ML20087F004
| ML20087F004 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 07/31/1995 |
| From: | Lyons D, Reddemann M Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9508150075 | |
| Download: ML20087F004 (11) | |
Text
-
+0 PSEG Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit i-(
August 10,1995 U. S. Nuclear Regulatory Commission Document Control Desk I
Washington,DC 20555 l
Dear Sir:
l 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 July 1995 are being forwarded to you with the l
l summary of changes, tests, and experiments that were implemented during July 1995 pursuant to i
the requirements of10CFR50.59(b).
.i s
Sincerely yours, Ih.
h ark ddemann General Manager-A Hope Creek Operations l
1
/
DL:RS:JC
(
Attachments I
Distribution ci.' '
f l U C13!l DeIxwwris in purlunds.
j 9508150075 950731
"'se m am PDR ADOCK 05000354 R
pg L
r.
. = = - - - - - - - - - - - - - -. - - - - - - - - - - - - - - - - - -. - - - - - - - - - - - -
wm=
INDEX NUMBER SECTION OF PAGES
. Average Daily Unit Power Level.......................................................................... 1 l
at ng D ata @ n......................................................................................... 2 '
Refueling Infonnation........................................................................................... 1 Monthly Operating Summary............................................................................... I l
Surranary of Changes, Tests, and Experiments...................................................... 4 1
1 1
J
.1 1
\\
l i
1 i
i l
1 l
l I
h i
l
i s
DOCKET NO.: 50-354 UNIT: Hope Creek DATE: Aneust 5.1995 COMPLETED BY: D. W.Lvons TELEPHONE: (609) 339-3517 AVERAGE DAILY UNIT POWER LEVEL MONTH JULY 1995 DAY AVERAGE DAILYPOWER LEVEL DAY AVERAGE DAILY POWERLEVEL (MWe-Net)
(MWe-Net) i 1
IQ21 17 Q
2 1Q22 18 Q
3 1Q31 19 9
4 1921 20 Q
5 1Q22 21 9
6 1022 22 9
7 22Q 23 Q_
8 Q
24 Q
9 9
25 22 10 Q
26 2Q2 11 Q
27 1010 12 0
28 1023 13 Q
29 1Q11 14 Q
30 221 15 Q
31 1021 16 Q
I
o.
1 4
DOCKET NO.: 50-354 UNIT: Hooe Creek DATE: August 5.1995 COMPLETED BY: D. W. Lyons TELEPHONE: (609)339-3517 t
OPERATING DATA REPORT OPERATING STATUS 1.
Reporting Period July 1995 Gross Hours in Report Period 744, 2.
Currently Authorized Power Level (MWt) 1291 Max. Depend. Capacity (MWe-Net) 1911 Design ElectricalRating (MWe-Net) 1067 3.
Power Level to which restricted (if any) (MWe-Net)
Eqas 4.
Reasons for restriction (if any)
This Month Yr To Date Cumulative 5.
No. of hours reactor was critical 353.7 45373 64473.2 6.
Reactor reserve shutdown hours M
M M
7.
Hours generator on line 325.6 4487.6 63491.0 8.
Unit reserve shutdown hours M
M M
9.
Gross thermal energy generated (MWH) 1021120 14551046 202965391 10.
Gross electrical energy generated (MWH) 330625 4834955 67262621
- 11. Net electrical energy generated (MWH) 309223 4625920 64279235
- 12. Reactor service factor d2d 392 314 9
13.
Reactor availability factor 47.5 89.2 8.jL4 14.
Unit service factor 43.8
!!l 34 1 15.
Unit availability factor 43.8 88.2 34 1
- 16. Unit capacity factor (using MDC) 40.3 88.2 82.6 17.
Unit capacity factor (using Design MWe) 39.0 312 79.8 18.
Unit forced outage rate 16 2 11 8 ij.
9.
Shutdowns scheduled over next 6 months (type, date, & duration):
Refueling Outage, November 11,1995,30 days 20.
