Text

Insp Repts 50-348/85-39 & 50-364/85-39 on 850923-27.No Violation or Deviation Noted.Major Areas Inspected: post-refueling Startup & Surveillance Tests
	ML20133Q263
Person / Time
Site:	Farley
Issue date:	10/17/1985
From:	Burnett P, Jape F; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20133Q261	List: IR 05000348/1985039; ML20133Q258;
References
	50-348-85-39, 50-364-85-39, NUDOCS 8511010464
	Download: ML20133Q263 (11)
	v • d • e

{{#Wiki_filter:_. .- . .. . . . . _ ____. _ _ - . - - - _ a . UNITED STATES

[an atougDo  NUCLEAR REGULATORY COMMISslON     :
*** *,^
%    REGION 11      ' *'

101 MARIETTA STREET N.W.

j - 2 ATLANTA, GEORGI A 30323

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f Report Nos.: 50-348/85-39 and 50-364/85-39 Licensee: Alabama Power Company 600 North 18th Street

!  Birmingham, Al 35291 Facility Name: Farley 1 and 2
, Docket Nos.: 50-348 and 50-364   Licensee Nos.: NPF-2 and NPF-8   i i
, Inspection Conducted: September 23-27, 1985       -

Inspector: h P.T. Burnett n m yd>Y /*ht) Os /.9&S~ Date Signed Approved by: / F. Jape, Section Chief

   /~.-    /e/7/.r5 Date' Signed Engineering Branch Division of Reactor Safety
'

SUMMARY Scope: This routine, unannounced inspection involved 34 inspector hours onsite in i the review of post-refueling startup tests and surveillance tests, i Results: No violations or deviations were identifie t

t 8511010464 e51021 PDR ADOCK 05000348 G PDR I _ . - _ . _ . - , ., . - - _ _ _ _ _ - _ . . . __ -._

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, REPORT DETAILS l i , Persons Contacted Licensee Employees

E. W. Carmak, Reactor Engineering Supervisor ,

  *R. D. Hill, Manager of Operations
  *W. MacDonald, Reactor Engineer

, *R. Marlow, Technical Supervisor

  *D. N. Morey, Assistant Plant Manager - Operations
  * G. Ware, SAER Supervisor J. D. Woodard, Plant Manager
:   Other licensee employees contacted included office personne NRC Resident Inspectors
  * H. Bradford, Senior Resident Inspector-l   *B. R. Bonser, Resident Inspector
  * Attended exit interview.

Exit Interview l The inspection scope and findings were summarized on September 27, 1985 with i those persons indicated in paragraph 1 above. The licensee identified as proprietary only that information contained in the Westinghouse reports reviewed by the inspector. That information is not repeated in this repor The licensee made one commitment that will be tracked as an Inspector ~

Followup Item:

348/364/85-39-01, Procedures will be modified to confirm a negative moderator temperature coefficient above 70% rated thermal power following maasurement of the AR0, zero-power, moderator temperature coefficient - paragraph 5.a(2). * Licensee Action on Previous Enforcement Matters Not inspecte . Unresolved Items Unresolved items were not identified during this inspection.

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5. Post-Refueling Startup Tests (61702,,61708,61710, 61711,72700) Unit 1, Cycle 7 The following completed engineering test procedures (ETP) and

surveillance test procedures (STP) were reviewed

 (1) FNP-1-STP-112 (Revision 9), Rod Drop Time Measurement, demonstrat-ed that each rod cluster control assembly (RCCA) had a drop time to dash pot entry less than the 2.2 second maximum allowed by Technical Specification 3.1.3.4.

,

 (2) FNP-1-ETP-3601 (Revision 2), Zero Power Test Procedures, was
'

performed on May 25-26, 198 The measured all-rods-out (AR0) critical boron concentration was 2066 ppmB, which was in good agreement with the predicted value of 2072 ppmB.

