ML20070R543
| ML20070R543 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 04/30/1994 |
| From: | Hovey R, Zabielski V Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9405200252 | |
| Download: ML20070R543 (20) | |
Text
,
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O PSEG 1
i Pubhc Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Hope Creek Generating Station April 16, 1994 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 April are being forwarded to you with the summary of changes, tests, and experiments that were implemented during April 1994 pursuant to the requirements of 10CFR50.59(b).
Sincerely yours, R. J. Hovey l
General Manager -
}DR:.
Hope Creek Operations Y
S.JC Att achments C
Distribution l
[l' f
The Energy Peop:e I
9405200252 940430
- 93.,,7ms. min, PDR ADOCK 05000354 R
t g-INDEX NUMBER ~
i SECTION OF PAGES Average Daily Unit F0wer Level.
-1 Operating Data Report 3
Refueling Information.
1 Monthly Operating Summary.
1 Summary of Changes, Tests, and Experiments.
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5 4
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OPERATING DATA REPORT DOCKET NO..
50-354 UNIT Hone Creek DATE 5/6/94 COMPLETED BY V.
Zabielski TELEPHONE (609) 339-3506 OPERATING STATUS 1.
Reporting Period April 1994 Gross Hours in Report Period 712 2.
Currently Authorized Power Level (MWt) 3293 Max. Depend. Capacity (MWe-Net) 1031 Design Electrical Rating (MWe-Net) 1067 3.
Power Level to which restricted (if any) (MWe-Net)
None 4.
Reasons for restriction (if any)
This Yr To Month Date Cumulative 5.
No. of hours reactor was critical 134.3 1649.9 54472.9 6.
Reactor reserve shutdown hours 0.0 0.0 0.0 7.
Hours generator on line 90.9 1604 1 53636.9 8.
Unit reserve shutdown hours 0.0 0.0 0.0 9.
Gross thermal energy generated 165703 5106897 171070266 (MWH)
- 10. Gross electrical energy 37370 1704190 56668144 generated (MWH)
- 11. Net electrical energy generated 27184 1621361 54149045 (MWH)
- 12. Reactor service factor 18.7 57 1 84,4
- 13. Reactor availability factor 18.7 57.3 84.4
- 14. Unit service factor 12.6 55,7 83.1
- 15. Unit availability factor 12.6 55.7 E3 t1 i
- 16. Unit capacity factor (using MDC) 3.7 54.6 81.4
- 17. Unit capacity factor 3.5 52.8 78.6 (Using Design MWe)
- 18. Unit forced outage rate 0,0 asQ 4.3
- 19. Shutdowns scheduled over next 6 months (type, date, & duration):
None
- 20. If shutdown at end of report period, estimated date of start-up:
N/A
OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO. 354 UNIT Hone Creek DATE 5/6/94 COMPLETED BY V.
Zabielski TELEPHONE (609) 339-3506 MONTH April 1994 METHOD OF SHUTTING DOWN THE TYPE REACTOR OR F= FORCED DURATION REASON REDUCING CORRECTIVE NO.
DATE S= SCHEDULED (HOURS)
(1)
POWER (2)
ACTION / COMMENTS 1
4/1 S
628.1 C
4 RFOS
AVERAGE DAILY ~ UNIT POWER LEVEL DOCKET NO.
50-354 UNIT Hope Creek DATE 5/6/94 COMPLETED BY V.
Zabielski TELEPHONE (609) 339-3506 MONTH April 1994 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(MWe-Net) 1.
0 17.
0 2.
0 18.
0 3.
0 19.
0 4.
0 20.
0 5.
0 21.
0 6.
0 22.
0 7.
0 23.
0 8.
0 24.
0 9.
0 25.
0 10.
0 26.
0 11.
0 27.
110 12.
0 28.
231 13.
0 29.
570 14.
0 30.
645 15.
0 31.
N/A 16.
0 l
l REFUELING INFORMATION DOCKET NO.
50-354 UNIT Hope Creek 1 DATE May 6.
1994 COMPLETED BY V.
Zabielski TELEPHONE (609) 339-3506 MONTH April 1994 1.
Refueling information has changed from last month:
Yes No X 2.
Scheduled date for next refueling:
9/16/95 3.
Scheduled date for restart following refueling:
10/31/95 4.
A.
Will Technical Specification changes or other license amendments be required?
Yes No X
B.
Has the Safety Evaluation covering the COLR been reviewed by the Station Operating Review Committee?
