ML18092A757

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SER Re Nss Uprating - 3423 Mwt
ML18092A757
Person / Time
Site: Salem PSEG icon.png
Issue date: 08/06/1985
From:
Public Service Enterprise Group
To:
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ML18092A756 List:
References
NUDOCS 8509050149
Download: ML18092A757 (48)
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Text

ATTACHMENT # l

ATTACHMENT.# 1 PUBLIC SERVICE.ELECTRIC AND GAS COMPANY SALEM UNIT 1 NSSS UPRATING - 3423 MWt SAFETY EVALUATION REPORT TABLE .OF CONTENTS -

OBJECTIVES v r

CONCLUSIONS vi

1.0 INTRODUCTION

1-1 1.1 Salem Unit 1 Uprating Program History 1-1 1 .2 NSSS Engineering Verification l -1 1 .3 Safety Evaluation l-2

  • l .4 Conclusions l -3 2.0 OPERATING PARAMETERS 2-1 2.1 Original Design Parameters 2-1 2.2 Uprated Operating Parameters 2-1 2.3 Comparison of Parameters 2-1 2.4 Conclusions 2-2 3.0 ACCIDENT ANALYSIS 3-1
3. l Introduction 3-1 3.2 Classification of Plant Conditions 3-1 3.3 Initial Power Conditions Assumed in the 3-2 Present Salem FSAR Accident Analysis
  • 3.4 Conclusions 3-2

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  • 2737e:ld/032285 .

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PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM UNIT 1 NSSS UPRATING - 3423 MWt SAFETY EVALUATION REPORT TABLE OF CONTENTS EQUIPMENl REVIEW 4-1 4.0 4., Design Transients 4-1

. 4.2 Primary Plant Components 4-1 4.2.1 Reactor Vessel 4-1 4.2.2 Reactor Vessel Internals 4-2 4.2.3 Reactor Coolant Pumps and Control Rod Drive 4-3 Mechanisms 4.2.4 Reactor Coolant Piping 4-3

  • 4.2.5 Pressurizer 4-3 Steam Generators 4-4 4.2.6 4.3 Auxiliary Equipment 4-4 4.3.1 Auxiliary Valves 4-4 4-4 4.3.2 Auxiliary Pumps 4.3.3 Auxiliary Heat Exchangers 4-5 4.3.4 Auxiliary Tanks, Demineralizers and Filters 4-5 BALANCE OF PLANT 5-1 5.0 5.1 Introduction 5-1 5.2 Mass and* Energy Release 5-1 5.3 Auxiliary Feedwater System 5-1 5.4 Source Terms for Offsite Dose Evaluations 5-1 5.5 Spent Fuel Pit Decay Heat Loads 5-2 5.6 Steam System Design Transients 5-2 5.7 RCS Loop Pipe Loads, Thermal Displacements 5-2 and Design Data
  • 2737e:1d/032285 ii

PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM UNIT l NSSS UPRATING - 3423 MWt SAFETY EVALUATION REPORT.

TABLE OF CONTENTS 5.8 Condensate and Feedwater Systems 5-2 5.9 Main Steam System 5-2 5.10 Component Cooling Water System 5-3 5 .11 Turbine 5-3 FLUID SYSTEMS DESIGN 6-1 6.0 6 .1 Results 6-1

    • 7.0 8.0 NUCLEAR FUEL REVIEW
7. l Results FSAR CHANGES 7-1
  • 1-1 8-1 8.1 FSAR B-1 8.2 Technical Specifications *a-2*

LIST OF TABLES iv LIST OF FIGURES iv LIST OF APPENDICES iv

  • 2737e:ld/032285 iii

PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM UNIT 1 NSSS UPRATING - 3423 MWt SAFETY EVALUATION REPORT LIST OF TABLES 2-1 ,Comparison of Reactor Coolant Systems Parameters 2-3 3-1 Nuclear Steam Supply Systems Power Ratings 3-4 Summary of Initial Condition and Computer Codes 3-5 3-2 Used

  • Figure LIST OF FIGURES 2-4 2-1 Reactor Coolant Temperatures versus Percent Rated Load APPENDICES BA FSAR Page Revisions BB Technical Specification Revisions
  • 2737e:ld/032285 iv

OBJECTIVES Public service Electric and Gas_Company (PSE&G) is applying to the Nuclear Regulatory Col1lfl'!ission for approval to operate Salem Unit l at a licensed power rating of 3423 MWt. Salem Unit l is currently licensed to operate at 3350 MWt. PSE&G has authorized Westinghouse to perform a safety evaluation of NSSS designs, operations, and analyses to provide the following information relevant to that application:

l. A des~ription of the proposed change in the licensed power rating of Salem Unit 1.
2. An assessment of the impact of that change on NSSS equipment designs, safety analyses, and systems operations.
3. Technical information to be used by PSE&G in support of its application for the increase in licensed power rating.
4. A technical basis for establishing that the proposed increase in power rating does not involve an unreviewed safety question in accordance with requirements of 10 CFR 50.59.

This report summarizes the results of the safety evaluation performed by Westinghouse, and presents the conclusions based upon it .

  • 2737e:ld/0322B5 v

CONCLUSIONS

  • The proposed increase in the licensed power rating of Salem Unit 1 has been reviewed in detail with respect to its impact on the following aspects of NSSS design and operation:
1. The co~sequences of ~ccidents postulated in the FSAR.
2. The capability of systems and equipment to meet design bases specified in the FSAR.
3. The capability of equipment to maintain ~tructural integrity under conditions defined in the FSAR.
4. Definition of NSSS/BOP safety related interfaces.
5. Operating limits and conditions contained in Technical Specifications.

that are impacted by the power rating increase.

This review has demonstrated that Salem Unit 1 is capable, in its present design configuration, of operating at the proposed power rating within the compliance specifications of the design criteria or safety limits contained in the FSAR for NSSS systems and equipment, providing that the*plant is operated in accordance with the Technical Specification changes proposed by Westinghouse. The review has verified th.at:

1. The probability of a malfunction of NSSS equipment importa~t to safety previously evaluated in the FSAR will not be increased at the proposed power rating.
2. The consequences of a malfunction of NSSS equipment important to safety previously evaluated in the FSAR will not be increased at the proposed power rating .
  • 2737e: 1d/03228'5 vi
3. The possibility of a malfunction of NSSS equipment important to safety
  • 4.

different from any already evaluated in the FSAR is not created by operation at the propo~ad power rating.

The margin of safety as defined in the bases to any technical specification will not be reduced by operation at the proposed power rating.

Therefore, it has been concluded that operation of Salem Unit at the increased pow~r rating does not reduce the NSSS safety margins, and does not involve an unreviewed question as defined by 10 CFR 50.59 .

  • 2137e:1d/0322B5 vii

i...us tor.:er l\ei -::rer.c~ i1G 1 s,.