, The reactivity computer checkout was based upon two measurements of positive periods, having corresponding reactivities or 19.8 and 37.3 percent millitho (pcm), and one negative period with a reactivity of -16.2 pc The ARO moderator coefficient was derived from the measurement of the isothermal temperature coefficient measured at a boron concen-tration of 2060 ppmB. The resulting moderator coefficient of pcm/ degree F was less than the Spcm/ degree allowed by Technical , Specification 3.1.1.3.a, but more than the predicted value of 4 '

l pcm/ degre Technical Specification 3.1.1.3.a further requires that the moderato" ccefficient be negative above 70% power. Discussions with licensee personnel revealed that procedures did not address this requirement. Nevertheless, the requirement had not been ignored. The fuel : vendor, Westinghouse, at the request of the licensee had provided rod withdrawal limit curves to reduce boron concentration and assure a non-positive coefficient at 70% powe a licensee engineer then evaluated the curves and limiting boron concentrations, and d(monstrated that, even with the most rapid startup, power defect and xenon buildup would make an unacceptable combination of rod position and boron concentration impossibl At the exit interview the licensee agreed with the position that l assurance of conformance to a limiting condition of operation ! should not rest on the initiative of an individual, but should be l a part of a procedure even when there is no corresponding l surveillance requirement. To that end, the licensee made the following comaitment to be implemented prior to the next operating __ __ _ _ _ . _ _ . . _ . _ __.. _ _ _ _ _ _ . _ ,

    . _ . _ .

i* -

cycle: Procedures will be modified to confirm a negative moderator temperature coefficient above 70% rated thermal power following measurement of the AR0, zero-power, moderator temperature coefficient (inspector followup item 348/364/85-39-01).

To perform rod bank worth measurements, the licensee first desig-nated control bank B, the anticipated highest worth bank, as the reference bank. Starting from the AR0 condition, boron dilution was initiated and bank B was inserted incrementally to compensate for the continuously increasing reactivity. The reactivity change associated with each increment of bank B insertion was measured using the reactivity computer. That is, the incremental reactivity changes were determined graphically from the reactivity computer recorder trace of reactivity versus time. In reviewing the chart record and performing an independent analysis of the reactivity increments, the inspector concluded that the reactivity computer had been used within its calibrated range. Some ending,

; after incremental motion, reactivities were more negative than the-16.2 pcm limit of the reactivity computer calibration. However, the majority of the trace used to determine the slope and inter-cept fell within the calibrated span. A photocopy of a portion of the chart record (attachment 1) is enclosed with this report to

, illustrate that poin A spreadsheet from the SUPERCALC 3 (Release 2) microcomputer program was used to evaluate the raw data obtained from the chart record by the inspector. The spreadsheet converted observations of reactivity and rod bank position to reactivity increments, rod increments, and differential and integral reactivity worths. The licensee's values of differential were added to the spreadshee Then the program was used to graph the results of the differential worth measurements. The spreadsheet (attachment 2) and graph (attachment 3) are enclosures to this report.

(3) FNP-1-STP-121 (Revision 16), Power Range Axial Offset Calibration l was performed on June 2, 1985. Using a least squares fit spreadsheet with the SUPERCALC 3 program, the inspector performed an independent analysis of the licensee's correlation of chamber currents to flux axial offset. A graph from that analysis is provided as attachment 4 to this report. A table comparing the licensees results with those obtained from the spreadsheet is 4 provided in atth(. bent 5. The agreement between results is acceptable for each chambe (4) FNP-85-0904, Unit 1, Cycle 2 Startup Report. That report was an accurate summary of the test results, and confirmed that all acceptance criteria had been satisified.

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 (5) WCAP-10795 (Proprietary), The Nuclear Design and Core Management of the Joseph M. Farley Unit 1 Power Plant, Cycle 7, March 1985.

This document was the source of the numerical acceptance or ' performance criteria used to establish acceptable startup test resul ts.

I Unit 2, Cycle 4

Most of the startup program had been witnessed as reported in '

;  inspection report 364/85-14. This inspection was limited to' confirming   l

, that the remaining startup program was completed and an acceptable i ! startup report issued. The documents reviewed included: '

 (1) FNP-2-ETP-3605, Power Ascension Procedure,
;
 (2) FNP-2-STP-121, Incore-Excore Detector Calibration,
'
 (3) FNP-2-STP-115.1, RCS Flow Measurement, and

.