Yes No
)
If no, when is it scheduled? Hgt scheduled ygtz 5.
Scheduled date(s) for submitting proposed licensing action:
Est scheduled yet.
6.
Important licensing considerations associated with refueling:
ELA 7.
Number of Fuel Assemblies:
A.
Incore 764 B.
In Spent Fuel Storage (prior to refueling) 1240 i
C.
In Spent Fuel Storage (after refueling) 1472 8.
Present licensed spent fuel storage capacity:
4006
)
Future spent fuel storage capacity:
4006 9.
Date of last refueling that can be discharged 5/3/2006 to spent fuel pool assuming the present (EOC13) licensed capacity:
(H2gs allow for full-core offload)
(Assumes 244 bundle reloads every 18 months until then)
(Does Dgt allow for smaller _ reloads due to improved fuel)
t HOPE CREEK GENERATING STATION MONTHLY OPERATING
SUMMARY
April 1994 Hope Creek entered the month of April in a refueling outage which started on March 5,1994.
The unit returned to service on April 27,1994.
As of April 30 the unit was on line for 4 consecutive days.
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SUMMARY
OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION
-l April 1994 The following items 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 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 effluent releases and did not alter the existing environmental impact.
The 10CFR50.59 Safety Evaluations determined that no unreviewed safety or environmental questions are involved.
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. Desian Changani Dmmary of. efsty 11aluations 4EC-0154:
This Design Change replaces the existing analog feedwater control system at Hope Creek with a microprocessor based, fault tolerant, critical component redundant Digital Feedwater Control System (DFCS).
Included in the scope is the replacement of the minimum flow recirculation control system associated with each reactor feedpump, the level 8 trip system and its associated test panel, and the original Woodward Position and speed control System.
The previous Feedwater Control System had been a source of problems for both operations and maintenance staff.
Besides operational difficulties the system was directly attributed to nine SCRAMS during the last eight years.
This modification does affect the facility as described in the SAR.
The new Feedwater Control System is a programmable microprocessor based digital system.
The system described in the SAR is an analog based system.
This affects the description in UFSAR section 7.7.1.3 which will be revised.
The DFCS is a non-safety related not important to safety system.
The failure modes and effects (minimum and maximum demand) of the new system are no different than the existing system.
The results of failures are already considered in the UFSAR section 15.0, Accident Analysis.
Therefore, there are no margins of safety affected.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
4HE-0065: This Design Change added a startup drain line for the HPCI Steam Supply header.
This will be used during initial warm-up to eliminate condensation buildup and water hammer in the steam supply line.
This modification does not change the fit or function of the HPCI pump or turbine.
This modification does not change the Facility or structures, components and systems described in the UFSAR.
However this modification requires that figure 6.3-1 in the UFSAR be updated.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
4HE-0083: This Design Change installs 1 inch vent valves in various location on the Residual Heat Removal Piping System (RHR) to improve the draining and venting.
The Vent Valves will be closed during normal operation.
The design basis of the RHR system will not be changed.
However, figure 5. 4 -13 (P&TD M51-1) will require revision to indicate the installation of the vent valves in the RHR system.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
4HE-01111 This modification installs two - 2 inch plug valves on locations of the 2 inch Circulating Water Pump Lube water return i
It also installs.three - 3 inch Plug Valves on locations of the 3 inch Circulating Water Pumps Lube water supply header.
This will allow each individual circulatinc Water Pump cooling water supply and/or return headers to be isolated.
This Design change will enhance the circulaving water system operational capabilities by allowing mainter.ance work in sections of the lube water header without affecting the normal flow path.
This modification does not change the fit or function of the circulating water pumps.
However, UFSAR Figure 10.4-3 required to be updated with the locations of the new valves.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and i
does not involve any Unreviewed Safety Question.
4HE-0113: This modification provides for the redistribution of dead weight load in the vicinity of valve HV-4964 via an additional vertical spring added adjacent to the valve.
This new spring hanger will reduce the dead weight load imbalance existing across the valve thus eliminating the potential seat binding due to dead weight imbalance which will enhance the operation and Local Leak Rate Testing of the valve.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
4EC-01021 (Various Pkas):
This design change removes various snubbers from various systems.
In certain circumstances the snubbers are replaced with struts.
Any deleted snubber that is part of the current ISI sample population will be functionally tested.
The Snubber' Reduction Program fully documents the analysis with calculations the margin of safety concerning the snubber removals.
