IR 05000338/1988010
| ML20153E716 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 05/02/1988 |
| From: | Burnett P, Jape F, Scott Sparks NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20153E714 | List: |
| References | |
| 50-338-88-10, 50-339-88-10, NUDOCS 8805100085 | |
| Download: ML20153E716 (14) | |
Text
UNITED STATES
'
[,.stico 'o; NUCLE AR REGULATORY COMMISSION
,
J%s
/*
REGION 11 k
101 M ARIETT A STREET. N W.
I "
,
'
- 1 g
ATLANT A.GEORGt A 20323
'.
%, v /
.....
-
!
Report Nos.:
50-338/88-10 and 50-339/88-10
Licensee:
Virginia Electric and Power Company I
Richmond, VA 23261 Docket Nc:.:
50-338 and 50-339 License Nos.: NPF-4 and NPF-7
'
Facility Name: North Anna.1
' 2
-
,
Inspection Conducted:
Apri
), 1988 Inspectors:#9IN e. wF N
P. T. Burnett'~'
Date Signed f tw/W.
'.'
f-2
'
~
5. E. Spark p
Date Signed
-
Approved by:
6A4'
-</ A_
2./
'
'
F. Jape, 5ection Chief f
Date 51gned
'
Engineering Branch Division of Reactor Safety
'
SUMMARY Scope:
This routine unannounced inspection addressed the areas of post refuel-ing startup tests, shutdown margin surveillance, reactor coolant system leakage
j surveillance, and thermal power monitoring, f
j Results:
No violations or deviations were identified.
!
,
i l
l
l
!
I
}
l
!
~
8805100085 880503 j
PDR ADOCK 05000333 i
Q DCD l
-~
_
-
-
-
-
.
.
.
,
.
REPORT DETAILS 1.
Persons Contacted Licensee Employees
- P. 8. Boulden
- R.F.Driscoll,PlantEngineering Manager, Quality Assurance
- L. L. Edwards, Superintendent, Nuclear Training
- R. O. Enfin
- R. Garver, ger, Assistant Station Manager, Operations and Maintenance Reactor Engineer
- S. Hamill, Supervisor, In-Service Inspection Engineering
- E. Hendrixson, Acting Supervisor, S & T Engineering
- M. R. Kansler, Superintendent, Maintenance
- J. H. Lebesstein, Engineer
,
- D. B. Roth, Nuclear Specialist Other licensee employees contacted included engineers, technicians, opera-tors, security force members, and office personnel.
NRC Resident Inspectors
- J. L. Caldwell, Senior Resident Inspector L. P. King, Resident Inspector
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on April 8, 1988, with those persons indicated in paragranh 1 above.
The inspectors described the areas inspected and discussed in detail the inspection findings.
No dissenting comments were received from the licensee.
Proprietary material was reviewed in the course of the inspection, but is not included in this report.
One commitment was received from the licensee:
Inspector Followup Item 338/339/88-10-01: Institute, by October 31 a program to perform Chi-Squared Tests on the source range nuclear, 1988, instruments to ensure their proper functioning prior to and during fuel loading and initial criticality following refueling - Paragraph 5.
3.
Licensee Action on Previous Enforcement Matters Thissubjectwasnotaddressedintheinspection.
4.
Unresolved Items No unresolved items were identified.
_
_ _ _ _ _ - _ _ _ _ _ _ _ _ - _ - _ _ _ _
.
...
.
.
'
<
.
5.
Post-Refueling Startup Tests (72700, 61708, 61710)
The records of the.most recent series of po'st refueling startup tests were reviewed for each unit.
Essentially identical procedures are used.
Initialcriticalitywasattainedfollowing: procedure 1(2)-0P-1.5as appropriate.
Subse accordance with 1(2)quent testing and power escalation was performed in-PT-94.0, Refueling Nuclear Design Check Tests.
From review of these tests and discussions with licensee personnel, it was determined that no tests are performed on the SRNIs to assure they are responding primarily and 3roportionally to neutrons before being used to monitor fuel loading or tie succeeding startup This is not accomplished by the surveillances required by the Technical Specifications since those tests exclude the neutron detectors and amount to no more than setting bistables in response to clean test signals.
The licensee agreed that a more certain method of determining SRNI reliability was desirable and committed to establish a program of Chi-Squared tests to that end.
