ML19290A374
| ML19290A374 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/06/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19290A373 | List: |
| References | |
| NUDOCS 7911060613 | |
| Download: ML19290A374 (16) | |
Text
.s UruTED STATES NL' CLEAR REGULATORY COMMISSION METROPOLITAN EDISON COMPMiY JERSEY CENTRAL POER AND LIG;IT COMPANY PENNSYLVA' IIA ELECTRIC CCMPX;Y DOCKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATICN, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 7 License No. DPR-50 1.
The Nuclear Regulatory Co= mission (the Cocmission) has found that:
A.
The application for amendment by Metropolitan Edison Company, Jersey Central Power and Light Company, and Pennsylvania Electric Co=pany (the licensees) dated April 16, 1975, and supplement dated May 29, 1975, comply with the standards and require =ents of the Atomic Energy Act of 1954, as a= ended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;
^
B.
The facility will operate in conforuity with the application, the provisiens of the Act, and the rules I
and regulations of the Cocmission,;
5 C.
There is reasonable assurance (i) that the activities authorized by this amend =ent can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Cocmission's regulations; and D.
The issuance of this a=endment will not be inimical to the coc=on defense and security or to the health and safety of the public.
2.
Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.c.(2) of Facility License No. DPR-50 is hereby amended to read as follows:
1556 190 g ooo 6/3
"(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised, are hereby incorporated in the license. The licensees shall oporate the facility in accordance with the Technical Specifications, as revised by issued changes thereto through Change No.
7."
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COSMISSION m
L A. Giambusso, Director Division of Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Change No. 7 to the Technical Specifications Date of Issuance:
JUN 6 G75 1556 i91 4
J'
JUN 6 55 ATTAC1 DENT TO LICENSE AME?DMD;T NO. 7 CHANGE NO. 7 TO THE TECICTICAL SPECIFICATIONS FACILITY OPERATING LICE!SE NO. DPR-50 DOCKET No. 50-289 Replace pages 3 3-6, 3 3-8, 3-35, and 3-35a 36 with the attached revised pages.
(No change has been made on pages 3-5 and 3-8.)
Replace figures 3.5-2A - 3.5-2B, 3.5-2C - 3.5-2D and 3.5-2E with the attached revised figures. Add figure 3.5-2F.
t 1556 192 I'
z 0'
i
The pressure li:it li en Tigare 3.1-1 his 'ocen sel t$d such that the reactor vessel stress resultint, fis: i~.t :n21 pressure vill not exceed 15 percent yield strength censide-ic; the follosing a.
A 25 psi errcr in =eas=ed pressure b.
Systen pressure is =etsured in either loop P.axi um differe: tie.1 press =e betvcen the point of syste= press.ure c.
measure =ent :i reactor vessel inlet for all operating pu p cocbinations For adequate conservatis=. in lieu cf crtions of the Fracture Tcushness Testinc ?tq.ir - ents cf ths pro;: sed.:.;_:eniin C to 10. CF? 30, a maxinun pressure of 550 psic a.:1 a :_xi
'--a:up rite of 50 7 in any ene nee has l5 been imposed belev 275, F as sncvn en Tigre 3.1-1.
The spray to:perature diffsrence restrictisn. based on a strens cr.r_ lysis of the sprcy line no::lc is is;csed to sli..tain the ther=al stresses v. the pressu-izer spray line no::le ':elev -he desica licit.
Te perature require-ments for the stec:: genercior corres;:..d th the seasured I.LM for the shell.
- (2)
AS:C Boiler a.:1 Pressure Cede,Section III, I; kl5 (3)
TSAR, Section L.3.10.5 (h)
PJAR, Section b.3.3 (5)
P3AR, Section h.k.5 (6)
TSAR, Sections h.1.2.8 c.d k.3.3
?0DR ORGINAL 1556 193
~
I 3-5 gee.
e,s 4
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i 3.1.3 MINI.%, CC::DITIO:iS FOR CRITICALI'I'l Appli c ab ili ty Applies to reactor coolant syste:s conditions required prior to criticality.
Objective a.
To limit the =c.gnitude of any pcVer excursions resu' ting frc= recctivity insertion due to coderator pressure and moderator te:Perature coefficients, b.
