ML20199E964
| ML20199E964 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 06/18/1986 |
| From: | FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML20199E948 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.2.10, TASK-TM GL-82-16, NUDOCS 8606240159 | |
| Download: ML20199E964 (18) | |
Text
.
ATTAC} MENT A FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 Docket No. 50-302/ License No. DPR-72 Technical Specification Change Request No.145A LICENSE DOCUMENT INVOLVED:
Technical Specification PAGES:
2-2 2-6 B2-6 DESCRIPTION OF REQUEST:
This submittal requests that the specification for the Reactor Protection System Instrumentation be changed to raise the Reactor High Pressure Trip setpoint from 2300 psig to 2355 psig.
REASON FOR REQUEST:
This request is submitted based on conclusion from Babcock & Wilcox (B&W)
Topical Report BAW-1890, " Justification for Raising Setpoint for Reactor Trip on High Pressure."
The f1RC staff has reviewed this report, and in its Safety Evaluation forwarded by letter dated April 22, 1986 from Dennis M. Crutchfield to J.
H.
Taylor, found it to be acceptable for referencing in License Applications.
Raising the Reactor High Pressure Trip setpoint at Crystal River Unit 3 (CR-3) to 2355 psig will return the setpoint to its originally licensed val ue.
This will provide more margin for plant operation, more time for operator action, and thereby potentially avoid some reactor trips t'uring minor plant upsets.
86062 % [g h 2
PDR P
SAFETY EVALUATION OF REQUEST:
Subsequent to the TMI-2 accident the staff required certain changes to reactor protection systems intended to reduce challenges to and opening of the power operated relief valve (PORV).
For B&W reactors those changes were 1) lowering the reactor High Pressure Trip setpoint from 2355 psig to 2300 psig, 2) raising the PORV setpoint from 2250 psig to 2450 psig, and 3) implementing a safety grade automatic Anticipatory Reactor Trip for, among other things, a turbine trip. These modifications were implemented and were accepted as meeting the NRC guidelines that PORV opening should occur less than E% of the time for all anticipated transients, and that the contribution to the probability of a small break LOCA (SBLOCA) from a stuck open PORV is insignificant.
While these modifications have met the objectives of reducing challenges to and opening of the PORV, they have increased the frequency of reactor trips and the attendant challenges to plant safety systems.
The referenced Topical Report provides justification that a number of high pressure transients would not have resulted in a reactor trip if more margin had been available to the High Pressure Trip setpoint.
The analyses presented demonstrate that when Reactor High Pressure Trip setpoint is raised to 2355 psig and +he arming threshold for Anticipatory Reactor Trip on turbin trip is raised to 45% power, a reduction in total reactor trip frequency of about 10% is expected. Reductions in reactor trip frequency will contribute to overall plant safety as well as plant availability.
Furthermore, NRC guidelines regarding the J
PORV, that the probability of SBLOCA due to stuck open PORV must be less than
.001 per reactor year and that less than 5% of the High Pressure Trips are allowed to open the PORV, continue to be met following these changes.
The Topical Report is based on the POWERTRAIN computer code which was previously accepted for use by the NRC staff by letter dated November 28, 1983.
The results of the reported analyses apply to all B&W 177 fuel assembly plants including CR-3.
FPC personnel have reviewed the Topical Report and the NRC's Safety Evaluation and have verified that they are applicable to CR-3.
FPC finds that the return of the high pressure trip setpoint to its originally licensed 1
value of 2355 psig, and the establishment of the arming threshold for ART on reactor trip at 45% reactor power, are safe and acceptable modifications for CR-3.
SHOLLY EVALUATION OF REQUEST:
I Florida Power Corporation proposes that this amendment does not involve a significant hazards consideration.
The Commission has provided guidance concerning the application of standards for determining whether a significant hazards considerations exists by providing certain examples (51 CFR 7751) of amendments that are considered not likely to involve significant hazards consideration.
Example (iv) relates to a relief granted upon demonstration of acceptable operation from an operating restriction that was imposed because acceptable operatio1 was not yat demonstrated.
This assumes the operating restriction and the criteria to be applied to a request for relief have been established in a prior review and that it is justified in a satisfactory way that the criteria have been met.
In this case, the proposed change described above is similar to example (iv).
