ML19322D934
| ML19322D934 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/07/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19322D920 | List: |
| References | |
| NUDOCS 8003110258 | |
| Download: ML19322D934 (8) | |
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O ATTAC ENT TO LICENSE AMENDMhNT NO. 6 FACILITY OPERATING LICENSE NO. OPR-73 DOCKET NO. 50-320 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated. The revised pages are identified by A.end:ent nurt$er and contain vartical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain docnent cocpleteness.
1 Paces r
2-2 2-3 2-5 2-6 2-7 2-8 8 2-1 8 2-2 B2-3 8 2-6 B 2-8 3/4 2-13 3/43-3 3/4 3-13 3/47-2 3/47-3 3/4 7-la (added)
B 3/4 7-1 3/46-5 3/4 7-17 9
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, REACTOR PROTECTI0t1 SYSTEM ItlSTRUMEf1TATI0tl TRIP SETPOIt{TS g
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flucloar Overpowar based on Pinnp Monttors{j) willi three pumps operating iilth tlirco pumps operatingf g
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< 57.18% of RATED TilERilAL POWER j
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< 0% of RATED TilERMAL POWER with
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Two pump operating in one loop and 5 0 pumps operating in one loop and no pump operating in the other loop no pump operating in the other loops a
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< 0% of RATED TilERMAL POWER with
< 0.28% of RATED TilERMAL POWER with iio pumps operating or only one pump iio pumps operating or only one pump.
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L pyTrip may be manually bypassed when RCS pressure < 1720 psig by actuating Shutdown Bypass provided that:
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The Ituclear Overpower Trip Setpoint is < 5% of RATED TilERMAL POWER
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The Shutdown Bypass RCS Pressure - !!!gh Trip Setpoint of 1 1720 psig is imposed, and p
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l 2.2 LIMITING SMETY SYSTEM SETTINGS
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2.2.1 REACTOR PROTECTION SYSTEM INSTRUMEKTATION' SETPOINTS ' ' ' ~
The Reactor Protection System Instrumentation Trip Setpoint specified in Taole 2.2-1 are the values at which the Reactor Trips are set for eacn paraceter.
The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a trip setpoint less conservative than its Trip Setpoint but within its specified Allowable Value is accept-able on the 1, asis that each Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
The Shutdown Bypass provides for bypassing certain functions of the Reactor Protection System in crder to permit convol rod drive tests, zero power PHYSICS TESTS and certain startup and shutdown procedures.
The purpose of the Shutdown Bypass RCS Pressure-High trip is to prevent normal operation with Shutdown Bypass activated.
This high pressure trip setooint is lower than the normal low pressure trip setpoint so that the reactor must be tripped before the bypass is initia ed. The Nuclear Overpower Trip Setpoint of < 5.0% prevents any significant reactor power from being produced.
Sufficient natural circulation would be available to remove 5.0% of RATED THEPJ%L P0'n'ER if none of the reactor coolant pumps were operating.
Manual Reactor Trio The Manual Reactor Trip is a redundant channel to the automatic Reactor Protectica System instrumentation channels and provides manual reactor trip capability.
Nuclear Over:cwer A Nuclear Overposer trip at high power level (neutron flux) provides reactor core protection against reactivity excursions which are too rapid to be protected by tenperature and~ pressure protective circuitry.
During nomal. station operation, reactor trip is initiated when the reactor power level reaches 105.5% of rated power.
Due to calibration and instrJeant errors, the maximum actual power at which a trip would be actuated could be 112%, which was used in the safety analysis.
THREE MILE ISL"dD < UNIT 2 B 2-4 e.-eeem e
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Tne A'XIAL P0iER IMBALANCE boundaries are established in order to prevent reactor therral limits fraa being exceeded.
These thennal limits are either power peaking kw/f t licits or DNBR limits.
The AXIAL POWER IMBALANCE reducer, the power level trip produced by the flux-to-flow ratio such that the brundaries of Figure 2.2-1 are produced.
