ML19206A772
| ML19206A772 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/19/1978 |
| From: | Silver H, Varga S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19206A761 | List: |
| References | |
| TASK-TF, TASK-TMR NUDOCS 7904210187 | |
| Download: ML19206A772 (4) | |
Text
/
SAFETY E','ALUAT!0ii ?Y THE CFFICE CF hUCLEAR REJ.CTCM REGULATICM
.S..UP.P_d___T.h_.C_ A:D_ US.EN._T _:.U
_._4T3 FACIE _ TTY _19E__XTKCTTCE_NSE ;U. UPR-73 FETRCPCLITAM EDISU: C@PANY O,. ~.,.,d JEiiSEY CENTRAL PU'..En d L'IGHT CC PANY
0d Mii."SYLVANI A uECTRIC__COLMUY UCCKET NO. 50-320 THREE MILE ISLA!;D NUCLEAR STATICN, UNIT 2 1.
Sodiun Hydroxide (fo.CH) Injection Sigrgl, Introduction By letter cated itay 10, 1973 transmitting Technical Specification Cnanr;e Request No. 007, Metropolitan Edison Cenpany (Met Ed) recuested arencrent of Appendix A to Facility 0;erating License No. DPR-73 for Three i'ile Island huclear Station Unit 2 (ThI-2).
The requested change would anend the Technical Specifications to pernit avoiding injection cf NaCH into the reacter coolant systea during inadvertent actuaticns of the energency core ccoling systen (ECCS).
. Discussion HaCH is injected into the Sorated Water Storage Tank discharge line in the event of a LCCA to provide corrosion control and to enhance iodine renoval capability of the reactor building spray systen.
At present, either of two signals open the sodiun hycroxide injection valves:
1600 psig Reactor Coolant System (RCS) pressure or 4 psig reactor building pressure. As result of two recent occurrences where scdiun hydroxide was inadvertently and unnecessarily injected into the prinary system when the RCS pressure went celow 1600 csig, the applicant has proposed nedifyinn the soddun hydroxide injection valve actuation signal to recuire that the valve ce actuated by a decreased level of the Borated Water Storage Tank (3bST) sinultaneously with an RCS pressure below 16C0 psig, or by a 4 psig reacter builcing pressure. This ncdification would allow a pericd cf tire for safety injection without sodiun hydroxide addition for those events where reactor building pressure cces not rise above 4 psig cefore the ShST level initiation, and would allow the operator to nanually prevent hydroxide addition fnr those situations where it was nct recuirec.
The license has presented an analysis which cercluces that this recification will reduce the prceability of spurious MACH injection witbcut dccracine the functional capability of tne systen. Chenistry control for corros cn 7904210187
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,) g. s. aov e nw a=v== inn =a o p ric ai t e7e - eas.a ma prctection w:ll be naintainde, and. the plant resrcase to any accident in which to 4 Ch injection..culu ce nccessary to recuce any offsite dcsas to within acceptacle lhiits, v.culd rcrain unchangcd.
In cddition tc the above. in a letter dated "ay 12, 1978 Fet Ec stated tnat the redundant suitches which generate the new BWST level sinral have apprcpriate accu..cy cnd repeatacility characteristics, tt:at they are qualifiec to scismic Categcry I rcquirenents, and that the reduncant signal caoles are "nuted in separate safety related acd saismicall-/
cualifico raceways.
..et Ed further comitted to uccating tne F3M in a future attendtent to reflect these changes.
Eval uat_t en.
he have reviewed the information provided by tFe licensee, and find tnat the propcsed change is cestrable in that it will reduce the probability of unnecessary injectico of NaOH into the reactor colant syste.m. end that the prepcsed change will nut cegrade the functienal capability of the systcm. We further find that the conccnents provided are appropriate and redundant and conform with seismic Cateccry I recuircrcnts.
Based on the above, we conclude that the proposed chance in the initiation of NaOH injection is acceptaolo, and that the facility operating license can be anenced by changing the Technical Specifications as shown in the attach:aent to this license arendcent.
2.
Cuadrant Perer Tilt Intreducticn Sy letter dated May 10, 1973, Fetropolitan Ecisen Cc=pany (Net Ec) requested anenc=ent of Accendix A to Facility Crerating License ::c.
CPR-73 for Three Mile Island.'<uclear station, Unit 2 (THI-2). The recuestec change wculd arend the Technical Specifications to recuce the maxinun allowable value of neutrnn flux tilt as reasured in each quadrant of the reactor core ey in-ccre er cut-cf-ccre neutron cetectcrs.
Discussion Babccck and Wilcox (34W) perfonned the initial error analysis fcr cuacrant tilt and axial ircalance cerived frca incere signals basec on data cbtained frcn prctotype detectors in 1974 As a result cf ct.servaticrs Of operating characteristics of these detectors in a eratinc reacters a re-evaluation of the errer analysis has recently been cerforced by EaW.