To assure that the reac or ecolant systen vill not go solid in the event of a rod withdraval or startup accident.
Socc i fient ion 3.1.3.1 Tne reactor coola. temperature she.11 be abeve 525 F except for portiens of low psver physics testing when the require-ments of Specification 3.19 shall apply.
3.1.3.2 Ecactor ecolant tenperature shall be above tcT.+10 F.
3.1. 3. 3 When the reae cr coolant tenperr.tu e is belov the tir.itu:.
tc:perature specified in 31.3.1 above, except fcr poMicns of Icv pcVer physics testing vnen the require ents of Specification 3.1.9 shall apply, the rer ctor shall be sub-critical by an neount equsl to or creater than the c tculated,I reactivity insertica due to depressurization.
3.1. 3. 4
'Ihe reactor shall be esintained suberitical by at lec.st one percent l}./h until a sten bubble is fo:-.ed and en indicated water level betvcen c0 and 365 inches is established in the pressuri:cr.
3.1.35 Safety rod groups shall be fully withdrawn prior to any other reduction in shutdevn targin by debcration or regulating red withdre. val during the approach to criticality with the folleving exceptions:
(a)
Inoperable red per 3.5.2.2.
~
(b ), physics testing per 3.1 9 (c)
Shutdovn cargin ecy not be reduced belev 1% Ak/k
- per 3.$.2.1.
(d)
Exercising r'ods per k.1.2.
Tollowing safety rod vitihdrawal, the regulating rods shall be positioned within their position limits as Jefined by specification 3.5.2.5 prior to deboration.
7 P]OROR8fa 1556 m 0
Bases At the beginning of life of _.te initial fuel cycle, the =oderator te=perature with the operating configurati:= of control rods.(1)perating tc=peratures coefficient is expected to be slightly positive at o Calculatiens show that above 525 F the positive ::derator coefficient is acceptable.
Since the =oderator te=perat ce coefficient at lever te=peratures' vill be less negative or =cre positive than at operating te=perature.(2) startup and operation of the reacter vien reactor coolant te=perature is less that 525 F is prchibited exce ; where necessary for lcv pcVer physics tests.
The potential reactivity ins e -ion due to the moderator pressure coeffi-cient(2) that could result fr:: depressurizing the coolant fro 2100 psia to saturation pressure of 9C psia is approxi=ately 0.1 percent t.k/k.
During physics tests, specia-Operating precautions vill be taken.
In addition, the strong negative.cppler coefficient (1) and the sna11 integrated dk/k would lini- '-- ~gnitude of a power excursion resulting fro = a reduction of =oderater density.
The require =ent that the rea:: r is not to be =ade critical belev C*T
+10 F provides increased ass esnees that the proper relationship between primary ecolant pressure a.i temperatures vill be =aintained relative to the NDTI of the pri=ary cocla system. Heatup to this te=perature vill
. be acec=plished by operating.he reactor coolant pu=ps.
If the shutdevn =argin required by Specification 3.5.2 is =aintained, there is no possibility of an acci:ie=tal criticality as a result of a decrease of coolant pressure.
The require =ent that the safety rod groups be fully withdrawn before criti-cality ensures shutdown capalility during startup. This does not prohibit rod latch confirmation, i.e.. vithdrawal by group to a =aximu= of 3 inches
, withdrawn of all seven groups prior to safety rod withdraval.
7 The require =ent for regulati=; rods being within their rod position lir.its ensures that the shutdown =argin and ejected rod criteria at hot zero power are not violated.
REFERENCES (1)
FSAR, Section 3 (2)
FSAR, Section 3.2.2.1.4
Arn1'i eab ility
. Applies to the =axi=u= reacter coolant syste= activity permitted during reacter operation.
Objective To li=it the whole body dose at the site boundary in the event of a double ended rupture of a stea= generater tube.
Srecification 3.1.L.1 The total activity of the reacter ecclant due to nuclides with half lives lenger than 30 =inutes shall not exceed 130/E =icrc-curies per =1 whenever the reae cr is critical. E is the average (=ean) beta plus the average (=ean ga==a energies per disintegration, in MeV, veighted in prcpertien te the =easured activity of the readienuclids in reactor cociant sampled.
