ML19210B128
| ML19210B128 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 09/05/1975 |
| From: | Arnold R METROPOLITAN EDISON CO. |
| To: | US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| GQL-1255, NUDOCS 7911040063 | |
| Download: ML19210B128 (8) | |
Text
.
S NRC DdTRIBUTION FOR PART 50 DOCKE1 MATERIAL (TEMPORARY FO3M)
FILE:
t]E DATE OF DOC DATE REC'D LTR TWX RPT OTHER FROM P nna R.C. Arnold 7-9-75 7-12-75 XXX
^A'"
TO:
SENT LOCAL PDR CLASS UNCLASS PROP lNFO INPUT NO CYS REC'D DOCKET NO:
XXXX 1
50-289 DESCRIPTION:
ENCLOSURES:
Ltr.Re our ltr. of 3-14-75 & 6-13-75.. & their Re-evaluation of ECCS cooling performance cal-Ltr. of 4-19-75...... w/ attachments.......
culated in accordance with an acceptable eval-uation model......
( 1 cy. Encl. Rec'd)
PLANT NAME: Three Mile Islan
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6 Director a
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"M Division of Reactor Licensing U. S. Nuclear Regulatory Commission C;
y#
Washington, D.C.
20555
Dear Sir:
Pursuant to the Commission's Ordei for Modirication of License for Three Mile Island
- Nuclear Station Unit 1 (TMI-1) dated December 27, 197h, a re-evaluation of Emergency Cort Ncling System (ECCS) cooling performance calculated in accordance with an acceptable evaluation model which conforms with the provisions of 10CFR50, Section 50.h6, has been ecmpleted. However, the proposed changes to the. Technical Specificati,ons made necessary as a result of this re-evaluation are not included in this submittal in that they are still in required committee review. Met-Ed will operate TMI-l within the most restrictive wid con-servative limits of the proposed Technical Specification supplied to us by Babcock and Wilcox (B&W) and our present Technical Specification limits. Met-Ed will submit within 30 days a completed Technical Specification change request consistent with the re-eve.luation.
"'he evaluation model utilized in performing the re-evaluation of ECCS cooling performance is described in B&W non-proprietary Topical Report BAW-101Ch, "S&W's ECCS Evaluation Model".
The results of the evaluation for 35W 177 fuel assembly units with a lowered-loop arrangement are described in non-proprietary Topical Report BAW-10103, "ECCS Evaluatien of B&W's 177 Fuel Assembly Lowered Loop NSS".
The analysis presented in BAW-10103 for the B&W 17T fuel assembly units with a lowered-loop configuration is generic in nature since the parameters used in this analysis are conservative for TMI-1.
The parameters associated with TMI-1 are bounded by those utilized in the generic *analysist and thus 3AW-10103 provides a conservative evaluation of ECCS performance for TMI-1.
The results presented in 3AW-10103 demonstrate the conformance of TMI-1 to the criteria of 10CFR50, Section 50.h6, under the operating conditions specified in the prcposed Technical Specifications which will be submitted within 30 days.
EI lrn'.0D
"( J J nr ui
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In addition to the above and in accordance with your letter of June 13, 1975, the following information is also provided.
1.
Break Spectrum and Partial-Loop Operation It has been demonstrated using the FAC guidelines that peak cladding temperatures were significantly lower for partial pump operation than for 4 reactor coolant pump operation. The proposed technical specification limits for partial pump operation are based on minimum shutdown margin and ejected rod worth criteria.
It has been shown by additional analysis, using the FAC guidelines that the minimum shutdown margin and ejected rod worth criteria are still limiting. This analysis will be issued by July 23, 1975.
2.
Potential Boron Precipitation The requested information was provided by Met-Ed in our letter of April 19, 1975 in response to your letter of March 14, 1975.
3.
Single Failure Analysis As requested, a single failure analysis for manually-controlled, electrically-operated ECCS valves has been performed. The results of this analysis are contained in Chapter 6 of the TMI-l FSAR and are supplemented in Attachment 1 to this letter.
Based on the information provided in the TMI-1 FSAR and Attachment 1, it is concluded that no credible single failure or operator error affecting any manually-controlled, electrically-operated ECCS valve could significantly adversely affect ECCS performance.
4.
