect the RCS against overpressure in the event of co=plete loss of hea't sink.
The 3&W topical report on overpressure protection (3AW-10043) identifies th'e flow capaci:7 safety =argin as two for the pressurizer cafety valves and 6% for the steam generstar safety valves.
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7 The staff considers the report to have de=cnstrated an adequate design approach to relief capacity for the pri=ary and secondary systems for the events considered therein.
The peak RCS pressure following the worst transient is limited to the ASME code allcwable (110% of the design pressure) with no credit taken for operation of RCS pressurizer level control system, or pressurizer spray. Four upsets were censidered (3A'4-10043), control rod withdrawal, turbine trip, co=plete loss of power, and loss of feedwater flev. Conservative assu=ptiens were =ade for the analysis of each upset. All *. rip setpoints were asau=ed ac =aximum tolerances.
The control rod withdrawal transient is = ore severe at zero or low power producing a peak pri=ary pressure of about 2665 psig which is 85 psi below the 110 limit.
Turbine trip fres =axi=us overpower conditions produces the = cst severe combined pressure transient on both the. steam system and the reactor coolant system. The steam syste= pressure rising =ccentarily to within approx 1=ately 15 psi of the 110% design pressure.
"eactor vessel overpressure protection during startup and shutdown conditiens has not been addressed in the TMI-2 FSAR. There have been several reported incidents of reactor vessel overpressurization in FWR facilities in which 10 C7R Part 50, Appendix G li=itaticus have been exceeded. The staff has initiated discussiens with 8O i 9' 9
-e-Babcock & Wilcox on a generic basis and also with the applicant relative to overpressure protection during these condit'_ons.
The plant contains features which reduce the probability of such an event occurring. This topic *. rill be developed further in 4 supple-ment to this SER.
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5.3 Ther=al-Hydraulie System Desizn The thernal and hydraulic design bases are discussed in Section 4.4 5.5 Cecoonent and Subsysten Des 12n 5.5.1 Reactor Coolant Pures and Motors The reactor coolant pu=p is designed to provide adequate core eccling flow and hence sufficient heat transfer to =aintain a DN3R of at least 1.3, Sufficient pu=p rotational inertia is provided by the flywheel to provide centinued flew following a loss of,pu=p power such that the reactor neutron power can be reduced before UN311=its are exceeded.
5.5.2 Steam cederator The steam generator is a vertical, straight-tube-and-shell heat exchanger and produces superheated steam which is controlled to
=aintain a constant throttle pressure over the power range.
The pri=ary reactor coolant enters the stern generator he=ispherical head and flcus downward inside the tubes giving up heat to generate steam en the shell side secondary loop.
The tube and tube sheet boundary have the same design pressure and te=perature as the rest of the prisary coolant boundary.
As the steam generators provide the heat sink for the primary reactor coolant system, the system is physically arranged to assure natural circulatien for decay heat removal.
e 86 243
5.5.7 Decay Heat de=cval System The decay heat re= oval systen CpERS) is designed to re=cve decay heat and sensible heat from the RCS and core during the latter stages of cooldown. The system also provides au: ciliary spray to the pressuri:er for co=plete depressuri:ation, =aintains the reactor coolant ta=perature during refueling, and provides the =eans for filling and draining the refueling cavity.
In the event of a 1CCA, the decay heat re= oval pu=ps are used for icw pressure injection of borated water into the reactor vessel for e=ergency core cooling.
The decay heat re= oval systes is placed.into operation approxi=ately six Lours af ter initiation of plant shutdown when the te=perature and pressure of the RCS are below 2SO F and 260 psig, respectively.
Assu=ing that two pu=ps and coolers are in service, and that each cooler is supplied with ec=ponent cooling water at design flow snd te=perature, the DHRS is designed to reduce the RCS ta=perature to 14d ? within 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />. If one of the two pu=ps or one of the two coolers is not operable, safe cooldown of ihe plant is not co= promised; however, the :ime required for cooldown is extended. The applicant has she,c that, assi-4 g only one train is available, the plant can be shut down to belcw 212 F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The decay heat re= oval system for this plant was designed in accordance with Criterion 6 of the'1967 General Design Criteria which called for
"... reliable... decay heat re= oval syste=s...." not specifically recuiring the application of single failura cri:aris. The staff had expressed the concern : hat the singic active failure of either of rao isolation valves would prohibit initiation of the RHR system during a nor=al shutdown.
