ML18082A295

Cycle 2 Startup Test Rept.

ML18082A295
Person / Time
Site:	Salem
Issue date:	05/01/1980
From:	Germann R, Heller H, Nichols J Public Service Enterprise Group
To:
Shared Package
ML18082A294	List: ML18082A293 ML18082A295
References
NUDOCS 8005050214
Download: ML18082A295 (62)

Text

_,_ - 80 05050 ll'f

• O SALEM GENERATING STATION UNIT 1

.'1-1:

I PS~G CYCLE 2 STARTUP TEST REPORT The Energy People 1-1*

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• I PUBLIC SERVICE ELECTRIC AND GAS COMPANY I SALEM NUCLEAR GENERATING STATION UNIT 1 CYCLE 2 I STARTUP TEST REPORT I

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I I* Written By:

Reviewed By:

I Reviewed By:

I Nuc ear Fue Techno1ogy Group

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I* Approved*By:

Station*Manager c .._;*.*

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.I I TABLE OF CONTENTS I

I Part I Zero Power Physics Tests Pow~r Ascension Tests Part II I Part III Augmented Surveillance Program I Appendix A Startup Test Program Description I Appendix B Appendix C Safety Evaluation Augmented Surveillance Program Conunitment I

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I I LIST OF TABLES I

I - 1 Startup Tests I I - 2 Summary Of HZP Reactivity Measurements I I II -

3 1

Zero Power Flux Maps At Power Flux Map Results I

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I I .-.----- *-*****--* - *- *--,~--*-******

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I I LIST OF FIGURES I

Map-173 I I - 1 Zero Power Assembly Zero Power Assembly F~H F~H Map-174 I - 2 1* I - 3 Zero Power Assembly F~H Map-175 I III - 1 A-C ROD INSERTION AND WITH DRAWL I III .- 2 A-C POWER INCREASE III - 3 A-C DILUTION I III - 4 A-C

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BO RATION I III . . . 5 III - 6 At Power Differences In Reaction Rate Integrals At Power Differences In Reaction Rate Integrals Map-185 Map-186 I III - 7 At Power Differences In Reaction Rate Integrals Map-187 I

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PART I I

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I I Part I ZERO POWE.R PHYSICS TESTS I

I Introduction I

From November 30 through December 24, 1979 the zero power physics tests I were conducted as part of the overall refueling outage and return of Salem 1 to servic~. The purpose of the test program was to assure that I the as-loaded reactor core performed in accordance with design para-meters and safety limits.

I The program content was similiar to the initial zero power tests and was based on past experience, Technical Specifications, and NRC commit-I ments.

1* Our commitment to dropped rodlet surveillance required two extra flux maps and a shutdown margin test.

I Table I-1 presents the tests conducted for the zero power test program.

I Appendix A provides a brief description of each of the tests.

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I I TEST RESULTS I Table I-2 summarizes the results of the control rod worth, boron endpoint, isothermal temperature coefficient, and shutdown margin verification I measurements. From these results, the following points are apparent:

I 1. The measured control rod worth agreed within approximately 1% of prediction. The rod banks were measured individually (dilution) and in overlap (boration).

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2. The boron endpoints were consistent with predicted valves except a I positive bias was noted. The design tolerance was all rods out reference case.

~50ppm on the The measurement was +56ppm (+5.2%)

!I above design.

3. The ARO isothermal temperature coefficient measurement agreed with I design predictions. The D bank .in measurement was within acceptable tolerance from design but did not move in the expected direction I (more negative) . This may be due to an uneven rate of cooldown and heatup during the D in measurement.

I 4. The shutdown margin verification was performed because the two zero

.I power flux maps failed the design acceptance criteria (see Appendix C).

The results indicate that the "stuck" rod (F-14) is worth 13.7% less than prediction. This is acceptable since it is in the conservative I direction. The measured critical boron concentration for all rods in less the most reactive rod out, agreed within 1.3% of prediction.

I The measurement was based on the results of 3 boron samples.

Table I-3 and Figures I-1 through I-3 show the results of the three zero I power flux maps. The maps were of concern for two reasons: 1) two large depressed areas were observed and 2) F N exceeded the 100% technical I specification limit.

xy Because assembly power exceeded the design specifi-cation, the shutdown margin verficiation test was initiated. N With F xy I exceeding the 100% limit, a program of taking flux maps at each 20% in-crement of power escalation was also initiated. Appendix B documents a

• 1 review of the above concerns.

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~~~~~--~~--- Table I-1 STARTUP TESTS DESCRI'PTION COMMENTS Rod Drop Every rod must be drop tested (53 rods).