Calculated stress levels and usage factors for postulated break locations are summarized in the Snubber Reduction Program.
Although the stress values changed because the calculations were reanalyzed for snubber optimization under the Snubber Reduction Program, all stress levels are still within original design basis allowables.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
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. 4EC-3008 Pka 6:
This Design Change Package replaces the Containment Atmosphere Control (GS) System containment isolation l
valves with upgraded Butterfly valves that are designed with metal seats for tight shutoff, extended wear and minimum maintenance.
This modification does not have any impact on the function of the system other than enhancing performance and having better leak rate testing characteristics (LLRT).
The specific description of I
the valves does not appear in the UFSAR but the description of the i
testing required per table 6.2-30 is changed and the UFSAR change will follow.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
i AEC-3102:
This Design Change involves the replacement of solenoid valves inside the drywell.
Rebuilt spare SRV Pilot Assemblies with new model solenoid valves are to be furnished by Target Rock for installation.
Design and model of the solenoid valves as described in section 5.2.2.4.2.1.1 of the UFSAR is changed by this modification and the UFSAR change will follow.
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The new solenoid valve is seismically and environmentally qualified.
The modification does not have any impact on the function of the system or equipment.
This model is designed to withstand pneumatic line pressure of 250 psig as compared to 135 psig of the previous design.
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Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
4EC-3061 Pka-3:
This Design Change will install an enclosure to house a Bitlogger.
It will also breach the barrier of the steam tunnel to run thermocouples.
Existing 3" diameter conduit will be opened and a 1.5" diameter conduit will be routed through the barrier.
The two penetrations will be sealed in accordance with standard procedures.
Temperature data will be used to extend the qualified life of EQ equipment where service conditions are lower than the design temperature.
It will also monitor for potential hot spots inside the steam tunnel.
The temporary temperature monitoring system installed under this change will not impact or cause malfunction of any. safety related equipment.
The cables will be routed so as not to interfere with any equipment or personnel.
The Bitlogger system is not address in the UFSAR or the Tech Specs.
There are no sections affected by this change.
However it modifies existing penetrations causing changes to the facility as described in the SAR.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
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T 4EC-3112 Pkc2: This Design Change replaces existing class HGC mortar lined carbon steel ASME III class 3 Service Water System (SWS) spool pieces with new material (HZC, 6% molybdenum stainless steel) procured ASME III class 3.
Industry experience has demonstrated that this new piping material has excellent
- resistance to corrosion /orrosion.
Additionally a 36" inspection port will be added to SWS loop "B" common discharge header located in the service water intake structure basement.
Two sets of break flanges will be added to the existing 1" copper-nickel lube water piping connections at each SWS Pump (B&D).
This will eliminate the need to cut socket welds for SWS pump removal.
UFSAR Figure 9.2-2 will be revised due to the installation of the inspection port.
The piping modifications and material upgrades do not modify the operation or function of the SWS or any other interfacing systems.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
AEC-3285 Various Pkos:
This Design Change modified the flanges of various pressure relief valves with a double Parker-Hannifin O-Ring Gask-O-Seal on the discharge side.
This will allow Type B Testing {LLRT) after relief valve lift set testing (IST Program) eliminating the requirement for a Type A Test (ILRT).
The modification does not change the facility its structure, components, and systems, described in the UFSAR.
However, various UFSAR sections figures are required to be updated to show the new test connection on the discharge flanges of the relief valves.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
This Modification will allow the installation of blind flanges.in place of the existing relief valves on air / nitrogen accumulators.
This changeout will eliminate the rigorous testing and maintenance of the relief valves every 18 months to meet the zero leakage requirement.
UFSAR Section 5.1 Figure 5.1-3 will be revised to show relief valve removed from accumulators.
The modification intends to minimize the amount of testing on valves and personnel exposure.
The PCIGS Compressors and receivers will maintain their over-pressurization protection devices installed to protect the system.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
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l 4EC-3316 Ekg21 This Design Change modifies the Plant Annunciator System (PAS) to eliminate Main Control Room (MCR) Overhead Annunciators (OHA) that are continuously alarmed.
Package 1 deleted all radiation monitor inputs to the RMS Computer (RM-11)
MCR OHA C6-A2, " Radiation Monitoring Alarm /Trbl", and also made i
possible a modification to the original RMS annunciator philosophy.