Raising the reactor high pressure trip setpoint is removal of a NUREG 0737 operating restriction that was imposed to reduce PORV challenges.
Subsequent analyses by B&W have demonstrated operation with the high pressure trip setpoint at its original value of 2355 psig continues to meet the NUREG 0737 Guidelines for PORV challenges while at the same time potentially reducing reactor trips.
The NRC Staff has reviewed the B&W analyses and issued its safety evaluation 4
accepting the analysis results for use in license amendment applications.
The Safety Evaluation is dated April 22, 1986.
FPC personnel have reviewed the B&W analyses and the Staff's Safety Evaluation and have verified they are applicable to CR-3.
FPC has concluded there is no significant hazards considerations involved with raising the reactor high pressure trip setpoint to 2355 psig.
Based on the above, FPC finds that the amendment will not:
1.
Involve a significant increase in the probability or consequence of an accident previously evaluated.
Rigorous analyses applicable to CR-3 have been performed which demonstrate that the guidelines on which the reduction of reactor high pressure trip setpoint was mandated, - will continue to be met at the higher (originally licensed) setpoint.
2.
Create the possibility of a new or different kind of accident from any
{
accident previously evaluated.
This change returns the reactor high pressure trip setpoint to the value with which the plant was initially licensed to operated.
a
]
3.
Involve a significant reduction in the margin of safety.
Reductions j
in %e margin of safety, in the fonn of PORY challenges, have been 3
evaluated and found to be acceptable by B&W and the NRC Staff.
Any increase in PORY challenges is offset by reduced reactor trips and the attendant reduced safety system challenges.
i I
4 i
'i l
4
.m m
. m.
__.___.________,______.__m.m_
~
2400 RCS PRES!'JRE HIGH TRif RC OUTLET TEMP
._g 2200 HIGH TRIP J
ACCEPTABLE 0PERATION 5
=
E 5
2000 SATETY LIMIT 4
i
/
/
Y UNACCEPTABLE OPERAT10N 1800 RCS PRESSURE LOW TRIP i
580 600 620 640 Reactor Outlet Temperature,*f FIGURE 2.1-1 j
REACTOR CORE SAFETY LIMIT CRYSTAL RIVER UNIT 3 2-2 Amendment No. M, #
~. - -
. _ _ = - -
O TABLE 2.2-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPOINTS e
M
]
FUNCTION UNIT TRIP SETPOINT ALLOWABLE VALUES E
H w
8.
Pump Status Based More than one pump drawing More than one pump drawing on Reactor Coolant Pump
< 1152 or > 14,400 kw
< 1152 or > 14,400 kw.
Power Monitors (1) 9.
Reactor Containment Vessel Pressure High 1 4 psig 5 4 psig 10.
Anticipatory Reactor Main Turbine Control Oil Main Turbine Control Oil Trip - Main Turbine (2)
Pressure > 45 psig Pressure > 45 psig
[
11.
Anticipatory Reactor Pump Ccntrol Oil Pump Control Oil Trip - Both Main Pressure > 55 psig Pressure >-55 psig Feedwater Pumps (3)
(1) Trip may be manually bypassed when RCS pressure 11720 psig by actuating Shutdown Bypass provided that:
a.
The Nuclear Overpower Trip Setpoint is < 5% of RATED THERMAL POWER, b.
The Shutdown Bypass RCS Pressure - High Trip Setpoint of f 1720 psig is imposed, and c.
The Shutdown Bypass is removed when RCS Pressure > 1800 psig.
(2) Trip bypassed below 45% of RATED THERMAL POWER.
(3) Trip bypassed below 20% of RATED THERMAL POWER.
~
LIMITING SAFETY SYSTEM SETTINGS i
BASES
=.
The AXIAL POWER IMBALANCE boundaries are established in order to prevent reactor thermal limits from being exceeded.
These thermal limits are either power peaking kw/ft limits or DNBR limits.
The AXIAL POWER IMBALANCE reduces the power level trip produced by the flux-to-flow ratio such that the boundaries of. Figure 2.2-1 are produced. The flux-to-flow ratio reduces the power level trip and associated reactor power-reactor power-imbal ance boundaries by 1.08% for a 1% flow reduction.