PO #928644 Westinghouse Reference No(s). _

(Change Control or RFQ As Applicable)

  • GO #NA42059

-WESTINGHOUSE NUCLEAR SAFETY EVALUATION CHECK LIST

1) NUCLEAR PLANT(S) Sa 1em Unit 1
2) CHECK LIST APPLICABLE TO: 2~~ Power Uprating (Subject of Change) -----------------------

3°) The written safety evaluation of the revised procedure, design change or modificat~or required by 10CFRS0.59 has been prepared to the extent required and is attached. If a safety evaluation is not required or is incomplete for any reason, explain on Page Parts A and B of this Safety Evaluation Check List are to be completed only on the basis of the safety evaluation performed.

CHECK LIST - PART A (3. l ) Yes No x A change to the plant as described in the FSAR?

(3.2) Yes No

-xx A change to procedures as described in the.FSAR?

(3.3) Y~s No A test or experiment nat described in the FSAR?

(3.4) Yes x No A change to the plant.technical specifications (Appendix A to the Operating License)?

4) CHECK LIST - PART B (Justification for Part B answers must be included on Page 2.)

( 4. 1) Yes No X Will the probability of an accident previously evaluated in

- - the FSAR be increased? *

(4.2) Yes No X Wi 11 the consequences of an accident previou*s ly eva 1ua ted in

- - the FSAR be increased?

(4.3) Yes No X May the possibility of an accident which is different than any

- - already evaluated in the FSAR be created?

(4.4) Yes_ No~ Will the probability of a malfunction of equipment important t:

safety previously evaluated in the FSAR be increased?

(4.5) Yes No x Will the consequences of a malfunction of equipment important

- ----- to safety previously evaluated in the FSAR be increased?

( 4. 6) Yes No X May the possibility of a malfunction of equipment important to

- ----- safety different. than any already evaluated in the FSAR be created?

( 4. 7) Yes No X Will the margin of safety as defined in the bases to any

  • - - technical specification be reduced?

Page l of 2

If the ans*,.,,ers to any of the above questions are unknown, indicate under 5) REAAR:<s and explain below.

If the answer to any of the above questions in 4) cannot be answered in the negative based on written safety evaluation, the change cannot be approved without an applica-tion for license 9mendment submitted to NRC pursuant to 10CFR.50.90.

5) REMARKS:

None The following sur.mari~es the justification upon the written safety evaluation, (l) for answers given in Part 8 of the Safety Evaluation Check List:

The proposed 2~ uprating of Salem Unit 1 from 3350 MWt to 3423 MWt has been reviewed in detail This uprating will license Unit 1 to operate at the same power level as Unit 2, which has been operating safely at a power output of 3423 MWt. The Unit 1 uprating has been thoroughly evaluated in the areas of accident analyses, NSSS Systems, NSSS components, NSSS/BOP inter-face and FSAR and Technical. Speci.fication impact. Based on this revi.ew, it is concluded that a 2~ uprating is acceptable and involves no unreviewed safety question. Salem U~it 1 can be safely operated at a power output of 3423 MWt without undue risk to public hea]th and safety. *

~1 )Reference to document(s) containing w~itten sa~ety evaluation: __::;S~a~le~m~U~n~it.:::......:---~

NSSS Uprating 3423 MWt Safety Evaluation FOR FSAR UPDATE Section: Page(s): _ _ __ Table(s) : _ _ __ Figure(s): - - - - -

Reason for/Description of Change:

Prepared by {Nuclear Safety):

Coordinated ~ith Engineer(s):

  • rdinating Group Manager(s):

Nuclear Safety Group Manager:

Page 2 of 2

SECTION 1 INTRODUCTION Public Servic~ Electric and Gas Company (PSE&G) is engaged in a program to increase the electrical output of Salem Unit 1. The program is directed toward obtaining approval from the USNRC to operate the plant at a slightly increased power level. At *present. Salem Unit 1 is licensed to operate at an NSSS power rating of 3350 MWt. PSE&G is applying for an amendment to the operating license that will permit operation at 3423 MWt. an increase of 2.2%.

As a.part of the program.to uprate Salem Unit 1. PSE&G authorized Westinghouse to perform a review of NSSS systems and equipment designs to verify their capability to meet requirements for operation at 3423 MWt. That review was conducted in accordance with groundrules and criteria put forth in Westinghouse topical report WCAP-10263, "A Review Plan for Uprating the Licensed Power of a Pressurized Water Reactor Power Plant 11

  • A summary of the major guidelines followed in the NSSS design review follow:
l. Scope of Review The review encompassed all. aspects of the Salem Unit 1 NSSS design and operation that were impacted by the power increase.
2. Safety Review Acceptance Criteria NSSS designs hav~ been reviewed to verify compliance at the increased power rating with licensing criteria and standards currently required by the Salem Unit 1 operating license. In addition. a review has been made as defined in 10 CFR 50.59 to identify any potential unreviewed safety question that might occur as a result of the increased power rating .
  • 2737e:ld/032685 1-1
3. Structural Review Acceptance Criteria The structural design 0£ NSSS equipment was reviewed to assure that compliance has been* maintained at the increased power rating with industry codes and standards that applied when the equipment was originally built.
4. Functional Capability A revjew has been made to verify that NSSS components and systems will continue to meet functional requirements specified in the FSAR at the increased power rating.
5. Analytical Techniques Current NRC approved analytical techniques have been used for analyses performed at the increased power rating.

&. Balance of Plant Interfaces Information provided by Westinghouse to other design groups has been reviewed and revised when impacted by the increase in power rating .

  • 2737e:.ld/032285 1-2

2.0 OPERATING PARAMETERS 2.1 ORIGINAL DESIGN PARAMETER~

Salem Unit 1 was originally designed to operate at a guaranteed NSSS power rating of 3350 MWt with a steam pressure of 805 psia. NSSS operating parameters for these conditlons are listed in Table 2-1.

2.2 UPRATED OPERATING PARAMETERS At uprated power conditions, Salem Unit 1 will operate at 3423 MWt with 805 psia steam pressure. Table 2-1 lists primary plant parameters for the uprated power operating conditions. Note that the uprated Salem Unit 1 parameters are identical to that of Salem Unit 2.

2.3 COMPARISON OF PARAMETERS A comparison of the operating parameters listed in Table 2-1 shows that operating conditions at the uprated power are identical to the Salem Unit 2 design conditions. A detailed comparison of the more significant parameters indicates the following:

1. NSSS Power The uprating will increase the NSSS thermal power by about 2.2% when compared to the Salem 1 original design conditions.
2. Reactor Flow rnermal design calculations at the current power of 3350 MWt are based on a reactor inlet flow of 349,200 gpm. Thermal design calculations for the uprated conditions were based on the same reactor inlet flow of 349,200 gpm. This is the same thermal design flowrate used for Salem Unit 2 .
  • 2737e:ld/032685 2-1
3. Reactor Coolant Temperatures As shown by Table 2-1, reattor coolant t~mpcratures for the uprated condition~ do not differ significantly from those for the current 3350 MWt power level. As would be expected, the higher power level is reflected by

~ slightly greater temperature rise in the coolant as it passes through the reactor vessel. Figure 2-1 provides a graphical comparison of the reactor vessel cold leg, hot leg and vessel avera~e ~emperatures. It shows that there is little difference between these parameters for the current ~nd uprated operating modes throughout the power range.