This commitment was confirmed at the exit interview with a completion date of October 31, 1988, which is prior to the next scheduled refueling outage.
Appropriate times for performing the Chi-Squared tests would be after loading the source-bearing assemblies during fuel loading, prior to pulling shutdown banks or renormalizing the inverse count rate ratio during the first startup on a new core, and any time those activities are interrupted for eight hours or more, a.
Unit 1 Test Results t
The test to determine the critical baron concentration for the all rods out condition (1-PT 94.0, Sequence Step No. 4)l when compared was performed 6-29-87.
The measured value of 1969 ppm agreed wel with the design value of 1995 i 50 ppm.
i The checkout of the reactivity computer satisfied the acceptance criterion that the reactivity derived directly:from period measure-ment and use of the inhour equation agree within four percent of the reactivity computer solution for both positive and negative periods.
The test to determine the isothermal temperature coefficient for the i
all rods out condition (1-PT-94.0, Sequence Step No. 5) was performed 6-29-87.
The measured value of -0.4f pcm/?F at a boron concentration
'
of 1965 ppm agreed well with the design value of -0.67 i 3.0 pcm/?F.
In addition, the measured value was within the Technical Specification 3.1.1.4 value of less than or equal to 6.0 pcm/?F.
,
i Control bank and shutdown bank worth measurements were performed 6-29-87.
The measured and design value worths were as follows:
,
.
.
-
--
--
.,
-
-
.
. _ -
'
.
.
.
.
.
.
Control Bank Heasured Worth Design Value Worth A
343 pcm-321 t 100 pcm B
1323 pcm 1338 1 134 pcm C
766 pcm 780 i 117 pcm
766 pcm 807 1 121 pcm Shutdown B6nk Measured Worth Design Value Worth A
1054 pcm 1056 i 158 pcm B
902 pcm 930 i 140 pcm The measured value for the total rod' worths of 5154 pcm compares favorably with the design value of 5232 i 523 1cm.
All measured control bank and shutdown bank worths were wit 11n the design value worths.
,
The test to determine the hot zero power, boron worth coefficient was performed 6-29-87.
The measured value of -7.27 )cm/ ppm compares well with the design value of -7.25 i 0.73pcm/ ppm..Tlat was determined
'
during the reactivity worth measurement of control bank B during continuous boron dilution.
(All other rcd bank reactivity worths'
were determined by rod swap with control bank B, the reference bank).
-
In addition, the inspectors performed an independent review of control bank B worth based on a detailed review of test data from the reactivity computer strip chart records.
Attachment 2 provides a graphical comparison of the licensee's differential rod worth data to inspector generated data and shows excellent agreement.
However the inspectors noted that the licensee's published results as contain,ed in the Cycle 7 Startup report for the differential worth of-control bank B do not coincide with the original results generated during the test.
There is an apparent smoothing out of the integral worth
,
profile at approximately rod position step 150, with no accompanying explanation as to why this smoothing out was performed.
Although i
this has an no effect on the integral worth of-control bank B or the other bank worths determined by comparison with it, the ins)ectors expressed concern about the publishing of test results whici do not coincidewithactualtestdatawithoutexplanationoftheadjustments
made.
This concern was discussed with the licensee, as an example of R
Joor practice, at the exit interview.
By changing one rod insertion
,
)y two steps from the annotated value on the strip chart the inspectorswereabletoobtainthesamesmoothingoftheltdata.
b.
Unit 2 Test Results Unit 2 startup testing was performed in the period October 19, to l
December 3, 1987.
,
.. - -
_ -.
. _,
_,,.... _ -. _.
_. _.... - _. - -..,.,..,
-.. -,
.
.
,
-
.
Following criticality with D bank partially inserted,.the checkout of the reactivity computer satisfied the acceptance criterion that the reactivity derived directly from period measurement and use of the inhour equation agree within four percent of.the reactivity computer solution for both positive and negative periods.
The AR0 critical boron concentration was 1982 ppmB, which was in good agreement with the design value for the actual conditions of 1994 1 50 ppmB.
The isothermal and moderator temperature coefficients for ARO were-0.6 pcm/ F and 1.13 pcm/ F respectively/ F.The predicted ITC for the
.
actual core conditions was -0.94f3.0 pcm The maximum MTC allowed by Technical Specification 3.1.1.4 is +6.0pcm/ F.