RCS Pressure - Low, High, and Variable Low The High and Low trips are provided to limit the pressure range in which reactor operation is permitted.
During a slow reactivity _ insertion startup accident from low power or a slow l
reactivity insertion from 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 l
below the safety limit, 2750 psig, for any design transient.
The RCS Pressure-High trip is backed up by the pressurizer code safety valves for RCS over pressure protection is therefore, set lower than the set pressure for these valves, 2500 psig. The RCS Pressure-High trip also backs up the Nuclear Overpower trip.
The RCS Pressure-Low, 1800 psig, and RCS Pressure-Variable low, (11.59 g
Tout F -5037.8) psig, Trip Setpoints have been established to maintain the DNB ratio greater than or equal to 1.30 for those design accidents that result in a pressure reduction.
It also prevents reactor operation at pressures below the valid range of DNS correlation limits, protecting against DNB.
I Due to the calibration and instrumentation errors, thg safety analysis used a RCS Pressure-Variable Low Trip Setpoint of (11.59 Tout F -5077.8) psig.
l l
CRYSTAL RIVER - UNIT 3 B 2-6
ATTACHMENT B FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 Docket No. 50-302/ License No. DPR-72 Technical Specification Change Request No.145B LICENSE DOCUMENT INVOLVED:
Technical Specifications PAGES:
2-5 3/4 3-2 3/4 3-3 3/4 3-6 3/4 3-7 B2-7 DESCRIPTION OF REQUEST:
This submittal requests that the specifications for the Reactor Protection System Instrumentation be changed to add two new reactor trips. These new trips are:
a)
Articipatory Reactor Trip - both main feedwater pumps, and b)
Anticipatory Reactor Trip - main turbine The Anticipatory Reactor Trip ( ART) on trip of both main feedwater pumps will be armed whenever reactor power is equal to or greater than-20% FULL POWER and the main turbine trip will be armed whenever reactor power is equal to or greater than 45% FULL POWER.
_ REASON FOR REQUEST:
This request is made in response to NUREG-0737, Item II.K.2.10 and Generic Letter 82-16 dated September 20, 1982.
SAFETY EVALUATION OF REQUEST:
Subsequent to the TMI-2 accident the staff required changes to Reactor Protection Systems intended to reduce challenges to and opening of the power operated relief valve (PORV).
Two of those changes required at Crystal River Unit 3 (CR-3) were the establishment of safety grade automatic Anticipatory Reactor Trips ( ARTS) for trip of both main feedwater pumps and for main turbine trip.
These ARTS are intended to anticipate plant transients which may ultimately result in reacter High Pressure Trips and thereby eliminate some PORV challenges.
The proposed Technical Specifications are in accordance with the sampl e Technical Specifications given in Generic Letter 82-16, except for the arming threshold of the turbine ARTS.
The arming threshold for the turbine ARTS is based on Babcock and Wilcox Topical Report, BAW-1893, " Basis for Raising Arming Threshold for Anticipatory Reactor Trip on Turbine Trip."
This Topical Report was reviewed by the NRC and accepted for use in license applications by the staff's Safety Evaluation forwarded by letter dated April 25, 1986, from Dennis i
M. Crutchfield to J. H. Taylor.
FPC personnel have reviewed the Topical Report and the Safety Evaluation and has determined their results to be applicable to CR-3.
As demonstrated in the Topical Report establishing the arming threshold for the ART on turbine trip at 45% FULL POWER with reactor High Pressure Trip set at 2355 psig will continue to meet NUREG 0737 guidelines regarding PORV challenges and PORV opening.
There will be an overall reduction in reactor trips and the attendant challenges to safety systems with these reactor protection setpoints.
These are therefore safe and desirable operating limits for the Reactor Protection System.
SHOLLY EVALUATION OF REQUEST:
Florida Power Corporation proposes that this amendment does not involve a significant hazards consideration.
The Commission has provided guidance concerning the application of standards for determining whether a significant hazards considerations exists by providing certain examples (51 CFR 7751) of
amendments that are considered not likely to involve significant hazards consideration.
Example (ii) relates to a change that constitutes an additional limitation, restriction or control not presently included in the Technical l
Specifications.
In this case the change described above is similar to example (ii).
