ML18078B018

Requests Amend of Safety Tech Specs (App a) Re Axial Flux Difference,Heat Flux Hot Channel Factor,Reactor Core & Nuclear Enthalpy Hot Channel Factor

ML18078B018
Person / Time
Site:	Salem
Issue date:	03/02/1979
From:	Librizzi F Public Service Enterprise Group
To:	Schwencer A Office of Nuclear Reactor Regulation
References
LCR-79-06, LCR-79-6, NUDOCS 7903150284
Download: ML18078B018 (22)

Text

e PS~G Ref. LC:R 79-06 Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201 /430-7000 March 2, 1979 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. Albert Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors Gentlemen:

~!g~~r~Y ~~~:T~~~Li~~iNsE DPR-~EGUL~TORY DOCf\ET FJL£ COPYJ UNIT NO. 1 SALEM GENERATING STATION DOCKET NO. 50-272 Salem Unit No. 1 is currently in its first cycle of operation with a refueling outage scheduled to commence on March 31, 1979.

Cycle 1 operation will be terminated within a cycle burnup range of 14,400 to 15,600 MWD/MTU. Startup of cycle 2 is expected to occur in mid June 1979. This letter is to advise you of PSE&G's review of and plans regarding the Salem No. 1, cycle 2 reload core.

The cycle 2 reload core will consist of forty (40) new Westing-house 17 x 17 fuel assemblies. Two (2) of these will be of the optimized fuel assembly design and will be inserted as part of Westinghouse's "Optimized Fuel Assembly Demonstration Program" (WCAP-9286).

The Salem No. 1, cycle 2 reload core was designed such that those incidents analyzed and reported in the Salem FSAR which could potentially be affected by the fuel reload have been reviewed for the cycle 2 design. This review was performed in accordance with the Westinghouse reload methodology as outlined in the March 1978 Westinghouse topical report entitled "Westinghouse Reload Safety Evaluation Methodology" (WCAP-9272). The small break LOCA is presently being reanalyzed to confirm the ECCS analysis for cycle

2. The results of this reanalysis will be submitted to you after it is completed, which is anticipated to be during the latter part of March 1979. PSE&G has reviewed in detail the bases of the reload analysis and the Westinghouse Reload Safety Evaluation (RSE) Report with Westinghouse. The review of all incidents completed to date has demonstrated and the small break LOCA reanalysis is expected to demonstrate that the results of all the postulated events are within allowable limits.

The Energy People

T

.. .e Ref. 79-06 3-2-79 The reload core design will be verified during the startup physics testing program.

• This program will include, but not be limited to, the following tests:

1. Control rod drive tests and drop time

2. Critical boron concentration measurements

3. Control rod.bank worth measurements

4. Moderator temperature coefficient measurement

5. Power coefficient measurement, and

6. Startup power distribution measurements using the incore flux mapping system.

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 analysis of the changes to Facility Operating.License DPR-70 for Salem Generating Station Unit No. 1.

This request consists of proposed changes to the Safety Technical Specifications (Appendix A), pertaining to the following areas:

Axial. Flux Difference, Heat Flux Hot Channel Factor - FQ(Z),

Nuclear Enthalpy Hot Channel Factor - FgH, and Reactor Core.

This change request package is deemed to involve several Class III changes (each involving a single safety issue and deemed not to involve a significant hazards consideration) and, therefore, is determined to be a Class IV amendment as defined by 10CFR 170.22.

A check in the amount of $12.,300 is enclosed.

This submittal includes three signed originals and 40 copies.

Very truly yours, Fra*~

General Manager -

Electric Production

Ref. LCR 79-06 U.S. NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-272 PUBLIC SERVICE ELECTRIC AND GAS COMPANY FACILITY OPERATING LICENSE NO. DPR-70 NO. 1 UNIT SALEM GENERATING STATION Public Service Electric and Gas Company hereby submits proposed changes to Facility Operating License No. DPR-70 for Salem Gen-erating Station, Unit No. 1. This change request relates to Safety Technical Specifications (Appendix A) of the Operating License, and pertains to changes required for cycle 2 operation.

Respectfully submitted, PUBLIC SERVICE ELECTRIC AND GAS COMPANY BY:~&J~ FREDERICK W. SCHNEIDER VICE PRESIDENT

Ref. LCR 79-06 STATE OF NEW JERSEY)

) SS:

COUNTY OF ESSEX )

FREDERICK W. SCHNEIDER, 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 signed the request for change to FACILITY OPERATING LICENSE NO. DPR-70.