Hence, all acceptance criteria were satisfied.
The reference bank, control bank B measured reactivity worth of 1282 pcmwaslessthanthepredictedvalueof 1367 1 137 pcm but-satisfied the design tolerance of f 101 Thesumofallrodworths, the remainder were measured by rod swap, was less than 6% below the predicted sum, and hence, satisfied the design tolerance.
The measured and predicted boron worth coefficients were -6.82 pcm/ ppm and -7.27 pcm/ ppm respectively, and the design tolerance was satisfied.
During power escalation, flux maps were taken using the moveable detector system at 28, 47, and 100% RTP.
In all cases, Fn satisfied Technical Specification 3.2.2, F satisfied Technical S 3.2.3,andQPTRsatisfiedTechni$1 Specification 3.2.4.pVcification No violations or deviations were identified.
6.
Determination of Reactor Shutdown Margin - Units 1 and 2 (61707).
The inspector reviewed the Unit 1 performance test procedure, 1-PT-10, Determination of Shutdown Margin, completed 11-23-87.
A recently com-pleted procedure for Unit 2 shutdown margin was not available, however,
determination of reactor shutdown margin for both units is essentially the The review consisted of verification of technical adequacy, same.
com)11ance with procedural requirements, and compliance with station
'
Tecinical Specifications.
The inspector verified that xenon worth curves, rod worths, boron worths, and temperature defect values were properly transcribed from the correct version of the Station Curve Book.
The calculated shutdown margin of -3207 pcm satisfied the Technical Specification shutdown margin of equal to or more negative than -1770 pcm.
!
-
-
__ _ ---,
,
.
.
,
'
.
In addition, the inspector reviewed operating procedure 2-0P-10, Estimated Critical Position, for the initial startup of Unit 2 Cycle 6,: completed 11-3-87.
The arecedure allows for the estimated critical condition to be calculated eitler manually or by a computer program.
The estimated critical condition for this startup was determined manually.
For a boron concentration of 1928 ppm, the predicted control rod positions, which include an administrative span of 400 pcm were as follows:
Bank C Bank D Worth 197 steps 6TTEeps-876 pcm 228 steps 194 steps-76 pcm Actual critical conditions of 1922 ppmB, Bank C at 228 steps, and Bank D at 98 steps, agreed well with the predicted rod positions.
7.
Reactor Coolant System Leakage Measurements (61728)
The microcomputer program RCSLK9, which was developed by the NRC Indepen-dent Measurements Program, is described in NUREG-1107, RCSLK9: Reactor Coolant System Leakage Determination for PWRs.
To customize the program for use at North Anna, plant-specific parameters for each unit were obtained from review of the following documents: Updated FSAR, vendor manuals for steam generators and pressurizers, station curve books and internal memoranda.
The parameter list for Unit 2 (Unit 1 is identical)
is given in Attachment 3.
To obtain data for use with RCSLK9 and TPDWR2, which is discussed below, the licensee established Group Review 11 on the plant computer to monitor and print out the required data at fifteen minute intervals.
Because of makeup to the VCT, the longest span of time for the calculation was 1.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />.
Over that period, the results from RCSLK9 were acceptable.
The licensee's calculational program, which was performed in parallel with the inspection activities, uses ten-minute-averaged data for the beginning and ending points.
Using the averaged data provided by the licensee, the results from RCSLK9 were in good agreement with the licensee value of 0.78 and 0.26 gpm identified and unidentified leakage respectively.
The output from RCSLK9 is given in Attachment 4.
'
No violations or deviations were identified.
8.
Thermal Power Determination (61706)
The NRC independent measurement program for determination of reactor thermal power is described in NUREG-1167, TPDWR2: Thermal Power Determin-ation for Westinghouse Reactors, Version 2.
To customize the program for use at North Anna 1 and 2 the necesiary system 3arameters were obtained by review of the documents listed in paragraph 7 a)ove.
The insulation losses were adjusted to duplicate the licensee' parameters for s measured
-
-
.
.
,
.
-
.
.
'
.
,
losses on Unit 1. To obtain data for use with the microcomputer program TPDWR2, the inspectors again made use of the Group Review 11 output.
The data obtained, although sufficient for use in TPDWR2, were not in the order or, in all cases, in the units required for input to that program. A SUPERCALC3 spreadsheet was created to facilitate ordering and conversion of the data for input to TPDWR2.
