ML18092A755
| ML18092A755 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 08/06/1985 |
| From: | Corbin McNeil Public Service Enterprise Group |
| To: | Varga S Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18092A756 | List: |
| References | |
| LCR-85-09, LCR-85-9, NUDOCS 8509050145 | |
| Download: ML18092A755 (18) | |
Text
Public Service Electric and Gas Company Corbin A. McNeill, Jr.
Vice President -
Public Service Electric and Gas Company P.O. Box236, Hancocks Bridge, NJ 08038 609 339-4800 Nuclear U.S. Nuclear Regulatory Commission Off ice of Nuclear Reactor Regulation Division of Licensing Washington, D. c. 20555 Attention:
Mr.
Steven A.
Varga, Chief Operating Reactors Branch 1 Division of Licensing Gentlemen:
REQUEST FOR AMENDMENT FACILITY OPERATING LICENSE DPR-70 UNIT NO.
1 SALEM GENERATING STATION DOCKET NO.
50-272 August 6, 1985 Ref:
LCR 85-09 In accordance with the Atomic Energy Act of 1954, as amended and the regulations thereunder, we hereby transmit copies of our request for amendment and our analyses of the changes to Facility Operating Licehse DPR-70 for Salem Generating Station, Unlt No.l.
This amendment request consists of changes to those sections of the Technical Specifications, and to the Facility Operating License, to accommodate an increase in RATED THERMAL POWER.
This change will result in identical power ratings for both Salem Units.
In accordance with the fee requirements of 10CFR170.21. a check in the amount of $150.00 is enclosed.
(. -*8509050145 850806 l..
- PDR ADDCK 05000272 p
PDR*
Mr.
Steven A.
Varga 8-6-85 Pursuant to the requirements of lOCRFS0.91, a copy of this request for amendment has been sent to the State of New Jersey as indicated below.
This submittal includes three (3) signed originals and forty (40) copies.
Enclosure C
Mr. Donald c. Fischer Licensing Project Manager Mr. Thomas J. Kenny Senior Resident Inspector Mr. Samuel J. Collins, Chief Projects Branch No. 2, DPRP Region 1 Sincerely, Mr. Frank Cosolito, Acting Chief Bureau of Radiation Protection Department of Environmental Protection 380 Scotch Road Trenton, N.J. 08628 Honorable Charles M. Oberly, III Attorney General of the State of Delaware Department of Justice 820 North French Street Wilmington, Delaware 19801
Ref:
LCR 85-09 STATE OF NEW JERSEY
)
)
SS.
COUNTY OF SALEM
)
Corbin A. McNeill, Jr., being duly sworn according to law deposes and says:
I am a Vice President of Public Service Electric and Gas Company, and as such, I find the matters set forth in our letter dated August 6, 1985, concerning our Request for Amendment to Facility Operating License DPR-70, are true to the best of my knowledge, information and belief.
Subscribed and Sworn to before me this ~TH day of AUGU'ST
, 1985
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PROPOSED CHANGE TO TECHNICAL SPECIFICATIONS SALEM UNIT NO. 1 Description of Change:
Ref: LCR 85-09 Increase the Unit l licensed core power from 3338 MWT to 3411 MWT.
The necessary Tech. Spec. changes are as follows:
- a.
Section 1.25, change RATED THERMAL POWER from 3338 to 3411 MWT
- b.
section 2.2, change RTS setpoints for core flow from 88,500 gpm/loop to 87,300 gpm/loop
- c.
section 3.2.5, change DNB parameters for RCS Tavg from 58.l °F to 582°F The changes to section 1. 25 and 3. 2. 5 are a direct result of the power uprate.
The change to Section 2.2 for core flow is for consistency with the core flow requirements of Section 3.2.5.
section
.3.2.5 requires a flow of 349,200 gpm total for four loops (which equals 87,300 gpm per loop).
The value of 87,300 gpm is consistent with Unit 2 Technical Specifications and with the design flow used by Westinghouse *for both plants.
Reason for Change:
Unit 1 and Unit 2 are essentially identical units.
This change will allow Unit 1 to operate at the same power as Unit 2 which is licensed for a core power of 3411 MWT.
Operating th~ units at the same power will allow greater standardization between the units and an increase in electrical output from Unit 1.
Significant Hazards Consideration Evaluation The attached evaluation documents the review don~ by PSE&G and Westinghouse to establish that operating Unit 1 at a core power of 3411 MWT does not represent a significant hazards consideration.