Tests & The slowest drop time rod must be drop tested 10 times.

Current The fastest drop time rod must be drop tested 10 times.

Profiles RPI RPI coil stacks are very temperature sensitive and Calibration equilibrium conditions are difficult to achieve.

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Table I-1 (cont'd)

STARTUP TESTS

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I DESCRIPTION COMMENTS I Reactor 1) Pull Control Banks and make "l/M" plot Startup 2) Dilute to Criticality (100 ppm/hr) (2000 ppm to 1200 ppm)

.3) Establish Equilib. boron concentration

. 4) Check out reactivity computer Zero Power 1) ARO - Boron Endpoint test Physics 2) ARO - Temp Coefficient meas H

Testing 3) ARO - 2 Flux Maps - Flux Maps Analysis and Results

4) "D" Rod bank worth meas.

5) "D" bank in - Boron endpoint test

6) "D" bank in - Temp~ coefficient meas

7) "C" Rod bank worth meas.

8) "C" bank in - Boron endpoint test

9) "B" Rod bank worth meas.

10) "B" bank in - Boron endpoint test

11) "A" Rod bank worth meas.

12) "A" bank in - Boron endpoint test

13) Shutdown Verification test (Involves diluting all shutdown banks into* core and "swapping" most reactive rod with shutdown banks)

14) Control Bank Overlap worth meas

15) SORC Review

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Table I-1 (cont'd)

STARTUP TESTS DESCRIPTION Incore/ Internal gain adjustments on NI channels so that excore channels Excore reflect the incore map results from above Adjustments Power Turbine load changes of 4 to 5% to make changes in ~T and T avg Coefficient Tests H

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I Table I-2

SUMMARY

OF HZP REACTIVITY MEASUREMENTS I

I Control Bank Worths Bank Measured Predicted M-P Measured Predicted M-P (pcm) (pcm) (pcm) (ppm) (ppm) {ppm}

I D 1041 1128 -87 104 118 -14 c 938 896 +42 96 94 + 2 I B A

534 1163 557 1130

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.+33 114 53 57 117 4

3 I Total 3676 dilution 3590 boration 3711 3711

-35 367 386 -19 I Boron Endpoints Condition Measured (ppm) Predicted (ppm) M-P I ARO D-In 1140 1036 1084 966 56 70 I DC-In DCB-In 940-887 872 815 68 72 DCBA-In 773 . 698 75 I Isothermal Temperature Coefficient I Condition Measured (pcm/°F) Predicted (pcm/°F) M;..P ARO -6.05 -6.4 +0-35 I D-In -5.78 -8.l +2. 32 .

Shutdown Margin Verification I Condition Measured. (pcm) Predicted (pcm) M-P 53 RODS IN 7360 -

I 52 RODS IN 5900 5190 +710 F-14 2170 I Condition Measured (ppm) Predicted {ppm) M-P I 52 RODS IN 538 531 +7 I

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Table I~3 ZERO POWER FLUX MAPS QUADRANT FNH FQz B/U TILTS MAP DATE POWER BANK COMMENT CB I Wll0 J'IW NE SW SE VALUE LOC VAL LOC 12  ! '

173 79 ~ 12078 .228 ZPPT 1135 1,0156 0.9589 1. 0084 1. 01F2 1.5199 Kl4ED 4.055 K4DE 174 1~92 0 12078 228 ZPPT 1135 1,0133 0.9576 1.0254 l.003P 1.5338 013EO 4.073 MlODM H 175 1~923 0 12078 139 ZPPT 1025 1,0231 0.949 1. 0132 l.014P 1. 6395 JlODM 3.885 J'lODM I

..,J F F Fxy AVE IFxy xy xy iuNROD, AXIAL

.UN ROD. RODDED RODDED AVE

~OP.

PEAKG.

BOT. TOP BOT. AXIAL FACTOR

!VAL LOC VAL VAL OFFSET LOC LOC VAL LOCI

1. 721 M3MC 1.8151 BllEN - - -. - 2.737 76.3

1. 766

• El4NE 1. 821 BllEN - - - - 2.7215 76.l

i. 7.94 M.3MC 1. 878 M3MC 1. 9632 F9NM - - 1. 7558 40.9

DA TE 12077Y PAGE 259

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DAH 121279 PAGE 261 REASURED AND .PERCEN1. Dlff. OF fDHN SALE" lNCOAE MAP174. ox POW[A*BANK D i 228,113Spp"* o"wDIR1u,PAR1.1.12-2 R P N .M * *L * *

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• 979. 1.039. 1.144. 1.338.