The matters set forth in said change request are true to the best of my knowledge, information, and belief.

~u)~

FREDERICK W. SCHNEIDER Subscribed and sworn to before me this  :;t:f day of V~L_, 1979.

LNotary Public of New Jersey My commission expires on fLHJ f, 19f3 BARBARA VALLEE A NOTARY PUBLIC OF NEW JERSEY My Commission F.x~irr 0 ' '*-:3

PROPOSED CHANGE AXIAL FLUX .DIFFERENCE TECHNICAL SPECIFICATION SALEM UNIT NO. 1 Description of Change During the first 2700 MWD/MTU of cycle 2 operation, the indicated axial flux difference will be restricted to less than +7.5% at rated thermal power. This allowable axial flux difference will increase by 1.0% for each 1.0% re-duction in power level.

Reason for Change This restriction is necessary to ensure that the core peak-ing factor limits are met during cycle 2 operation.

Safety Evaluation This change is required to ensure that the peaking factor limits assumed in the bases of the technical specifications and in the FSAR are met during cycle 2. Therefore, this change does not involve an unreviewed safety question.

(C1 3/4.2 POWER DISTRIBUTION LIMITS AXIAL FLUX DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE shall be maintained within a

+5% target band (flux difference units) about the target flux difference.

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APPLICABILITY: MODE l ABOVE 50% RATED iHERMAL POWER*

ACTION:

a.

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With the indicated AXIAL FLUX DIFFERENCE outside of the ~ 'taP!!t MA~ ataewt tRe iaPge'&--- fl tm di ffeF8Aie and with THERMAL PuwER:

ol I .

3. 2.. I a.bewe.. *

1. Above 90% of RATED THERMAL POWER, within 15 minutes:

a) Either resto.re the indicated AFD to within the target band limits, or b) Reduce THERMAL POWER to less than 90% of RATED

(~ THERMAL POWER.

~

2. Between 50% and 90% of RATED THERMAL POWER:

a) POWER OPERATION may continue provided:

1) The indicated AFD has not been outside of the +5% Ii.. : .$ e I '

3 .~. / .~e.. t.rget ~a Ad- for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> pena 1ty de vi a-ti on cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and

2) The indicated AFD is within the limits shown on Figure 3.2-1. Otherwise, reduce THERMAL POWER to less.than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip Setpoints to < 55% of RATED THERMAL POWER within the next 4 liours.

b) Surveillance testing of the Power Range Neutron Flux Channels may be perfonned pursuant to Specifi-cation 4.3.1.1.1 provided the indicated AFD is maintained within the limits of Figure 3.2-1. A total of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> operation may be accumulated with the AFD outside of the target band during this testing without penalty dev~ation.

• C:J *See Special Test Exception 3.10.2 SALEM - UNIT 1 3/4 2-1 Amendment No. 5

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued) b.- THERMAL POWER shall not be increased above 90% of RATED THERMAL POWER unless the indicated AFD is within the + &; ta~~et ha~ '

a.bcv~ and ACTION 2.a) l). above has been satisfied.- L: .. :t. .s of' 3. 2, I

c. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER.unless the indicated AFD has not been outside of the L;~.i~

of3,2.lai""'! 5% target ea"d for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty deviation I

cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above l5% of RATED THERMAL POWER by:

a. Monitoring the indicated AFD for each OPERABLE excore channel: ~=-}

1. At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and

2. At least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the AFD Monitor Alarm to OPERABLE status.

b. Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 30 minutes thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable .

.The logged values of the indicated AXIAL FLUX DIFFERENCE shall .

be assumed to exist during the interval preceding each logging.

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~~~~al*2 The indicated AFD shall be considered outside of its ..... 51 tar~et

. w'hen at least 2 of 4,or~2 ~f. ~ PP!RABLE excore channels are indicating the AFD to be outside the 'taFget3aane: -Penalty deviation outside of the

+ 5% tirgat baA~ shall be accumulated on a time basis of:

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a. One minute penalty devialio'} f~ _£i_C 2ne minute of POWER OPERATION outside of the ** ~ ~ ' at~'fHERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and

b. One-half minut: penalty ,P,e~JJt.,.i,P~,fr... ~~c)l one minute of POWER ,

OPERATION outside of the target ~" at IHERMAL POWER levels bel 50% of RATED THERMAL POWER. .__

SALEM - UNIT 1 3/4 2-2

Proposed addition to page 3/4 2-1 In addition the following restriction will be required during the first 72 EFPD (2700 MWD/MTU) operation of cycle 2.