The flux-to-flow ratio reduces tne power level trip and associated reactor power-reactor power-imbalance boundaries by 1.05" for a 1% flow reduction.
RCS Pressure - Low, Hich and Variable Lcw Tne High and Low trips are provided to limit the pressure range in which reactor operation is penaitted.
During a slow reactivity inser cica startup accident from low power or a slw reactivity insertion frca high power, the RCS pressure-High setpoint is reached before the Nuclear Overpower Trip Setpoint.
The trip setpoint for RCS Pressure-High, 2355 psig, has been established to maintain the system pressure below the safety limit, 2750 psig, for any design transient.
The RCS Pressure-High trip is backed up by the pressurizer code safety val ?cs fcr RCS over pressure protection, and is therefore set y lower thu' cne set pressure for these valves, 2500 psig.
The RCS Pressure-l High tvip also backs up the Nuclear Overpower trip.
Tne RCS Pressure-Low,1800 psig, and RCS Pressure-Variable Low, (13.00 T tain the M F-5887) psig, Trip Setpoints have been established to main-l c
3 ratio greater than or equal to 1.30 for those design 3
accidents that result in a pressure reduction.
It also prevents reactor i
oper!tio. at pressures below the valid range of DNS correlation limits, protacting against DN3.
Due to the calibration and instru:nentation errors, the safety analysis used a RCS Pressure-Variable Low Trip Setpoint of (13.00 T F-5927) psig. j out
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Nuclear Over ower Based on Pumo Monitors In conjunction with the power / imbalance / flow trips the Nuclear Over-powcr Based Gn Pump Monitors trip prevents the minimum core DNBR from decraasing below 1.30 by tripping the reactor due to the loss.of reactor coolant pu=p(s). The pump monitors also restrict the power level for the nu:aber of pu:ps in operation.
THREE MILE ISLAND - UNIT 2 B 2-6 Amendment No. 6 J
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TABLE 3.3-1 (Continued)
TABLE NOTATION
- With the control rod drive trip breakers in the closed position and the control rod drive systen capable of rod withdrawal.
- When Shutdown Bypass is actuated.
!The provisions of Specification 3.0.4 are not applicable.
!!High voltage to detector may be de-energized above 10-10 amps on both Intermediate Range channels.
(a) Trip ray be manuelly bypassed when RCS pressure 5,1820 psig by l
actuating Shutdo$<n Bypass provided that:
(1)
The Nuclear Overpower Trip Setpoint is < S% of RATED THERMAL PUdER.
(2)
The Shutdewn Bypass RCS Pressure--High Trip Setpoint of 5,1820
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psig is icposed.
(3)
The Shutdesn Bypass is removed when RCS pressure > 1900 psig.
l (c) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.
ACTION STATEMENTS ACTION l With the number of channels OPERABLE one less than required by the Minimum Channels OPERASLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or he in at least HOT STANDSY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or cpen the control rod drive trip breakers.
ACTION 2 With the number of OPERASLE channels one less than the Total Number of Channels STARTUP and/or POWER OPERATION cay proceed provided all of the following conditions are satisfied:
a.-
The inoperable ch;annel is placed in the tripped condition within one hour.
b.
. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance tes. ting per Specification 4.3.1.1.1, i
l THREE MILE ISLAND - UNIT 2 3/4 3-3 Amendrent No. 6 M
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' TABLE NOTATION Trip function may be bypassed in this MODE with RCS pressure below 1920 psig. Bypass shall be autocatically removed when RCS pressure exceeds 1950 psig.
3 channels per Automatic Actuation Logic, Each R. B. Pressure High Channel tri as one Safety Injection Channel and one R. B. Cooling &
Isolation C1annel.
3 channels per Automatic Actuation Logic, R. B. Spray Valves are actuated by R. B. Cooling and Isolation.
- Trip function may be bypa', sed in this code with steam generator pressure 1 800 psig.
