ML20070L211
| ML20070L211 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 02/28/1991 |
| From: | Hagan J, Zabielski V Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9103190346 | |
| Download: ML20070L211 (18) | |
Text
..
t wp pgggg Public Servico Electric and Oas Company P O. Dox 236 hancocks Bridge, Now Jon,ey 08030 Hope Crook Generating Station March 15, 1991 U. S. Nuclear Regulatory Commission Document Control Desk-
-Washington, DC 20555.
Dear Sir MONTHLY OPERATING REPORT HOPE CREEK GENERATION STATION UNIT 1 DOCKET No.- 50-354 In compliance:with Section 6.9, Reporting Requirements for the Hope. Creek Technical-Specifications, the operating.
statistics for February are being forwarded to you with the summary of changes, tests, and experiments-for February _1991
, pursuant to the. requirements of 10CFR50.59 (b).
i 3
Sincerely yours, o
ayAO % AMH J. J.' Hagan-General' Manager -
-Hope Creek Operations L
RARild Attachments C. Distribution 9103190346 910228 PDR ADOCK 05000354
?...
"fhh l
The Energy People i
fj, m auworn
~
. _ _. -..... _ _.. _. _ _.. _ _ _. _ _ ~ _ _. -.....
l
-l
- =,
c. --
l
.1 l
INDEX-q NUMBER-4-
SECTION OF PAGES
--Average-Daily Unit' Power Level..
1 Operating Data Report.
2
~
Refueling Information.
l'
- Monthly operating Summary.
11 Summary of Changes,; Tests, and Experiments.
.11 l
l l -.
9
?
1 w
I ;g. T t
t T
J
n.
g
,,I
?+:
+.
.g W,,
- AVERAGE UnIL'." UNIT POWER LEVEL
($;
r i
DOCKET NO.
50-354 UNIT LLQRR,, creek DATE-3/15/91-COMPLETED BY V.
Zabielski-TELEPHONE (609) 339-3506'
-MONTH February 1991-DAY' AVERAGE DAILY-POWER LEVEL DAY AVERAGE DAILY POWER LEVEL'-
-(MWe-Net)
(MWe-Net) 1.
lac 17. -
H
- 2. t A
18.
2 3.-
'2-19.
11
'4i-A-
20.
2 5.-
=A:
21.-
in
'-6..
2 2 2. --
J111
-7, p
~23.
115
.),-
<g.
24.
LA 9.1 A
-25.
A 1
- 10.
12 --
2 6. -
- a-
=
i p
- 11..
l2 --
27.
- g --
- 12.
A 28.-
-A
'13.
-A.
29.
M
.y
[14.
-A=
30.
. H],4-15..
A 31' U~
= 16. --
9.
I..
L we,+
-o.
s'.,,
,,,y
,~ e
-m..
P 6
-l
\\
OPERATING DATA REPORT DOCKET NO.
50-354
{
UNIT Eqpe Creek -
DATE 3/15 i
COMPLETED BY V.
Za 1 ski TELEPHONE- (609) 339-350%
' OPERATING STATUS
- 1. _ Reporting: Period-February 1991-Gross Hours in Report Period 52?
".=
2..
Currently-Authorized Power Level (MWt) 3293
-Max. Depend. Capacity 4'MWe-Net) 1031 Design Electrical Rat tg (MWe-Net) 1067 l
3.- LPower Level to which r'stricted (if any) (MWe-Net)
None
-4.-_
Reasons for restriction (if any)-
j Month CR femulative i
. 5. -- No. of hours reactor was critical li5d 146.0 29.927.5 6.--Reactor' reserve shutdown hours M
M a
j 17_.;-
Hours-generatorJon line=
80.7 80.7 29,373.8 8.'
Unit reserve shutdown hours M
M M
a 9,
Gross thermal energy generated 101,544-101.5G 92.643.95f l (MWH) _
.-10.: Gross electrical-energy 25,250 25.250 30.646.923-
- generated-(MWM)
- 11. Net' electrical' energy generated 14.429 6.141 29;262.825
_- j
'(MWH);
^
12.LReactor service factor 21.7 10.3 81.4 13.cReactor. availability factor 21.7.
10,3 81.4
'14."Unitiservice factor 12.0
. 5_.2, 79.9 4
1$.-Unit: availability factor 1M 54 79.9
- 16. Unit; capacity-factor (using-MDC) 2d
'M 22d.