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• l'IEAS 15 9.4. 3.8. .9. 1 .3. 2. 1..

• 4. 2 .o *

• Dl f f

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-I I

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PART II I

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_I

I I Part II POWER ASCENSION TESTS I

I Introduction I On December 24, 1979, a SORC meeting was held to review the results of the

,1 zero power physics testing program. It was decided the plant could safely proceed to Mode 1 operation and continue the test program.

I Reactor power was increased to 20% and. another flux map was taken. Since the F measured at 20% was higher than the limit at rated thermal power xy I but below the limit at 20%, another map was scheduled for 40% power as per the Technical Specification requirement.

I Reactor power was increased to 30% and held for a short period to collect heat balance data for nuclear instrumentation adjustments.

I Following the adjustments, power was increased to 40% and held constant I for xenon stabilization. Upon reaching equilibrium xenon, the following tests were performed:

I 1. Incore Flux Map

2. Incore-Excore Detector Calibration I 3. Power Coefficient Measurement I F xy as measured at 40% was above the rated thermal power limit but below the limit at 40%. Another map was scheduled for 60% *.

I The Incore-Excore Detector Calibration was performed* at 40% as a pre-liminary alignment of the excore detectors. This test was scheduled to be I re-performed at 95% power.

I I

II-1 I.

I I The results of the 60% power flux map F xy measurement indicated a map was required at 80%.

• I The final power plateau for testing was at the 95% level. Xenon was I allowed to stabilize after which the following tests were performed.

I 1.

2.

Incore Flux Map Statepoint Data

3. Incore-Excore Detector Calibration I 4. Power Coefficient Measurement I The overall adceptance criteria for the startup program are listed in Appendix A.

I The results of the flux maps are summarized in Table II-1 and the power coefficient measured values are as follows:

I POWER LEVEL MEASUREMENT DESIGN DIFFERENCE I 40% -13.67pcm/% -14.0pcm/% 2.4%

I 95% -13.15pcm/% -13.7pcm/% 4.0%

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I I II-2

- .. - - - -*.. - - - - - - -*- - - ... - I I

i i

Table II-1 AT POWER FLUX MAP RESULTS

~~

N HAP DATE POWER BANK ~CB QUADRANT Fil FQz MTU .D TILTS H I NW NE SW SE VALUE LOC VAL LOC 12-29 1176 79 20 12089 165 853 1.0181 o. 9724 1. 0060 1. 0034 1. 6474 M3MC 2.211 JlODM 177 o.L 40 12123 169 845 1.0186 0.9706 1. 0081 1. 0027 1. 6099 M3MC 2.207 H5MN

-6 180 . 0 60 12203 163 690 1. 0221 0.9697 1. 0049 1. 0034 1. 4967 N4NM 1. 8816 G6MD

'H 181 08 8*0 12236 172 712 1. 0211 0.9695 1. 0016 1. 0078 1. 4715 N4NM 1. 7939 H5NM

• .H I 11.84 in14 95 12388 191 614 1. 0268 0.9738 1. 0016 0.9978 1.4884 M3MD 1. 799 M6NM w

q

~p AVE AVE

.Fxy Fxy i'xy Fxy AXIAL AXIAL UN,8~ UN ROD RODDED RODDED BOT TOP BOT PEAKING FACTOR OFFSET I VAL Luc VAL Luc VAL LOC VAL LOCI 176 1.851 .MJMC 1.882 MJMC 1. 786 JlODM - - 1. 285 12.5 1177 1. 8141 MJMC 1. 8153 M3MC 1. 776 H5MN - - 1. 232 10.05 1180 1. 6704 . P6QA 1. 7151 P6QA 1.7381 G6MD - - 1.1152 -0.297 1181 1.643 Fl4QA L 6695 Fl4QA 1. 7256 H5NM - - 1.1479 -2.004 1184 1. 6558 M3MD 1. 6824 P6QA -- - - L 162996 o. 573

• Map 181 was not taken at equilibrium Xenon conditions.