 (4) FNP-85-0516, Unit 2, Cycle 4 Startup Repor (5) WCAP-10674 (Proprietary), The Nuclear Design and Core Management of the Joseph M. Farley Unit 2 Power Plant, Cycle 4, November 1985 (Revision 1 incorporated August 1985). This document was the source of the numerical acceptance or performance criteria used to j  establish acceptable startup test results.

f All procedures were found to be complete and the startup report was an i accurate representation of the results obtained.

l No violations or deviations were identified in the inspection of the l post-refueling startup prcgrams.

l Surveillance Tests (61702,61707,61708) l The following completed surveillance tests were reviewed for technical

adequacy, compliance with procedure requirements, and compliance with
Technical Specification requirements for frequency of performance during the
current operating cycl '

l Unit 1 '

 (1) FNP-1-STP-29.1, Shutdown Margin Calculation (T-average at 547),
 (2) FNP-1-STP-29.2, Shutdown Margin Calculation (T-average at 547),

!

 (3) FNP-1-STP-110, Determination of Limiting Hot Channel Factors, and (4) FNP-1-STP-115.1, RCS Flow Measurement (Heat Balance Method).

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__

5 Unit 2 ! ,

 (1) FNP-2-STP-29.1, Shutdown Margin Calculation (T-average at 547),

. (2) FNP-2-STP-29.2, Shutdown Margin Calculatior. (T-average _ 547),

 (3) FNP-2-STP-110, Determination of Limiting Hot Channel Factors, l  (4) FNP-2-STP-111, Overall Core Reactivity Balance,
,
 (5) FNP-2-STP-114.1, Moderator Temperature Coefficient for Boron i  Concentration Less Than 300 ppmB, and (6) FNP-2-STP-115.1, RCS Flow Measurement (Heat Balance Method).

No violations or deviations were identified during the inspection of these surveillance test . Followup on Inspector Identified Item (92701)

(0 pen) Inspector Followup Item 348/364/84-12-02: Evaluate use of chi-squared test for source range nuclear instruments (SRNI). The licensee has insti-i tuted the use of the test to confirm operability of the SRNIs prior to starting a refueling outage. Currently the licensee is considering more i frequent use of the test to assure operability of the SRNIs during those j periods when reactor safety is particularly dependent on them : during refueling, during startup of a new core, and during periods when the vessel

water level had been lowered for maintenance and inspectio Attachments: Reactivity Computer Trace Table of Rod Worth Graph of Rod Worth N41 Least Squares Test Nuclear Instrument Correlation

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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . ___ _ _ . __ FAR1_EY UNIT 11 Beginning of Cycle 7 Inspect ion Report 50-348/364 4 5-39 WMM2 Rod 5tcps Reactivity (pcm> Het Change Differcntiol Flot et Lacenses Table of Rod Worth Start, Finish Start Finish Rho (pce) Steps (pcm/ctep) Step Different Control Rod Bank 8 0 228 22 .O 4 .O 4 .O 1.64 214 1.60 20 .O .O .O 3.00 199 3.50 19 .0 1 .8 1 .0 3.03 196 3.00 19 .0 1 .0 2 .0 3.48 191 3.50 18 .0 .8 1 .0 3.83 186 3.80 18 .0 .0 2 .0 4.07 181 4.10 17 .0 .9 2 .0 4.22 175 4.20 17 .5 1 .2 2 .5 4.07 169 4.10 16 .0 1 .2 3 .5 4.68 163 4.70 16 .0 1 .9 2 .0 4.85 157 4.80 15 .O 1 .O 3 .O 5.15 151 5.10 14 .0 1 .0 3 .0 5.33 145 5.30 14 .0 1 .3 3 .0 5.42 139 5.40 13 .O 1 .4 3 .O 5.88 133 5.80 1 .2 2 .O 6.30 *