This re-evaluation has resultec in an increase in the neasure art uncertainty for tilt and inbalance with a consecuent necessity fcr 3 meg, svamaur >
SATE D NRC PCMt 313 (9 76) NROI 0 +0 W un s. aovenwuswr pen <a omcas s,n - eawu
2 altering alarn setpoints for these quantities. dy letter dated Fay 11,1973, 22W has submit.ed a repcrt on this re-evaluation. This report was usec es tne casis for the evaluation of tne recuest for the Technical Specificaticn enange.
Br.1 las perfer ed a statistical analysis of the neasurerent of cuadrant tilt anc axial i:2alance and established an error which assures that the alam setpoint will be reached or exceeded 257. of the tire uhen the reasured quantity is et its li. nit value. Tne analysis was perfer ed by the ?'onte Carlo technicua.
Individual cetecter signal cerponents (chociuri signal, cackground signal, etc.) were encsen frca distributions of these quantities which were obtainec f rc:.1 critical expericents and operatinr; reactors. Conservative incivicual uncertainty cenponents were used.
Lis:iting valves of the "real' tilt and irbalance were assurac and "neasurec" values v.ere obtainec Oy perferning the same calculations that are done by the online corputer. The error value was then chosen so that the alarn setpoint was reached or excceded 957.
cf the time.
The !!onte Carlo analysis was perforned with 5000 trials.
84W has also investigated the effect of the new data base on inccre detector uncertainty on the ceasurcrents of F3 and 6, Ccaparisens were made between calculations and ceasurenents of these quantities in several operating reactors. The measurement uncertainty was inferred from these cccparisons by assuming a conservatively snall calculational uncertainity and ascribing the rest of the cifference to neasurenent uncertainty. The results shewed that the presently used uncertainties (57, for F and 7.5". for F are conservative.
3 e
Evaluation Based un our review cf this dccument we conclude that the rethod of analysis is acceptable. We further conclude that the values of alarn setpoints for cuadrant tilt anc axial irbalarce recorrended fcr T'il-2 are acceptable.
We note that our review of the 5&W subnittal of "ay 11, 1970, has not been fully ccepleted, but that it has progressed sufficiently so that we have been able to evaluate and find acceptable the specific changes in alarn setpoints for THI-2, as stated abcve.
Since the precosed changes to the TMI-2 Technical Specifications falicu the applicable 35W reccrrendations, v.e conclude that the recuestuc changes are likewise acceptacle.
..e further conclude that no chances are necessary in the uncertainty values assignec te ncasurecents of F;_ and F, and that the facil';y cperating licence can be arenced by changing tne Technical Specificaticns as shown in the 4ttachr.ci.t
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';iill oct result in any significant envircrrental irpect.
-:cvi m nade this deteminatica, t 3 l']ve further corcluded that tha ir rc en involves an action which is insignificant f rcn the starepair.: c" envircmetal incdct and, cursuant to 10 C9 [.51.5(o)(c), that en envirenrental incact s:cremnt er necative ceclarc ier anc c:virc* ar sl inpact Lp,;raisal r.eed not Lc prepared in connec;icn 'ith ' e i ss..~cc of this crendrent.
.C onc l u si o.n.
.le i: ave conclucr.d, basec on the consicerations ciscussed c"cvu, that:
(1) tecause the ar.encnent does not involve a signific. int 1"cre;>
in the prcbability cr consecuences of accidents previously cc.msive~.:
and does not involve a sicnificant decrease in a safety Trrin, tt e arendrent dces not involve a significant hazards consideration, (2) there is reascnable assurance that the health and safety cf t' e public will not be endangered by c::eration in the prcposed ranner, and (2) such activities will be corducted in cenpliance with the Comissian's regulations and the issuance of this anandment will not be inimical to the corron defense and security or to the health ar.c safety of the public.
H. Silver, Project t'anager Light Water P.ecctors 3rar.ch t'a. 2 Division of Project id:r'acerent 0;7+a! W-d I:7 Steven A. '!arga, Chief Licht Water Recctors Brarch No.
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NETRCPOLITAN EDISCN COMPANY JERSEf CENTRAL POLER & LIGHT CCMPANY PENNSYLVANIA ELECTRIC COMPANY DOCKET NO. 50-320 THREE MILE ISLAND NUCLEAR STATION, UNIT 2 AMENDMENT TO FACILITY CPERATING LICENSE Amendnent No. 4 License No. DPR-73 1.
The Nuclear Regulatory Cconission (the Commission) ha fcund that:
A.
