ML19220B248
| ML19220B248 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/26/1978 |
| From: | Boyd R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19220B246 | List: |
| References | |
| FRN-780526, NUDOCS 7904250539 | |
| Download: ML19220B248 (9) | |
Text
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U:!ITED ST3TES CF Ai' ERICA NUCLEAR PEGULATOW/ C0l l!I SS ION
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In the Matter of
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Docket No. 50-320
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Three Mile Island Nuclear Station, Unit 2
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3 DER FOR t'0DIFICATION OF LICE'lSE I.
The "etropolitan Edison Company, et al (the licensee or Met Ed), is the holder of Facility Operating License No. DP9-73 which authorizes the operation of the nuclear pcuer reactor known as Three Mile Island Nuclear Station, Unit 2 (the f acility or TMI-2), at reactor core power levels not in excess of 2772 negawatts thernal (rated power). The facility, using a Babcock & Wilcox Conpany designed pressurized water reactor (PWR), is located at the licensee's site in Dauphin County, Dennsylvanir.
II.
In accordance with the requirements of the Ccrnission's ECCS Acceptance Criteria,10 CFR P 3rt 50.46, the licensee submitted on Narch 31, 1975, an ECCS ev3luation for the f acility. The ECCS evaluation subritted by the licensee was based upon an ECCS Evaluation Podel developed by the 42bcock & Wilcox Company (3aW), the designer of the nuclear stean supoly systen for this facility.
The B&W ECCS Evaluation "adel had been oreviously founc to confer-74-123 7904250538
. 3 to the requirerer.
the Comission's E"C5 Acceptance Criteria,10 CFR P3rt 50.46 and Appendi<
The evaluation indicated that with the linits set forth in the f acili'.y's Techetical Specifications, the CCI coolia.g perforaance for the f acility vould confor 1 wi th tiie cri'.eria contained in 10 CFR Part 50.16(b) which govern calculate 1 pe ak clad temperature, caxinun cladding oxidation, maximun hydrogen generation, coolable gecc.etry and lang-ter.n cooling.
On bril li, 19 78, E s.l i n for 'ed *.he *:RC ' nat it h of fetornined that in the event of a snalI break loss-of-coolant acciden' (i.0CA) on the discharge side of a reactor coolant purp, high pressure injection (HFI) flow to the core could be reduced sonewhat.
Subsequent calculations indicated that in such a case ".he calculated peak clad ter;erature nicht exceed 0
2200 F.
Pewious s all he?ak analyses for 3.D 177 foal assembly (FS) lowered loop pl ants had iden'.ified the limiting snall areak to be in the suction line of the reactor coolant purp.
Recent analy;ns h3ve shoun that the discharge line break 'is more liniting thei the suction line breM.
The Three ' tile Is!and :uclear Station,ilni' ?, has an
~"S configuration
. A i t: b causi;* s of ' ro high pressure injel.ilo traias.
~ scii trai n has
.in @I pump Ind the train injects into two of *:.e 'our redoc calant
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systen (RCS) :lld lejs an the discharje si1 + 4 also 3 third HP: pu:,p i nsta l l e i. ) The t'vo :arallel
'P trains are connectea W
+c3 :et isolated by :anual valres (knc.so as the cr>ss-c;nnect /alves)
'a t are nornally closed. Upon receivig 3 safej injec'ino signal the HPI puip; r4 s.irted and talves in the 'our iojection lines are opened.
Ass iling loss of of fsite power and th e vorst single f ailure (f ailure of diesel to start) only one iPI pump ' soul f M aiailable and two of t'n four injection valves would fail to g en.
If a snall areak is postulated to occur in the PCS piping bet.seen the RCS pomo discharge ind the reactor vessel, the high pressure injection flou injecteJ into this line (aMut half of the out, nut of one high, ores-sure injection pu1p) could flow not the bras.
Ther2 fore, for the worst conbination of break location and single f ailure, only one-hal f of the flou rue of a single hign pressure injection purp would contribute to maintaininc the c]olant inventory in the reactor vessel.
