ML20245G017
| ML20245G017 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  (NPF-03-A-131, NPF-3-A-131) |
| Issue date: | 04/25/1989 |
| From: | Hannon J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20245G021 | List: |
| References | |
| NUDOCS 8905030051 | |
| Download: ML20245G017 (9) | |
Text
_ _ _ _ _ - - _ - _ _ _ - _ _ _ - _ _ -
4
/pt,* nicoq'o, UNITED STATES 4
< ',g NUCLEAR REGULATORY COMMISSION
?
3 3
1 E WASHINGTON, D. C. 20555 f
o
%...../
TOLED0 EDISON COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION. UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.131 License No. NPF-3 r
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Toledo Edisen Company and The Cleveland Electric Illuminating Company (the licensees) dated May 4, 1987 as modified April 29, 1988 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate In conformity with the application, 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 cmendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common 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 Commiss%n's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NFF-3 is hereby i
amended to read as follows:
1 3905030051g904b46 o
PDR ADOCK pgg
?
E i
t l'
i 1
2 (a) Technical Specifications The. Technical Specifications contained in Appendices A ana B, as revised through Amendinent No.131, are hereby incorporated in the license. The Toledo Edison Company shall operate the facility in..
dCCordahCe With the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall-be implemented not later than June 9,1989.
FOR THE NUCLEAR REGULATORY COMMISSION t/ v n y /p M
(.
A - John N. Hannon, Director d
Project Directorate III-3 Division of Reactor Projects - III, IV, V, & Special Projects Office of Nuclear Reactor Regulation 1
Attachment:
Changes to the Technical Specifications Date of Issuance: April 25, 1989 1
4 I
l l
1 1
i l
l
4 ATTACHMENT TO LICENSE AMENDMENT NO.131 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identifica by amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to raintain document completeness.
Remove Insert 3/4 7-4 3/4 7-4 3/4 7-5 3/4 7-5 3/4 7-Sa 3/4 7-Sa B 3/4 7-la B 3/4 7-la I
r I
1 1
1
a
\\',
a, W.c
)%
1 ev i
l
(
av G
N e
I h
T g
g t
T i
i E
s s
f S
p p
o S
w-G T
0 0
s N
F 5
0 n
I I
0 1
o T
L 1
1 i
t i
E i
S dn T
o F
c I
L tn E
e 1
V i
L b
7 A
m 5
V a
4 Y
oe e
T t r l
E u
b F
d s a
A ns T
S oe pr E
sp N
e I
R rd L
O rn T
oa M
A c
A R
e E
E l r w
T N
l u S
E at G
h a N
s r I
M e
A A
ep M
E r m T
ue S
st s
R eg E
rn P
pi t
R 2
7 ga E
nr B
i e M
t p U
t o N
a b
e sl a t n i m fi l on e
ht Ta s
i t E a a C g. E we w Rs. Yw
@a&ar Z*.
C.
~
PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2.
Two independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3.
ACTION:
a.
With one Auxiliary Feedwater System inoperable, restore the inoperable system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within i
the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
With any Auxiliary Feed Pump Turbine Inlet Steam Pressure Interlocks inoperable, restore the inoperable interlocks to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
l' SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each Auxiliary Feedwater dysten naii ve demonstrated OPERABLE:
l a.
At least once per 31 days on a STAGGERED TTST BASIS by:
1.
Verifying that each steam turbine driven pump develops a differen-tial pressure of > 1070 psid on recirculation flow when the secondary steam supply pressure is greater than 800 psia, as measured on PI SP 128 for purrp 1-1 and PI SP 12A for pump 1-2.
The provisions of Specification 4.0.4 are not applicable for entry into Mode 3.
l 2.
Verifying that each valve (power operated or automatic) in the flow path is in its correct position.
3.
Verifying that all manual valves in the auxiliary feedwater pump suction and discharge lines that affect the system's capacity to deliver water to the steam generator are locked in their proper position.
4.
Verifying that valves CW 196, CW 197, FW 32, FW 91 and FW'106 l
are c1rsed.
b.
At least once per 18 mc'ths by:
1.
Verifying that each automatic valve in the flow path actuates to its correct position on an auxiliary feedwater actuation test signal.
2.
Verifying that each pump starts automatically upon receipt of an auxiliary feedwater actuation test signal. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
DAVIS-BESSE, UNIT 1 3/4 7-4 Amendment No. 9E.122,131
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 3.
Verifying that there is a flow path between each auxiliary feedwater pump and each steam generator by pumping water from the Condensate Storage Tank to the steam generator.
The flow path to the steam generator shall be verified by either steam generator level change or Auxiliary Feedwater Safety Grade Flow Indication.
Verification of the Auxiliary Feedwater System's flow capacity is not required.
c.
The Auxiliary Feed Pump Turbine Steam Generator Level Control System shall be demonstrated OPERABLE by performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a CHANNEL FUNCTIONAL TEST at least once per 31 days, and a CHANNEL CALIBRATION at least once per 18 months.
d.
The Auxiliary Feed Pump Suction Pressure Interlocks shall it demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days, and a CHANNEL CALIBRATION at least once per 18 months.
e.
After any modification or repair to the Auxiliary Feedwater System that could affect the system's capability to deliver water to the steam generator, the affected flow path shall be demonstrated available as follows:
1.
If the modification or repair is downstream of the test flow line, the auxiliary feed pump shall pump water from the Condensate Storage Tank to the steam generator; and the flow path availability will be verified by steam generator level change or Auxiliary Feedwater Safety Grade Flow Indication.
2.