I I

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I 11

1 i

'I I

I

• I PART III

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I

I I Part III I AUGMENTED SURVEILLANCE PROGRAM (DROPPED RODLETS)

I The purpose of the augmented surveillance program for Salem 1 Cycle 2 I was to detect the presence of dropped rodlets. This part summarizes equipment, procedures and results of our pro.gram. (see Appendix C)

I The equipment used is shown in the attached photograph. It shows five I (5) 2 pen* recorders and the reactivity computer.

played are:

The parameters dis-I 1) Power average of the 4 power range nuclear channels

2) Power one channel of delta temperature (fl T)

I 3) T error the difference between actual moderator temperature and reference moderator temperature I 4) 5)

T avg Reactivity all 4. loops of average moderator temperature I 6) Flux Difference (fl I)

The ranges of the instruments were as follows:

I Power 0-100% 1 division = 1%

I T error

-5°F to +5°F 1 division = .1 o F T

avg 540°F to 580~F 1 division = .4°F reactivity -50pcm to +50pcm 1 division = lpcm flux difference -5% to +15% 1 division =

• 2 9-,

0 The chart paper width is 10 -inches, therefore each division equals .1 I inch. The chart speed on all charts except the power range was .2 inch per minute; the power range was 1 inch per hou~.

I Collection of data started on December 26, 1979 with the start of power ascension and has been continuous as of the date of this report.

I I III-1

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RESEARCHAND TESTING LAB.

AUDIO VISUAL DEPARTMENT 1*,

PAT NO. c ,,&(__/(a 13 ,:. 7>

I

I I PROCEDURE I The average worth of one rodlet is approximately lOpcm. There are 53 rod control cluster assemblies (RCCA) and each RCCA has 24 rodlets.

,. The reactivity computer was set up so that 10 pcm represented 1 inch of chart paper .

. The moderator temperature coefficient at full power is estimated to be I -12 pcm/°F therefore a 1°F change represents 1.2 inches of chart paper -

motion on the reac~ivity pen.

I The delta flux trace (~I) was used to check for intended rod motion.

I By observing known transients; ie, borations, dilutions, power changes I and rod motion, responses can be "finger printed" for comparison to later observations.

I The following figures show some normal transients; ROD INSERTION AND WITHDRAWL, POWER INCREASE, DILUTION, BORATION.

• 1*

Figure III-lC clearly shows the effects, of rod insertion and withdrawl.

The reactivity trace shows 1 step of control bank motion worth about 5-7 pcm. The corresponding delta flux (~I) trace shows a positive change (flux moving toward top of core) for rod withdrawl and a negative I change for rod insertion.

I Figure_III-2C demonstrates the constant reactivity additions from rod withdrawls needed.during a power increase. The corresponding shift in I delta flux (flux moving toward bottom of core) is expected on a power increase from the increased ~T across the core; ie, increase in tempera-ture gradient or colder water in bottom of core.

'I Figures III-3A-C shows that dilutions (primary water) is a relatively long term effect and has little effect on reactivity or delta flux.

I

.I III-2

I I A small increase in turbine power drops the average reactor temperature slightly in Figure III-3C. The 100 gallon dilution adds sufficient reactivity to restore the reactor temperature to its original value.

Figure III-3B shows no rod motion.

Figure III-4A demonstrates the reactivity effects of swapping boron for I control rod withdraw!. The boron.addition initiates the event in figure III-4B by showing a negative reactiviiy change. Rod withdraw! follows I with a positive reactivity addition. The average reactor temperature III-4C shows a decrease in temperature after the boron addition. With the control rods withdrawn further out of the core, the slow swing in I delta flux toward the bottom of the core is reduced.

remained relatively constant.

Reactor power has I

.t I

I I

I I

I I

I.

I* *III-3

I I* RESULTS t Three zero power (<5%) fl'ux maps were taken. Because the design acceptance criteria was exceeded, control rod worth to the N-1 condition t was measured_. __ The results of these measurements and a safety evaluation are included in zero power test section and Appendix B.

After achieving full power, differences in reaction rate integrals I between consecutive flux maps were evaluated.

changes that exceeded 5%.

There were no differential (see Figures III-5-7 for results)

I At steady state power levels, all traces are straight lines, therefore changes are very easy to observe. This is not true however during unit t startup when all parameters are changing making interpretation of charts somewhat more difficult. Startups and large load changes represent a I small fraction of total running time.

All of the recorder traces have been reviewed and it has been determined I that no unexplained events have occurred.

I We have determined that there was a reasonably high probability that dropped rodlets would have been detected by our test program as outlined above. In addition, our review of flux maps and on-line monitoring does

.:t not indicate the presence of dropped rodlets

• We plan on continuing our surveillance program for at least 1 more month.

Any indications of dropped rodlets would be included in a supplemental I report.

t Although the program was designed to detect a single dropped rodlet the primary concern was to protect against-a. loss of shutdown margin or violation of hot channel factors. Shutdown margin was verified by actual

':~

measurement to the N-1 condition; and periodic flux maps are taken to evaluate FQ and F xy

'I III-4

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I I Appendix A STARTUP TEST PROGRAM DESCRIPTION I SALEM UNIT 1 CYCLE 2 I General:

I The startup program is written such that there is a mandatory hold point at the zero, 50, and 95%

testing plateau. The Station Operating -Review I Committee reviews the data and makes a decision as to continued escalation of power. Deviations I between measurement and design beyond established tolerances could call for a review of test data, a I repeat of the measurement or a review of the safety analysis. Operation is always within FSAR and I Technical Specification limits.