4. Steam Pressure The steam pressure is being maintained at 805 psia and is identical to the steam pressure 9f Salem Unit 2.
5. Steam Flow Steam flow at the 3423 MWt conditions has increased over th~ 3350 MWt condition roughly in proportion to the thermal power increase of 2.2 percent.

2.4 CONCLUSION

S Comparison of the data presented in Tabli 2-1 and in figure 2-1 leads to the conclusions that conditions proposed for future 3423 MWt operation are:

o Not significantly different from those for the original plant design

.at 3350 MWt with 805 psia steam pressure.

o Identical to currently licensed operating conditions for Salem Unit 2.

2731e:ld/032285 2-2

    • TABLE 2-1 COMPARISON OF ~EACTOR COOLANT SYSTEM PARAMETERS Salem Unit l Salem Unit 2 Design Conditions Design Conditions Current Uprated Current L1 cense License License 3350 3423 3423 NSSS Power, MWt 3338 3411 3411 Reactor Power: MWt Reactor Coolant Pump Heat, MWt 12 12 12 349,200 349,200 349,200 Reactor Flow, Total, gpm Reactor Flow, Total, million lbm/hr 132.3 132. 2 132. 2 2250 2250 2250 Reactor Coolant Pressure, psia Reactor Coolant Temperature, °F 61l.8 613. 7 613.7 Core Outlel 609.1 610.8 610.8 Vessel Outlet 579.8 581 . 0 581.0 Core Average 576.8 577.9 577.9 Vessel Average 544.4 545.0 545.0 vessel/Core Inlet 544.2 544.8 544.8 Steam Generator Outlet Steam Generator 519.0 519.0 519 .0 Steam Temperature, °F 805 805 805 Steam Pressure, psia 14.47 14.86 14.86 Steam Flow, Total, million lbm/hr 547 547 547 Zero Load Temperature, °F 4.5 4.5 4.5 Core Bypass, percent 17x17 17x17 17x17 Fuel Design
  • 2737e:ld/03268~

2-3

Figure 2-1 REACTOR ~OOLANT TEMPERATURES VERSUS PERCENT RATED LOAD 620 REACTOR VESSEL

/ HOT LEG TEMPERATURE 600

0 L&J C::.

~

~

c::

L&J 580 h

/

REACTOR VESSEL AVERAGE TEMPERATURE c..

E:

L&J D

I- Q 560 f-------=-=-='---=---=--=-=-=-=-=-::....;.;*= REACTOR VESSEL COLD LEG TEMPERATURE 540 UPRATED PRESENT 0 20 40 60 80 100

% RATED THERMAL LOAD

  • 2-4

3.0 ACCIDENT ANALYSIS

3. 1 INTRODUCTION The following discussion summarizes the safety analysis evaluation performed to assess the effect of increasing the rated thermal power for Salem Unit 1 from 3350 MWt to 3423 MWt with respect to Salem Generating Station FSAR Chapter 15.

3.2 Classification of Plant Conditions r

Since 1970 the American Nuclear Society (ANS) classifica~ion of plant conditions has been used which divides plant conditions into four categories in accordance with anticipated frequency of occurrence and potential radiological consequences to.the public. The four categories are as follows:

0 Condition I: Normal Operation and Operational Transients 0 Condition II: Faults of Moderate Frequency 0 Condition II I: Infrequent Faults 0 Condition IV: Limiting Faults The basic principle applied in relating design requirements to each of the conditions is that the most probable oc~urrences should yield the least radiological risk to the public and those extreme situations having the potential for the greatest risk to the public shall b~ those least likely to occur. Where applicable, reactor trip system and engineered safeguards functioning is assumed to the extent allowed by considerations, such as the single failure criterion, in fulfilling this principle.

2737e:ld/032285 3-1

3.3 INITIAL POWER CONDITIONS ASSUMED IN THE PRESENT SALEM FSAR ACCIDENT.

ANALYSIS Table 4-1 lists the principal power ratfog values which were assumed in analyses performed in the Salem FSAR. Two ratings are given:

l. The guaranteed Nuclear Steam Supply System thermal power output: This power output includes the guaranteed core thermal power generation and the thermal power generated by the reactor coolant pumps.
2. The Engineered Safety Features design rating: The Westinghouse supplied Engfoeered Safety Features are designed for a thermal po\-1er higher than the guaranteed value in order not to preclude realization of future potential power capability. This higher thermal power value is designated as the Engineered Safety Features design rating. This power output includes the thermal power generated by the reactor coolant pumps.

Where initial power operating conditions are assumed in accident analyses, the "guaranteed Nuclear Steam Supply System thermal power output" plus allowance for errors in st~ady state power determination is assumed. Where demonstration of adequacy of the containment and Engineered Safety Features are concerned, the "Engineered Safety Features design rating" plus allowance for ~rror is assumed. The thermal power values for each transient analyzed are given in Table 3-2.

3.4 CONCLUSION

S The currently docketed.Salem FSAR LOCA and non-LOCA ac~idents that are not zero po\'l!er transients were performed at ~ 3423 M\<lt NSSS power 1evel.

Therefbre no additional analysis needed be performed. Revisions to the Salem FSAR and technical specifications are required. These page changes are presented in Section 8.

2737e:ld/032685 3-2

Operation of Salem Unit l at the increased power rating of 3423 MWt docs not reduce the NSSS safety margins, and does not involve an unreviewed question as defin~d by 10 CFR 50.59. Sale~ Unit l is c~pable, in its present design configuration~- of operating at 3423 MWt within the compliance specifications of the design criteria or safety limits specified in the Salem FSAR.

2737e:ld/032285 3-3

TABLE* 3-1*

  • NUCLEAR STEAM SUPPLY SYSTEM POWER RATINGS Guaranteed Nuclear Steam Supply System thermal power output 3423 MWt The Engineered Safety Features design rating (maximum calculated turbine rating) 3577 MWt Thermal power generated by the reactor 12 MWt coolant pumps Guaranteed Core Thermal Power 3411 MWt
  • Tabl~ 15.1-1 from Salem FSAR .
  • 2737e:ld/032285 3-4
  • TABLE 3.2* (Sheet 1 of 4)

SUHHARY OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED INITIAL NSSS MOOERATOR(l) THERMAL POWER OUTPUT

. ' I COMPUTER TEMPERATURE DENSITY ASSUMED CODES USED (llk/9F) (AK/gm-cc) OOPPLER( 2) (MWt)