17.-Unit capacity factor M
M 74.6
-(Using Design MWe)
- 18.-' Unit forced; outage rate 42d 67.1 sd-19.-Shutdowns scheduled over next 6 months (type, date, & duration)
None
- 20. If. shutdown at end of report period, estimated date-of start-up:
gl 3/2/91 1
I
OPERATING DATA REPORT Ul' ?"
3IUTDOWNS AND POWER REDUCTIONS DOCKET Nt 50 354 UNT'r llooe creek i
be." E
[3/15/91
~
1 COMPLETED L.Y L_Zpbielski J
TEJ.EP110NE 1609) 338-3506 MONTH February 1991 I
)8ETi!OD OF 6rIUTTING DOWN THE TYPE REACTOR OR F= FORCED DURATION REASON REDUCING CORRECTIVE
+
NO.
DATE
'S= SCHEDULED (HOURS)
(1)
POWER (2)
ACTION / COMMENT 6 I
S 431 C
4 3rd Refueling Outage 2
2/1 3
2/19 F
44 A
3 Failure of Reactor Level Control LER 354/91-005 4
2/23 F
120.5 A
9 Shutdown to Fix H 2 Leak on Main
=
Generator i
Summary k
l
I REFUELING INFORMATION DOCKET NO.
30-3S4 UNIT uppe Creek DATE
_2 /15 / 9 :
COMPLETED BY Su11011:.ngsworth TELEPHONE (609) 339-1051 MONTH February 1991 1.
Refueling information has changed from last month:
Yes X
No 2
Scheduled date for next refueling:
9/6/92 3.
Scheduled date for restart following refueling:
10/21/92
-4.
A.
Will Technical Specification changes or other license amendments be required?
Yes No X
B.
Has the reload fue"
.esign been reviewed by the Station Operating Review Cunmittee?
Yes No X
If no, when is it scheduled?
not currentiv scheduled 6.
Scheduled date(s) for submitting proposed licensing action:
Elh 6.
Important licensing considerations associated with refueling:
- Amendment 34 to the Hope Cree % Tech Specs allows the cycla specific operating limits to be incorporated into the CORE OPEsATING LIMITS REPORT; a submittal is therefore not required.
j 7.
Number of Fuel Assemblies:
A.
Incore 211 B.
In Spent Fuel Storage (prior to refueling) 121 C.
In Spent Fuel Storage (after refueling) 212 8.
Present licensed spent fuel storage capacity:
4006 Future spent fuel storage capacity:
4006 9.
Date of last refueling that can be discharged July 22. 2007 to spent fuol pool assuming the present licensed capacity:
1
l l
HOPE CREEK GENERATING STATION MONTHLY OPERATING
SUMMARY
FEBRUARY 1991 At the beginning of February, Hope-Creek remained shutdown for the third refueling outage.
The outage concluded on February 18th after a duration of 54.9 days.
On February 19th at 10s10 AM, the reactor automatically scrammed cn low level due to a relay failure in a Startup Level Control Valve.
The unit was again synchroniten to the grid on February 21st.
On February 23rd, the unit was shutdown to repair a hydrogen leak on the Main Generator.
The unit remained shutdown at the end of the month.-
s b
I I
i i
i 4
SUMMARY
OF C)lANGES, TESTS, AND EXPERIMENTS FOR Tile ilOPE CREEK GENERATING STATION i
FEBRUARY 1991 t
]
- _. _ _,. = _.
The following Design Change Packages (DCP's) have been evaluated to determine 1.
If the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or 2.
If a possibility for an accident or malfunction of a different type than any evaluated previously in the ssfety analysis report may be created; or 3.
If the margin of safety as defined in the basis for any technical specification is reduced.
The DCP's did not creat's a new safety hazard to the plant nor did they affect the safe shutdown of the reactor.
The DCP's did not change-the plant effluent releases and did not alter the existing environmental impact.
The Safety Evaluations determined that no unroviewed safety or environmental questions are involved.
i
1 QqE Descriotion of Desian Chapae PackaSA 4EC-0032 This DCP replaced the alternate feeds for the
'A' and
'B' Reactor Protection System Buses.
The new regulating transformers will suppress voltage dips caused by the starting of large motors.
4EC-1002/15 This DCP added a Control Rod Drive Storage Rack, a radiation shield wall for the storage rack an extension to the existing access platform In the Control Rod Drive Removal Area, and a non-lt electrical power supply for the Control Rod Drive Storage Rack.
It also relocated an overflow drain line and capped an equipment drain.
This DCP will enhance the Control Rod Drive removal and assembly cleanup offorts and provide storage for the control Rod Drive Assemblies prior to rebuild activities.