Bypass shall be removed when steam generator pressure > 800 psig.
The provisions of Soccification 3.0.4 are not applicable.
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THREE MILE ISLAND - UNIT 2 3/4 3-13 Amendment No.' 6
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UNITED STATES NUCLEAR REGULATORY COMM!ss;ON I
W ASHINGToN. D. C. 20555 W'
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AUG 171978
%,..... f SAFETY EVALUATION BY THE U N ICE OF NUCLEAR REACTOR REGULATION W
SUPPMllhG Ar2H0? INT NO. 6 TO FACILITY OPERATliiG LICENSE NO. OPR-73
%j f-METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY DOCKET NO. 50-320 THREE MILE ISLAND NUCLEAR STATION, UNIT 2 g,,-
1.
Containment Air Lock Seal Leak Rate Testing g;
M' Introduction By letter dated May 19, 1978 transmitting Technical Specification Change Request No. 009, Metropolitan Edison Company (Met Ed) requested amendment of Appendix A _to Facility Operating License No. CPR-73 for Three Mile Island Nuclear Station, Unit 2 (TMI-2).
lhe requested change would amend
@h the Technical Specifications to permit a more effective method of seal leakage verification.
Discussion The present wording of the THI-2 Technical Specifications requires that contairment air lock seal leak rate testing be performed "by pressure decay when the volume between the door seals is pressurized to > 10 psig..."
The acceptance criteria specified, (Leakage 2 0.01 La) translates to a pressure drop of 10 psig in a period of 10 seconds.
This is inconsistent wit!h the additional requirement to maintain door seals pressurized to > 10 h.-
psig for at least 15 minutes.
In addition because the manufacturer has indicated that the volume between the door seals should not be pressurized above 10 psig, and because the volume between the door seals is quite small (s 0.02 cu. ft.), Met Ed states that it is not possible to perform the surveillance using the pressure decay method. The proposed wording of the Technical Specifications would allow measurement of leak rate testing by another method (e.g., the flow monitoring method). This proposed char.ge deletes the requirement to measure seal leakag.e by a pressure drop test method, and specifies the pressure at which the seal leak rate is to be N'.
detemined using a flow meter.
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The basis of the surveillance requirement is to provide assurance that the containment leakage rates of Limiting Condition for Operation 3.6.1.2 are E
pK not exceeded as a result of seal damage occurring during door usage.
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DUPLICATE DOCUMENT 78
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,' The proposed. Technical Specification Change does not change the limit M
established for allowable leakage through the door seals.
This limit k
remains j 0.01 La (1 percent of the total allowable coatainment leak f4 rate).
Evaluation We have reviewed the infomation provided by the licensee as well as additional information from our Office of Inspection and Enforcement and find that the proposed method of seal leakage measurement is effective h.
(in fact, for this particular application, it is more et 7 ative than that required by the present Technical Specifications), satisfies the intent mA and the basis of the Te:hnic al Specification, and therefore provides equal or greater assurance that seal leakage will be within acceptable limits.
Based on the above, we conclude that the proposed change permitting an alternate method of measuring containment air lock seal leakage is acceptable, and that the facility operating license can be amended by changing the Technical Specifications as shown in the attachment to this license amendment.
2.
Ultimate Heat Sink Temperature Introduction w
By letter dated June 5,1978 transmitting Technical Specification Change M-Request No. Oil, Metropolitan Edison Company (Met Ed) requested amendment 0
of Appendix A to Facility Operating License No. DpR-73 for Three Mile L
- Island Nuclear Station, Unit 2 (TMI-2).
The requested change would amend the-Technical Specifications to permit plant operation with temperature of the Susquehanna River (the ultimate heat sink) in excess of the present W@
limit of 85 9.
Discussion
.i The present Technical Specifications require shutdown of TMI-2 within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> if heat sink (Susquehanna River) temperature exceeds 850F. Past a2 operating experience shows that river water temperature may exceed 85 9 TAS-during the suncer months.