The issuance of this license amendnent ccmplies with the standards and recuirenents of the Atomic Energy Act of 1954, as amenced (the Act) and the Cornissien's rules and regulations set forth in 10 CFR Chapter l',
B.
The facility will operate in conformity with the license, as amended, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, arid (ii) that such activities will be conducted in compliance with the Ccomission's regulations; D.
The issuance of this amendment will not be inimical to the connon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Connission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the amended Facility Operating License No. DPR-73 is hereby amended by changing the Technical Specifications as indicated in the attachnent to this license amendnent.
Paragraph 2.C.(2) of amended Facility Operating License No. DPR-73 is hereby amended to read as follows:
d 277
=
"2.C.(2)
Technical Scecifications The Technical Specifications contained in App'endices A and B, as revised through Amendment No. 4 are hereby incorporatec in the license.
Metropolitan Edison Company sh:ll operate the facility in accordance with the Technical Specifications."
3.
This license amendment is effective as of the date of its issuance.
Conformance with the revised provisions of the Technical Specifications in the attachment dealing with chemical. injection may be delayed for 10 days from the date of issuance of this amencment.
FCR THE NUCL'AR REGULATORY CCFitISSION
,)
A eY Eehl.V a
Light Water Reactoq Sranch No. a Division of Project"1anagement
Attachment:
Changes to the Technical Specifications Date of Issuance:
May\\% 1978 61 278
ATTACHMENT TO LICENSE A!1ENCf!ENT NO. 4 FACILIT OPERATING LICENSC NO. DPR-73 D0CKET tl0. 50-3200 Change the following pages of the Ascendix "A"
Technical Specifications with the enclosed pa3es as indicated.
The revised pages are identifjed by Amendrent number anc contain vertical lines indicatinc the area of change.
The corres;:anding overlea f pages are also previ5ed to maintain document completeness.
Pages 3/4 2-11 3/4 3-12 3/4 3-17 3/4 3-18 3/4 3-19 3/4 3-22 G E79
TABLE 3.2-1 00ADRANT POWER TILT LIMITS STEADY STATE TRANSIENT MAXIMUM LIMIT LIMIT LIMIT Me a s u r emen t Independent QUADRANT PCWER TILT 3.69 9.74 20.0 QUADRANT POWER TILT as Measured by:
Symmetrical Incore Detector System 2.30 7.71 20.0 Power Range Channels 0.96 5.88 20.0 Minimum Incore Detector System 1.72 3.71 20.0 THREE MILE ISLAND - UNIT 2 3/4 2-11 Amendment No, a 61--280
TABLE 3.3-3 (Continued),
EflGIfit LRED SAFETY _ FLATURE ACTUAflufQYSit M_IfisiltilMEril AT IOtt 5
m MlfilMUM l0TAL flu.
CllAfiflCLS CHAriflELS APPL ICAllLL 35 f _ull_C I_l_0flAl_ _Ull.i._f Of CilA_ft_il.E_LS
_1_0 IRIP _
OP_E_R_A_B_L E_
MO_ DES _
_AC T_I Of_t 3.
REACIOR HUILDIflG W
SPRAY b
a.
Reat. tor ihsildirig Pres sure Ilich-liigh 3***
2***
2***
1, 2, 3 10#
?5
- i b.
Autcmatic Actuation logic 2
1 2
1, 2, 3 11 no 4.
REACIOR BUILD!flG SUMP SilCIl0ft u
d.
UWSI LOVel-low 1/trdin 1
1/trdin 1, 2, 3, 4 9
D 5.
FLLDWATER LATClllflG w
a.
Main Steam Pressure Low 4/St. Gen 1/St. Line 2/St. Line 1, 2, 3****
9 (2/ Main St. line) 6.
ITEDWATER LINE RUPTURE p
DiTECT10!1 u
a.
feedwater/ Main Steam N
Lina Differential CO Pressure Low 1/St. Gen 1/St. Gen 1/St. Gen 1, 2, 3 9
P
TABLE 3.3-4 (Continued).
EilGillEERED SAFETY FEATURE ACIUATION SYSTEMS INSTRtlMENTATION TRIP SETPOINTS
??
~
FUNCTIONAL. UNIT TRIP SETP0 INT At10WABLE VALUES 7.
1OSS Of POWER continued T-};
b.
4.16 kv Einergency llus Undervoltage (Degraded a
i Voltage)*
E Q
l.
Emergency flus #2-lE and 2-2E
(
+
) valts with a
(
L
) volts with a m
) second time delay
(
+
) second time delay
(
+
2.
Emergency Bus #2-3E and 2-4E
(
+
) volts with a
(
+
) volts with a
(
[
) second time delay
(
1
) second time delay
'i'.
8.
CllEMICAL ADDITION SIGNAL d.