The MCR OHA new philosophy is to have the "HI" alarms alarm at the applicable dedicated MCR OHA and the " ALERT" Alarms at the RM-11 console.
Therefore the design changes of Package 2 involved deletion of all Radiation monitor inputs to the RMS computer (RM-11) annunciator C6-A2, and the regrouping of all Tech Spec and safety related RMS annunciator inputs with the exception of the SACS A, SACS B, and RACS radiation monitors, so that they alarm the applicable dedicated MCR OHA on the "HI" radiation alarms instead of the
" ALERT" radiation alarms.
A review of the accidents listed in the UFSAR Section 15.0 has determined that these modifications do not involve any situations or circumstances described.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
4EC-3344 Ekg i This design change replaced the existing carbon steel p p ng with 2 1/4 Cr-1Mo piping as part of the Iron Reduction Program for Hope Creek.
The piping included all three of the Feedwater Heater Strings.
There are no changes to the function of the system, components, structures or facilities with these modifications.
However the UFSAR will be updated to reflect the material changes as described in table 12.2-133 and figure 10.2-4 sheet 1.
The modification will not increase the probability of an accident previously described in the SAR because the new piping material is superior to the replaced carbon steel in errosion and corrosion resistance.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed safety Question.
4EC-3372: This Design Change was performed in support of the Station Blackout Isolation Requirements.
The Drywell Sump and Equipment Drain lines should be closed or verified closed locally.
This change is to modify the air operated valve to a failed closed valve.
The modification does not change the operation or function of any Liquid Radwaste System described in the UFSAR.
However two P&ID's included in the UFSAR Figure 9.3-7 sheet 1&3 of 3 must be revised to show the changes.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
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l 4EC-3375:
This Design Change raised both the alarm level and i
transfer pump start level of the Emergency. Diesel Generators (EDG) day tanks to be in compliance with Reg Guide 1.137.
ANSI 195-1976 i
Soc 6.1 (ANS-59.51) states that "Each Diesel shall be equipt with day or integral tank or tanks whose capacity is sufficient to malntain at least 60 minutes of operation at the level where oil is automatically added to the day tank or integral tank or tanks".
Although not committed to by Hope Creek or required by the Reg Guide ANSI /ANS 59.51-1989 clarified this wording by stating that this level is "the low level alarm setpoint".
During a recent EDSFI for the Standby Emergency Diesel Generators at Hope Creek it was noted that the present level when fuel oil is added setpoint is not in compliance with the Reg. Guide.
This Design Change is being instituted to bring the facility into compliance with the Reg. Guide.
This is considered a change to the facility as described in the UFSAR because the revision has to be made to an engineering calculation that forms the basis for the level setpoint calculation used to determine where the low level alarm and transfer pump start are located.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
4EC-3388: This Design Change installs vent valves on locations of the CRD's drive water headers where end caps are presently 1
installed, and inctalled a nipple for threaded end caps in lieu of a second isolation valve to minimize impact on weight.
This modification does not change the Facility, its Structure, Components, and System as described in the UFSAR.
However UFSAR Figure 4.6-5 will be updated to show the new valves installed on l
locations where end caps were shown.
l Therefore, this DCP does not increase the probability or i
consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
4EC-3394: This Design Change installed 4" manual operated gate j
valves in the RWCU return line to Feedwater Header. Currently the RWCU system must be removed from service during the Feedwater LLRT.
The addition of the manual valves will allow the continuous use of RWCU while the Foodwater line can be individually tested, and the now valves will isolate RWCU from Feodwater.
This modification does not change the Facility, its Structure, Components, or System as described in the UFSAR.
However, UFGAR Figures 5.1-3, 5.4-18, and 6.2-27 will be updated to show the new valve locations.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
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. 4EC 341:
Ekg 1.2;. This Design Change installs a toe and a normally closed
. solation valve in each of the two pump discharge headers in RHR AP202 Discharge Header and RHR Pump CP202 (A&C RHR Pumps).
Package 2 installs a full flow cross tie between the two pumps A and C discharge headers.
Additionally a manual keylock switch to override the interlock between the Suppression Pool suction valve position and the RHR Pump "C" stop circuit that prevents the RHR pump "C"
from operating when the HVF004C valve is closed.
This interlock will allow "C" Pump to operate when the pump is aligned to the alternate suction from the Reactor Pressure Vessel.
A review of the various sections of the UFSAR shows that this modification constitutes a change to the facility.