I I

I I.

I I

I I A-1 I

I I STARTUP TEST PROGPJ..."~ DE.SC RI F'Tl o~;

SALEt-~ UN IT l CY CLE 2 I

I Test Initial Criticality Initial Conditions:

I 0 Mode 3, Tave= 547 F, CB= 2000??;;,

I Shutdown Banks ~ithcrawn Control Banks insertea I Test

Description:

I Pull control banks to D at 160 ste?s.

Dilute to criticality.

I Deterr..ine point of adding heat.

Checkout reactivity cowputer.

I Acceptance Criteria:

I Design:

I Reactivity computer reacings are ~ithin 4% of doubling time measurements.

I.

FSAR/T.S.

I Reactor must achieve criticality with the control banks above the zero power I. insertion limits.

I I

I A - 2 I

I I STARTUP TEST PR~GRAM DESCRIPTJON SALEM UNIT 1 CYCLE 2 I

Test: Flux Mapping I Initial co~oitions:

I HZP, ARO I Test De~cription:

I Operate Flux Mapping System in accordance with Part 13 of the Reactor Engineering Manual "Incore Flux Map~ins Syste~

I Operation" I Acceptance Criteria:

Design: Assembly Power I Design +lOt for assembly Power >.9 I Design +lSi for asserr~ly power <.9 I FSAR/T.S.

Not applicabie below Si power I

I I

I I

I A - 3 I

.I

1-STARTUP TEST PROG?..k.~ DESCRlPTlO~;

1 SALEM UNIT 1 CYCLE 2 I Test Roe Worth Measurements I Initial Conditions:

I* EZP This test is run several times during the startup program. It is scheduled to measure I all control bank worths. Based on the results of ..these measurements, adcitional shutdown

\

I ban~s will be measured. All control banks will be measured in overlap on the way out.

I Test

Description:

I Start boron dilution* at approx SOOpcrr./hr.

maintain criticality by adjusting the controlling I bank. Reactivity computer will monitor rea::tivity as a function of rod position.

I Acceptance Criteria:

I Design:

I **Design value + 10% on total banks worth

• • Design value + 15~ on any individual bank I Design values are listed on attachec Table.

FSAR/T.S.

I Worth of all rods less most reactive stuck rod must be 1.6% AK/K.

I

• Boron addition for measuring b~nk worth in overlap.

I **If either of these acceptance criteria are not met, the worth of Shutdown Bank A will be measured. If its measured worth is I not within :t, 15% of its design, the minimum Shutdown Margin anc Stuck Rod Worth. test will be performed.

A - 4 I

I STARTUP TEST PP.OGRr-"~ DE5Cr.J PTlON I SALEM UNlT 1 CYCLE 2 Test Boron Endpoint I

Initial Conditions:

I HZP I Endpoints are run for each of the following roe configurations:

I ARO: D in; D, C in; D, C, B ir.;

D, C, B, A in.

I ..

Test

Description:

I Adjust RCS boron to near the just critical en=~~i~~

configuration. Move rods to endpoint while I measuring additional worth on the reactivity c~~?~t~r.

I Acceptance Criteria:

I Design:

Design* + 50pprn 1*

F'SAR/T. S.

I Design* + lOOpp~

I *Design values on attached Table.

I I

I I A - 5 I

I I STARTUP TEST PROGRA..V. DESCRIPTION I SA.LE~ UNIT 1 CYCLE 2

'I Test Isothermal Temperature Coefficient Initial Conditions:

I HZP I The coefficient is measured at each of the follo~ing rod configurations:

I ARO; D in.

Test

Description:

I Starting with T ave

= 547°F, cooldo~n the primary systern approx. s°F.

I Heat the prirnary system back up to 547°F.

The X-Y recorder ~ill plot reactivity vs T ave I

Acceptance Criteria:

I Design:

ARO -~- ~ +/- 3 pcm/'F I D in - ~-I ... 3 pc;..~F FSAR/T.S.

I MTC * < OxlO - 4 [jK/K/°F and > -2.9 x 1 o- 4 LK/1V°F I *The design value of Doppler coefficient is addec to the isothe:mal measurement to derive the moder a tor temperature I coeffh:ient.

I I

I A - 6 I

I.