CONDITION 11 Uncontrolled RCC assembly Bank Withdrawal WIT-&, FACTRAN Lower 0 from a Subcritical Condttton Uncontrolled RCC Assembly Bank Withdrawal LOFTRAN 0 and 0.43 lower and 342'3 at Power upper RCC Assembly Misaltg11111ent TH I NC, TURTLE, 0 upper 3423 LOFT RAN Uncontrolled Boron Oilutton NA NA NA NA 0 and 3423 Partial Loss of Forced Reactor Coolant PHOENIX, LOFTRAN 0 upper 2391> and Flow THINC, FACTRAN 3423 Start-up of an Inactive Reactor Coolant HARVEL, THINC 0.43 lower 231>9 Loop Loss of External Electrical Load and/or LOFT RAN 0 and 0.43 upper 3423 Turbine Trip Loss of Nonnal Feedwater BLIWUT NA NA 3571 Loss of Off-Site Power to the Plant BLKOUT NA NA 3423 Auxiliaries (Plant Blackout)

  • Table 15.1-2 of Salem FSAR 2137e: ld/032285 3-5

TABLE 3.2 (Sheet 2 of 4)

SUHHARY OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED INITIAL NSSS MODERATOR(l) HODCRATOR(l) THERMAL* POWER OUTPUT COMPUTER TEHPERAlURE DENSITY ASSUMED CODES USED (Ak/*f) (ftK/gm-cc) DOPPLER( 2) (MWt)

CONDITION II (Cont'd.)

Excesstve Heat Removal Due to Feedwater HARVEL 0.43 lower 0 and 3423 System MaHunct tons Excesstve Load Increase LOFT RAN 0 and 0.43 lower 3423 Acctdental Depressurtzatton of the LOFT RAN 0 upper 3423 Reactor Coolant System Acc1dental Depressurtzatton of the HARVEL Function of Mod- -2.2 pcm/PF 0 Ma1n Steam Sys~em erator Denstty (Subcrittul)

See Sec . 15. 2 . 13 (Fig. 15.2.41)

Salem FSAR Inadvertent Operatton of £CCS During . LOFTRAN 0 lower 3423 Power Operation CONDITION Ill Loss of Reactor Coolant from Small WF~ASH, LOCTA-R2 3571 RuP.tured Pipes or from Cracks in Large Pipe whtch Actuate Emergency Core Cooling 2137e: 1d/D3~285 3-6

  • TABLE 3.2 (Sheet 3 of 4)

SUMMARY

OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED INITIAL NSSS HODERATOR(l) MODERATOR(l) THERHA~ ,PqwER OUTPUT COMPUTER TEMPERATURE DENSITY ASSUMED CODES USED UkrFl UK/gm-cc) OOPPLER( 2) (HWt)

CONDITION Ill (Cont'd.)

Inadvertent Loading of a Fuel Assembly LEOPARD, TURTLE NA NA 3423 into an Improper Position Complete Loss of Forced Reactor Coolant PHOENIX, LOFTRAN 0 upper 23% and Flow THINC, FACTRAN 3423 Waste Gas Decay Tank Rupture NA NA NA 3517 Single RCC Assembly Withdrawal at TURTLE, THINC NA NA 3423 Full Power LEOPARD CONDITION IV Major rupture of ptpes contatntng reactor SATAN Function of Function of coolant up to and Including double-ended LOCTA-R2 Moderator Fuel Temp.

rupture or the largest ptpe tn the Reactor density See See Section Coolant Syste~ (Loss of Coolant Accident) Section 15.4.1 15.4.l Salem FSAR Salem FSAR Major secondary system pipe rupture up HARVEL, THINC Function of -2.2 pcmrF 0 0

to and Including double-ended rupture Moderator (Subcrttlcal)

(Rupture of a Steam Pipe) Density See Section 15.2.13 (Fig. 15.2-41)'

Salem FSAR

/

( 3-7 2l31e: 1d/OJ2205

TABLE 3.2 (Sheet 4 of 4)

SUHHARY OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUHED lNlT IAL NSSS MODERAlOR(l) MODERATOR(l) THERMAL.POWER OUTPUT COMPUTER TEMPERA TUR[ DENSITY ASSUMED CODES USED (Ak/9f) (AK/9111-cc) DOPPLER( 2) (HWt)

CONDITION IV (Cont'd.)

Steam Generator Tube Rupture NA NA NA NA 3577 S1ngle Reactor Coolant Pump Locked PHOENIX, LOFT RAN 0 upper 239~ and Rotor THI NC, FACTRAN 3423 Fuel Handling Accident NA NA NA 3577 Rupture of a Control Rod Mechanism TWINKLE, FACT RAN -1 pcmrF BOL Consistent 0 and 3423 Housing (RCCA Ejection) LEOPARD -2& pcm/9F EOL with lower limit shown Fig. 15 .1-5 NOTES:

(1) Only one 1s used in an analysis i.e., either moderator temperature or moderator dens1ty coefftcient (2) Reference Figure 15.1-5 of Salem FSAR 3-8 2137e:ld/OJ2285

4.0 EQUIPMENT REVIEW

  • 4.1 DESIGN TRANSIENTS Design transients originally used for Salem Unit 1 and 2 components are contained in Westinghouse System Standard Design Criteria 1 .3 Rev. 1, Nuclear Steam Supply_ System Design Transients. These transients were conservatively based on a thermal power level in excess of the 3423 MWt rating and, therefore, remain bounding for the Sa1em 1 uprating. The above statement is valid based on the following:
1) The original licensing basis remains applicable;
2) The same assumptions remain applicabl*; and
3) The same calculational techniques are applicable.

4.2 PRIMARY PLANT COMPONENTS 4.2.1 REACTOR VESSEL To assess the impact of the uprating on the reactor vessel design and opera-tion, the vessel design specification, stress report, and fracture mechanics analyses were considered.

The reactor vessel design specification was reviewed and the changes associated with the increased temperatures and power rating were incorporated.

The reactor vessel stress report was reviewed and where appropriate updated to reflect the duty cycle associated with the uprating. In all instances the updated data for stress intensity range and usage factor remain in compliance with ~he original design bases limits.

A review of the reactor vessel fracture mechanics evaluation revealed that the currend end-of-life fluence calculation was based on a core power of 3483 MWt. Therefore, the uprated power level fluence is bounded by the current analysis and the currently docketed data remains applicable .

  • .2737e:1d/032685 4-1

4.2.2 REACTOR VESSEL INTERNALS Review of the Unit 1 reactor vessel -lower internals and comparison with Unit 2 revealed a des~g~ difference in the baffle barrel region former flow holes.

Unit 1 was designed with flow holes of varying diameter in each former and from one former level to the next. Unit 2 was designed with uniform diameter holes in each former and at.each former level.

Because of this difference it was necessary that a thermal-hydraulic analysis of the internals be performed to evaluate the effects of the power uprating; The results of this hydraulic review indicated negligible differences in core bypass flow and pressure drops and therefore the current data remain bounding for operation of the Unit 1 internals at the Unit 2 power level. In addition, hydraulic lift forces and the RCCA scram time changed insignificantly.

A thermal stress analysis was performed for the lower internals to assess the.

Impact of the higher core radiation heat generation rates associated with th~

uprating.* The increased component gamma heating was accounted for by increasing the heat generation rates to the level corresponding to the uprated power level. The reiults indicate that the radial, axial and azimuthal temperatures change by less than 2°F in the core baffle plates and core barrel. This results in a negligible effect on the performance of the internals components.