The customized plant parameters for Unit 2 (Unit l's are identical) and a typical set of input data are given in Attachment 5.
Both units have both feedwater and steam flow venturis, and the latter are used in the licensee's calculation of thermal power.
TPDWR2 was first run in its designed mode of using feedwater flow, and the agreement in its result was within 0.1% of licensee values on average.
The input data were then adjusted to simulate the steam flow venturi data, and the results were in even better agreement with licensee values.
Typical results for Unit 2, correspondin Attachment 5, are given in Attachment 6.g to the input data in No violations or deviations were identified.
Attachments:
'
1.
List of Acronyms and Initialisms 2.
Unit 1, Bank B, Differential Reactivity Worth 3.
RCSLK9 Parameters for Unit 2 4.
RCSLK9 Results for Unit 2 5.
TPDWR2 Parameters and Data for Unit 2 i
6.
TPDWR2 Results for Unit c i
l i
W
l
.
-
.
'
.
Attachment 1 Acronyms and Initialisms AR0 - All Rods Out F
- Nuclear Enthalpy Rise Hot Channel Factor dH F
- Total Heat Flux Hot Channel Factor F9AR-FinalSafetyAnalysisReport gpm gallon per minute ITC - Isothermal Temperature Coefficient MTC - Moderator Temperature Coefficient OP
- Operating Procedure pcm - Percent Millirho (unit of reactivity)
ppm - Parts Per Million ppmB - Parts per Million Boron PT
- Periodic Test QPTR - Quadrant Power Tilt Ratio RTP - Rated Thermal Power SRNI - Source Range Nuclear Instrument VCT - Volume Control Tank i
i
N c'
O W
uc w
w
O LO
$.,
W Z
'
ca.
w
$
O U
'
e a
-
cc E3
'
o k/
'25 N w.
N
<
CQ o-a e--
~.
%%
r,'
h s-E
%s
,
s Y
w O9
__p V
g v
i
,
D V'
Z
-<
&%
n d
~ '~ a m
//
rif
,_ b d$
H h$
M M
~s O
y%
a (m Q f
,w Q
D Q
,
x g
F,N C
-?d
% s.,
.
e&O l
u, o
G)
O
-
d1LS/NDd
.
.
-
.
.
_
_,
.
'
Attachment 3
'
'
PARAMETER LIST
'
Unit Identification:
Plant Name NORT". ANNA Unit Number
Docket Number 50-339 Nuclear Steam System Supplier Westinghouse Vessel and Piping:
Volume 8557.2 cubic feet Pressurizer:
Level Units
%
Temperature Compensated No Calibration Curve Slope 354 pounds per %
Upper Level Limit 100 %
Lower level Limit 0%
Relief Relief Tank Volume Control Tank:
Level Units
%
Calibration Curve Slope 116.7 pounds per %
Upper Level Limit 100 %
Lower level limit 0%
Geometric Method Available No Drain Tank:
Level Units
%
Calibration Curve Slope 70 pounds per %
Upper Level Limit 76.5 %
Lower level limit 32.5 %
Geometric Method Available No Relief Tank:
Level Units
%
Calibration Curve Slope 921 pounds per %
Upper Level Limit 70 %
Lower level limit 30 %
Geometric Method Available No
' ~
..
.
.__
~
Attachment 4
.
NRC INDEPENDENT MEASUREMENTS PROGRAM REACTOR COOLING SYSTEM LEAK RATES STATION: NORTH ANNA TEST DATE APRIL 6, 1988 UNIT
- 2 START TIME:
DOCKET : 50-339 DURATION 1.333 hours0.00385 days <br />0.0925 hours <br />5.505952e-4 weeks <br />1.267065e-4 months <br /> TEST DATA Initial Final System Parameters Pressure, psia 2257.32 2257.32 T Ave, degrees F 586.75-586.66 Water Levels Pressurizer, %
62.54 62.85 Relief Tank, %
50 Volume Control Tank, %
47.77 40.34 Drain Tank, %
28.98 36.67 Water Charged = 0 gal Water Drained = 0 gal
'
TEST RESULTS Change in Water Inventory in pounds:
Vessel & Piping
Relief Tank (1)
O Pressurizer 110 Drain Tank (1)
538 Volume Control Tank (1)
-867
Less: Water Charged
Collected Leakage 538 Plus: Water Drained
______
Cooling System-697 Leak Rates in gpm (3):
Gross 1.05 Identified 0.81 Unidentified 0.24 (1)
Determined from tank calibration curve.