This situatinn had not beea previcusly analyzed and M.J '13d imiicated that the limits spec i fi ed i n 11 ^,r'
' 3r'
~,0. If e exceeded.
r RW has statti 'Mt they have analyzed i spectro l of small br Mks in the pump discharge line and have deterniqed ' bat to icet *he li ii t s of 10 CP D art 50.46, operator action is required to open the *mn ma nu al l y-c;; era t ed cross-connec*. valves and to manually open the two.,9t w ir, en isolati;n valves which had fliled to 7 pen and align all four isol+t on ealves.
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This would allow the flow from the ore HP! ot c to feed ali four react s-ccolant legs. B AW has assured that 30 percent of the flcw would be 105t through the break and 70 percent would contribute to recosering tne core.
S&W has prepared a surnary entitled ' Analysis of Sna113reaks in the Reactor Coolant Punp Discharge Piping for the B&.; Lowered Loop 177 FA Plants," May 1,1978 (the S?h Sunnary), which describes tne rethods used and the results cotained in the above analysi?. The analysis models operator action by assuning a step increase in flew to the reactor vessel (with balanced flow in the three intact loops) ten minutes after the LCCA reactor protection system trip signal occurs.
By letter dated May 5,1978, f:et Ed subnitted a copy of the B&W Sumary for our review.
In their subnittal Net Ed stated that they had reviewed the B&W Sumary and deter lined that the results were applicable to IM!-2 and that oceration of TMI-2 up to 2563 megawatts thernal acu!d be in full conformance with 10 CFR Part 50.d6.
They also stated that adoitional analyses will be available to the Ccmissto" for power levels up to 100 cercent power (2772 mecawatts ther'al) by June I, i'U In their submittal nf May 5,1978, Net Ed also stated that they had mcdified certain plant crocedures to provide the necessary crerator actions on a time scale consistent with tnat assured in the analysis, 4a,-
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- r and that they had conducted drills to verify th3t the assu~ed coerator resoonse tire was achievable.
The Cconission's U?fice of Inspection and Enforcement has confirned that appropriate procedures are in place and that drills were perforced which verified operator resconse tite. Met Ed also cornitted to submit as soon as possible a request for amendment of the TMI-2 Technical Specifications as aopropriate to reflect acoption of these procedures, and cconitted to submit a proposal for a permanent solution to this problem by August 5, 1978.
In their letter of hay 11, 1973, Met Ed provided additional information clarifying aspects of the proposed manual actions.
In the event of a small break and a limiting single failure, nanual action will be taken to begin ocening the crosscannect valves and the isolation valves within five minutes and have then crened and an 3dequate ficw split obtained within 10 minutes.
To f acilitate this oneration the licensee has connitted to maintain one of the series-connected, manually-operated cross-connect valves nornally open. The analyses perfor ed by B&W assumed that the flow split aas established at 650 seconds by operatcr action.
We conclude that the analyses are a reasonable approximation of the cuerator action that actually will be taken, since scecific rrccedures have been prepared and drills performed to verify the adequ3Cf of the procedures and to train the plant operators.
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In thei r analysis, BS'.1 states that 3 0.13 f t cischarga
oe
?reak, witn the 3 fore entioned ocerator actions, is the mos' linitino case. To arrive at this conclusion, B&W has perfor ed analyses at break sizes
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o f 0.04, 0.0 7, 0. l, 0.13, 0.15, 0.17, 0.2, a nd 0. 3 f t.
The results, which.ere cbtained using an approved Apcendix K rodel for blowdoun, incicate core 2
uncovery for about 300 seconds for the 0.13 ft break.
For inis break size B&W has conservatively calculateq the coak claa tenperaturn 0
to be accroxinately 1551 c, well below the limits cf 10 CFR Part 50.46(b), #cr i power level of 2568 megawatts thernal.
Based on review of the B3W Sunnary we find that the calculations supcort
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the conclusion that a 0.13 f t-discharge line break is the nest limiting case.
However, the B3',1 Sunnary does not demonstrate that the assunptions emcloyed in sucolying heat inputs to the FOAM rortion of the calculations were conservative.
e are also reviewing whether use of simplifiec input in the F0AM calculations satisfies the requirement for calculation using an aperoved nodel. Accordingly, we cannot conclude at this time inat operation of TMI-2 at 2562 negawatts thernal would be fully : n conf arnance with 10 CFR Part 50.16.