I START'lJP TEST PROGRJ._"-~ DESCRlP'TlO?;

I Sh.LEM tml T l CYCLE 2 I Test Power Coefficient Initial Conditions:

I Reactor power is not changing. The Xenon I concentration is within 3t of its value. This test is run twice during the e~ui l il::.r i ur..

I start~~ program.

at 100?..

First at SOt power, then I Test

Description:

I Take a heat bal~nce. Change turbine 2t. and 41 with the control rods in manual.

po~er bet~een The I T ave anc bT recorders will monitor chances in these para~eters as a f~nction of time.

I Acceptance Criteria:

I Design:

I Design value *+ 30%

I FSAR/T.~ *.

Mwst fall between upper and lower curve I assWT1ed in FSAR accident analysis.

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A - 8 I lklmt 1~ Cycle 2

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I APPENDIX B I

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I Appendix B I NFTG 79-142 RPG/BEH I December 24, 1979 I

To the Chairman - SORC Committee I. SAFETY EVALUATION FOR SALEM 1, CYCLE 2, WITH OBSERVED FLUX DEPRESSIONS I Flux depressions were observed in the Salem 1, Cycle 2, HZP maps. The Fuel Supply Department has reviewed the data and concurs with the Westinghouse conclusion (see attached I letter) that the plant can safely proceed. with power ascension.

Following are our comments on this:

PSE&G and Westinghouse agreed at a December 17, 1979 I meeting, in Pittsburgh, that calculations should be done assuming. broken rodlets were the cause of the flux depressions observed in Salem l, Cycle 2, BOL HZP .

I Maps 173 and 174. The presence of broken rodlets as a cause of the depressions could not be eliminated with existing data. These calculations and others indicate that the existing INCORE constants are adequa~e to I describe the peaking factors even if broken rodlets were present.

I The ppm bias observed, while slightly over the design acceptance criterion, is not by itself a reason for concern. The temperature coefficient is still negative.

A possible cause is that the 2-D analysis model is I underpredicting the reactivity of the HZP core with its extreme top oriented power distribution. . This bias would be expected to reduce with power escalation.

I Westinghouse also aµalyzed the impact of the observed power distribution on the validity of the current RSE and concluded that it is still valid for power operation I and the *plant can safely proceed to Mode 1 operation.

Based upon numerous telecons, - a meeting in Pittsburgh, I and indep~ndent analysis, the Fuel Supply Department is satisfied with the* Westinghouse efforts **and concurs with the attached transmittal. It should be noted. that these conclusions are based only upon HZP data. Each phase of I start-up testing must stand upon its own merits.

/!Ci'. /,,_./:;. >~

I R. A.~

General Manager - Fuel Supply I RPG:rmg Attachment B - l I

I I,

I Westinghouse Electric Corporation Water Reactor Divisions Box355 Plttsll.lrgh Pennsylvania 15230 February 6, 1980 FP-PS-296

• I *Mr. R. A. Uderitz General Manager - Fuel Supply Public Service Electric and Gas Company I 80 Park Place Newark, New Jersey 07101

Dear*Mr. Uderitz:

I PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION I _ _ _FXY TECHNICAL SPECIFICATION The attached discussion of Fxy Technical Specification for Salem Unit 1 is provided in response to a request by PSE&G during our meeting of I January 21, 1980 in Monroeville. The intent of this discussion is to provide guidance in determining whether Fo(z) is withir its limit in the event the measured F~Y. exceeds the Tech. Spec. limit Fxy* and as such I supersedes that proviaed informally to PSE&G by telecopy on December 31, 1979. .

I An advance transmittal of the enclosed was telecopied to the Salem site on January 31, 1980.

Please contact me if any questions arise~

I Very truly yours, I !JvtJ~

D. W. Wi 11 i ams NFD Fuel Projects I /vyb cc:

• R. J. Gennone I R. P. Gennann H. J. Midura J. A. Nichols E. Rosenfeld I C. F. Barclay I-I B - 2 I*

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DISCUSSION OF FXY TECHNICAL I SPECIFICATION FOR SALEr1 UNIT l I The Technial Specification on Fxy (TS 4.2.2.2) applies only to MODE l operation (reactor power greater than 5% rated thermal power) and as _such allows credit to be taken for power levels less than RTP (rated thermal power). in determining I the limit,

= FRTP [ 1 + 0.2 (l-P)]

I xy where P = fraction RTP I and F~~P = Fxy limits for RTP I The Technical Specification is written in a manner that requires additional flux

. maps to measure Fe~

x at a power level within 20% of the previous power level when ever F~Y exceeds Fx~P. This is required even though F;Y satisfies the pov1er I dependent limit Fiy*