Bases The above specification is based en li=iting the consequences of a pcstulated accident invciving the dedble-ended rupture of a stea= generater tube.
The rupture of a steam generater tube enables reacter coolant and its associated activity to enter the secondary system whera volatile ise cpes ceuld be dis-charged to the at=csphere thrcugh condenser s acuum pumps and txrcugh stea=
safety talves (which =ay lift =c=entarily). Since the major portion of the activity entering the secendary system is due to neble gases, the bulk of the aqtivity would be discharged to the at=osphere.
The activity release con-tinues until the operator stops the leakage by recucing the reactor ecolant system pressure below the set point of the stean safety valves and isolates the faulty steam generater.
The operater can identify a faulty stea= generater by using the off-gas =enitors en the condenser vacuum pu=p lines ; thus he can isclate the Caulty steam generater within 3L =inutes after the tube break oc curred.
During that 3h =inute period, a =axi=u: of 27c3 ft3 of het reacter ceolant leaked into the secondary syste=.
'The centrolling dose for the steam' generate'r ' tube rupture acciden't is'the" vhcle-bcdy dose resulting frc= i==ersion in the cloud of released activity.
To ensure that the public is adequately protected, the specific activity of the reacter coolant vill be limited to a value which vill ensure that the whole-body dese at the site boundary vill not exceed 0.5 re= should a stea=
generator tube rupture acciden,t cecur.
Although only velatile isotcpes will be released fro = the secondary syste=,
the following whole-body dose calculatien conservatively assu=es that all of the radicactivity which enters the secendary system with the reactor ecclane is released to the at=osphere. 3cth the beta and ga=na radiatien frc= these isetcpes contribute to the whole-body dose. The grnna dose is dependent en the finite size and eenfiguratien cf the cicud.
McVever., the analysis e= ploys the simple =cdel of a semi-infinite elcud, which gives an upper li=it to the pot ntial gn==a dese.
The semi-infinite elcud mode is applicable to the beta dose because of the shcrt range of beta radiation in air.
It is further assumed that =etecrolegical ccnditiens during the course of the accident cor-respond to Pasquill ?j7e F and 1 =eter per second vind speed, resulting in a X/Q value of 2 x 10-' sec/=3 l
3525 Centrol rod positions:
a.
Operating red group overlap shall not exceed 25 percent,
+ 5 percent, between two sequential groups except for physics tests, b.
Except for physics tests or exercising centrol reds, the centrcl red insertion /vithdrawal li=its are specified on Figures 3.5-2A (for up to the centrol rod interchange),
Figure 3.5-23 (frc= centrol red interchange up to LLO full 7
power days of operatien), Figure 3.5-2C ( for after LLC full pcVer days of cperation) for fcur pu=p cperatien, and Figure 3.5-2D for three or two pu=p cperatien.
If the centrol red positicn limits are exceeded, ecrrective =easures shall be taken i==ediately to achieve an acceptable centrol red positica.
Acceptable centrol red pcsitions chall be attained within four hours.
c.
Except for physics tests, power shall not be increased above the pcVer level cutoff (See Figures 3.5-2A, 3 5-23 7
and 3 5-2C) unless the xenen reactivity is within 10 percent of the equilibrit value for operatien at rated power and asy=ptotically approaching stability.
d.
Core i=balan~ce shall be =cnitored on a =ini=u= frequency of once every tvc hcurs during pcVer operatien abcve 40 percent of rated power.
Corrective =easures (reducticn of i= balance by AFSR :ove=ents and/cr reduction in reactor pover) shall be taken to maintain cperation within the envelope defined by Figure 3 5-2E.
If the i= balance is not within 7
the envelope defined by Figure 3 5-2E, corrective =easures shall be taken to achieve an acceptable i= balance.
If an acceptable L: balance is net achieved within fcur hours, reactor power shall be reduced until L= balance li=1ts are
=et.
N e.
Safety rod limits are given in 3 1 3 5 3 5 2.v The centrol red drive patch panels shall be locked at all ti=es with limited access to be authorized by the superintendent.
3.5.2 7 A power map shall be taken to verify the expected pcVer distri-butien at periodic intervals of approxi=ately 10 full pcver days using the incore instru=entation detection syste=.