If shutdown at end of report period, estimated date of start-up:
N/A e
T':
4 DOCKET NO.: 50-354 UNIT: Hooe Creek DATE: August 5.1995 COMPLETED BY: D. W. Lyons TELEPHONE: (609) 339-3517 OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS MONTH JULY 1995 METHOD OF SHlTITING DOWN THE TYPE REACTOR OR F= FORCED DURATION 2EASON REDUCING CORRECITVE NO.
DATE S= SCHEDULE (HOURS)
(1)
POWER (2)
ACTION / COMMENTS 1.
7/8/95 F-FORCED 418.4 HRS A-1 - MANUAL When LCO 3.7.2.a for TO EQUIP Control Room 7/25/95 FAILURE Ventilation Action Statement expired a unit shutdown was initiated and was completed at 00:18 hours on July 8,1995.
Troubleshooting determined that the root cause was a momentary interuption to the control circuit combined withlengthy cable runs which prevented reemtry within the 200 millisecond time delay.
The temperature switch mostlikely have caused the interuption was replaced. A modification was installed to reduce the effective length of the cable runs. A team has been formed to further review the event and make recommendations.
.l Start up commenced after resolution of this concern and several
)
emergent problems.
)
ri l
1;t l
4 DOCKET NO.: 50-354 UNIT: Hope Creek DATE: August 5.1995 COMPLETED BY: D. W. Lvons i
TELEPHONE: (609) 339-3517
. REFUELING INFORMATION l
MONTH JULY 1995 i
1.
. Refueling information has changed from last month:
Yes -
No X
'I
- 2.
Scheduled date for next refueling:
11/11/95
~
3.
Scheduled date for restart following refueling:
12/10/95
. 4A. Will Technical Specification changes or other license amendments be required?
Yes No X
l B. - Has the Safety Evaluation covering the COLR been reviewed by the Station Operating Review Committee (SORC)?
Yes' No X
If no, when is it scheduled? October 25.1995 5.
Scheduled date(s) for submitting proposed licensing action:
Nat reauired.
6.
Important licensing considerations associated with refueling:
j N/A 7.
Number ofFuel Assemblies:
1 A. Incore 764 t
B. In Spent Fuel Storage (prior to refueling)
D.4. _Q 4
C. In Spent Fuel Storage (after refueling)
.L422 8.
Present licensed spent fuel storage capacity:
4006 Future spent fuel storage capacity:
4006 9.
Date oflast refueling that can be discharged 5/3/2006 to spent fuel pool assuming the present licensed capacity; (EOC13)
(Dass allow for full-core off-load)
(Assumes 244 bundle reloads every 18 months until then)
(Does agl allow for smaller reloads due to improved fuel) i 7
I
I DOCKET NO : 10-J11
~'
UNIT: Hone Creek DA*E: Aurust 5.1995 COMPLETED BY: D. W. Lyons TELEPHONE: (609) 339-3517 i
MONTHLY OPERATING
SUMMARY
MONTH JULY 1995 The Hope Creek Generating Station began July 1995 operating at 100% power and e
remained there until 18:30 hours on July 7,1995. At that time a shutdown was initiated in accordance with the Technical Specification Limiting Condition for Operation 3.7.2.a because of continuing CREF inoperability. The Hope Creek Generating station went offline at 00:08 hours on July 8,1995. Reactor shutdown was complete by 00:18 hours on July 8,1995. Troubleshooting continued until the root cause of the CREF inoperability was determined. After resolution of this issue and emergent Emergency Diesel Generator and Chemistry concerns a start up was inhiated with criticality occurring at 06:38 hours on July 24,1995. The unit was syncitronized to the power grid at 10:35 hours on July 25,1995 and 100% power was achieved at 09:20 hours on July 26,1995, The unit operated at 100% power from July 26,1995 until the end of the month e
except:
On July 28,1995 when power was reduced 5% for starting of the "B" Reactor e
Feed Pump.
On July 30,1995 when power was reduced 13% for turbine valve testing and rod swaps then restored and reduced 5% later in the day to temporarily remove the 6A Feedwater heater from service.