Submerged Valves The following valves will be submerged when the entire contents of the BWST are discharged into the Reactor Containment Building:
ICV 1A and IB (Letdown Cooler shell side inlet isolation valves)
ICV 20 (RC Drain Tank cooler outlet isolation valve)
MUV 1A and IB (Letdown Cooler tube side inlet isolation valves)
MUV 2A and 2B (Letdown Cooler containment isolation)
ICV 2 (Intermediate Cooling containment isolation)
WDL V302 (RC Drain Tank recirculation)
WDL V305 (RC Drain Tank recirculation)
Only three of the above valves (ICV 2 and MUV 2A and 2B) are Engineered Safety Feature valves. These three valves are containment isolation valves and will have performed their safety function prior to being submerged.
None of the above valves are rer,uired to c..ange position for the short term or '.oug term ECCS function and therefore their submergence will not affect any ECCS function.
1586 256
Even if power were maintained or inadvertently applied, during or af ter submergence, to one of the above valves, this single failure has been determined to have no adverse effects on the remainder of the electrical system.
5.
Containment Pressure The containment pressure used to evaluate the performance capability of the ECCS has been calculated in accordance with the methods con-tained in Section 4.3.6.1 of BAW-10104 and the results are presented in Section 4.4 of B AW-10103.
Also, as requested on Page 7 of the staff's Safety Evaluation Report which accompanied the Order for Modification of License, as-built passive containment heat sink data has been compiled and is given in.
The hett sink inputs to the generic model are conservative compared to this as-built data compilation. Using the generic heat sinks, the containment pressure calculation is in accordance with Branch Technical Position CSB 6-1.
Sincerely, A
R. C. Arnold Vice President RCA:CES:tas cc: Office of Inspection and Enforcement, Region 1 Attachments 1 & 2 File:
20.1.1 / 7.7.3.1.1 1FD/
9E7 IJOU LJ/
In addition to that provided by Chapter 6 of FSAR ATTACHMEHT 1 VALVE NORMAL SIHGLE SYSTEM l'JSK NO.
DESCRIPTION POSITION FAILURE EVALUATICN HPI MU-V20 Seal Injection Line Open Closed Would only stop seal injection flow which in RB Isolator Valve being provided by one HPI pump. Ho effect on ECCS performance.
MU-V18 Isolation Valve in Closed Open Would cause slight increase in flow rete in Ucrmal Makeup Line one HPI string. If required, the flow rate can be reduced by throttling er closing HPI Valve MU-V16 A/B or by closing MU-V17.
MU-V12 Make-up Tank Closed by Open Check valve MU-V112 prevents reverse flev Isolation Valve Emergency of core cooling water into MU-T1.
Procedure LPI DH-V1,2,3 Decay Heat Drop Closed Open Valves are all in series and are also in Line Isolation series with manual normally closed valven Valve DH-V12 A/B. Unplanned operations of one of these valves vould have no effect en ECCS injection capability. These valves are also used for control of boron concen-tration during long term cooling and must be opened within 30 days after the accident.
A redundant flow path is available should any of these valves fail to open.
LPI DH-VSA or BWST Isolation Open Fails to If Dif-V5 A or 5B rail to close when injection DH-VSB Valve Close from the RB sump is to be establiched, check valve DH-V1hA/B will prevent reverse flow back to BWST.
DH-V6A or RB Sump Outlet Closed Fails At most, one LPI punp and one RB spray DH-V6B Valve during Open pump would be effected. The redundant initial pumps vould still be operable. Both strings portion of of HPI vould be available during 3nitial Ln accident injection from the BUGT.
Ca If either DH-V6A or B should fail open CJ' during initial LPI from the BWCT, ther. a RB pressure above %35 psig would prevent flov out of the BWST to that LPI pump nnd RB
(,
spray pump. These pumps vould therefore run Crj dry for the tgme required to fill the RB sump. A 5 ft LOCA vill release %2 times the sump volume in liquid form within the first 5 seconds.
---m----
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=
VALVE NORMAL SINGLE SIC"Mi MARK NO.
DESCRIPTION POSITION FAILURE EVALUATION Since the LPI pump and building spray pump probably are capable of operating dry for several minutes, failure of DH-V6A/B durint-the initial portion of the accident, would probably have no effect on ECCS performance 8
and at the vorst, vould enly affect cne of the redundant strings of LPI and building spray system.
DH-VTA/B Isolation Valve Closed Open Ucula increase flow rate of LPI pump by Ed&veen LPI and abouw 500 gpm since LPI pump would be HPI pumping to both the reactor vessel and also to the suction of one HPI purp. Uo adverse effect will occur and flow will be throttled if necessary to maintain acceptable flow rate.