These proposed changes may cause.
some increase in the consequences of a previously analyzed
~
PROPOSED CHANGE TO TECHNICAL SPECIFICATIONS SALEM UNIT NO. 1 Ref: LCR 85-0~
Significant Hazards Consideration Evaluation (Cont.)
accident or some slight decrease in a margin of safety; but, the results of the change on plant operation will remain clearly within the bounds of the FSAR analyses and within the guidelines of the Standard Review Plan Sections 4.3 and 4.4.
These changes also bring about consistency between the Technical Specifications for both Salem Units; the changes, therefore, correspond to examples (vi) and (i} of the guidance provided by the Commission on changes considered.
"Not Likely to Involve A Significant Hazards Con~ideration" in Federal Register 48FR14870.
~2-
DEFINITIONS PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 of the Updated FSAR, 2) authorized under the provisions of 10CFRS0.59, or 3) otherwise by the Connission.
PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fauit in a Reactor Coolant System component body, pipe wall or vessel wall.
PROCESS CONTROL PROGRAM ( PCP) 1.22 The PROCESS CONTROL PROGRAM shall be that program which contains the current for111Jla, sampling, analyses, test, and determinations to be made to ensure that the processing and packaging of. sol id radioactive wastes, based on denonstrated processing of actual or si111Jlated wet solid wastes, will be acco1T1>l ished in such a way as to assure compliance with 10 CFR Part 20, 10 CFR
~art 71 and Federal and State regulations and other requirements g:>verning the.
disposal of the radioactive waste.
PURGE - PURGING 1.23 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinerrent to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.
Q.JADRANT POWER TILT RATIO 1.24 Q.JADRANT POWER TILT RATIO shall be the ratio of the maxirrum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maxi111Jm lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for co1T"4Juting the average.
RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor cool ant of 3411 MWt.
SALEM - UNIT 1 1-5
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U1I TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT.
- 2. Power Range, Neutron Flux
- 3. Power Range, Neutron Flux, High Positive Rate
- 4. Power Range. Neutron Flux, High Negative Rate
- 5. Intermediate Range, Neutron Flux
- 6. Source Range, Neutron Flux
- 7. Overtemperature AT
- 8. Overpower Ill
- 9. Pressurizer Pressure--Low
- 10. Pressurizer Pressure--High
- 11. Pressurizer Water Level--High
- 12. Loss of Flow TRIP SETPOINT Not Appl 1cable Low Setpoint - < 25% of RATED THERMAL POWER -
High Setpo1nt - < 109% of RATED THERMAL POWER
< 5% of RATED THERMAL POWER with a time constant > 2 seconds
< 5% of RATED THERMAL POWER with a time constant > 2 seconds
~ 25% of RATED THERMAL POWER
~ 105 counts per second See Note 1 See Note 2
> 1865 psig
~ 2385 psig
~ 92% of instrument span
~ 90% of design flow per loQp*
- Design flow is 87,300 gpm per loop.
ALLOWABLE VALUES Not Applicable Low Setpoint -
< 26% of RATED THERMAL POWER -
High Setpoint - ~ 110% of RATED.
THERMAL POWER
< 5.5% of RATED THERMAL POWER with a time constant ~ 2 seconds
< 5.5% of RATED THERMAL POWER with a time constant ~ 2 seconds
~ 30% of RATED THERMAL POWER
~ 1.3 x 105 counts per second See Note 3 See Note 3
~ 1855 psig
~- 2395 psig
~ 93% of instrument span
~ 891 of design flow per loop*
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~
PARAMETER Reactor Coolant System Tavg Pressurizer Pressure Reactor Coolant System TABLE 3. 2-1 DNB PARAMETERS 4 Loops In Operation
< 582°f LIMITS
~ 2220 psia*
~ 349,200 gpm 3 Loops in Operation*
< 572°f
~ 2220 psia*
~ 278, 100 gpm
- Limit not applicable during either a THERMAL POWER ramp incease in exc.ess of 5% RATED THERMAL POWER
. per minute or.a THERMAL POWER step increase in excess of 10% RATED THERMAL POWER.