RedClor Ouilding Cooling and Isolation Initiation
(
same as
- 2. a/b/c
)
b.
Safety Injection and
(
same as
- 1. a/b/c/d/e
)
IlWST Level Initiation S3'9".+._ 2.9" 53'9" _+ 3" 1g 0
- llot required until the first refueling per U.S.N.R.C. letter dated August 26, 1977 from S. A. Varga, j', g)
Light Water Reactors Branch #4, Division of Project Management to Metropolitan Edison Company, subject:
m Transmittal of Staff Positions 222.46 and 222.47. A proposed change to the Technical Specifications 2
to incorporate trip setpoint and allowable values shall be submitted to the flRC at least 90 days N
prior to start up af ter the f' erst refueling.
(ON
TABLE 3.3-5 (continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE,[IME IN SECONDS 3.
Reactor Building Pressure--High-High a.
Reactor Building Spray Pumps 1 31 */31 '
4 Reactor Coolant Pressure-Low a.
High Pressure Injection
< 25*/25**
b.
Low Pressure Injection
< 25'/25" c.
Ccmponent Coolant Water System (1)
Decay Heat Closed Cooling
< 300*/300**
(2) Nuclear Services Closed Cooling i 95*/NA" d.
Service Water System (Nuclear Services River Water) 1 95*/95**
5.
Feedwater Latching a.
Main Steam Isolation
-< NA*/124.6**
b.
Feedwater Isolation (1)
FW-V30A/B
< NA*/ 9. 2" (2)
FW-V17A/B 7 NA*/32.6" (3)
FW-V25A/S 7 NA*/14.6**
(4)
FW-Vl9A/B
[NA*/32.6**
6.
Emergency Feedwater Pump Actuation a.
Turbine Driven Pump
< NA*/29**
b.
Motor Driven Pumps i 29*/12 "
TABLE NOTATION
- Diesel generator starting and sequence loading delays included.
Response time limit includes movemerd of valves and attainment of pump or blower discharge pressure.
- 0iesel generator starting and secuence loading delays not included.
Offsite power available.
Response time limit includes movement of valves and attainment of pump or blower discharge pressure.
(*)Re;ponse time applicable for Reactor Building cooling and isolation only.
THREE MILE ISLAND - UNIT 2 3/4 3-19 Amendmejit jto a
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UNITED STATES NUCLEAR REGULATCRY CC""ISSICN DCCKET N5. 50-320 METROPOLITAN EDISCN CCMPANY JERSEY CENTRAL POWER & LIGHT CUMPANY PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND BUCLEAR STATION, UNIT 2 NOTICE OF ISSUANCE OF Af!ENDMENT TO FrCILITY CPERATING LICENSE The U. S. Nuclear Regulatory Commission (the Commission) has issued Amendment 4 to Facility Operating License No. OPR-73, issued to the Metropolitan Edison Ccapany, Jersey Central Power & Light Ccnpany, and Pennsylvania Electric Company, for operation of the Three Mile !sland Nuclear Station, Unit 2 (the facility), located in Dauphin County, Pennsylvania. The amendmant is effective as of its date of issuance.
The license is amenced by revising certain Technical Specifications.
The application for the amendment complies with the standards and requirements of the Atcmic Energy Act of 1954, as amendec (the Act), and the Commission's rules and regulations.
The Commission has made aporopriate findings as required by the Act and the Ccmnission's rules and regulations in 10 CFR Chapter I, wnich are set forth in the license amendment.
The Cornission has cetermined that the issuance of this amendment will not result in any 3ignificant environmental impact and that pursuant to 10 CFR 51(d)(4), an environmental statement or negative declaration and environmental impact appraisal need not be prepared in connection with issuance of this 'mendment.
For further details sith respect to this action, see (1) Amendment No. 4, to Facility Operating License No. CPR-73, and (2) the Cccnission's related safety evaluation e oporting Amencment No. a to Facility Operatinc bl.' ?S5
.c
_2 3
License ilo. DPR-73.
These items are available for public inscection at the Commission's Public Document Roca, 1717 H Street, N. W.,-Washington, D.
C.,
and at the State Library of Pennsylvania, Comonwealth and Walnut Streets, Harrisburg, Pennsylvania 17126. A copy of items (1) and (2) nay be obtained upon request addressed to the U. S. Nuclear Regulatory Ccmission, Wasnington, D. C. 20555, Attention: Director, Division of Project Management.
Cated At Bethesda, Maryland, this \\h day of May 1973.
FOR THE NUCLEAR REC-ULATCRY LCMMISSION ll h
Y l
S.
\\t[IMll..(.#
Steveh n. Narga, Cntef
' Light Water Reactors 3r,ch No. 4 Division of Project Management 6.,I ~ 28Ib