4EC-3010/01 This DCP added manual isolation valves to the Chilled Water System Containment Isolation Valves.
It also added vent and drain valves.
Thin DCP will permit Inservice Inspection testing of the containment Isolation Valves in the Loss of Coolant Accident direction.
4EC-3061 This DCP relocated thermocouples inside the drywell to obtain actual equipment temperatures for installed Environmentally Oualified equipment.
The thermocouples will monitor for drywell hot spots.
4EC-3062 This DCP replaced a flow control valve with a manually throttled globe valve to control flow to the
'A' Reactor Recirculation Pump Seal.
4EC-3069 This DCP modified the High-High ?:ndenser Pressure and the High-High Reactor Feed Pump Discharge Pressure Trip Circuits to provide protection against a single logic module failure.
This DCP does not change the function of the trip circuits, it modifies them so a failure of a single module will only affect the associated Reactor Feed Pump Turbine.
4EC-3206/01 This DCP added temporary piping and valves for the Filter and Ion Excaange Test Program.
The Filter and Ion Exchange Test Skids will be installed in the future by DCP 4EC-3206/02.
4EC-3216 This DCP replaced a section of non-safety related Station Service Water / Safety Auxiliaries Cooling System Loop
'A' Discharge Piping, The pipe, located in the yard, was tenking and was replaced with concrete lined pipe.
I t
1
QgE Descriotion of Deslan Chance Package 4HC-0127/03 This DCP replaced contaminated pipe, added a clean out for pressure flushing to the new pipe, and modified drain lines.
These modifications will reduce radiation levels in the 132' elevation Reactor Building Equipment Area.
4HC-0195/03 This DCP modified the discharge path of the Crack Monitoring System from Clean Radwaste to the suction of the Reactor Water Cleanup Pumps.
This modification will substantially reduce the volume of water that is processed through Radwaste.
4HC-0206 This DCP replaced area temperature switchen aid the Reactor Water Cleanup differential flow measu7ement circuitry with microprocessor-based Leak Detection Monitoring equipment.
This DCP will improve the reliability and operating capability of the instrumentation.
4HC-0238/02 This DCP modified the
'A',
'C',
and
'D' Safety and Turbine Auxiliaries cooling System piping and instrument tubing for the flow measurement used to monitor the Safety Auxiliaries Cooling System to Turbine Auxiliaries Cooling System flow.
The
'B' Safety and Turbine Auxiliaries Cooling System piping and instrument tubing was previously modified by 4HC-0238/01.
This DCP will reduce the flow measurement errors that have resulted in spurious actuations, 4HC-0271/02 This DCP installed keylock swjtches to temporarily bypass functions of the Power Load Unbalance Trip Circuitry, the Backup overspeed Trip Circuitry, and the Lockout for overspeed Trip Circuitry.
It also added indicator lights to show when a bypass switch is in the " Bypass" position, mid-stroke indication for the bypass valves, a computer alarm, and indication if any of the condenser vacuum switches I
l have been activated.
This DCP also added overload protection for the power supply for the Thrust Bearing Wear Detector test motor.
4HC-0311 This DCP removed the inverters in the Emergency Core Cooling Systems and the Reactor Core Isolation Cooling System.
The AC power will now be provided I
by station instrumentation inverters.
This DCP will eliminate the problems with the original inverters and increase the reliability of the systems.
l 4HC-0320/06 This DCP replaced solenoid valves on Safety Relief Valves.
It also installed a different connector and cable assembly en the newly installed solenoid valves.
l
pgE pancriotion of Design Chance Packace 4HC-0320/07 This DCP replaced solenoid valves on Main Steam Isolation Valves.
It also installed a different connector and cable assembly on the newly installed solenoid valves.
4HC-0324/03 This DCP added inspection platforms adjacent to the Turbine Pedestal on elevation 126' of the Turbine Building.
These platforms will facilitate the inspection of the slide bearing assemblies.
4HX-0780 This test DCP verified the results of calculations made to determine the Reactor Water cleanup System's capacity to provide an Alternate Decay Heat Removal Path during Hot Shutdown.
i
,--,e
,w,
-,,,-i-
-,<-.,,w.,.
.-+--r,
-e
~
w-,
m-
r d
t 9
Tht following Temporary Modification Requests (TMR's) have been eva.uated to dctermines 1.
If the probability of occurrence or the consequences of an 4
accident or malfunction of equipment important to safety i
previously evaluated in the safety analysis report may be increased; or
+
i 2.