The indicated axial flux difference will be maintained less than +7.5% at RATED THERMAL POWER with the allowed axial flux difference increasing by 1.0% for each 1.0% reduction in power level.

/

PROPOSED CHANGES HEAT FLUX HOT CHANNEL FACTOR - FQ(Z)

TECHNICAL SPECIFICATIONS SALEM UNIT NO.* 1 Description of Changes

1. Revise the third line segment of the K(Z) curve (figure 3.2-2).

2. Revise the F limits as contained in -this section of the

. XY.f. .

technical speci 1cat1ons.

Reason for Changes

1. This revision is necessary due to violation of the F (Z) limits during cycle 2 operation with the current thi~d line segment of the K(Z) curve.

2. Recent analyses for Westinghouse plants which undertake reload cycle operations with an F technical specification show the need to revise these limY~s upward for other than cycle 1 operation. Salem No. 1 is one of the first Westing-house units with an F limit to reload and a revision in this limit is requireaYto avoid exceeding Fxy limits during the second and subsequent cycles.

Safety Evaluation

1. The revised K(Z) curve has been used in the design of cycle 2. Analysis of cycle 2 is expected to verify the validity of the revised K(Z) curve for cycle 2 operation.

Because of this change the small break LOCA analysis must be redone. This analysis, using currently approved Westing-house analysis models will be completed in March 1979, and is expected to confirm the validity of the revised K(Z) curve. The results of this reanalysis will be submitted after it is completed. If the results are as anticipated the LOCA analysis for cycle 2 will be confirmed using the revised K(Z) curve and therefore, this change will not constitute an unreviewed safety question. The revised third line segment will be reverified for each reload beyond cycle 2.

2. F (Z), which is the primary power distribution parameter in tRe Technical Specifications for LOCA protection, is deter-

mined by the product of the radial F and the axial F(Z) peaking factors. In the event the Fxy is exceeded, con-tinued operation is allowed providedxthe Fn(Z) limit is met.

Since reload cores exhibit flatter axial snapes, as in-dicated by lower F(Z), the revised Fxy limits will still result in*F Cz).being within allowable limits.

0 For operation of cycle 2 and subsequent cycles of the Salem No. 1 core, bounding values of the peaking factor F (Z) x (relative power) were calculated as a function of e~evation by assuming various load follow transients on the.reactor through insertion and withdrawal of control rod Banks C and D. The effects of the accompanying variation in axial xenon and power distributions were also considered as described in the Salem FSAR. Both beginning and end of .cycle conditions were included in the cycle 2 calculations, and several dif-ferent histories of operation were assumed in calculating effects of load follow transients on the axial power distri-bution. Results of these calculations demonstrate that the F Cz) limits will not be exceeded during operation of cycle 2.

0 Further evaluations indicate that the proposed F change will not cause the FQ(Z) limits to be exceeded i~Yprojected reload cycles beyond cycle 2. The FQ(Z) limit envelope will be reverified for each reload beyond cycle 2 to confirm this projection. Therefore, this change does not involve an un-reviewed safety question.

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2 4 6 8 10 12 14 CORE HEIGHT CFTI-FIGURE 3.2-2 KCZI - NORMALIZED F0 CZI AS A FUNCTION OF CORE HEIGHT L <..i

POWER UISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) b) At least once per 31 EFPD, whichever occurs first.

2. When the Fx; is less than or equal to the F~~P limit for the appropriate measured core plane, additional power distribution maps shall be taken and Fx; compared to F~~P and Fx; at least once per 31 EFPD.

e. The Fxy limits for RATED THERMAL POWER within specific core planes shall be:

~ ~~P <~.71 ~r a1~ore~1ane\ cont~ning'tithe~nk

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i ~.I (;.. fl,&. I, eJ. ~ 'cf,* i.1'e.n -Co r',..:? c-_ 3 /""f .2 - 7

2. FRxyTP < 4-: for all unrodded core planes .

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f. The Fxy limits of e, above, are not applicable in the fol-lowing core plane regions as measured in percent of core height from the bottom of the fuel:

1. Lower core region from Oto 15%, inclusive.

2. Upper core region from 85 to 100% inclusive.

3. Grid plane regions at 17.8 + 2%, 32.1 + 2%, 46.4 + 2%,

60.6 + 2% and 74.9 + 2%, inclusive. - - _

4. Core plane regions within + 2% of core height (+ 2.88 inches) about the bank demand position ot the bank 0 or 11 11 part length control rods.

g. Evaluating the effects of Fxy on Fg(Z) to detennine if FQ{Z) is within its limit whenever Fx; exceeds Fx;.