~
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Pag* l of ~
Date 6/14/85 Public Servfce ElectrWand Gas Campany P.O. Box 236 Hancocks Bridge, New Jersey 080~8 TITL&~ MISS PoWBit**UPRATING. TO 3423 MWt SALEM UNIT l SALEM NUCLEAR GENERATING STATION PORPOSB1 The purpose of this Evaluation is to demonstrate that the proposed changes to technical Specifications to accommodate an uprate in core power for Salem Unit l from 333& MWt to 3411 MWt does-not involve~
significant hazards consideration.
SCOPEi This proposed power uprate is for Unit 1 only1 however, the proposed power for Unit 1 is the p~esent licensed power of Unit 2.
The power capability parameters are listed in Table*l of this safety evaluation.
A core power of 3411 MWt corresponds to a NSSS power 9f' 3423 MWt.
The difference, 12 MW, is due to RCP pump heat.
This evaluation covers both the Nuclear Steam Supply System (NSSS) and the Balance-of-Plant (BOP) design.
ATTACHMENT.S:
- 1 Westinghouse Report *salem Unit 1, 3423 MWt NSSS Uprating, Safety Evaluation*. transmitted by PSE-85-552 dated April 2, 1985.
- 2 Draft FSAR Changes To Be Incorporated UPan Issuance Of The Salem Unit 1 Powar Uprate Amendment GENERAL DISCUSSION:
The primary and secondary plant operating parameters selected for the.unit 1 uprate are the existing parameters for Unit 2. *The rationale for selecting these specific values is two fold.
- First, as.will be demonstrated later in this section of the report, the units are essentially identical (Salem Units 1 and 2 form a twin unit station}.
Nearly all of the original equipment criteria and an~lys7s are common to both Units, and these common power related crite:ia an~ analyses.assumed the Unit 2 higher power.
- secondly, uprating Unit 1 to Unit 2's power will allow even greater standardization between the Units.
The selected primary and secondary plant operations parameters are listed in Table 1.
Page 2 of 9 Date:
6/14/85 The uprate was demonstrated by verifying that at the uprated condition, all current FSAR analyses remain applicable and by verifying that all equipment and systems for Unit 1 remain within their current specifications, acceptance criteria, and qualifications.
In some cases, the documentation for equipment had
- to*be.revised to reflect the uprated operating conditions.
Feasibility had earlier been demonstrated by verifying that all power related equipment in Unit 1 was identical or functionally identical to Unit 2 equipment.
A fundamental basis of this uprate evaluation fs that the present licensing criteria and acceptance standards for current Unit 1 operations remain applicable to the uprated Unit 1.
All equipment reviews and evaluations were likewise done to the current Unit 1 codes and standards.
The impact of the* uprated condition on the actual Unit 1 operating temperatures, pressures, and. flows is small.
These changes in actual operating conditions do not cause any of the systems to exceed it's design parameters.
The RCS T-hot increases 1.7° F; T-average increases 1.1° F; and T-cold increases 0.6° F.
The CVCS letdown line temperature correspondingly increases 0.6° F.
There is no increase in the RCS operating pressure and flow.
The main steam temperature and pressure do not change.
The feedwater temperature at the inlet of the steam Generator increases by 3.8° F.
Feedwater pressure does not change~
Main steam and feedwater flows increase by 2.7%.
Other systems connected to these
.systems, e.g. condensate, turbine drains, will also see similar increases in flow and minor changes in temperature.
The flows and pressures in the auxiliary cooling systems -
component cooling water, service water, and circulating water -
do not increase.
The heat load increase to the component cooling water and service water is very marginal; the heat load increase to the circulating water results in a 0.3° F increase in the return temperature to the river.
As part of the overall evaluation, Westinghouse was authorized to reevaluate at the uprated condition all the equipment, systems, and accident events which Westinghouse performed during the original plant design.
The Westinghouse report, Attachment 1, is in summary form incorporated into this evaluation.
The results of the review are diccussed below.
Impact on Accident Analyses All currently docketed FSAR accident analyses were reviewed.
These include Loss of Coolant Accident (LOCA) events, High Energy Pipe
.(.
Page 3 of 9 Date:
6/14/85 Breaks Outside Containment, Containment Analyses, Subcompartment Pressure Analyses, Main Steam Line Breaks, and other non-LOCA events.
All. the current accident analyses either _o'riginally assumed unit 2's power (or higher), or else they are insensitive to the power increase.
None of the analyses will need to be modified to implement the Unit 1 uprate.
The discussion below on accidents supplements the discussion in, Section 3.
High Energy Break Analysis - The power uprate does not change the pressure in any line and has only a small temperature increase in a few lines.