At the end of th'e month the unit had been on-line for 7 days.
I I
9 9
DOCKET NO.: 50-354 UNIT: Hone Creek DATE: August 5.1995 b COMPLETED BY:
D. W. Lyons TELEPHONE: (609)339-3517
SUMMARY
OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION MONTH: JULY 1995 The following items have been evaluated to deternune:
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 safety analysis report may be created; or
- 3. If the margin of safety as defined in the basis for any technical specification is reduced.
The 10CFR50.59 Safety Evaluations showed that these items did not create a new safety hazard to the plant nor did they affect the safe shutdown of the reactor. These items did not change the plant efIluent releases and did not alter the existing environmental impact. The 10CFR50.59 Safety Evaluations determined that no unreviewed safety or environmental questions are involved.
l l
l l
I A
Procedures Summary of Safety Evaluations
]tC.RP-EU.Z7 0002(O), Revision 7 - OPERATION OF PORTABLE WATER FILTRATION SYSTEMS This revision provides direction for use of temporary radwaste processing equipment. This procedure revision changes the facility as described in the UFSAR because it adds a radwaste processing flowpath that is not shown on either UFSAR Figure 9.3-7 for the South Radwaste High Conductivity Sump or UFSAR Figure 11.2-1 for the Waste Samale Tank. This temporary liquid radwaste processing system enhances the in-plant liquic radwaste filtration capacity and does not affect the Process Control Program (PCP). No plant operating parameters are affected. Spent filters and disposable pressure vessels with spent resm will be disposed of in accordance with Radiation Protection Department directions. A design review of the system to ensure compliance with Regulatory Guide 1.143 has been performed by Nuclear Engineering.
The only credible failures are leaks in the skid or connections. Any leakage would isolated quickly and contained on the 54' elevation of the Radwaste Building. The leakage would either drain back to the South Radwaste High Conductivity Sump, the source ofliquid for this processing system, or to the floor drains. Continuous monitoring while the system is in use will prevent any excessive input due to leakage. This monitonng will, also, allow any air leaks to be isolated quickly to prevent an adverse affect on the station air systems.
When processing is not in progress the isolation valves will be closed and one vessel vented.
This proposal does not change the ability of the floor drains to contain leakage and will not affect the analyzed flooding levels for these rooms.
The solid and liquid Radwaste processing systems have no safety-related functions (UFSAR Sections 11.2 & 11.4) and do not communicate mechanically or electrically with any safety related systems. Failure of this equipment could not prevent satisfactory accomplishment of the safety function specified for any safety related components.
This revision gives instructions for use of a portable liquid Radwaste treatment skid to be used on elev. 54' of the Radwaste Services Building. This skid meets the requirements of Reg. Guide 1.143. None of the anticipated operational transients or postulated design basis accidents are applicable. Therefore, this procedure revision does not increase the probability or consequences of an accident previously described in the UFSAR and does not involve any Unreviewed Safety Question.
t
Procedures Summary of Safety Evaluations (continued)
THC.CH-SO.AN-0005(O)
FILLING THE HOTWELL FROM THE e
DEMINERALIZED WATER SYSTEM This temporary procedure provides directions for filling the Main Condenser Hotwells from the Demmeralized Water (AN) system.
Water from the AN system meets the chemical purity requirements for water found in UFSAR Table 10.4.5.
Water transfer will be accomplished using hoses run from a manway on the Demineralized Water Storage tank to a flange connection downstream of drain valve 1 ADV-371 and a pump.
The installation will, also, include a backfill connection and an isolation valve on the suction portion, two check valves for backflow protection, and sampling capability. This method of filling the Main Condenser is not j
shown on UFSAR Figure 10.4-5.
However, since this is a temporary procedure an i
l UFSAR change for that figure will not be submitted.