Redundant LPI string is not effected.
DH-V61 A/B Isolation Valve Closed Open Upstream manual isolation valve CA-V256 in cauntic addition vould be closed. This single failure line to suction of therefore has no effect.
LPI pump CF CF-VlA/B CF Tank Isolation Open Not Tech. Spec. 3.3.1.2 requires breuker for Valve Credible valve operator to be open.
CF-V2A/B CF Tank Drain Closed Opens Redundant containment isolation valves Valve CF-V20A/B vould remain cloned preventing any effect as a result of this single failure.
CF-V3A/B CF Tank Vent C1csed Not The breaker for the valve operator vill be Valve Credible opened to ensure that this valve cannot open before or during CF tank discharge and result in a loss of CF driving pressure.
LD G3 ON LJ7 s4)
i AT? O C CT 2
- Ill A Cenparisen of ey Paranaters Enplcyed in the ocneric ji Eval':stion 'Jedel to Individual Plsnt Parr2eters
- i t li 4
(
Par 1neter Generic '5 del
- .et-II la t'
'\\
Reacter Building Free
?
Volune f 3 2.205x10 12.205x10 l
6 6
r
?
I The building is modeled with five heat sinks:
i c.
The reactor building valls including concrete vall, steel liner, and anchors:
f 2
Exposed area, ft
= 67,410.0 167.h10.0 0.00083
> 0.00083 Faint thickness, ft
=
.0550h I.0550h Steel thickness, ft
=
Concrete thickness, ft(**)=
4.0 4.0 1
b.
The reactor building done including concrete, steel liner, and anchors:
2 Exposed area, ft
= 18,375 0 118,375.0 0.00083
> 0.00083 Paint thickness, ft
=
.065L6
][.065h6 Steel thickness,.ft
=
3.0 1 3.0 l-Concrete thickness, ft
=
(*)
c.
Painted internal steel:
2 Exposed area, ft
= 249,000.0 12L9,000.0 0.00083
> 0.00083 Paint thickness, ft
=
0.03125
<_ 0.03125 Steel thickness, ft
=
Unnainted Internal teel n
krea, ft squared ***)
36,000 1
36,000
=
d.
Unpainted internal steel, stainless steel: (*)
Exposed area, ft2 10,000.0 110,000.0
=
0.03125 1 0.03125 Thickness, ft
=
e.
Internal conareter 100F350.L622 says these can be assuned 2
Enposed area, ft
= 160,000.0 11c0,0C0.0 0.00083
> 0.00083 l
Paint thickness, ft
=
1.0 11.0 t
Concrete thickness, ft
=
f.
Chernophysical Properties:
cm na nn Thern 1 Conductivity, D
D D
f Btu /h-ft _p e9 g
_ g]
_ g j
2 v
"aterial 0.92 1 0.92 Ccncrete
=
27.0
< 27.0 S; eel
=
9.1036 7 9.15;6 S;1.nless Steel
=
.6215
][
.6215 Paints
=
1536 200
(*)
Che total n.re; cf sinta C al D nhcald ce calculated r.nd aheten to te leza t'.cr ?95,000.0 ft.
("") Ooe IF Cocical Eencr: ICC91 Ce:.1 Pu ;es 3-lh '. follotcin; /owstion 25).
( m) Thi;. thici noss nc[ te r ar.nad, per C52 6-1 ':'?.2 ).
i.
~
~
e 4
f.
Ther:ophysical ?roperties:
(Cont'd)
D "" ^" D 7' D
' 'D hf f>
Zeat Capacity,
_suvuML
. eb
..\\ W ;a u
3 3tu/ft _y o.,,,
concrete
= 22.6
<22.6 7 8.8 53.3 5
Steel
=
7 h.263 5L.263 5
Stainless Steel
=
40.h2 Th0.b2 l
?aint
=
i g.
Delay Ti=es, Sec.
i 0.0
>0.0 Reactor Building Coolers
=
(No loss of off-site power)
(Go loss of off-site power) 65.0
>65 0 Reactor Building Sprays
=
{
i h.
Building Initial Coniitions:
Temperature, F
= 110
>110 13.7
> 13.7 Pressure, psia
=
Relative Humidity,.5
= 100
[100 Verificction of the above values can be obtained from EL*.I topical report 3r.1-10103 Section b.h.
i5L5 2o1
.