Unit 1 and Unit 2 also have differing control rod guide tube*arrangements which were determined to be inconsequential with respect operation at the uprated conditions.

Based upon the above results, the Salem Unit l uprating to 3423 MWt has been demon~trated to be acceptable and remain in compliance with the original design cr~teria and bases .

  • 2737e:ld/032685 4-2

4.2.3 REACTOR COOLANT PUMPS ANO CONTROL ROO DRIVE MECHANISMS For the 3423 MWt uprating program, the cold leg temperature at 100 percent

. power increases from 544.4°F to 545.0°F and the hot leg temperature increases from 609.l°F to 610.8°F. The component design ~ressure and temperature conditions and design transients are unchanged. *The proposed 3423 MWt operating tem~eraturcs are bounded by ~hose contained in the original equipment spec1fications and associated stress reports, therefore, no additional thermal and ~tructural analysis were required at 3423 MWt for the reactor coolant pumps and control rod drive mechanisms to demonstrate compliance with codes and standards applicable to Salem Unit 1. The 3423 MWt uprating changes are bounded by the existing thermal and structural analysis.

No equipment modifications or revised operating limits associated with these components are appropriate or necessary at the uprated conditions.

4.2.4 REACTOR ~OOLANT PIPING Review of the thermal analysis of the reactor coolant loop, piping and

  • supports was. conducted for all Class I piping systems in Westinghouse scope.
l. All piping stresses for the uprated conditions are enveloped by those documented in the current stress report. The original design contaJned adequate margin to envelope the uprated conditions;
2. All support loads are within engine~ring tolerances of the original loads and require no additional examination. No reanalysis was necessary for the implementation of the uprating.

4.2.5 PRESSURIZER Each pressurizer component analy~ed in the original stress report was reviewed to identify any modifications required for operation at uprated conditions.

Results of the evaluation showed that the existing pressurizer stress report satisfies all applicable ASME Code requirements when the Unit l parameters are 2737e:ld/032685 4-3

revised to reflect the enveloping plant operating parameters of the uprating.

The impact of the uprating on the pressurizer is bounded, and no equipment modificatiori was required.

A separate review was performed to assess the adequacy of the pressurizer spray, safety and power operated relief valves for operation at 3423 MWt.

Results indicate that the presently installed valve capacities were based on the Unit 1, 3423 MWt. power level and are sufficient for the uprated conditions.

4.2.6 STEAM'GENERATORS Since both Salem Unit 1 and 2 have Model 51 Series S/G's. the uprating of Unit 1 to the power level of Unit 2 is enveloped by the current design bases and a new stre~s analysis was not required. Thus the Unit 1 steam generator design is adequate for the uprated power level.

4.3 AUXILIARY EQUIPMENT The auxiliary equipment for Salem Unit 1 affected by the uprated parameters has been reviewed and found to be functi~nally or physically identical to Unit 2.

4.3.1 AUXILIARY VALVES The Salem 1 safety, POR, pressurizer spray and auxiliary valves were evaluated for a power-uprating to the Unit 2 power of 3423 MWt. *westinghouse records indicate the valves for both units are identical and hence the Unit 1 valves will operate acceptably at th~ Unit 2 parameters.

4.3.2 AUXILIARY PUMPS Auxiliary pumps for Salem Unit 1 were designed and supplied to the same design and functional requirements as Salem Unit 2. fhcrcfore, these pumps w111

  • 2737e:ld/032685 4-4

support a power uprating equivalent to the Unit 2 rating of 3423 MWt without modification or alteraiions.

4.3.3 AUXILIARY HEAT EXCHANGERS Auxiliary heat exchangers (e.g., regenerative, non-regenerative, seal water return, and RHR heat exchangers) for Salem*uniL 1 were designed and supplied to the same design and functional requirements as Salem Unit 2. Therefore, these heat exchangers arc adequate and capable of performing at Lhe uprated power of 3423 MWt.

4.3.4 AUXJLIARY TANKS, DEMINERALIZERS AND FILTERS Auxiliary tanks, deminerallzers and filters for Salem Unit 1 were designed and supplied to the same design and functional requirements as Salem Unit 2.

Therefore this equipment is not significantly affected by the Unit 1 upratin~ .

2737e:ld/032&85 4-5

S.0 BALANCE OF PLANl

5. 1 1NTRODUCf ION To coordinate *the NSSS review 1r1ith the Balance of Plant (BOP) rcvie'l!S, data provided by Westinghouse as input to the BOP design was examined to identify those areas where revisions.might be required. Westinghouse conducted a review to verify the bases and to confirm the continued applicability of data originally supplied to PSE&G considering the NSSS uprating to 3423 MWt. The following sections describe the more significant areas where design data was confirmed by'Westinghouse.

5.2 MASS AND ENERGY RELEASE The original loss of coolant accident mass and energy release data provided as Input for containment integrity was based upon a power rating of 3579 MWt.

This data bounds the power Increase to 3423 MWt. The main steamline break data was based on the event occurring at the no load condition which is

  • unchanged.

5.3 AUXILIARY FEEOWATER SYSTEM The original auxiliary feedwater system is identical to the auxiliary feedwater system in Salem Unit 2 which is currently operating at 3423 MWt/805 psia steam. Since the auxiliary feedwater requirements presented in the Salem FSAR cover both Salem Units 1 and 2 and were designed to 3577 MWt, the auxiliary feedwater system requirements r& Jii1.1 unchanged.

1 \~

t'efl\111.\"3 5.4. SOURCE TERMS FOR OFFSITE DOSE EVALUATIONS The data in the current Salem FSAR are based upon a core power of 3558 MWt.

The source terms are essentially only a function of core power and burnup.

The increase in core power from 3338 Mt*Jt to 3411 Mt-Jt is still within the bounds of the Salem FSAR. The current Salem FSAR data remain valid. *

  • 2737e:ld/03268S 5-1

5.5 SPENT FUEL PIT DECAY HEAT LOADS The decay heat_ of the fuel 1n the spent fuel *pH is a function of core power and burnup. Each Salem Unit has completely independent spent fuel pit cooling systems. The components and systems are identical for both units. Since Salem Un1t 2 is presently operating at 3423 MWt and the requirements presented in the Salem FSAR remain unchanged, the data employed in the original design remain unchanged.

5.6 STEAM S~STEM DESIGN TRANSIENTS The steam system transients provided in the Westinghouse Steam Systems Design Manual are unchanged.

5.7 RCS LOOP PIPE LOADS, THERMAL DISPLACEMENTS AND DESIGN DATA Based on a detailed review, it was determined that any changes in loadings and

  • piping/support thermal displacements and other design data are within the bounds of the original evaluation.