I J
Adding anticipatory reactor trips ( ARTS) on trip of both main feedwater pumps, or on trip of the main turbine is a reactor control function not presently i
l included in the Technical Specifications. The proposed Technical Specifications are in accordance with the guidance of Generic Letter 82-16 except for the arming threshold of the main turbine ART.
The main turbine ART arming threshold was chosen based on B&W analyses that have been reviewed and accepted by the i
Staff in its Safety Evaluation dated April 25, 1986.
FPC personnel have reviewed the B&W analysis and Staff Safety Evaluation and have verified they are applicable to CR-3.
FPC has concluded there is no significant hazards considerations involved with adding these ART specifications to the CR-3 1
Technical Specifications.
Based on the above, FPC finds that the amendment will not:
i l
1.
Involve a significant increase in the probability or consequence of an accident previously evaluated.
Adding these specifications places an additional restriction on the operation of CR-3 that will shut the reactor down in anticipation of a reactor high pressure condition that j
could exist due to a main turbine trip or both main feedwater pump l
trip.
The ARTS preclude either of these events from producing a i
challenge ta the reactor coolant system PORV.
l 2.
Create the possibility of a new or different kind of accident from any j
accident previously evaluated.
ARTS provide an additional safety i
function, two additional reactor trips, and offer 'no opportunity for creating a new kind of accident.
3.
Involve a significant reduction in-the margin of safety. ARTS provide j
an additional safety function which increases the margin of safety relative to transients with a probability of resulting in an over pressure condition. in the reactor coolant system.
I i
1
I TABLE 2.2-1 w
REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPOINTS is M
FUNCTION UNIT TRIP SETPOINT ALLOWABLE VALUES E
G w
1.
Manual Reactor Trip Not Applicable Not Applicable 2.
Nuclear Overpower
<104.9% of RATED
< 104.9% of RATED THERMAL POWER THERMAL POWER with four with four pumps operating pumps operating
<79.92% of RATED
< 79.92% of RATED THERMAL POWER THERMAL POWER with three with three pumps operating j
pumps operating
'?
3.
RCS Outlet Temperature -
High
<618 F
<618 F i
4.
Nuclear Overpower Trip Setpoint not to exceed Allowable Values not to exceed the l
Based on RCS Flow and the limit line of Figure limit line of Figure 2.2-1 2.2-1 AXIAL POWER )
IMBALANCE (1 l
5.
RCS Pressure - Low (1) i 1800 psig 1 1800 psig
< 2355 psig 6.
RCS Pressure - High
< 2355 psig 1
j 7.
RCS Pressure - Variable 1 (11.59 Tout F - 5037.8)
> (11.G9 Tout F - 5037.8) psig l
Low (1) psig l
l
TABLE 3.3-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION
- )
MINIMUM gj TOTAL NO.
CHANNELS CHANNELS APPLICABLE
-j FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION l
r-i gg 1.
1 1
1, 2 and*
8
!b 2.
Nuclear Overpower 4
2 3
1, 2 2#
3.
RCS Outlet Temperature - High 4
2 3
1, 2 3#
i l
[
4.
Nuclear Overpower Based on RCS 4
2(a) 3 1, 2 2#
Flow and AXIAL POWER IMBALANCE 5.
RCS Pressure - Low 4
2(a) 3 1, 2 3#
6.
RCS Pressure - High 4
2 3
1, 2 3#
7.
Variable Low RCS Pressure 4
2(a) 3 1, 2 3#
8.
Reactor Containment Pressure - High 4
2 3
1, 2 3#
4e 9.
Intermediate Range, Neutron Flux 2
0 2
1, 2 and*
4 A
and Rate
{
10.
Source Range, Neutron Flux and Rate A.
Startup 2
0 2
2 # # and*
5 B.
Shutdown 2
0 1
3, 4 and 5 6
- 11. Control Rod Drive Trip Breakers 2 per trip 1 per trip 2 per 1, 2 and*
7#
system system trip system
- 12. Reactor Trip Module 2 per trip 1 per trip 2 per 1, 2 and*
7#
system system trip system 13.
Shutdown Bypass RCS Pressure - High 4
2 3
2**,
3**,
6#
4**, 5**
1 from 2 2
1,2 25 14.