The licensee has stated that w.th one exception, sufficient margin exists
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in the design of all safety-related equipment which would reject heat to W'.
the ultimate heat sink, such that the equipment will operate acceptably
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with the heat sink temperature up to 95cF.
The exception is the control e
building air conditioning equipment, which can operate satisfactorily 1
with heat sink temperatures up to 900F.
With increased flow, which will V.
be available after replacement of the control building booster pump impellers, L==.-
this equipment will also operate satisfactorily at heat sink temperatures up to
&h 950F.
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.The licensee has provided results of their analyses and additional information to confirm that each safetyrelated component will operate within its design pJ.:.
parameters and will not suffer degradation of perfomance or impairment in perfomance of its safety function for tise proposed increased heat sink temperatures.
C Evaluation Based on our evaluation of the results of the licensee's analyses of component perfomance and on our calculations and estimates of system perfomance basted on component design temperatures, we find reasonable assurance that safety-related canponents will operate within their design parameters and will not suffer impaiment in perfomance of their safety functions at the proposed increased ultimate heat sink temperatures. We therefore find that operation at the. proposed heat sink temperatures will not cause a significant decrease in the perfomance margins of safety-related systems, and that such operation is acceptable.
Based on the above, we conclude that the facility operating license can be amended by changing the Technical Specifications as shown in the attachment to this license amendment.
yg 5.
3.
Orifice Rod and Burnable Poison Rod Assemblies L'(s c/
Introduction M '
By _ letter dated July 7,1978 transmitting Technical Specification Change a.o Request No. 014, Metropolitan Edison Company (Met Ed) requested amendment
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of Appendix A to Facility Operating License No. DPR-73 for Three Mile Island Nuclear Station, Unit 2 (THI-2).
The requested changes would amend F
the Technical Specifications to permit removal of all but two orifice D
rod assemblies (0RA's) and installation of retainers on the remaining two ORA's and on the burnable poison rod assemblies (BPRA's). These changes are proposed because of the concern over wear of the fuel assembly holddown latch assemblies as found in other plants, caused by levitation and vibration of the ORA.'s and BPRA's.
Additional changes covered by this change request are the following, which which are not related to ORA removal or BPRA installation:
hs Increase in RCS pressure - low trip setpoint by 100 psig and
.CV corresponding increase in the high pressure trip during startup b.1 in shutdown bypass.
kEk Correction of rod bow penalty to correctly reflect the NRC rod bow model.
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Addition of allowable values for Channel Functional Test to pasa account for instrument errors.
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'em Discussion
- NN' ORA'.s had been provided in guide tubes not containing control rod assemblies or axial power shaping rod assemblies to limit reactor coolant bypass flow through otherwise empty guide tubes. BPRA's are used to provide partial control of slowly ocurring. negative reactivity changes and to flatten the radial power distribution.
A burnable poison rod assembly (BPRA) was ejected from the core at one of B&W.'s Mark B plants. B&W analyzed the problem and determined it to
,T be caused by levitation of the BPRAs during four-pump operation and M
subsequent fretting wear in the holddown 1atching mechanism. After the initial inspection of fuel assemblies at the affected plant, B&W also observed visual indications of wear at other plants in the idential latching mechanisms of BPRAs, assemblies that held orifice rods (0 ras), and source or niodified orifice red assemblies (MORAs).
To resolve this problem in TMI-2, the licensee proposes installation of retainers on the BPRA's and.
on two modified ORA's, and removal of the remaining thirty-eight ORA's.
Infomation supporting this proposal, attached or referenced in the submittal of July 17, 1978, includes-m B&W 1etter to IRC dated June 7,1978, Taylor to Varga BAW-1496, "BPRA Retainer Design Report," May 1978 BAW-1497, " Justification for Removal of Orifice Rod Assemblies in Three Mile Island Unit 2, Cycle 1," June,1978.