4.2.2.3 When FQ(Z) is measured pursuant to specification 4.10.2.2, an overall measured Fg{Z) shall be obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

SALEM - UN IT 1 3/4 2-7 Amendment No. 9

Proposed addition to page 3/4 2-7

1. For all core planes containing bank "D" control rods:

a) F RTP ~ 1.92 for core elevations up to 2.0 ft.

xy b) p RTP.:;. 1. 89 for core elevations f rorn 2.0 to xy -

12.0 ft.' and

(_) 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integ-rity during Condition I {Normal Operation) and II {Incidents of Moderate Frequency) events by: (a) maintaining the minimum DNBR in the core >

1.30 during normal operation and in short term transients, and (b) 11miting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. In addition, limitin~ the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200°F is not exceeded.

The definitions of hot channel factors as used in these specifi-cations are as follows:

Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for man-ufacturing tolerances on fuel pellets and rods.

( -* .**

Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.

/fell ~ttr.,l t=. d µJ, ~,-(),, -t:o pt&.) c... t3 3/-1 ..2.. - I 3/4.2.l AXIAL FLUX DIFFERENCE (AFD)

The limits on AXIAL FLUX DIFFERENCE assure that the F0 (Z) upper bound envelope of 2.32 times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redis-tribution following power changes.

Target flux difference is determined at equilibrium xenon conditions with the part length control rods withdrawn from the core. The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions.

Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations ..

SALEM - UNIT l B 3/4 2-1

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POWER DISTRIBUTION LIMITS .J BASES AA/J) tf,1])1# L 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACTORS=

Fq ( Z~ aAa ~H p £ /J I\/¥6-F Fl C. T 0 tf5 - F q(.2.)1 Fa~ lti.~ J 0< y ( 2) t The limits on heat flux and nuclear enthalpy hot channel factors ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200°F ECCS acceptance criteria limit.

Each of these hot channel factors are measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure that the hot channel factor limits are maintained provided:

a. Control rod in a single group move together with no individual rod insertion differing by more than + 12 !teps from the group demand pas i ti on. * -

b. Control rod groups are sequenced with overlapping groups as described in Specification 3. 1.3.5.

c. The control rod insertion limits of Specifications 3. 1.3.5 and 3. 1.3.6 are maintained.

d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

The relaxation in F~H as a function of THERMAL POWER allows changes i~ the radial power shape for all permissible rod insertion limits.

F~H will be maintained within its limits provided conditions a thru d aoove, are maintained.

When an F measurement is taken, both experimental error and man-ufacturing tolQrance must be allowed for~ 5% is the appropriate allowance for a full core map taken with the incore detector flux mapping system and 3% is the appropriate allowance for manufacturing tolerance.

When FN is measured, experimental error must be allowed for and 4~

is the apprA~riate allowance for a full core map taken with the incore detection system. The specified limit for FNH also contains an 8% allow-a~ce for uncertainties which mean that normat operation will result in F~H ~ 1.55/1.08. The 8% allowance is based on the following considera-tTons:

SALEM - UNIT 1 B 3/4 2-4

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POWER DISTRIBUTION LIMITS BASES

a. abnonnal perturbations in the radial power shape, such as from rod misalignment. effect F~H more directly than FQ,

b. although rod movement has a direct influence upon limiting FQ to within its limit, such control 1s not readily available to limit F41NH, and ...

c. errors in prediction for control power shape detected during startup physics tests can be compensated for in FQ by restrict-ing axial flux distributions. This compensation for F:H is less readily available.

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3/4. 2.4 QUADRANT POWER TILT RATIO co f

,p1.i c. J3 3/.tf 2 - S-The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation.

The limit of 1.02 at which corrective action is required provides ONB and linear heat generation rate protection with x-y plane power tilts. A limiting tilt of 1.025 can be tolerated before the margin for uncertainty

• in FQ is depleted. The limit of l.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt._

The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and cor-rection of a dropped or misaligned rod. In the event such action does not orrect the tilt, the margin for uncertainty on FQ is reinstated by .

educing the power by 3 percent from RATED THERMAL POWER for each percent f tilt in excess of 1.0.

ALEM - UNIT 1 B 3/4 2-5 Amendment No. 9

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Proposed addition to page B 3/4 2-1 Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation z.

Proposed addition to page B 3/4 2-5 The radial peaking factor, F (Z), is measured periodically

. . . . xy f to provide additional assurance that the hot channel actor, F 0 CZ), remains within its limit. The Fxy(Z) limits were determined from expected power control maneuvers over the full range of burnup conditions in the core.