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s In response to this concern the applicant has provided the following assurances. Assu=ing a nor=al sequence to hot shutdown, the complete inventory of condensate can be used for decay heat removal. This would be a closed cycle = ode using the condensor for the syste=
heat sink. This system could operate for an unlisited ti=e period.
The condensate storate tanks and the desineralired storage tanks 0
contain an additional 1.3 x 10 gallons et water which could be used as =akeup if required or on a once through basis would per=it cooling through the steam generators for approx 1=ately four days. The nuclear service river water system can also be aligned to provide unli=ited river water to the steam generators.
The applicant has esti=ated that the repair of either valve can be accomplished in two shifts. This would include time to remove the valve from the line, perform repairs, inspect the repairs, and test the valves.
The single failure of either valve would delay achieving cold shutdcan but would not interrupt or terminate core cooling. The staff concludes that this is acceptable for this plant.
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/ 5.5.10 Pressuri:er The pressuri:er saintains the RCS pressure during steady-s ate operatica and li:1:s pressure changes during transients. I: contains a water volu=e, si:ed :o pr= vide the ability of the syste_s to experience a reac:or trip and not uncover the icv level sensors in the bot::= head and to nain:ain the pressure high enough so as not to activa:e the high pressure injecti:n system; and a volu=e of steas, sized to provide t.Se abili:7 of the system to eeperience a turbine crip and no: cover the level sensor 1 the upper head.
Electric heater bundles, located in the icver section, and a water spray no::le in the upper section nain:ain the steas and water a:
the saturatica te=perature which corresponds to che desired :2 actor coolant system pressure. During outsurges, as the RCS pressure decreases, sc=e of the water flashes to stea= and the elederic hea:ers restore the nor=al cperating pressure. During insurges, as RCS pressure increases, the va:er spray condenses stean :o reduce the pressure to the nor=al opera:ing level.
"wo ASME cede safety valves are connected to the upper pressurizer head :o relieve systes overpressure. A pilot-cperate.d relief valve is also provided to 14"d: the lifting frequency of the code safety valves. The sa'fety and relief valres discharge to the pressurizer quenen tank, loca:ed w1:hin centainnent.
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. 5.5.13 Safety and Relief valves The pressurizer safety valves are bellows sealed, balanced, spring-leaded safety valves which are provided with a supplenental back-pressure balancing piston for handling a bellows failure. The pressurizer relief valve is an electrically actuated, electrically controlled, pilot operated, pressure leaded relief valve.
5.3.14 Internal vent valves, The eight core support internal vent valves are located on a ec=ncn plane in the upper core support weldnent above the outlet no::les.
These valves provide a direct flow path between the upper core region and the inlet annulus in the event of a loss-of-coolant accident from an inlet (cold leg) line break. This flow path provides for pressure equalization by the venting of steam to the break and per=1ts the e=ergency coolant water to reflood at a higher rate.
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. 6.3 Ene:2ency Core Coolinz Svstem 6.3.1 Desizn Bases The applicant sta:es : hat the principal design basis for providing protection ever the entire spec::u= cf break sizes for the ECCS is General Design C 1:e-1:n 35. Very s=all breaks that do not actuate che engineered safety fea:ures nede of opera:1cn are ace===cdated by the nornal nakeup systa= in confor=ance vi:h General Design Crice:1cn 33.
Separate and independent flew paths are provided in the ICCS.
Redundancy in active ce=penen:s ensures,tha: the required functions will be perfor=ed if a single failure cecurs.
Separate energency.
power scurces are supplied :o :54 redundan: ac:ive c:=penents.
Separate instrenent channels are provided Oc activate :he sys: ens.
The range of coolan: systen ruptures and leaks the ICCS is designed to ni:1 gate extends fres leuks in excesa :f the capability of :he reactor coolant =akeup syste: up to and including a break equivalent in area to a double-ended rupture of the largest pipe in :he reae:::
m coolant sys:em.
6.3.2 Svstem Desi n The energency core cooling system for Three Mile Island Uni: No. 2 censists of the core fleeding tanks, high pressure inj ection system, and a low pressure injectica systes. The high and 1:v pressure injec:1on sys: ens take water fres a borated water s::: age during the injection phase folleving a less-of-coolan: accident.
As the borated 86 248
13 water supply nears deple:icn, the systen must be realigned to a recirculatica =cde taking water from the reactor building su=p.
Be applican: had previously proposed tha: realign =en: be acc:=plished
=anually. Ecwever, in response :o our concerns, the applicant recently co=si::ed to aut==atic opening of the reactor building su=p valves on lov level signal f:cm the berated va:ar storage tank to acce=plish this func:1cn. Se applicant has ec=sitted to satisfying staff concerns rela:ive to i=ple=en:ation. Section 7.6.1 p;cvides discussier. of :he inst:n=ent and centrol aspects.