Since there is no pressure increase in any line, there is no change to the pipe whip or jet impingement loads.
The changes to the service temperature of high energy lines analyzed for a break outside containment are limited to:
(1)
Feedwater (penetreation area) +0.3° F (2) eves Letdown (betwen the Regenerative heat exchanger and the Letdown heat exchanger) +0.3° F The increase in feedwater temperature has no impact on the Unit 1-High Energy Break Analysis.
A comparison of* FSAR section 3.6.5.2 and 3.6.5.3 shows that in all cases main steam line breaks are postulated in the same safety related areas where feedwater line breaks* are postulated.
The main steam line breaks remain limiting.
In fact, there is only one short run {approximately 15 feet) of feedwater piping in a safety area where there is not also main steam piping.
Refer to FSAR Figures 3.6-11 and 3.6-15.
However, due to the low stress values, there are no postulated breaks in this section of feedwater piping.
The increase in the eves letdown temperature, 0.3° F, is considered insignificant and well within the accuracy of existing calculations and conservatism.
Subcompartment Pressure Analysis -
The Salem subcompartment pressure analysis uses common assumptions which envelope both units; therefore, the report (Reference 64 of SNGS UFSAR Section 15.4) remains applicable for Unit 1 at the uprated condition.
Containment Analyses -
The containment transient analyses presented in the SNGS FSAR Section 15.4.8 are common to both units.
The transient analyses for RCS pipe breaks assumed a NSSS power of 102%
of 3570 MWt.
The transient analyses for main steam line breaks assumed a NSSS power of 102% of 3425 MWt.
The radiological consequences of the postulated design bases loss-of-coolant-accident assumed a reactor power of 3558 MWt.
The hydrogen production and accumulation analysis assumed a reactor power of 3575 MWt.
- I.
Page 4 of 9 Date:
6/14/85 Impact to Primary Plant Components The original equipment Specif icai:ions for the Unit 1 Reactor Coolant Pumps (RCP), Control Rod Drive Mechanism (CRDM), Reactor Coolant Piping, Pressurizer, and Steam Generators enveloped the proposed power uprate and only minor documentation changes were required to reflect the proposed, new operating conditions.
The unit 1 Reactor Vessel and some of its internals were originally analyzed for 3338 MWt power.
These analyses were revised for the uprated condition.
During the initial feasibility study it was determined that all Unit 1 primary components are essentially identical to the Unit 2 components except.for the following:
- 1)
The flow hole pa~terns in the lower reactor internals baffle barrel region formers are different, and
- 2)
The control rod patterns are different.
The flow hole and control rod patterns were evaluated as having no impact on the capability to uprate.
(As part of a standardization task separate from the Un~t 1 uprate, the control rod pattern on Unit 2 was modified starting with cycle 3 ~o more closely resemble Unit l's pattern).
A third difference, drainage capacity from the moisture separators in the Steam Generators, was identified during the feasibility review.
However, during Unit l's fifth refueling outage, the drainage capacity was increased to the same capacity as Unit 2's.
This modification was necessary to ensure that moisture carryover from the Steam Generators is kept below 0.25% when Unit 1 is operated at 3423 MWt.
This was the only hardware modification required for the Unit 1 NSSS power uprate.
Since the original End-of-Life (EOL) neutron fluence for Unit l's Reactor Vessel was based on a core power of 3483 MWt, a power higher than the proposed uprated core power, and because Salem 1 has shifted to a low neutron flux leakage core starting with Cycle 6, the original EOL n~utron fluence calculation remains bounding.
Impact to Piping The RCS piping and all other Nuclear Class 1 piping six.inches in diameter or greater was originally analyzed by Westinghouse.
Westinghouse in Attachment 1 reviewed the piping analyses and determined that the original analyses still envelope the uprated condition.
~-
Page 5 of 9 Date:
6/14/85' PSE&G reviewed the class 1 piping below six inches in diameter.
This review indicated that the transients used for PSE&G analyzed
-Class 1 piping likewise enveloped the proposed power uprate.
All ~uclear Class 2 and 3 piping and all non-nuclear piping was analyzed by PSE&G using criteria that enveloped Salem Units 1 and
- 2.
No reevaluation is necessary to any of these analyses.
The pressurizer safety and relief valve discharge piping on Unit 1 was analyzed for the relief valve's rated flow condition.