The submittal states that installation of the retainers would reduce reactor c5 by coolant system (RCS) flow by less than 1 percent and removal of the ORA's
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would increase bypass flow in the hot assembly by 1.6 percent.
To compensate for core flow distribution effects caused by the changes, the licensee proposed increasing the primary system flow rate (flow requirement increase of 2 percent).
The present margin in flow rate between measured and technical specification requirement (5 percent) would be reduced.
Because this operating margin is reduced from 5 percent to 3 percent, flow instrumentation
,v was evaluated to assure that its accuracy is within the range of the margin, fM The flow measurement system and its calibration were identified by the d
licensee 'to be identical to the system for Three Mile Island, Unit 1 which has previously been shown to have a measurement uncertainty of about 1.5 percent.
Q This uncertainty is within the 3 percent margin available.
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gg The limiting fuel assembly does not contain a BPRA during cycle 1 operation.
Though this would further increase flow in the hot assembly, no credit cl was taken for it.
The net effect of the increased flow and bypass penalties is a slight increase in DNBR's.
DNBR-limited transients were reanalyzed considering the increased flow, trip setting adjustments, uncertainties, and rod bow penalties (for cycle 1 f
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a OfiBR criterion of 1.41 accounts for rod bow effects).
DtiBR values NQ of 1.65 for the 4-pump coastdown event and 1.58 for a feedwater temperature N.
decrease event were calculated.
The 1-pump coast-dowa from 4-pump operation Nt was identified to be the most limiting flow transient because it is used to detemine the flux / flow trip set point.
Discussion with B&W indicates that the DNBR 'for the 1-pump coastdown is 1.43.
A retainer device has been designed and tested by B&W to ensure positive holddown of BPRAs, ORAs and MORAs during reactor operation. The design and the test results were reported to tGC in BAW-1495 and the above-referenced "v
letter of June 7,1978.
For continued operation of TliI-2, Metropol' tan Edison Canpany proposes to install the retainer devices on 68 BPRAs and 2 MORAs. All regular ORAs (38) will be renoved from the core.
These changes apply only for the remainder of the current cycle, Cycle 1, at which time BPRAs are usually withdrawn from the core.
The potential consequences of a retainer failure have also been addressed although failure is considered unlikely. The neutronic and thermal-hydraulic consequences are considered small.
Interference with control rod motion, for example, would not, according to analysis of stuck-out control rod gy transients for B&W 177-FA plants, prevent safe shutdown of the plant.
g-P The major concern associated with retainer failure is plant damage, A
primarily in the steam generators, and potential outages for repair.
This g fik damage should be precluded by the Loose Parts Monitoring System (LPMS).
The LPMS is designed to detect a failed retainer in either the reactor vessel or steam generator. Even though the retainer device is designed for only one cycle of operation, B&W has stated that it will recommend that A
surveillance inspections be made following retainer use.
This should N
provide additional confirmation of acceptable operation.
B&W has also
-N stated that definite plans regarding surveillance will be provided to tRC P
as they are fomulated.
k-+-
The first of the additional changes is the increase in RCS Pressure - low trip setpoint from 1800 psig to 1900 psig.
This change is being made for 6
greater operating flexibility and to increase the margin to high pressure injection (HPI) so that a rapid depressurization will not unnecessarily M.
3 -
cause HPI as frequently as would occur with less margin.
As a result 4,.
. of this increase in the RCS pressure - low trip setpoint, it is correspondingly N
necessary to increase the manual bypass by 100 psi to > 1820 psig to b
incorporate 1820 psig as the new high pressure trip during startup in shutdown bypass. This will enable startup to be performed more easily
{.N
((h and will continue to maintain the same margin previously used to allow f
for instrument errors.
W=
m
kg The rod bow penalty has been revised to correctly reflect the NRC rod bow
'ammi model.
The original Technical Specification was prepared using the B&W rod bow model and during investigation into the remov,al of the ORAs was
(*+7, discovered and is corrected herein.