Adequate N?S'd is to be provided to all ECCS pu=ps in all.cdes of opera-ica. Final staff app;cval that these requirements are satisfied is contingen upon addi:10 al docu=entatica based en preoperational tests and further analysis, ne staff has questioned the respense of the ECCS =c a core fleeding tank line break concurrent with the loss of the diesel povering the unaffected train. Be applicant has provided additicnal analysis, docu=en:atien, and changes in systa=s and precedures :o assure :he the staff cf the folleving:
(1) One high pressure injection pu=p will provide adequa:e core cooling.
(2) Only cne opera::: action is required. 21s occurs af ter the aut==atic switchae of the lcw pressura inj ection pt=cs :o the recirculatica :Me.
Se opera:or has a=ple ti=e to cpen a 'ralve fram the centrol ::cs provi*g the high pressure inj ectica pt=p suction vi:h the 1:v pressure injecti:n pu=o discharge.
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@ (3) In the recircula:ica node, the high pressure injecti n pu=p will have sufficient NFSH available evea though the icv pressure injectics pu=p supplies both the HFI pu=p and the break. Adequate core cooling vill also be retained in the event tha: the operator spli:s the low pressure inj eciten flev between the evo low pressure inj ection lines.
~he 1500 GPM supplied through the unb: ken line vill be adequa:e.
The question of whether the renaining high pressure injection pu=ps vill have adequate NPSE under these conditions has not been resolved. I: is no longer censidered a safety issue.
To '4'4'ize the potencial for water ha=er occurring due to ECCS injectics into dry lines, the applicant has stated that curing nor=al operation the ICCS lines vill be sain:aiaed full.
- he staff requires that the capability := -'d tain filled ECCS piping be observable prior to startup and that the venting provision constitute a periodic surveillance requirenent.
Specificaticus are required relative :o se :ings, lini:3, and surveilletce of ECCS ele =ents. including 3WST and c.lar=s.
We vill assure :h.
tesdcal specifications satisfy the above requirenen:s.
6.3.3 Perfor ance Evaluation
- he applicant has provided an analysis of energencv core cooling systen pericr a-ce for Three Mile Island Uni: No. 2 in accordance vi:h 10 Cy?. 50 paragraph.46.ed Appendix I refer-ing to 3abecek & Wilecx 86 25B\\
_ topical reper: 3X4-10103. The staff generic review of the ::pical report resulted in a reques: for addi:icnal break analysis and require =ents for plan: specific infor=ation relative to =ini=u=
ccatain=ent pressure, single failure analysis, sub=erged valves within contain=ent, and cen:rol of boren precipitati n during long-ter= cooling.
The applica - has satisfied staff require =en:s for plant specific infor=ation relative to centain=ent pressure, single failure analysis, and sub=erged valves. Staff review of the leng-ter= cooling codes resulted in a require =ent for a revisien to provide a positive flow indication in each =cde.
preepera icnal cases to de=enstrate each =ede are also required.
Outstanding issues relative to the ICCS evaluation are cen:ered en the additional break analysis and icng-ter= cooling issnes. The results of the staff reviev vill be provided in a supple =en: to this SER.
6.3.4 Tests and Insceetions The applicant will de=custrate che operability of the ICC3 by subjecting all systa=s and c =penents chereof :s preoperational tests, periodic testing, and in-service testing and inspection. The applican: has ce==1::ed to the i=cle=en:stien of :he bulk of the Regulator 7 Geldes applicable to inicial tes: p cgra=s.
In sc=e cases, the applican has provided acceptable alternates or justifica:icn f:r deviaticus. Relative :o Regula:Ory Guide 1. 79, the applicant has cc.- 1::ed :: perfor= a :::bination Of :asts and
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analysis to desenst:2:e vortex con:rol in the sung and to verify that adequate NPSE is provided to the punps drawing sue:ica fr==
the su.p.
Due to the developnental na:ure of the scaled suno :ests, this progras of tests and analysis vill be reviewed by the staff and reported in a supple =en: :o :his SER.
The staff will revieu :he technical specifica: ices :s dete =ine the adequacy of periodic testing, in-se: tice :esting, and inspectices.
6.3.5 Conclusiens The acceptability of the e=ergency core cooling systes :.a dependen:
upon resclu:1cn of the concerns noted in See:1cer 6.3.2, 6.3.3, and 6.3.4.