PROPOSED CHANGE NUCLEAR ENTHALPY HOT CHANNEL FACTOR - F~H TECHNICAL SPECIFICATIONS SALEM UNIT NO. 1 Description of Change Revise the F~H technical specification, incorporating the rod bow penalty curve that is currently approved by the NRC.

This curve has been documented as having generic application for Westinghouse 17 x 17 plants in a Westinghouse submittal, NS-TMA-1760, "Fuel Rod Bowing", T. M. Anderson (Westinghouse) to J. F. Stolz (NRC), dated April 19, 1978. This change would result in a reduction of the current rod bow penalty applied to Salem No. 1.

Reason for Change To reduce the rod bow penalty (especially for second cycle fuel) to accommodate expected F~H values that will be en~

countered at rated thermal power during cycle 2 operation.

Safety Evaluation This change only incorporates recent information on rod bow penalties and does not reduce the margin of safety as de-fined in the basis for this specification or as defined in the FSAR. Therefore, this change does not constitute an unreviewed safety question.

POWER DISTRIBUTION LIMITS

\

NUCLEAR ENTHALPY HOT CHANNEL FACTOR - ~H LIMITING CONDITION FOR OPERATION 3.2.*3 ~H shall be limited by the following relationship:

~H !.. 1. 55 [ 1

• 0 + 0. 2 (1-P )J [ I - R B PJ I

_ THERMAL POWER where P - r&\TEO THERMAL POWER RB P -

APPLICABILITY: MODE 1 lfcJ /do.., f>e.,,4/ t t

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I ACTION:

With y:MAH exceeding its limit:

a. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER

~ithin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Power Range Neutron Flux-High Trip Setpoints to < 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, -

b. Demonstrate thru in-core mapping that F~H is within its limit

~ithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the limTt or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and

~. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a. or b. abRve; *subsequent POWER OPERATION may

-proceed provided that FAH is demonstrated through in-core mapping to be within its limit at a nominal so: of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL power and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95% or greater RATED THERMAL POWER.

SALEM - UNIT 1 3/4 2-9 Amendment No. '

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0 6 10 16 20 26 30 REGION AVf:RAGE BURNUP (10 3 MWD/MTU)

FIGURE 3.2-1 3

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ROD BOW PENALTV.AS A FUNCTION OF BURNUP*

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• 1

PROPOSED CHANGE REACTOR CORE TECHNICAL SPECIFICATION SALEM UNIT NO. 1 Description* of Change The current technical specifications state that each fuel rod shall contain a "nominal total weight of 1743 grams of uranium". This should be revised to read a "maximum total weight of 1766 grams of uranium". This change is consistent with the wording contained in the June 15, 1978 version of the Westinghouse Standard Technical Specifications (STS).

Reason for Change PSE&G is going to be inserting two demonstration assemblies in Salem No. 1 during cycle 2. These assemblies are of the new Westinghouse fuel assembly design known as the optimized fuel assembly. These optimized fuel assemblies have a nominal total weight of uranium of about eight percent less than the standard Westinghouse 17. x 17 fuel assembly.

Safety Evaluation This proposed change will only allow for a reduction of the weight of uranium, while maintaining the same limits on enrichment and cannot adversely affect the safety of the unit. Therefore, this change does not involve an unreviewed safety question.

This wording change also has the generic approval of the NRC in the Westinghouse STS.

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• )

DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be main-tained for a maximum internal pressure of 47 psig and an air temperature of 271°F.

5.3 REACTOR CORE FUEL ASSEMBLIES

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• .* 5.3.l The reactor core shall contain 193 fuel assemblies with each f~el assembly containing 264 fuel rods clad with Zircaloy -4. Each rn~,,..'"'"",..,

fuel rod s~ll have a nominal active fypl length of143.7 inches and

.. contain a _-.:-~i total weight of ~"'Yrams uranium. The initial 17 ~b core loading shall have a maximum enrichment of 3.35 weight percent U-235.

I Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrictvnent of 3.5 weight percent U-235.

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CONTROL ROD ASSEMBLIES

\.J 5.3.2 The reactor core shall contain 53 full length and 8 part length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The part length control rod assemblies shall contain a nominal 36 inches of absorber material at their lower ends. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium.

All control rods shall be clad with stainless steel tubing. -The balance of the void length in the part length rods shall contain aluminum oxide.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

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SALEM - UN IT l 5-4