Reactor Coolant Pump Power Monitors 2 per pump or more pumps (a,b) per pump l
i
- 15. Anticipatory Reactor Trip 4
2(c) 3 1
3#
- 16. Anticipatory Reactor Trip 4 per pump 2 per pump (d) 3 per pump 1
3#
- Both Main Feedwater Pumps
9 TABLE 3.3-1 (Continued)
TABLE NOTATION With the control rod drive trip breakers in the closed position and the control rod drive system capable of rod withdrawal.
When Shutdown Bypass is actuated.
- The provisions of Specification 3.0.4 are not applicable.
I 1
- High voltage to detector may be de-energized above 10.10 amps 'on both Intermediate Range channels.
(a) Trip may be manually bypassed when RCS pressure less than or equal to 1720 psig by actuating Shutdown Bypass provided that:
(1) The Nuclear Overpower Trip Setpoint is less than or equal to 5% of RATED THERMAL POWER, (2) The Shutdown Bypass RCS Pressure--High Trip Setpoint of less than or equal to 1720 psig is imposed, and (3) The Shutdown Bypass is removed when RCS pressure greater than 1800 psig.
(b) Trip may be manually bypassed when reactor power is less than or equal to 2475 MWT and four reactor coolant pumps are operating.
(c) Trip automatically bypassed below 45 percent of RATED THERMAL POWER.
(d) Trip automatically bypassed below 20 percent of RATED THERMAL POWER.
ACTION STATEMENTS l
ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE 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 be in at least HOT STANDBY within the next six hours and/or open the control rod drive trip breakers.
ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channel s,
STARTUP and/or POWER OPERATION may i
proceed provided all of the following conditions are satisfied:
i a.
The inoperable channel is placed in the tripped condition within one hour.
I b.
The Minimum Channels OPERABLE requi rement is met; however, one additional channel may be bypassed for up to two hours for surveillance testing per specification 4.3.1.1.
1 CRYSTAL RIVER - UNIT 3 3/4 3-3
TABLE 3,3-2 REACTOR PROTECTION SYSTEM INSTRUMENTATION RESPONSE TIMES 2
vi jy Functional Unit Response Times B
{g 1.
Manual Reactor Trip Not Applicable
]
E 2.
Nuclear Overpower *
~-< 0.266 seconds
- 1 3.
RCS Outlet Temperature - High Not Applicable u,
4.
Nuclear Overpower Based on RCS Flow and j
AXIAL POWER IMBALANCE *
< 1.842 seconds 5.
RCS Pressure - Low
< 0.44 seconds 6.
RCS Pressure - Hi'gh
< 0.44 seconds a
7.
Variable Low RCS Pressure Not Applicable N
8.
Pump Status Based on RCPPMs**
< 1,44 seconds 9.
Reactor Containment Pressure - High Not Applicable 10.
Anticipatory Reactor Trip -
Not Applicable Main Turbine i
11.
Anticipatory Reactor Trip -
Not Applicable Both Main Feedwater Pumps I
Neutron detectors are exempt from response time testing.
Response time of the neutron flux signal portion of the channel shall be measured fran detector output or input of first i
electronic component in channel.
Time response testing of the RCPPM's may exclude testing of the current and voltage sensors i
and the watt transducer.
TABLE 4.3-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS l
E}
CHANNEL MODES IN WHICH -
Di 4
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED E3 1.
Manual Reactor Trip N.A.
N.A.
S/U(1)
N.A.
5
]
2.
Nuclear Overpower S
D(2) and Q(7)
M 1,2 l
z
- 3 3.
RCS Outlet Temperature--High S
R H
1,2 w
4.
Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE S(4)
M(3) and Q(7,8)
M 1,2 i
5.
RCS Pressure--Low S
R H
1,2 6.
RCS Pressure--High S
R M
1,2 7.
Variable Low RCS Pressure S
R H
1,2 ff 8.
Reactor Containment Pressure--High S
R M
1,2 a
9.
Intermediate Range, Neutron 02 4
Flux and Rate S
R(7)
S/U(1)(5) 1, 2 and*
l 10.
Source Range, Neutron Flux and Rate S
R(7)
S/U(1)(5) 2,3,4 and 5
- 11. Cons. :1 Rod Drive Trip Breaker N.A.
N.A.
M and S/U(1) 1, 2 and*
l
- 12. Reactor Trip Module N.A.