5.8 CONDENSATE AND FEEDWATER SYSTEMS The condensate and feedwater.systems in Salem Unit 1 are identical to the condensate and feedwater systems in Salem Unit 2 which is presently operating at 3423 MWt/805 psia steam. S~nce the r~quirements presented in the Salem FSAR cover both Salem Units 1 and 2, the condensate and feedwater systems requirements remain unchanged.

5.9 *MAIN STEAM SYSTEM The main steam system for Salem Unit 1 was designed at the higher* Salem Unit 2 design flowrate of 3.7 x 10 6 pounds per hour for each steam generator. The current Salem FSAR data remain valid .

  • 2737e:ld/032285 5-2

5.10 COMPONENT COOLING WATER SYSTEM Salem Units 1 and 2 have similar component cooling systems. They differ only in that Salem}Jnit 1 has one tube and shell type heat exchanger and one plate type heat exchanger whereas Salem Unit 2 has two tube and shell type heat exchangers. This difference is inconsequential as each type has a design heat transfer rate of 44.2 x 10 6-BTU/hr. Salem Unit 2 is presently operating at 3423 MWt/805 psia steam. The code requirements and minimum flow requirements presented in the Salem FSAR are applicable to both units. Therefore the component cooling system for Salem Unit 1 is acceptable for operation at 3423 MWt/805 psia'steam.

5.11 TURBINE

  • Based on a review~ it was determined that the assumptions, analyses, and evaluations performed to verify the operating characteristics and structural*

integrity of the turbine bound operation at 3423 MWt/805 psia steam .

2737e:ld/032685 ~3

6.0 FLUID SYSTEMS REVIEW 6.1 RESULTS The Salem Unit 1 uprating engineering implementation program included assessments of the following systems: the Reactor Coolant System, Safety Injection System, Chemical and Volume Control System, Residual Heat Removal System, Spent Fuel Pit Cooling System and Sampling System.

The systems ~esign documentation was reviewed and found to be consistent with the uprated power level of Unit 2. The review confirmed that the fluid system design parameters of Unit 2 (at 3423 MWt NSSS power) were in fact. used to size the systems and equipment for both units.

All of the above mentioned systems will perform their functions at the uprated power in accordance with all current Salem Unit 1 design basis and criteria.*

  • 2737e:ld/032685 6-1

7.0 NUCLEAR FUEL REVIEW 7.1 RESULTS The review to determine the effects of the Salem Unit l uprating to 3411 Mwt nuclear power on the fuel design was conducted an a basis consistent with PSE&G's desires to convert Unit l to 18 month fuel cycles. Both the fuel currently in the core and reload fuel were reviewed with respect to nuclear design, thermal hydraulic design and fuel performance.

r The nuclear design of the core at 3411 MWt will be addressed as part of the normal design process for Cycle 7.

The thermal hydraulic design of Salem Unit l is already being performed at 3411 MWt core power. No additional effort was required to verify acceptability.

An analysis of the impact of the Salem Unit l uprating has indicated that sufficient margin exists in the design of the fuel presently in the core to permit operation at the uprated power level. Reload fuel design analyses have been amended using the 3411 MWt core power design requi~ements. The fuel presently in the core is capable of being operated at either 3411 MWt or at the current 3338 MWt core power level.

2737e:ld/032685 7-1

S.0 FSAR CHANGES The following describes the required revisions to the FSAR to reflect the change in power rating of Salem Unit 1 from 3350 MWt NSSS power to 3423 MWt NSSS power. In accordance with PSE&G's request, handwritten mark-ups of the revised FSAR pages are included in appendices SA and SB.

S. 1 FSAR SECTION 4. 1

1. Table 4.1-lA (Sheets 1, 2, and 3) should be replaced with the attached sheets.
2. Table 4.1-lA (Sheet o of 6) contains table headings with no information. This page should be eliminated and the footnotes moved to the previous pages of the Table .
  • SECTION 4. 3
1. In Section 4.3.1 .1, Fuel Burnup, µnder Basis, the average region discharge burnup of 33,000 MWD/MTU should be changed to 3S,000 MWD/MTU to reflect the current contract discharge burnup. This revision is not a result of the uprating.
2. In Section 4.3.2.1, ~uclea~ Design Description, the average region discharge burnup of 33,000 MWD/MTU should be changed to 3S,000 MWD/MTU to reflect the current contract discharge burnup. This revision is not a result of the uprating.
3. In Section 4.3.2.2.o, Limiting Power Distributions, the average kw/tc should be changed from 5.33 kw/ft to 5.44 kw/ft. This is the resultant change in linear power density in uprating from 333S MWt to 3411 MWt .
  • 2737e:ld/0326S5 8-1

SECTION 4.4

1. Tablel 4.4-lA, 4.4-2A,-and 4.4-3A should be: replaced with the attached sheets.

8.2 TECHNICAL SPECIFICATIONS

1. In Section 1 .0 Definitions, subsection 1.3 Rated Thermal Power, 3338 MWt should be changed to 3411 MWt which incorporates the uprated power level.
2. In Table 3.2-1, DNB parameters (pg. 3/4 2-14), change Reactor Coolant System Tavg to 582°F. Delete 3 Loops 1n operation limits .
  • 2737e:ld/032285 8-2

APPENDIX BA FSAR PAGE REVISIONS HAND MARKED

  • 2737e:ld/032285 8-3

TABLE 4.1-lA (Sheet 1 of 6)

REACJOR DESIGN COMPARISOI~ TABLE ..

Salem Unit 1 Salem Unit 1 17xl7 Fuel Assembly 15xl5 Fuel Assembly With Densification Without Thennal and Effects Densificati6n Effects Hydraulic Design Parameters

~~411

~.3~ , , , ... l.

1. Reactor Core Heat Output, MWt
2. Reactor Core Heat Output, Btu/hr

"~~ x 106 t1*"~3 *H,393" x 10 6

97.4 97 .4

3. Heat Generated in Fuel , *;

2250 2250

4. System Pressure, Nominal, psia
5. System Pressure, Min. Steady 2220 2220 State, psi a
6. Minimum DNBR at Nominal Con- ~::caJ ditions Typical Flow Channel, Thimble (Cold Wall) Flow f .&'O

~

Channel

7. Minimum DNBR for Design

>l.30 >l.30 Transients Coolant Flow 0

~

132. l x 10 6 134.j° x 10 6

8. Total Thennal Flow Rate, 1b/hr
9. Effective Fl ow Rate for Heat .3 6 6 126.,4' x 10 128.0 x 10 Transfer, lb/hr
10. Effective Fl .ow Area for !Jea t 51.1 51. 2 Transfer, ft2
11. A'verage Velocity Along Fuel 15.5 15.3 Rods, ft/sec
12. Average Mass Velocity, 2.47 x 10 6 2.so x 10 6 1b/hr-ft 2
  • SGS-UFSAR Revision 0 July 22, 1982

TABLE 4.1-lA (Sheet 2 of 6)

REACTOR DESIGN COMPARISON TABLE Salem Unit 1 Salem Unit 1 17xl7 Fuel Assembly 15xl5 Fuel Assemblj Thennal and With Densification Without Hydraulic Design Parameters Effects Densification Effects Coolant Temperature, °F