The olant operating parameters will be affected by this proposal. None of the transients or accic ents evaluated m the UFSAR are applicable to this proposal. The Condensate and l
Make Up Demineralizer systems have no safety related function and will not compromise l
operation of safety related systems. Failure of the Condensate and Demineralized Water
)
Make Up systems is discussed in UFSAR sections 9.2.3.3 and 10.4.7.3 any postulated
{
failure would be bounded by total system failure. Credible failure modes include hose I
leakage, failure of Condensate or Demineralized Water Make Up systems, and cross contamination of the AN system. Hose leakeage will be collected by the floor drain system in accordance with plant design as analyzed in UFSAR Section 9.3.3. The hoses will be restrained to prevent whip and they will be monitored continuously during transfer operations. The Condensate and Make Up Demineralizer systems are not required for safe shutdown of the plant. The check valves should prevent cross-contamination of the Demineralized Water system from the Condensate system and the operator will be instructed to immediately isolate the hose if reverse flow is indicated.
l Therefore, this temporary procedure does not increase the probability or consequences of an accident previously described in the UFSAR and does not involve any Unreviewed Safety Question.
Temporary Modification Summarv of Safety Evaluation e 95-042 - JUMPER FOR #2 FEEDWATER HEATER HIGH/HIGH LEVEL SWITCII This Temporary Modification installed an electrical jumper across the #2 Feedwater Heater Hi-Hi Level trip switches and installed a temporary keep fill line to the low side of the level transmitters. This modification is performed due to spurious indications during power ascension and is removed at approximately 40% Reactor Power.
l This T-Mod does not increase the probability or the consequences of an accident listed in l
Table 15.0-2 of the UFSAR since the worst case would be for water induction into the turbine resultirg in a turbine trip. Because the start-up is complete, this T-Mod has been removed.
Therefore, this Temporary Modification does not increase the probability or consequences of an accident previously described in the UFSAR and does not involve an Unreviewed Safety Question.
=
Deficiency Report Summary of Safety Evaluation
. 950608149 - LEAK REPATR OF 1 AFV-045,6A FEEDWATER ITEATER DRAIN
'LINE VALVE This Deficiency Report addresses the leaking of IAFV-045, the 6A
^
Feedwater Heater Drain Line Valve. This is a normally closed valve but is leaking past its seat causing increased Radwaste. The valve is currently closed and was areviously leak repaired in May 1995. To stop the leakage the valve will be injected on t ae upstream of the seat with a sealant. Precautions will be taken to prevent injection of the sealant into the shell side of the feedwater heater or the feedwater stream. However, the affect of this sealant on the system has been analyzed and found to be acceptable. This repair does not affect any equipment that is important to safety nor are there any previously evaluated accidents that are applicable. The valve will be replaced during the next refueling outage.
No parameter of the extraction steam system or feedwater system is affected by this change. The effect of the injected material on the extraction steam system is that l AFV-045 will not be able to be opened. This will prevent use of this line for maintenance but will not effect system operation. The credible failure mode is failure to seal and stop the leak. If this happens, the valve will continue to leak and leakage will go to Radwaste for processing in accordance with design. This proposal does not affect any equipment that is important to safety.
Therefore, the disposition of this Deficiency Report does not increase the probability or conseguences of an accident previously described in the UFSAR and does not involve an Unreviewed Safety Question Other Summary of Safety Evaluation e UFSAR CITANGE NOTICE CN 95-21. SYSTEM MANAGER (SYSTEM ENGINEER) This change notice is administrative in nature and is a title change only with no changes in responsibility. The title change is from ' System Engineer" to ' System Manager"in UFSAR Sections 12.1.2.4,13.1.2.2.6, and 17.2.3. This title change does not change the position requirements and more accurately reflects the level of authority given to the position.
Because the requirements of the position are not being changed this proposal does not affect any operational transients, design basis accidents, or malfunctions of equipment important to safety previously evaluated in the UFSAR. Neither does this proposal increase the probabihty or consequences of an accident or a malfunction of equ,ipment important to safety previously evaluated in the UFSAR. Nor does it create the possibility of an accident or malfunction of a different type than previously evaluated in the UFSAR.
Therefore, this UFSAR change does not increase the probability or consequences of an accident previously described in the UFSAR and does not involve an Unreviewed Safety Question.