Bases The power-i= balance envelope defined in Figure 3 5-2E is based on IcCA analyses which have defined the =axi=u= linear heat rate (see Figure 3.5-2F) such that 7
the maxi =u clad tenperature vill not exceed the Interi: Acceptance Critaria.
Operation cutside of the power L: balance envelope alene does not consitu e a situation that vould cause the Final Acceptance Criteria to be exceeded should l7 a LOCA occur. The power i= balance envelope represents the boundary of operation 1556 197 j
j
limited by the Final Acceptance Criteria enly if the centrol rods are at the vithdrawal/insertien li=its as defined by Figures 3.5-2A, 3.5-23, 3.5-20, and 3 5-2D and if a L percent quadrant pcver tilt exists. Additional ecnservatis: is l7 introducted by application of:
a.
Nuclear uncertainty factors b.
Ther=al calibratien uncertainty c.
Fuel densificaticn effects d.
Hot rod manufacturing tolerance factors.
POORORSIN,0 1556 198 t
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3-35a
'Ite JU percent overlap cetween successive centr:1 rce groups is alleved since Ocntrol the_ verth of a red is lever at the up;er and icver part cf the streke.
rods are arranged in groups or banks defined as follows:
Group Ftnetien_
1 Safety 2
Safety 3
Safe ty 4
Safety 5
Re gulating 6
Re gulating 7
Xenon translent override 8
APSR (axial pcVer shaping bank)
Centrol red groups are withdrawn in sequence beginning with grcup 1.
Groups 5, 6 and 7 are overlapped 25 percent. The ner:a1 positien at pcver is for groups 6 and 7 to be partially inserted.
The =inimu: available red verth prevides fer achieving hot shutdevn by reacter trip at any ti=e assu=ing, the hi6 est worth centrol red re=ains in the full h
out positien(l).
Inserted rod groups during power operation vill not contain single rod worths greater than 0.65 percent ok/k. This value has been shown to be safe by the safety analysis of the hypothetical rod ejection accident (2).
Single inserted control rod worth of 1.0 percent ak/k at beginning of life, hot, :ero pcver would result in lower transient peak thermal power, and therefore, less severe f7 environ = ental consequences as a 0.65 percent ak/k ejected rod worth at rated power.
The plant cc=puter vill scan for tilt and imbalance and vill satis ^/ the tech-nical specification requirements. If the corputer is cut of service, then canual calculatien for tilt above 15 percent pcver and imbalance above 40 percent pcver cust be perfor=ed at least every two hours until the computer is returned to service.
The quadrant pcVer tilt limits set forth in Specification 3 5.2.h have been
. established within. the-ther=al analysis-design base using the defirdtion c '
quadrant pcVer tilt given in Technical Specificaticns, Section 1.6.
During the physics testing pregra=, the high flux trip setpcints are ad inis-tratively set as follevs to assure an additional safety =argin is provided:
Trie Set;oint Test Pcver L
1 5%
ko 50%
>C 60%
75 85%
>75 105.5%
Ru :..~EICES 1556 199 (1) FSAR, Section 3 2.2.1.2 (2) FS AR, Section 14.2.2.2 3-36
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Red inces is the percentage se*,of the withdra,wal of the operating greses.
2.
The additional restractions en vitadrawal (hannes areas) are in effect after tne centrol red interchange. The restrictions so witnera.al are fortner seeifies af ter eso f.li ee.er da, of eseratse. (see figure 3.sr2:
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2.
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j 1556 202
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1556 203-
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POWER LEVEL, 5 RESTRICTE0 102 REGION
-20.4
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100 80
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PERNISSIBLE OPERATING REGION
( _
_20
-40 20 0
+20
+40 Core Imbalance, 5 Ib 1556 204 OPERATIONAL POWER !MBALANCE ENVELOPE e
(
THREE MILE ISLAND NL' CLEAR STATION L?ilT 1 FIGUR E 3.5-2 I 7
20 18 E5 l
cc 16
_n E
E=
14 12 0
2 4
6 8
10 12 Axial 1.acation f rom Bottom Of Core,Ft.
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O 1556 205 LOCA LIMITED MAXtMUM ALLCWABLE LlHEAR HEAT RAT E THREE MILE ISLAND NUCLEAR STAT 10N L?iti :
FIGURE 3.5 2F l 7 i