I In the event that F~Y exceeds F~Y then the action is to determine i~ FQ(z) is with1n its limit according to Technical Specification 3.2.2, This may be done I by evaluating the effects of Fxy on FQ(z) to determine if Fq(z) is within its limit whenever F~Y exceeds F~y*

I The effect of an Fxy violation and the determination of whether Fq(z) is within I its li~it depends on the core elevation at which the violation occurs. The evaluation also depends on the available margin between the calculated envelope of Fq(z) x PRel points for load follow maneuvers and the FQ LOCA envelope, as stated in the normalized I K(z) curve on Figure 3.2-2 of the Technical Specification. The viola_tion in Fxy may fn fact occur at a core elevation where the Fq(z) x PRel is below the LOCA envelope I as illustrated in the attachment. In this situatJon no reduction in thermal power below RTP would be necessary.

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• I B - 3

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I ATIACHMENT I

I Illustration of Fxy Violation Evaluation for Salem Unit 1 I In the event F;Y* exceeds F~Y calc;ulate both ExY(z) and Mq(z)*as described below.

If the following relationship is satisfied for all appropriate values of z then I. it may be stated that Fq(z) is within its limit and no action to reduce thermal power as stated in TS 3.2.2 is required:

I I Calculate MQ(z):

I* MQ(z) = [Margin in Fd ]

- at Elevation z

= 2.32 K(z) - Fq

• PReJ(z)Max x 100%

FQ. PRe1(z)Max I where K(z) is shown in Figure 3.2.2 of Technical Specification (see amendment

1. 20 for Salem) and Fq

• PRel (z)Max represents th_e maximum values of Fq

• PRel I plotted versus elevation (z) for Salem 1 Cycle 2. These elevation dependent .

peaking factors and the FQ envelope were transmitted to PSE&G by FP-PS-249 I (reference 1) in conjunction with Revision 4 to the Reload Safety Evaluation, transmitted by.FP-PS-245 (reference 2). It should be noted that a copy of the I attached Figure was included by PSE&G in License Change Request (LCR)79-068 to the NRC on August 9, 1979.

I Calculate Exy(z):

F~y(z)

I Exy(z,P) = [ftfount by which F~y exceeds] = [ """"L,...

Fxy at Elevation z *

- F;y(z,P)J

____....___ x 100% .

Fxy(z,P)

I where F~Y(z) is based on the measured Fxy at elevation z as described in Salem Unit 1 Technical Specification 4.2.2.2.b and F~Y(z,P) is the limit on Fx*y at I elevation z and relative power P (the fraction of Rated Thermal Power) as described in TS 4.2.2.2.C.2.

II I For Salem l Cycle 2 the limit for unrodded pl~nes based on Technical Specification I.

I Amendment 20 for applicable core planes (defined in TS 4.2.2.2.f) is as foll0\*1s:

I L FcY <z < s.o ft., P) = 1.67 [l + 0.2(1-P)]

I Fxy (z ~ 6.0 ft., P} ,;. 1.65 [l + 0.2(1-P)]

B 4

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An illustration of an Fxy violation at an unrodded elevation of 4.9 feet for a I core power level of 95% RTP is given below.

I (1) assume F~.(4.9') = 1.75 I calculate F~Y(z<6' ,.95) = 1.6867 FC - FL calculate Exy (4.9',.95) = xt = 3.753%

I Fxy xy x 100%

I (2) c*alculate MQ(z):

I MQ(4.9') = 2

• 32 :.~i~- 2

• 019 x 1003 = 14.91%

I (3) FQ is within its limit* since the relationship ExY(z) MQ(z) is satisfied.

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I -~-----------------------------

Reference (1) FP-PS-249 to PSE&G dated 8/7 /79 I Reference (2) FP-PS-245 to PSE&G dated 7/26/79 I

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I B - 5 1.

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I SALEM I C.~c.. I~ ~

I Maximum [FQT.pRel] Versus Axial Core Height

• Curing Norma 1 Operation "' 0 r9

') ~.3k)(f,O-""* -x/006;(= /L/.°t I MG,CL/.i -::::. .;;,01~ . . 0

• 2.4 I

I 2.2 I r.'

I 2.0 I

I ,...., 1.8 I

• 0..

I- CT LI..