'Je vill report en these satters in a supplenen: :o this SER.
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. 15.0 ACCIDE';T ANALYSIS 15.1 General The safety analysis sub=itted evaluates the ability of Three Mile Island 2 to operate without undue ha:ard to the health and safety of the public. The events pertinen: :o safety as discussed herein are divided into two groups; abcor=al transients and postulated accidents.
15.2 Abnormal Transients The criterion, adopted to assure that the reactor coolant pressure boundary integrity is =aintained, is that the systes pressure shall re=ain below the code pressure limits set forth in ASME Code See:icn III (110% of RCS design pressure). The criterien adopted to ensure that no fuel da= age has occurred is that the DN3R sust be greater than 1.30 throughou: the transient.
The applicant has submitted analyses of abnormal transients and has shown that the integrity of the RCS houndary has been naintained and that the min' - DN3R exceeded 1.30 for al3 analyzed transients.
The pressure transient which produced the highest reactor coolant system pressure was identified (3AW-1C043) as the control rod withdrawal at lev powar conditions, resulting in a peak RCS pressure of about 2565 psig. The most severe secondary side pressure transient was the turbine trip f ca overpower conditions, resulting in a maxi =us steam generator pressure of about 1140 psig, 86 253
. The applicant was notified in a July 12, 1976 letter that the Anticipated Transients Without Scram issue is being pursued on a generic basis and schedule independent of the operating license s chedule.
The conputer code "p0WER TRAIN",used for several abnor=al transients in the FSAR, is currently under review by the staff. Should
=odificaticus to the code be required, the effect of these changes on the Three Mile Island 2 analyses =ust be censidered.
The evaluation of abnormal transients indicated that the trcnsients presented do not lead to enacceptable consequences and are acceptable for i.ssuance of the operating license.
15.3 Accidents The applicant has evaluated a broad spectrum of accidents that =1ght resulu from postulated failures of equipment, or other i= proper operation. These highly unlikely accidents have bees analy:ed in detail by the applicant and are repre:entative of the spectrum of types and physical locations of postulated events involvina the various engineered safety feature systa=s.
Loss-of-Coolant Accident Analysis for Three Mile Island Unit 2 were submitted in 3abcock & Wilcox topical report 3AW-10103. The staff generic review resulted in a request for additional break analyses.
Eapplicant had co _.itted to submit the analysis in DeEe=ber 1976.
~
The re's'Eit's of the staff' resiew -iill be presented in a supple =ent to the SER.
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@ The locked rotor accident was analy:ed by postulating an instanzaneous seizure of one reactor coolant punp. Reactor coolant flow decreased rapidly and reactor trip occurred as a result of high power-to-flow signal. The ei-us DN3R (1.3) was exceeded for less than two seconds. Cladding surface tenperature was calculated to reach 1800 7 at the hot spots on.57T of the rods. No fuel cladding failure or significant zirconium-watet.aaction would be expected during such a short period.
If the conservative assu=ption that the 11:1:ed percentage of fuel would fail when the DN3R limit is exceeded, the radiological consequences of this event would be bounded by other events considered.
Various issues have arisen relative to less-of-secondary-coolant accidents (steam line breaks and feedwater line breaks). The appli-cant has agreed to analy:e a spectrum of steam line breaks inside of contain=ent to identify worst single active fai~ures affecting core behavior and contain=ent integrity and to propose system or procedural changes if required. Additional analysis of a feedwater line break presented in response to question S-3-21.49 was recently submit:ed and is currently being re-iewed by the staff.
The adequacy of the analysis and possible changes in the secondary systems will be reported in a supplenent to thd.e Safety Evaluation Feport. Discussions to this point have resulted in a co._-.:: lent to J
auto =ata the closing of nain steam isolation valves on a steam line break signal.
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References 1.
BAW-10043 Overpressure Protection for Babcock & Wilcox Pressurized Water Reactors", May 1972.
2.
BAW-10099, "36W Anticipated Transients Without Scra malysis" December, 1974 3.
" Status Report on enticipated Transients Without Scras for Babcock &
Wilcox Reactor" NRC, December 9, 1975.
4 Letter fres Karl Kniel to R. C. Arnold, November 21, 1975.
5.
Letter fres R. C. Arnold to Karl Kniel, February 26, 1976.
6.
Letter fres R. C. Arnold to Karl n'aiel, June 25, 1976.
7.
Letter from Roger S. Boyd to R. C. Arnold, July 12, 1976.
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