Since neither the performance of systems require to mitigate the consequences of an accident is being changed, nor are the initiating event mechanisms being changed because of this evaluation, the probability of any accident previously evaluated in the UFSAR is not changed.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and l
does not involve any Unreviewed Safety Question.
I 4EC-3430 Ekg 11 This Design change installs a hand rail around the centerline of the torus in the reactor building room 4102 for purposes of personnel safety.
Previously the station personnel working and testing equipment on the Torus had to work with safety belts on the arched Torus shell.
This modification will give an 4
added measure of safety to maintenance activities.
l This addition introduced a new load on RB 102 floor structural steel altering seismic calculations therefore this constituted a change to the facility.
The now handrail design and installation satisfies seismic II/I design criteria.
This design does not interfere with the function of any adjacent systems.
1 Therefore, this DCP does not increase the probability or.
consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
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, Procedure Chancesi Summary gf Safety Evaluations i
HC2 OP-GP.SM-0001(O) Rev2:
This procedure is for " Defeating the NSSSS Isolation Signal for Shutdown Cooling".
The Procedure Changes consisted of:
a) allowing bypass of one of the isolation trip system, during mode 4 and 5 provided Shutdown Cooling System integrity is maintained; b) permits bypassing of the high pressure (82#) isolation function in modes 4 and 5 for operable trip systems; c) adds a requirement that the RHR S/D Cooling suction header pressure high annunciator is operable and clear; d) adds a prerequisite for two operable L3 channels per operable isolation system; e) adds a prerequisite that this procedure will not be used during RPV IST Hydro; f) adds a prerequisite that an open path to containment atmosphere must be maintained while in this procedure; g) adds a precaution step to immediately secure RHR pump and close F008, F009, F015A, & F015B valves for SDC suction feeder pressure high alarm; h) adds a precaution step to immediately secure ECCS if an invalid ECCS initiation occurs and to check for SDC suction high pressure alarm; i) adds signature blocks for I&C Supervisor for two L3 annunciator, S/D Cooling header high pressure annunciator, and two L3 Channels per isolation trip system requirement; j) adds a prerequisite that the 1BCV-043 & 1BCV-133 are closed.
The 82 psig function is not required in mode 4 or 5 per Technical Specifications providing at least one operable isolation trip system (two L3 channels per trip system) capable of isolating RHR-SDC from the Reactor Coolant System exceeds Tech Spec required SDC isolation instrumentation requirement and is consistent with Rev 4 of the BWR 4 Standard Technical Specifications.
The margin of safety is not reduced.
Therefore, this Procedure revision does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
ILQ_, C11-S A. H A-0001 ( r ) : This procedure " Sampling the Offgas System" added section 5.2 to provide the technician with an alternate means of sampling the Offgas System if the sample panel pump ~(10-P-397) or the panel becomes inoperable.
A review of UFSAR Section 11.5-1 revealed that the pump which is installed as part of the alternate sampling method is not shown on the drawing (sheet 2).
A review of various related Tech Specs indicate no differences i
between acceptance limits and a failure point or system limitation as defined in the Basis for any Tech Spec.
-l Therefore, this Procedure revision does not increase the probability or consequences of an accident previously described in j
the SAR and does not involve an Unreviewed Safety Question.
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. Temporary Modificationsi Summary pl Safety EyAluAtiong T-Mod 94-009: 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.
This T-Mod does not increase the probability or the consequences of an accident listed in Table 15.0-2 of the UFSAR since the worst case would be for water induction into the turbine resulting in a turbine trip.
Therefore, this Temporary Modification does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
T-Mod 94-010:
This Temporary Modification addresses the jumpering out of battery cell number 13 in 125vdc C-1E battery 1DD447 and justified the continued operability of the battery with 59 cells.
Cell #13 was disconnected from the battery bank and a jumper was installed between batteries 12 and 14.
An Engineering Evaluation was performed providing justification for operating battery 1DD447 with 59 cells.
The battery was capable of providing design bases voltage to all class 1E loads and does not create the possibility of any potential safety hazard.
Therefore, this Temporary Modification does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
T-Mod 94-011: This Temporary Modification diverts Service Water normally flowing to the Cooling Tower Basin or the Tower Bypass Line to a manhole in the yard during operational condition 4.
This Modification will permit repairs to yard piping for the Service Water Air Release valve piping.
The margins of safety as discussed in the Tech Specs will not be impacted by the utilization of the temporary piping.