        .28 6.30 13 .O 12 .0  1 .6 2 .0 6.63 124 6.60 12 .0  1 .5 2 .0 7.00 120 7.00 11 .0  1 .8 2 .0 7.10 116 7.10 11 .0  1 .7 3 .0 7.50 112 7.50 11 .0  1 .2 3 .0 7.68 108 7.60 10 .0  1 .8 3 .0 7.65 104 7.60 10 .0  1 .7 3 .0 8.40 100 8.40 9 .0  1 .0 3 .0 9.08 96 9.00 9 .0  1 .0 1 .0 9.35 93 9.30 9 .O  2 .9 3 .O 9.73 90 9.60 8 .O  1 .9 1 .O 9.90 87 9.80 8 .0  1 .8 2 .0 10.20 85 10.30 8 .0  1 .2 2 .0 10.10 83 10.00 8 .O  1 .8 2 .O 10.15 81 10.30 8 .0  1 .2 2 .0 10.35 79 10.50 7 .0  1 .9 2 .0 10.55 77 10.50 7 .0  1 .1 2 .0 10.30 75 10.30 7 .0  1 .4 2 .0 10.40 73 10.30 7 .0  1 .3 1 .0 9.80 71 9.80 7 .0  1 .0 2 .0 10.00 69 10.00 6 .0 .9 2 .0 10.50 67 10.50 6 .0  1 .2 2 .0 10.60 65 10.50 6 .0 .5 2 .0 10.50 63 10.50 6 .0 .2 2 .0 10.65 61 10.50 6 .O .5 2 .O 10.35 59 10.30 5 .O .1 2 .O 10.00 57 10.00 5 .0  1 .7 3 .0 9.68 54 9.60 5 .0  1 .0 3 .0 8.65 50 8.60 4 S. 0   4 .5 -1 .5 .63 46 7.60 4 .O  1 .O 2 .O 6.33 42 6.25 4 .0 .8 1 .0 4.90 38 4.88 3 .0  1 .7 2 .0 3.45 33 3.42 3 .O  1 .1 1 .O 2.98 27 3.00

' 2 .0 .9 1 .0 1.98 21 1.83 1 .0 .5 .0 1.25 16 1.25 1 .7 - .7 1 .55 7 .57

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ATTACHMENT 3

GRAPH DTTUlTTDRTIT Control Rod Bank 3 Differential Worth (pcm/siep) 15-x Licensee ,

 (10-O
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 $

Y s CS T

 $ 5-
  /

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m N O i , , , , , i i Q 0 25 51 76 101 127 152 177 203 228 Average Bank Position (steps)

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_ _ _ _ _ _ _ - __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ . _ _ _ _ _ _ ATTACllMENT 5

NUCLEAR INSTRUMENT CORRELATION FARLEY UNIT 1: Inspection Report 348/85-39 _ Results of Incore-Excore Nuclear Instrument Correlation Beginning of Cycle 7 FULL POER CWWWER CURRENT = IZ< current at zero offset) + Be[RXIAL OFFSET] IZ IZ IZ uncertainty 8 B B uncertainty CMWSER LICENSEE SUPERCHLC 3 SUPERCALC 3 LICENSEE SUPERCALC 3 SUPERCALC 3 H41 TOP 192.811 192.690 33.000 .989 .997 .184 N41 BO .442 191.580 32.810 -1.247 -1.259 .175 N42 TOP 188.370 188.243 36.410 1.068 1.077 .204 H42 BO .255 106.404 21.475 -1.323 -1.335 .118 N43 TOP 188.329 188.209 29.300 .995 1.003 .164 N43 BO .670 200.819 39.880 -1.346 -1.359 .203 N44 TOP 181.742 181.620 8.715 1.036 1.046 .051 H44 BO .008 187.154 42.039 -1.283 -1.294 .229 The full power chamber currents reported here are only about 2/3 that reported at similar facilities. However, the licensee re-ports no difficulity with systes performance as a omsequence.

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v t e Inspection Report - Farley - 1985039
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \|
1985 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 8(0) 11(0) 15(0) 17(0) 19(0) 20(0) 21(0) 22(0) 23(0) 24(0) 25(0) 26(0) 27(0) 28(0) 30(0) 31(0) 32(0) 33(0) 34(0) 35(0) 36(0) 37(0) 38(0) 39(0) 40(0) 41(0) 42(0) 43(0) 44(0) Security Operator Exams Emergency Planning Other

IR 05000348/1985039

Text

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