The safety and relief valves are identical between the two units.
Impact to Reactor Fluid Systems The following system design documentation was reviewed:
Reactor Coolant Safety Injection Chemical and Volume Control Residual Heat Removal Containment Spray Spent Fuel Pool Cooling Sampling It was determined that these systems were originally designed using common (Unit 2) _power par~meters.
Impact to Reactor Auxiliary Equipment All power related Reactor Auxiliary.Equipment was reviewed.
All Unit 1 equipment was found to be functionally or physically identical to the Unit 2 equipment except for the seal water injection filters.
The filter capacity of Unit 1 is 300 gpm and for Uni~ 2 it is 350 gpm, both of which are more than adequate for handling the seal water flow rate.
The equipment reviewed includes valves, pumps, heat exchangers, tanks, demineralizers, and.filters.
Accordingly, all the auxiliary equipment is considered capable of performing satisfactorily at the uprated power of 3423 MWt.
NSSS/BOP Interfaces -
The original Balance-of-Plant interfaces specified by Westinghouse are identical for both units and were based.on Unit 2's higher power rating.
Accordingly, any BOP equipment/component affected by the Unit 1 uprate should remain within its original* design criteria *
. Impact to secondary Plant Components and Systems There are minor differences between the Unit 1 and 2 turbine, but as a result of modifications done during earlier outages, the Unit 1 turbine can operate at loads in excess of the expected load at the
' Page 6 of 9 Date:
6/14/85 uprated condition.
At the uprated conditions, the turbine will remain below the original maximum calculated load.
The turbine will not require any modification.
The Westinghouse generator on Unit l*and the General Electric generator on Unit 2 are of equal rating.
The rest -of the secondary plant (i.e., main steam, auxiliary feedwater, condensate, feedwater, turbine drains, steam dumps, circulating water, turbine service water, etc.} is considered identical.
Based on a review of the PSE&G purchase specifications, all power related equipment purchased directly by PSE&G was determined to be identical between the two units.
The equipment for the two units was bought at the same time, from the same vendors, to the same criteria using purchase specifications.common to both units.
The sole exception is the purchase specification for the main steam isolation valves which specifies slightly higher flow rates for the unit 2 valves than for the Unit 1 valves.
The valves supplied by the vendor are, however, identical for both units.
The criteria of the common purchase specification were based on the higher' requirements of Unit 2.
one modification, upgrade of the condensate pumps, has been identified as a desirable modification for plant reliability.
Th~s modification will decrease feedwater pump low suction pressure trips during secondary plant transients.
The modification has been authorized for both units.
The Unit 1 modification is scheduled to be implemented during its sixth refueling outage *. The modification was done on Unit 2 during its second refueling outage.
Impact to Electrical Plant Since all electrically operated equipment in Unit 1 was originally specified for the higher power rating of Unit 2, no changes are required to any motor, cable, or switchgear.
There will be no increase in the calculated loading to the emergency generators.
There will be only a negligible increase in the normal electrical requirements for Unit 1 (The condensate pump upgrade will impact the electrical plant, but that evaluation was done as part of the pump upgrade).
The electrical system from the generate~ to the grid can handle the increased electrical output estimated at 26 MWe.
Impact to Controls and Instrumentation There is no impact on the controls and instrumentation other than a number of setpoints and calibrations will have to be done for the higher power.
A review indicated that recalibration will be required on the Nuclear Instrumentation System and on the Feedwater and Main Steam flow instrumentation and controls.
Page 7 of 9 Date:
6/14/85 Impact to Equipment Environmental Qualifications All equipment environmental qualifications are common to both units, and these assumed ~he higher power level of Unit 2.
No documentation changes were necessary in this area for the Unit 1 power uprate.
Impact to.Nuclear Fuels The Unit 1 Cycle 6 Reload Safety Evaluation.(RSE) was performed assuming the present power limit of 3338 MWt; however, a preliminary review indicates that the power uprate would not have a significant impact on the core.
The cycle 7 core will be designed for 3411 MWt.
The RSE will address the core's safety parameters.
Environmental Impact The Unit 1 uprate will not cause the Salem Nuclear Generating Station to exceed its allowable heat discharge rate to the river (16.3 x 109 BTU/hr, total both units) nor the maximum allowable difference between river intake and discharge temperatures (27.5° F)
- No revision is required to the source t"erm for the radioactive waste management (FSAR, Chapter 11) and for radiation protection (Chapter
- 12) since it is based on a 3558 MWt core power.