I § 1.6

-E x

f'O I  :?::

I 1.4 I

1.2 I

I 1.0 I Core Height (feet)

I PS~G I Public Service Electric and Gas Company 80 Park Place Newark. N.J. 07101 Phone 201 '430-7000 NFTG 80-9

• BEH I

January 9, 1980 I

I To the Reactor Engineer - Salem QUADRANT POWER ASYMMETRIES I The Fuel Supply Department has reviewed the attached Westing-house letter, PGD-80-003, dated January 7, 1980, and concurs I with the position expressed in that letter: The excore detectors should be calibrated to remove asymmetries when the reference condition is established, the safety analysis basis is confirmed, and the steady state peaking factors [F~H, Fxy I (z), and Fq (z)] have been verified to be within limits. As per the Westinghouse letter FP-PS-246, a review of the validity of safety analysis should be done if an INCORE tilt, in excess of + 2%, is observed during Acceptance Criteria Testing. This I safety review has been performed, and adequate margin exists in the RSE to accommodate the observed asymmetries. The peak-

_ing factors are within limits, and it is appropriate to cali-I brate the excore detectors such that the asymmetries are removed.

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'2 i) //7 11 r/

._;j ll

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'-!-1,/,'I I~~ VVVL~v~

A/IA,;_,,,... ,

v R. P . Germann Reactor Fuel Engineer I BEH:rmg Fuel Supply Department Attachment I cc . General Manager - Fuel Supply

.Nuclear Fuel Cycle Engr. - Electric Prod.

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I l!f0,1*m;'H B - .6 I lE\llSOf St:n\ln:

! . 95-2001 (200"11 2-75 I

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APPENDIX C I

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I . Appendix C I , AUG~ENT~D SURVEILI.>.NCE PROGRAM SALEM NO. l CYCLE 2 I The purpose of the augmented surveillance program for Salem l cycle 2 is to detect the presence of dropped rodletso **The program relies on sevPr~l t.echniques, which when evaluated can I d~tect the occurrence of ~ropped rodlets. ~

Two power distribution measurements, using the incore flux I mapping system, will be taken at beginning of life hot zero power conditions (where one is normally taken). Flux map an~lysis will pay particular attention to flux depressions. If the design accepta~ce criteria for flux maps is exceeded (as outlined in the I standard startup program) all rod worths will be measured to the N~l condition. ,~he results of the above tests will be repor~ed in the ~alem l cycle 2 startup test report.

I During normal power operation, periodic power distribution measurements are performed. These flux maps will be closely analyzed for significant flux depressions. A change in the I difference between measured and predicted reaction rate integrals *

.on the order of five to ten percent, between consecutive flux maps, could be an indication of dropped rodlets.

I To compliment the flux map results, the following plant para-meters will be continuously monitored:

I l.,..

~

0 The reactivity computer which receives its input frorr.

the four excore power rang~ neutron flux detectors, con-tinuously calculates and displays reactivity.* The reac-I tivity computer has sufficient sensitivity to detect the worth of one rodlet (approximately 10 pcm). Whether a dropped rodlet will be detected by this means, depends on I the core location of the rodlet and the amount it perturbs the core leakage flux.

I 2~ A aropped rodlet which adds negative reactivity .to the core, will cause the primary coolant temperature to drop due to the moderator temperature defect. By monitoring each loop's average temperature (Tavg) this effect can be seen. The I number of rodlets detectable depends on the value of th~

existing moderator* temperature coefficient. It is expected that this instrumentation will detect the worth cf one I rodlet at the beginning cf life and three rodlets at the end cf life.

I 3. ~o distinguish between expected and unexpected reactivity changes, control rod motion must be monitored. That is, reactivity chan9e due to turbin,- load, xenon; boration, er dil~tion resulting in demand control rod motion must be I separated from reactivity changes due to dropped rodlets.

I t---- *-- ---------**--------

C-1

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• 1 I Delta flux (~!) is very sensitive to control rod motion and therefore, can be used to detect control bank motiono I Additional control room recorders will be used for this program.

'l'.he recorders will be scaled to obtain adequate resolut5.on of the parameters being monitored. Chart speed will be sufficiently high to distinguish short term (dropped rodlets) from long term I (xenon) ~ransients.

• When an unexplained in3ication appears on the reactivity computer I and Ta~g recorders, with no control bank motion, a full core flux map will be ta1'.en as soon as core conditions allow

• All indications cf dropped rodle~s will be analyzed to detenr.ine I their effects on safety and appropriate corrective action will be taken. In addition, all indications of dropped rodlets will be

~*

re?orted.

I After a period of approximately three months (during cycle 2 operation) an e.Jaluation of the test program will be performed to I determine if the program should the cycle. A summary report of pared at this point.

be continued for the remainder of this test program will be pre-I I

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'-* -- . ~- ....... ,. .. ***-

C-2