Also incorporated in this change is the addition of the allowable values for the Channel Functional Test which previously were not included in the Technical Specifications. These values have been added to account for instrument calibration error, instrument drift and instrument error.
Evaluation h
WA We have reviewed the effect on core flow of installation of retainers on the MORA!s and BPRA's and on core bypass flow of removal of the ORA's.
Based on our review of the submitted data and on our calculations on similar changes previously approved for Davis Besse Unit 1, we find that the calculated reduction in core flow of 1 percent and increase. in bypass flow of 1.6 percent are reasonable and acceptable.
Based on the similarity of flow instrumentation to that on TMI-1, and our N
previous evaluation of the flow measurement uncertainty for TMI-l, we f
find that Unit 2 flow measurement instrumentation accuracy is expected to be within the 3 percent operating margin between measured and technical
%n specification flow rates. We have reviewed the adequacy of the additional b
2 percent RCS flow to compensate for the 1.6 percent inen.ase in core bypass introduced by the core modifications.
It is estimsted that the RCS flow increase would provide an additional 1.8 percent RCS flow through the core, which is greater than the 1.6 percent reduction because of the.0RA bypass.
Since there is no significant reduction in safety margins,
~
s-e we find the proposed core modifications acceptable.
g.1 We have reviewed the DNBR evaluations for the 4-pump coastdown, the F l feedwater temperature decrease event, and the one-pump coastdown from 4-pump operation, and find that the results for these limiting transients are above the cycle 1 DNBR criterion of 1.41, and are acceptable.
w With regard to the ORA and BPRA retainers, based on (1) design analyses
. M-and test results on the retainer device, (2) analyses which indicate that r'
failure of the retainers, however unlikely, would not prevent shutdown and 9' '
(3) failure detection capability of the Loose Parts Monitoring System, Q
we find that there is reasonable assurance that the retainers will provide k
adequate holddown force on the BPRAs and MORAs and that the proposed use
$1 of the retainer devices in TMI-2 is acceptable.
M.
We have reviewed the proposed increase in the RCS pressure-low trip set
[.-i 4
g point and the associated increase in the high pressure trip during startup, t
and find that since these increases result in the same or larger safety margins, they are acceptable.
l
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han.
%p The original calculations for a rod bowing penalty had been performed with f
.a B&W rod bow model that we found unacceptable. We have verified that the revised rod bow model :s presented in the change request ccnfor:ns with the F
liRC-approved rod bow equation for B&W plants. Therefore, we find this change acceptable.
The allowable values for channel functional test in Technical Specification Table 2.2-1 reasonably account for various instrument errors, and we therefore find these changes acceptable.
qQ In sumary, we have evaluated the proposed changes in Technical Specification g
Change Request tio. 014, and having found them all acceptable, we conclude y
that the faciliuy operating license can be amended by changing the Technical
.i -
Specifications as shown in the attachment to this license amendment.
4.
Main Steam Safety Valves Intreduction
'54 By letter dated July 24, 1978 transmitting Technical Specification Change w,
Request f{o. 015, Metropolitan Edison Company (Met Ed) requested amendment be of Appendix A to Facility Operating License flo. DPR-73 for Three Mile Island fiuclear Station, Unit 2 (TMI-2). The requested change would amend 7,-
the Technical Specifications to permit replacement of the original 12 dual discharge port main steam safety valves with 20 scxnewhat smaller single t
discharge port valves.
{
piscussion u.3 0.%.
During a previous event at THI-2, some of the original main steam safety valves failed to close at an appropriate pressure after actuation. Efforts
[
to modify the valves to eliminate the problem were unsuccessful, and the
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licensee elected to replace the valves. These Technical Specifications changes are required to reflect this design change.
g The new safety valvhs provide a relief capacity of 120 perceat of the total h
secondary steam flow compared with 114 percent provided by the original valves.
g.
The licensee states that all system modifications conform with requirements gj of appropriate. sections of the ASME Code and with criteria previously Pf -
accepted in the Final Safety Analysis Report (FSAR).