~s'ls.o 544 * ..-s~s.o

13. Nominal Inlet
14. Average Rise in Vessel .6A-rl- (, s. 1 ~ (,5,/

~ hS.7 ~ 6?.i

15. Average Rise in Co re
16. Average in Core .i-t9.8 Si/.l) SH.1 sao.4

~ ,~"17.S"

16. Average in Vessel S:f-6.7'"" 517. 9 Heat Transfer
18. Active Heat Transfer, Surface Area, ft 2 59, 700 52, 200 2 res, 700 lt'11o" 2~'0 ~17J..o.;;
19. Average Heat Flux, Btu/hr-ft
20. Maximum Heat Flux for Nonnal "(4/IJ:J-0()

Operation, Btu/hr-ft 2 36, geie [ b J 580 ,000

21. Average Thennal Output, kw/ft ~s ........ ..e-.ae- 7. () 3
22. Maximum Thennal Output for Nonnal Operation, kw/ft 18.8
23. Peak 1 inear power for deter-mination of protection set-points, kw/ft
  • la.oCdJ
24. Heat Flux Hot Channel Factor 2.32[c]

FQ

  • SGS-UFSAR Re vision 0 July 22, 1982

I . .

TABLE 4.1-lA (Sheet 3 of 6)

REACTOR DESIGI~ COMPARISON TABLE Salem Unit 1 Salem Unit 1 17x17 Fuel Assembly 15x15 Fuel Assembly Thennal and Hydraulic With Densification Without Design Parameters Effects Oensification Effects Fuel Centra1 Temperature, °F 3+oa

25. Peak at 100 Percent Power -B5(t 4250
26. Peak at Maximum Thennal Output for Maximum Overpower Trip Point 4150 Core Mechanical Design Parameters Fuel Assemblies
27. Design RCC Canl ess RCC Canless
28. Number of Fuel Assemblies 193 193
29. U0 2 Rods per Assembly 264 204
30. Rod Pitch, in. 0.496 0.563
31. Overall Dimensions, in. 8.426 x 8.426 8.426 x 8.426
32. Fuel Weight (as U0 2 ), pounds 222,739 215,400
33. Zircaloy Weight, lbs. 50 ,913 48,250

.34. Number of Grids per Assemoly 8-Type R 7-Type L

35. Loading Technique 3 region non-unifonn 3 region non-unifonn Fuel Rods
36. Number 50,952 39*, 372
37. Outside Diameter, in. 0.374 0.422
38. Diarnetral Gap, in., Regions 1, 2, (and 3} 0.0065 o.ou75 (O.ooas)

SGS-UFSAR Re vi sic n 0 July 22, 1982

TABlE ._4.1-lA (Sheet 6 of 6). * .....

\ / "\

REACTOR DESIGN COMPARISON TABLE I i \

\

\

Salem Unit j I

sa*1 em Unit \

\

\l7x17 Fuel As~embly 15x15 Fuel Assembly I I , I

"'- I ~ith Densiftcation Without \

l~uclear DesignI Parameters* \ Effects Densification Effects

./

[a] Previously, the value of 2.09 for a limiting typical channel was quoted only since the thimble (cold wall) DNB tests were incomplete.

[b] This limit is associated with the value of FQ = 2.32.

[c].Includes the effect of fuel densification.

[d] See Section 4.3.2.2.6 *

  • SGS-UFSAR Revision 0 July 22, 1982

Condition Ill incidents shall not cause more than a small fraction of the fuel elements in the reactor to be damaged, although sufficient fuel element dama~e might occur to-preclude immediate resumption of opera-tion. The release of radioactive material due to Condition Ill.inci-dents should not be sufficient to interrupt or restrict public use of these areas beyond the exclusion radius. Furthennore, a Condition III incident shall not, by itself generate a Condition IV fault or result in a consequential loss of function of the Reactor Coolant System or reactor containment barriers.

Condition IV occurrences are faults that are not expected to occur but are defined as limiting faults which must be designed against. Con-dition IV faults shall not cause a release of radioactive material that results in an undue risk to public health and safety.

The core design power distribution limits related to' fuel integrity are met for Condition I occurrences through conservative design and main-tai ned by the action of the Control System. The requirements for Condition II occurrences are met by providing an adequate protection system which monitors reactor parameters. The Control and Protection Systems are described in Chapter 7, and the consequences of*condit}on II, III and IV occurrences are given in Chapter 15.

4.3.1.1* Fuel ~urnup Basis The f~el rod design basis is described in Section 4.2. The nuclear design basis is to install sufficient reactivity in the fuel to attain

, .3Booo an average region discharge burnup of 33,000 MWD/MTU. The above, along with the design basis in Section 4.3.1.3, Control of Power Distribution~*

satisfies GDC-10.

SGS-UFSAR 4.3-2 Revision O July 22, 1982

fuel on the core periphery, with depleted fuel moved inward. The cores

  • will normally operate approximately 11,000 MWD/MTU per year. The enricllnents for the first cores-are shown in Table 4.3-1.

The core average enricllnent is detennined by the amount of fissionable material required to provide the desired core lifetime and energy 3ioOO requirements, namely a region average discharge burn up of 3-3,000 MTD/MTU. The physics of the burnout ~rocess is such that operation of the reactor depletes the amount of fuel available due to the absorption of neutrons by the U-235 atoms and their subsequent fission. The rate of U-235 ~epletion is directly proportional to the power level at which the reactor is operated. In addition, the fission process results in the formation of fisson products, some of which readily absorb neu-trons. These effects, depletion and the buildup of fission products, are partially offset by the buildup of plutonium shown in Figure 4.3-2 for the 17 x,17 fuel assembly, which occurs due to the non-fission absorption of neutrons in U-238. Therefore, at the beginning of any cycle a reactivity reserve equal to the depletion of the fissionable fuel and the buildup of fission product poisons over the specified cycle life must be "built" into the reactor. This excess reactivity is con-trolled by removable neutron absorbing material in the form of boron dissolved in the primary coolant and burnable poison rods.

The concentration of boric acid in the primary coolant is varied to provide control and to COlilpensate for. long-term reactivity requir~

ments. The concentration of the soluble neutron absorber is varied to compensate for reactivity changes due to fuel burnup, fission product poisoning including xenon and samarium, burnable poison depletion, and the cold-to-operating moderator temperature chanye. Using its normal ma~eup path, the Chemi~al and Volume Control System (CVCS) is capable of inserting negative reactivity at a rate of approximately 30 pcm/min when the reactor coolant boron concentration is 1000 ppm and approximately 35 pcm/min when the reactor coolant boron concentration is*lOO ppm. The peak. burnout rate for xenon_ is 25 pcm/min. Rapid transient reactivity

  • SGS-UFSAR. 4. 3-11 Re vision 0 July 22, 1982

Allowing for fuel densification effects the average kw/ft for Unit No. 1 u.-v:l-fA,,,J /l/o.J. ,, S.9-4 ""'/'l'f, t-9 &.33 k:w/f L and 5.44 fer YAit Ho. 2. From Figure 4.3-20, the conser-vative upper bound value of nonnalized local power density, tncluding allowances ~or densificatio~ effects, is 2.32 corresponding to a peak local power density of *H?.6 kw/fi 111Hi 12.9 kw/ft at 102 percent power for Unit No. 1 and Unit No. 2. -respeecluci:;c-To detennine React-0r Protection System set points, with respect to power distributions, three categories of events are considered, namely rod control equipment malfunctions, operator errors of colllllission and operator errors of omission.