Therefore, this Temporary Modification does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
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.Reficiency Reports:
Ellmmary of Safety EvaluatioDH QB HTE-94-089: This Deficiency Report which.was disposition "Use-1 As-Is" described the flooding of the reactor cavity with the drywell head installed.
The Drywell Head is the uppermost part of the Primary Containment.
Per UFSAR Section 6.2.1.1.4.1 the Containment is designed to withstand an external to internal differential pressure of 3 psi.
However this pressure is a uniform gas pressure and not the hydraulic gradient experienced by the Drywell Head.
The Head has been evaluated for the hydraulic forces from the flooding and found to be unaffected.
Therefore, this Deficiency Report does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
RE !!MD-94-094:
This Deficiency Report which was disposition "Use As Is" was written after a Drywell walkdown identified a 2' portable tubelight and a 10' power cord were left in the space between the Reactor Pressure Vessel (RPV) and the inner wall of the biochield near penetration N17B.
Several attempts were made to remove the light and cord but further attempts were stopped due to ALARA considerations.,
It is highly unlikely the tube light could migrate out of the space between the RPV and the bioshield.
However, this may occur during a pipe break inside the biological shield if jet forces push the light up and out of the biological shield.
Should the tube light be deposited into the suppression chamber, the small size and. weight would be too small to cause significant damage or blockage of the suction strainers.
Therefore, this Deficiency Report does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
RE HMD-94-096:
This Deficiency Report which was disposition "Use-As-Is" evaluates effects of leakage identified from "B" Recirculation Pump at the end of the Fifth Refueling Outage.
There is a breakdown bushing in the Recirculation Pumps seal design that will limit the leakage through the seal even after total failure of the seal.
Seal Failure is mentioned in the UFSAR Section 1.8.1.29 and the conclusion there is that the leakage is sufficiently small so that no safety concern exists.
Therefore, this Deficiency Report does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
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.Other Chances: Summe_ry p_f Safety Evaluati2Ds SAE Chance 93-431 In Table 3.2-1 on page 14 or 42, the Reactor Auxiliary Cooling System (RACS) is describes in the Hope Creek UFSAR.
A typographical error is included in this description under the description of RACS Heat Exchangers.
The Heat Exchanger is described as being manufactured in accordance to TEMA Standard Class C.
It Is actually supplied in accordance with TEMA Standard Class R.
A DCP (4HZ-04665)-and this change notice where initiated to correct this erroneous TEMA Class designation in Table 3.2-1 and to eliminate the discrepancy within the SAR itself.
Therefore, this SAR Change does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
EAR Chance 93-44:
In Table 10.4-7 on page 1 of 6 the condensate System is described in the Hope Creek UFSAR.
A typographical error is included in this description of the Condensation Drain Tank.
The UFSAR Table 10.4-7 and DITS 10855-D3.4 indicates that the Condensate Drain Tank was a 250 gallon tank.
The Vendor Documents 10855-m100-92-1 and 10855-pm100-113-2 describe the tank as being designed, manufactured and supplied as a 1000 gallon (min) tank.
This Change Notice was initiated to correct the inconsistency.
4 Therefore, this SAR Change does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
RB Chance 94-06:
Each Standby Diesel Generator (SDG) is equipped with two independent air start systems.
The UFSAR states that i
cach air receiver can supply starting air for a minimum of five consecutive engine starts with a beginning air receiver pressure of 325 psig.
The SDG factory tests have shown that both air receivers operating in the normal design configuration will supply adequate starting air for a minimum of five consecutive starts with a beginning air pressure of 325 psig.
This change revises the UFSAR Section 9.5.6.2 to clarify the as-built capability of the starting air system of the SDG.
Therefore, this SAR change does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
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' Safety EvaluatiSD SCM0009:
This Safety Evaluation justifies the use of the offline P1BACK computer programs a substitute means for the calculation and monitoring of MCPR LHGR, MAPLHGR, and LPRM g
gain adjustment factors (GAFs).
This is restricted to periods i
when the NSSS plant process computer is inoperable.
Separate provision has been made in Hope Creek operating procedures'to j
correct P1BACK calculations of CMAPR in MAPLHGR evaluations for i
single loop operation.
Any inherent limitations or restrictions i
in the use of the NSSSS process computer will also apply to the
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use of the P1BACK.
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Therefore, this Safety Evaluation does not increase the probability or consequences of an accident previously described in 4
the SAR and does not involve an Unreviewed Safety Question.
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