N.A.
M 1, 2 and
- l l
13.
Shutdown Bypass RCS Pressur.e--High S
R M
2**,3**,4**,5**
l 14.
Reactor Coolant Pump Power Monitors S
R(9)
M 1,2 l
- 15. Anticipatory Reactor Trip - Main Turbine S
R M
1 16 Anticipatory Reactor Trip - Both Main Feedwater Pumps S
R M
1
i LIMITING SAFETY SYSTEM SETTINGS BASES
=
- = - -
Reactor Containment Vessel Precsure - High The Reactor Containment Vessel Pressure-High Trip Setpoint less than or equal to 4 psig, provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the containment vessel or a loss-of-coolant accident, even in the absence of a RCS Pressure - Low trip.
Reactor Coolant Pump Power Monitors In conjunction with the power / imbalance / flow trips, the Reactor Coolant Pump Power Monitors trip prevents the minimum core DNBR from decreasing below 1.30 oy tripping the reactor due to more than one reactor coolant pump not operating.
A reactor coolant pump is considered to be not operating when the power required by the pump is greater than or equal to 262% (14,400 kw) or is less than or equal to 20.9% (1152 kw) of the operating power (5500 kw).
In order to avoid spurious trips during normal operation, the trip setpoints have been selected to maximize the operating band while assuring that a reactor trip l
will occur upon loss of power to the pump.
The 20.9% trip setpoint and response time are based on the maximum time within which an RCPPM-RPS trip must occur to provide DNBR protection for the four pump coastdown.
Florida j
Power has agreed to take credit for the pump overpower trip in order to assure that certain potential faults (such as a seismically induced fault high signal) will not prevent this instrumentation from providing the protective action (i.e., a trip signal).
Thus, the maximum setting, approximately 262%
(14,400 kw), was selected.
Anticipatory Reactor Trips The " Main Turbine" and both " Main Feedwater Pump" Anticipatory Reactor Trips are. intended to reduce the consequences of undercooling transients that result in a pressure increase in the reactor coolant system.
The trips " anticipate" a certain class of pressure increasing transients (i.e., loss of heat sink on the secondary side), and therefore reduce challenges to the PORV.
The Main Turbine is considered to be not operating when the turbine control oil pressure monitor indicates less than or equal to 45 psig.
A Main Feedwater pump is considered to be not operating when the pump control oil 1
pressure monitor indicates less than or equal to 55 psig.
4 CRYSTAL RIVER - UNIT 3 B 2-7
r j
STATE OF FLORIDA COUNTY OF PINELLAS Rolf C. Widell states that he is the Manager, Nuclear Operations Licensing and Fuel Management for Florida Power Corporation; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statemerits made and matters set forth therein are true and correct to the best of his knowledge, information, and belief.
. r\\ f l
k Rolf C. ' (idell N
Manager$uclear Operations Licensing and Fuel Management
^ Subscribed and sworn to before me, a Notary Public in and for the State and County above named, this 18th day of June,1986.
af Notary Ip6blic
/
Notary Public, State of Florida at Large, My Commission Expires: July 22,1989
r UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN THE MATTER OF
)
)
DOCKET No. 50-302 FLORIDA POWER CORPORATION
)
CERTIFICATE OF SERVICE Rolf C. Widell deposes and says that the following has been served on the Designated State Representative and the Chief Executive of Citrus County, Florida, by deposit in the United States mail, addressed as follows:
- Chairman, Administrator Board of County Commissioners Radiological Health Services of Citrus County Department of Health and Citrus County Courthouse Rehabilitative Services Inverness, FL 32650 1323 Winewood Blvd.
Tallahassee, FL 32301 A copy of the request for amendment to Operating License No. DPR-72 (Change Request No.145).
FLORIDA POWER CORPORATION
.A I
kr 4
Rolf C. Tidell
\\
Managej) Nuclear Operations Licensing and Fuel Management SWORN TO AND SUBSCRIBED BEFORE ME THIS 18th DAY OF JUNE 1986.
/
Notary EFublic i
Notary Public, State of Florida at Large notaar rustic start or rtcaron My Commission Expires:
nr canissica exP auty 22.1989 an:to inau stardat Ins. uta.
l (NOTARIAL SEAL)
Y 1
- - - - - -