The first category comprises u*ncontrol led rod w*i thdrawal (with rods moving in the nonna l bank sequence) for full 1ength rod banks. A1 so included are motions of the full length rod banks below their insertion limits, which could be caused, for example, by uncontrolled dilution or primary coolant cooldown. Power distributions were calculated through-out these occurrences assuming short tenn corrective action, that is no transient xenon effects were considered to result from the malfunction.

The event was assumed to occur from typica*l nonnal operating situatJons which did include nonnal xenon transients. It was further assumed in detennining the power distributions that total power level would be limited by reactor trip to below 118 percent. Since the st~dy is to detennine protection limits with respect to power and axial offset, no credit was taken for trip set point r~duction due to flux difference.

Results are given in Figure 4.3-21 in units of kw/ft. The peak power density which can occur in such events, assuming reactor trip at O*r:'

below 118 percent, is thus limited to 18.0 kw/ft including uncertainties and densification effects. The second category, also appearing in Fi~ure 4.3-21, assumes that the operator mis-positions the full length rod bank in violation of the insertion limits and creates short tenn conditions not included in nonnal operating conditions *

  • SGS-UFSAR 4.3-25 Revision 0 July 22, 1982

TABLE 4.4-lA (sheet 1 of 2)

REACTOR DESIGN-COMPARISON TABLE SALEM UNIT .l 17 x 17 With 15 x 15 Without Thenna 1 and Hydraulic Design Parameters Densification Densification W&- 3+11 ~ 3+-JI Reactor Core Heat Output; MWt

// (,4-Z. /l{,f-2.

Reactor Core Heat Output, Btu/hr 11,J9'3-x 106 11,~g3 x 106 Heat Gene rated in Fue 1 , 97.4 97.4*

System Pressure, Nominal, psia 2250 2250 System Pressure, Minimum Steady 2220 2220 State, psi a Minimum DNBR at Nominal Conditions Loo t'3'T 2 .2.+- ~[a]

Typical Flow Channel Thimble (Cold Wall) Flow Channel ~ /. 'io Minimum DNBR for Design Transients >l. 30 >1.30 11 DNB Correlation 11 R-Grid (W-311 11 R-Grid (W-3 with modified with modified sp.acer factor) spacer factor)

Coolant Flow 0

'l,.

Total Thermal Fl ow Rate, lb/hr 132.,i x 106 134,,.( x 106 Effective Flow Rate for Heat 3 Transfer, 1b/hr 126.A' x 106 128.0 x 106 Effective Flow Area for Heat Transfer, ft2 51.1 51.2

+

Average Velocity Along Fuel Rods, ft/sec 15.,1 15.$ '

Average Mass Velocity, lb/nr-ft2 2.47 x 106 2. 50 x 106 CoolI ant Temperature i44.4 so/.s.r; 544. 4 Sc/~.()

Nominal Inlet, °F

~ 6~.J Average Rise in Vessel, °F.

Average Rise in Core, °F ~ ~7-B

  • SGS-UFSAR Revision O July 22, 1982

TABLE 4.4-lA (sheet 2 of 2)

  • REACTOR DESIGN COMPARISON TABLE SALEM UNIT.1 Thermal and Hydraulic Design Parameters 17 x 17 With Densification 15 x 15 without Densification Average* in Core, °F 579. B- sF1.o 679. l Sfo. if Average in Vessel, °F s?s. 1 s 77.Cr ..e?o. J- s11. s-Heat Transfer Active Heat Transfer, Surface Area, ft2 59,700 52,200 li'17ao l../?~oo Average Heat Flux, Btu/hr-ft2 185,7oe -H-r,600 Maximum Heat Flux, for nonnal 4+oioo operation Btu/hr-ft2 4aa ,9oe[bJ 580 ,000 S"*6i-I '?.t;3 Average Thennal Output, kw/ft ~ +.-88 Maximum Thennal Output, for 12.,

nonnal operation, kw/ft ~[b] . 18. 8

  • Peak Linear Power for detennination of Protection Setpoints, kw/ft Fuel Central Temperature Peak at 100 Power, °F 18.o[c]

.J4-dO

-3350' . 4250 Peak at Thennal Output Maximum for Maximum Overpower Trip Point,* °F 4150 Pressure Drop Across Core, psi 24.7 + 2.s[dJ 32.6[e]

Across Vessel, including nozzle, psi 49.8 + 5.0 52.0

[a.] Previously", the value of 2.09 for a limiting typical channel was quoted only since the thimble (cold wall) DNB Tests were incomplete.

[b] This limit is associated with the value of FQ = 2.32.

[c] See Section 4.3.2.2.6.

[d] Based on best estimate reactor flow rate of 95,600 gpm/loop.

[e] Previously, a conservatively.high value of pressure drop was used to detennine vessel loop flow rates *

  • SGS-UFSAR Revision O July 22, 1982

r TABLE 4.4-2A THERMAL-HYDRAULIC DESIGN PARAMETERS FOR ONE OF FOUR COOLANT LOOPS OUT OF SERVICE SALEM UNIT 1 Total Core Heat Output, Mlilt ~zJ;f Total Core Heat Output, 10 6 Btu/hr ~ 1/~CJ Heat Generated in Fuel. 97.4 Nominal System Pressure. psi a 2250 Coolant Flow Effective Thennal Flow Rate for Heat Transfer, 106 lbs/hr 90.6 Effective Flow Area for Heat Transfer, ft 2 51.1 Average Velocity along Fuel Rods, ft/sec 10. 9 Average Mass Velocity, 10 6 l b/hr-ft 2 1. 77 Cool ant Temperature, °F Design Nominal Inlet t>38.8" ~39,/

~ 61?*2..

Average Rise in Core Average in Core ~ s7if.7 Heat Transfer Active Heat Transfer Surface Area. ft 2 59,700

~verage Heat Flux, Btu/hr-ft2 130 .ooo Minimum DNB Ratio at Nominal Conditions . > l-r86- /.?~

Minimum DNB Ra~i~ for Design and Anticipate? Transients > 1. 30 SGS-UFSAR Revision 0 July 22, 1982

TABLE 4.4-3A

  • VOID FRACTIONS AT NOMINAL REACTOR CONDITIONS WITH DESIGN HOT 0-IANNEL FACTORS SALEM UNIT l Average Maximum 0*11 Core 9':1'5'%

13.&

Hot Subc tranne 1 ..z..-2--

  • SGS-UFSAR Revision 0 July 22, 1982
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