@t
>6',
-2 Evaluation
.!e have evaluated the information provided by the licensee and find that since the relieving capacity of the new main steam safety valvas exceeds L.
that originally provided, the proposed change is in the conservative direction and is therefore acceptable.
F l *.' '?
.F4*sessensas2@rawswSc!h;wwwss
3.,
e'.
i Ps We further find acceptable the licensee's statements regarding confonnance y
of all modifications with ASME Code and FSAR. criteria.
l Among the changes proposed by the licensee is a revision of the equation on page B 3/4 7-1 of the Technical Specifications for determining reduced Nuclear Overpower Trip Setpoint for inoperable safety valves.
The proposed equation is essentially identical numerically to tne original, and we do not find sufficient justification to make the proposed change.
- Eff, Based on the above, we conclude that the proposed changes in the Technical g
Specifications covering the new main steam safety valves are acceptable, except as noted above, and that the facility operating license can be f
amended by changing the Technical Specifications as shown in the attachment to this license amendment.
Environmental Consideration We have detennined that the amendment does not authorize a change in i
-/
effluent types or total amounts nor an increase in power level and will
- h..
not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an th action which is insignificant from the standpoint of environmental impact p'
and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be 1:
prepared in connection with the issuance of this amendment.
j Conclusion
..f.
We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable i
assurance that the health and safety of the public will not be endangered gl by operation in the proposed manner, and (3) such activities will be k
conducted in compliance with the Cocmission's. regulations and the issuance EE.,a of this amendment will not be inimical to the common defense and security C
or to the health and safety of the public.
/
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x ev'enA.(V
. 5'i'Ivd,kro et i anager ga M
LightjamerReactorsBranchNo.4
,LightWaterReacto%BranchNo.4 g
Divisio6 of Project Management Division of Project danagement t'
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5; AUG 171978 5
hnb bir.41&nMUd&%M & MNO-@ M M*SNS
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g NUCLEAR REGULATORY COMI Reference 41 7.
E WASHINGTON. D. C. 20555 r
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AUG 171773 Docket No: 50-320 Metropolitan Edison Company 1E BLE C0%
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ATTN:
P.r. John G. Herbein Vice President P. O. Box 542 Reading, Pennsylvania 19503 Gentlemen:
SUBJECT:
TrREE MILE ISU.ND NUCLEAR STATION, UNIT 2 - ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSE The Nuclear f.e9ulatory Coaafssion has issued the enclosed Amendment No. 6 to Facility Csera:ing Licanse No. DPR-73 which is effective as of the date of issuar.ce.
Amendment No. 6 is in response to the following Technical Specification Change Reques:s to acend Appendix A of Facil'ty Operating License DPR-7,3:
Cher.oe hecuest No.
Date 0]9 May 19, 1978
~
011 June 5,1978 01 4-July 7, 1978 015 July 24,1978 Tne amen 6ent consists of the following:
1.
Changes in license Paragraph 2.C.(2), and in Appendix A, Technical Specifica:f oru.
We have deterained that Acendment No. 6 does not authorize a change in I
effluent types or total arounts nor an increase in power level and will not result in any significant environmental impact. Having made this deter.aination, we have further concluded that the amendment involves an action which is insignificant frcm the standpoint of environmental impact, and pursuant to 10 CFR 451.5(d)(4), that an environmental impact statement, negative declaration t1d environmental impact appraisal need not be prepared in connec-ion with the issuance of this amendment.
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4 VI30, fl COPY SENT REGION /
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- v Copies of the FEDERAL REGISTER Notice of Issuance and the safety evaluation 5:
supporting !c.endment No. 6 are also enclosed.
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1 Sincerely, E
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k k".\\ r 'g a C e Light Water Reacters Branch 4 E.
Division of Project anagement if
Enclosures:
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1.
Aaendaant No. 6 i
E' 2.
Federal Register Notice E'
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3.
Safety Evaluation cc: w/ enc 1.
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