If the modification or repair is upstream of the test flow line, the auxiliary feed pump shall pump water through the Auxiliary Feedwater System to the test flow line; and the flow path availability will be verified by flow indication in the test i
flow line. (see note below)
This Surveillance Testing shall be performed prior to entering MODE 3 if the modification is made in MODES 4, 5 or 6.
Verification of tha Auxiliary Feedwater System's flow capacity is not required.
l Note:
When conducting tests of the Auxiliary Feedwater System in MODES 1, 2, and 3 whi.h require local manual realignment of valves that make the system inoperable, a dedicated individual shall be stationed at the valves (in communication with the control room) l able to restore the valves to normal system GPERABLE status.
I 9
DAVIS-BESSE, UNIT 1 3/4 7-5 Amendment No. /3,63,96,722,131
i 1
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) f.
Following each extended cold shutdown (> 30-days in MODE 5), by:
j 1.
Verifying that there is a flow path between each auxiliary feedwater pump and each steam generator.by pumping Condensate Storage Tank water to the steam generator.
The flow path-to the steam generator. shall be verified _ by either steam generator level change or Auxiliary Feedwater Safety Grade Flew Indication.
Verification of the Aur.iliary Feedwater System's flow-capacity is not required.
4.7.1.2.2 The Auxiliary Feed Pump Turbine Inlet Steam Pressure Interlocks shall be demonstrated OPERABLE when the steam line pressure is greater than
'r 275 psig, by performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days, and a CHANNEL CALIBRATION at least once per 18 months. The CHANNEL FUNCTIONAL TEST shall-be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding 275 psig during each plant startup, if the test has not 'been performed within the last 31 days, l
DAVIS-BESSE, UNIT 1 3/4 7-Sa Amendment No. pg,131
I l
3/4.7 PLANT SYSTEMS I
BASES
)
3/4.7.1 TURBINE CYCLE
{
l 3/4.7.1.1 SAFETY VALVES
!j The OPERABILITY of the main steam line code safety valves ensures that the l
secondary system pressure will be limited to within 110% its design pressure of 1050 psig during the most severe anticipated system opera-tional transient. The maximum relieving capacity is associated with a
}
turbine trip from 100% RATED THE!UfAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).
The safety valve set pressures and relieving capacities are in accordance e
with Section III of.ASHI Boiler and Pressure Vessel Code, 1971 Edition.
The Code requires the following:
1.
At le'ast two pressure relief valves are required to provide relieving capacity for steam systems.
2.
The capacity of the smallest pressure-relief valve shall not be less than 50 percent of.that of the largest pressure-relief device.
3.
The set pressure of one of the pressure-relief devices shall not be greater than the maximum allowable working pretsure of the system at design temperature.
4.
Total rated relieving capacity of the pressure-relief devices shall prevent a rise in pressure of more than 10 percent above
(
system design pressure at design temperature under any pressure j
transients anticipated to arise.
I These requirements are, respectively, met as follows:
1.
Nine safety valves are instal 3ed per steam generator.
2.
The relief capacity of two of the cine safety valves per steam I
generator is 583,574 lbs/hr each, and the capacity of the remaining seven is 845,759 lbs/hr each.
i 3.
A minimum of two OPERABLE safety valves per steam generator, with a combined total relief capacity of at least 1,167,148 lbs/hr, one with a setpoint not greater than 1050 psig (+/-1%),
l and one with a setpoint not greater than 1100 psig (+/-1%).
4.
The total relieving capacity of all safety valves on both main steam lines is 14,175,000 lbs/hr which is 120 percent of the total
,tondary system flow of 11,760,000 lbs/hr at 100 percent of rat thermal power. A maximum safety valve setpoint pressure of 1100 psig (+/-1%) assures main steam system pressure remains below 110 percent, or 1155 psig.
DAVIS-BESSE, UNIT 1 B 3/4 7-1 Atendment No. 48.117 l
1
3/4.7 PLANT SYSTEMS BASES I
I 3/4.7.1.1 SAFETY VALVES (Continued) l l
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable i
within the limitations of the ACTION requirements on the basis of the l
reduction in secondary system steam flow and THERMAL POWER required by the i
reduced reactor trip settings of the High Flux channels.
The reactor trip setpoint reductions are derived on the following bases:
1 (X) - (Y)(V)
Sp, xW 2
where:
SP = recuced Trip Eetpoint in percent of RATED THERMAL POWER (Not to Exceed W)
V = maximum number of inoperable safety valves per steam generator W = High Flux Trip Setpoint for four pump operation as specified in Table 2.2-1 X = Total relieving capacity of all safety valves per steam generator in lbs/ hour, 7,087,500 lbs/ hour Y = Maximum relieving capacity of any one safety valve in lbs/ hour, F45,759 lbs/ hour 2 = Required relieving capacity per steam generator in lbs/hr, 6,585,600 lbs/hr.
3/4.7.1.2 AUXILI ARY FEE 0!!ATER SYSTEMS The OPERABILITY of the Auxiliary Feedwater Systems ensures that the Reactor Coolant System can be cooled down to less than 280*F from normal operating conditions in the event of a total loss of offsite power.
The OPERABILITY of the Auxiliary Feed Purrp Turbine Inlet Steam Pressure Interlocks is required y
only for high energy line break concerns and does not affect Auxiliary Feedwater System OPERABILITY.
I Each steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 600 gpm at a pressure of 1050 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate j
feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than.280"F where the Decay Heat Removal j
System may be placed in operation.
l
)
i 1
DAVIS-BESSE, UNIT 1 B 3/4 7-la Amendment No. JJ7,JZL 131 Nert page is 3 3/4 7-2 1
b