The uprate should not have an impact on the volume of radioactive wastes because these are dependent on maintenance and operating events.
The slight power
. change will not in any significant manner alter *maintenance and operating events.
Licensing Changes The FSAR changes identified are listed in Attachment 2.
These changes include those identified by Westinghouse in Attachment 1.
The Unit 1 uprate will not cause the SNGS to exceed its environmental (thermal) discharge limits.
If changes to the.Technical Specifications for the core physics parameters are required, they will be addressed by a fuel Safety Evaluation as discussed previously.
None are forseen.
The Unit 1 Technical Specification changes that have otherwise been identified are limited to:
Date:
6/14/85 Section Title
- 1. 25 Rated Thermal Power 2~2 RTS Setpoints 3.2.S DNB Parameters Pag of 9 Change from 3338 to 3411 MWt Core flow from 88,500 gpm/loop to 87,300 gpm/
loop RCS Tavg from 581° F to s02° F The changes to Section 1.25 and 3.2.S are a direct result of the power uprate.
The change to section 2.2 for core flow is for consistency with the core flow requirements of Section 3.2.S.
section 3.2.S requires a flow of 349,200 gpm total for four loops (which equals 87,300 gpm per loop).
The value of 87,300 gpm is consistent with Unit 2 Technical Specifications and with Westinghouse primary plant parameters, Table 1, for both units.
NO change will be required in the overtemperature T and overpower-T setpoints.
The Unit 1 and Unit 2 values are already equal.
CONCLUSION:
rhe Salem Unit 1 core power uprate from 3338 MWt to 3411 MWt has oeen thoroughly reviewed by both Westinghouse (Attachment 1) and PS8&G to establish the following:
- 1.
No equipment has to be added or deleted.
- 2.
8xcept as noted in paragraph 3 below, all systems and equipment affected by the power uprate were either initially specified to conditions which bound the proposed power uprate, or it has been determined by revising the documentation to reflect the proposed power uprate, that the systems and equipment will continue to meet all their current specifications, acceptance criteria, and qualifications.
- 3.
As a result of the Unit 1 Steam Generator moisture separator drain modifications done during the fifth refueling outage, all power related NSSS equipment between Unit 1 and Unit 2 is either identical or functionally identical.
Similarly, after the Unit 1 condensate pumps are upgraded during Unit l's sixth refueling outage, all power related BOP equipment between the units will again be either identical or functionally identical (the Unit 2 pumps were upgraded during Unit 2's second refueling outage.
This change is a plant improvement project which will increase the reliability of normal feedwater to
-~".,. -.
Page 9 of 9 Date:
6/14/85 steam generators.
It is not* a requirement for meeting the original design bases).
- 4.
The functions and requirements for all systems and equipment remain unchanged.
- 5.
None of the accident analyses for Salem 1 have to be modified.
- 6.
The licensing bases and environmental impact for the unit as defined by the FSAR and the Technical Specifications are not altered.
In addition further assurances of safe operation of Salem Unit 1 at 3411 MWt have been provided by the operation of Salem Unit 2 at 3411 MWt.
Accordingly," the proposed uprate in Unit 1 core power starting with fuel cycle 7 does not represent a significant hazards consideration.
TABLE 1 SALEM UNIT 1 POWER CAPABILITY PARAMETERS NSSS Power, MWt Core Power, MWt Thermal Design Flow, Loop gpm Reactor Flow, Total, 106 lbm/hr Reactor coolant Pressure, psia Reactor Coolant Temperature, °F core outlet vessel outlet Core Average vessel Average vessel/Core inlet Steam Generator Outlet Steam Generator Steam Temperature, °F Steam Pressure, psia Steam Flow, 106 lbm/hr Total Feedwater Temperature, °F
.Present Power 3350 3338 87,300 132.3 2250 611.8 609.1 579.8 576.8 544.4 544.2 519.0 805 14.47 429.0 547 Zero Load Temperature, °F Percent Tube Plugging Core Bypass.Per~ent Fuel Design 17 x 0
4.5 17 STD 1132 Gross Electrical Output, MWe
- These parameters are identical to the Unit 2 values.
Uprated Power*
3423 3411 87,300 132.3 2250 17 x 613.7 610.8 581.0 577.9 545.0 544.8 519.0 805 14.86 432.B 547 0
4.5 17 STD 1158