If a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report way._be created; or
-t 3.
If the margin of safety as defined in the basis for any.
technical specification is reduced.
The TMR's did not create a new safety hazard to the plant nor did they affect the safe shutdown of the reactor.
The TMR's did not change the plant effluent releases and did not alter the existing environmental impact.
The safety r aluations determined that no unreviewed safety or environmental questions are involved.
J.
j B
t j
l s
er+,w,,--
gw--,~m, l w.-, rgm,.or.--.-
y-,, e r a,n aamr..---.am,.,-m,.e..wn,,,
Snw, - -,,,, - -,.,,.,...,,,.
-,,,.e, - -, ---, -, -, - - -. -
an.=,
,,-.w
.e
..,,.e,.'
- i i
~
4 o
THE
_ Descrintion of Tamporary Modification Reaumst 91-003-Thjs TMR installed electrical jumpers _across the #2 Feedwater Heater High High Level Trip Switches.
These switches cause spurious high level trip signals during low power levels.
The jumpers will be removed after the level signals stabilize.
i 91-007 These THR's removed the overload neaters from the -
i 91-013 breakers for the Reactor Water cleanup Discharge to Condenser Va.1.ve and the Reactor Water Cleanup Discharge to Equipment Drain Valve.
Removing the overload; heaters from the breakers will prevent the valves from inadvertently opening-during an Appendix R fire.91-008 This TMR installed an insert into the dischargo I
flange of the
'A' Control Rod Drive Pump to allow mating of-the pump-dischargo flange to the discharge flange of the piping.91-010 These THR's unpinned a sway strut hanger on the 91-012 High Pressure Coolant' Injection Steam Supply Line to prevent the strut =from bending or-breaking 1
during the High Pressure Coolant Injection. System Steam Line warmup.
The TMR's were removed when-the steam line warmup was completed.
9 1
l
-r e
% e e-m v-+v + - m mer,
w p -e-i-v-e aa w e-
~
- ww-
- e-o---o,--e-w,--+*
-w=w+r--inn-er-w-*
w i-+.e
-~~*---w---u-
- - e v w1+m-->-e-u-e-t-w--*w$e-weve=-3--
a--*-*
e The following Deficiency Report (DR) has been evaluated to determinet 1.
If the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or 2.
If a possibility for an accident or malfunction of a different type than any evaluated previously in the uafety analysis report may be created; or 3.
If the margin of safety as defined in the basis for any technical specification is reduced.
The DR did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor.
The DR did not change the plant effluent releases and did not alter the existing environmental impact.
The Safety Evaluation determined that no unroviewed safety or environmental questions are involved.
I l
DE Descriotion of Deficiency Report HTE-91-102 This DR addresses the inability of the Reactor i
Auxiliaries Cooling System Heat Exchanger Cooling Water Discharge Isolation Valve to close completely to isolate Service Water flow from the Reactor Auxiliaries cooling System Heat Exchangers.
There is a manual valve in the same line that may be used to isolate Service Water flow frota the Reactor Auxiliaries cooling System, if needed Therefore, t
. this_DR was dispositioned "use-as-is".
h a
b k.
J G
c,,w'n,.
- e~,-
s
.,.n-x e.,,,w,,_-
ew,
,ve.,
,.we+.,,w-.w,..,,-g.-
m em,,,
.,,,,,e, m-,4
-,m..-,-w.m,y,n e
,- p - te ww
, e y
n.
The following procedure revision has been evaluated to determine:
1.
If the probability of occurrence or the consequences of an accident or malfunction of equipment important te, safety previously evaluated in the safety analysis report may be increased; or 2.
If a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or 3.
If the margin of safety as define
- in the basis for any technical specification is reduced.
The procedure revision did not create a new safety hazard to the plant nor did it aff et the safe shutdown of the reactor.
The procedure revision did not change the plant effluent releases and did not alter the existing environmental impact.
The Safety Evaluation determined that no unreviewed safaty or environmental questions are involved.
Procedure Revision Descrietion of Procedure Rgylgig.L HC.OP-SO.GB-0001(Q)
This Safety Evaluation addresses the use of Rev. 10 the Chilled Water System in support of an Alternate Decay Heat Removal Path.
This revision is based on the successful completion of 4HX-0780, a test design change to show the Reactor Water Cleanup System can provide an Alternata Decay Heat Removal Path in operational Condition 4, Hot Shutdown.
e i
1 1
,, - - +,-
.p-,,-
n.
p,v.
m-
--