On the other hand, the range of calcuiaticns new available shows that for operation of this f acility at co..er levels uc to 2568 neg3uatts thermal, ECCS Derformance calculations for the limiting small break indicate that this break has a very substantial margin on peak clad tencerature below the limits of 10 CFR Part 50. '6 (b) if accropriate ocerator action is procerly taken (is describeo ]bove).
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_7 Therefore, until we have had the ccccrtunity to fully issess the Sin
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calculations, the staff cannot determine that oreratic, of T'tI_2 at full cowcr unaer the conditions of tne revised calculations cy ESW applicaole to this f acility confarns fully to the requirements of 10 CcR Part 50.26.
Hcwever, operation of TMI-2 at power levels up to 2563 megawatts thermal a m' in accordance with acpropriate operating procedures will ensure that the ECCS will conform to the perfcrnance criteri) of 10 CFR Part 50.06.
Therefore, until Eau calculations applicable to this f acility are ccmpleted to assure full coroliance with 10 CFR Part 50.J6, the peak clad temperature margins provide reasonable assurance that operaticn of the facility at PGwer level s up to 2568 megawatts thernal with anpropriate oyerating crocedures specified herein will not endanger life or property or the connon defense and security.
With the procedures described in the licensee's letters of May 5 and 11,197?,
the staff believes that the licensee's actions are arprocri3te and that these actions shculd be confirued by NRC Crder.
In the course of our review of this matter, a related issue arose: the need to apply greater uncertainties to the measured values cf neutron flux in each quadrant of the reactor core.
34W recently reported to het Ed that on the oasis of oceraticnal ex;erienco anc a reevaluation of measure ent error statistics and error 'recaga.lon,
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greater uncertainties should be applied to the t'easured values of cuacrant flux tilt.
This greater uncertainty was necessary to 3ssure that the actual flux tilt did not ' exceed the limiting value assuned in the evaluation of pustulated accidents including a LOCA. A description of the reevaluation and recccrended reduced limits on allowable measured flux til* were presentec in a B5'1 report submitted to the staff en May 11, 1973. By letter datec "ay 10, 1978, Met Ed requestec anendrent of the "11-2 Technical Scec1 1catiors to reflect the more conservative linits.
c!e have reviewed the Ba'.' recort and the Pet Ed request relative to this matter and have concluded that the linits requested for TMI-2 are acceptable. Use of these !inits is being authorized by Amend ~ent No.' 4 to the TMI-2 Operating License No. CPR-73, issued on May 19, 1978.
III.
Copies of the following docurents are available for inscection at the Comnission's Public Docunent Rocn at 1717 H Street N.W., nashington, D. C.
20555, and are being olaced in the Connission's local public cocument rcon at the State Library of Pennsylvania, Ccrronwealth arc cialnut Streets, Harrisburg, Pennsylvania, 17126:
1.
Letter fren J. G. Herbein to S. A. Varga, Chief, Light later Reactors Branch Mo. 4, dated Day 5, 1973.
2.
Letter frcn J. G. Hercein to S. A. Varga, Chief, Light '.eter React rs Branch No. 4, dated "ay il, 19'3.
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IV.
Accordingly, pursuant to the Atenic Energy Act of 1952, as arended, and the Ccenission's R _les and Regulations in 10 CFR rarts 2 and 50, :T :S ORDERED THAT Facility Ocerating License No. OPR-73 is hereby conditioned by acding the following new provisions:
(1)
As soon as possible, the licensee shall subnit a reevaluation (wholly in conformance with 10 CFR Part 50.25) of ECCS cooling performance calcu13ted in accordance with the Br.i Evcluation !!cdel for cceration with operating procedures described in its letter of May 5,1973, (2)
Until further authorization by the Connission, the ;cwer level shall not exceed 2563 regawatts thermai, and (3) Until furthe-authoriz3 tion by the Ccnnissian, the licensee shall ocerate in actnrdance with the procedures described in its letters of itay 5 and May 11, 1978.
FCR THE NUCLE.5R REGULAINY CCUMISSION c
her s. s,
a Rcger S. Boyd, Directcr Division of Project Management Office ;f Nuclear Reactor Regulation Dated at Bethesda, Maryland, this 25th day cf Pay 1973.
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