(2)
Determined from tank dimensions.
(3)
The density used for converting inventory change to leak rate was 62.31 pounds / cubic foot based on standard conditions.
,
l t
, - -
_ _
_
_
.
'
'
Attachment 5 i
.
'
IIAfBALAICIDA7A 10171AIIA2 4-688 PLAlfPARABIftlS:
IIACf01000 Lilt $fSfit IlfLICfillIISOLifl08 Pusp Power (BW each)
5.9 lasideSarfaceArea(stit)
26,000 PuspIfficiency(1)
90.0 leat loss Coeffieleat (Bf0s/hr sq ft)
110.00 PressuriserInsideDiaseter(inches)
84.0 30llIFLICfiftIISBLifl01 Sfl&HGilliif0lS InsideSerfaceArea(sqft)
9,423 DoselisideDiaseter(inhes)
168.50 Thickness (inches)
LO Ilser Oatside Disseter (lacha)
56.75 ThernalCondectivity(if0s/hritT)
0.100 InsberofIlsers
Eolature Cany-over (1) in A 0.100 LICIISIDfillHALPOVil(HWt)
2893 HolstereCartyover(1)inP, 0.100 foistereCarr.v-over(1)isC 0.100 DATA:
Sif1 Sif2 Sif1 Sif2 fill 1059 1108 fill 1959 1108 SfillGIlitif0tA StillGilllATORB SteasPressere(psia)
895.2 ISt.1 SteasPressure(psia)
890.7 890.0 feedsaterfler(Illb/br)
4.286 4.297 Teedsaterllos(16lb/hr)
4.269 4.272 feediaterfesperatore(1)
439.7 439.9 Teediaterfesperatore(F)
441.1 441.3 SarfaceBlovdova(ges)
0.0 0.0 SurfaceBlovdova(gps)
0.0 0.0 BottesBlondova(gr.)
30.0 30.0 BottonBloedown(gps)
30.0 30.0 Water Level (liches)
63.7 62.7 WaterLevel(inches)
85.1 84.1 S!!!HGillllf0IC SteasPressure(psla)
885.6 885.9 feediaterflov(16lb/hr)
4.231 4.235 Teediater fesperature (F)
441.1 441.2 SurfaceBlordova(gre)
0.0 0.0 BottonBlondova(gps)
30.0 30.0 WaterLevel(1ches)
62.4 62.8
,
l l
LifDOWILill CHAIGIIGLIII flee (gps)
85.8 85.6 Flos(gts)
41.8 41.4-fenperatore (f)
553.8 553.7 fesperature(f)
483.8 483.8 l
i FIISSilllit 1110f0I Pressure (psia)
2245.5 2246.1 7are(f)
586.7 586.7-VaterLew!(liches)
244.1 245.7 fcold(i)
553.5 553.5
- - _ __ __ __ _ ___
-
-
,
_
Page 1 of 2:
,
-
-
Attachment 6
.*
,
\\^
HEAT BALANCE NORTH ANNA 2 4-6-88 DATA SET 1 OF 2 ENTHALPY FLOW.
POWER POWER j
1059 hours0.0123 days <br />0.294 hours <br />0.00175 weeks <br />4.029495e-4 months <br /> (BTUs/lb)
(E6 lb/hr)
(E9 BTUs/hr)
(MWt)
STEAM GENERATOR A I
Steam 1195.9 4.277 5.114 l
Feedwater 419.0-4.286-1.796 Surface Blowdown 525.9 0.00000 0.00000-Bottom Blowdown 470.9 0.01200 0.00565 l
_ _ _ _ _ _..
Power Dissipated 3.3241 973.5 STEAM GENERATOR B Steam 1196.0 4.260 5.095
I Foodwater 420.6-4.269-1.796 i
Surface Blowdown 525.2 0.00000 0.00000
Bottom Blowdown 471.3 0.01199 0.00565
_______
Power Dissipated 3.3049 967.9 STEAM GENERATOR C Steam 1196.2 4.218 5.045 Feedwater 420.6-4.231-1.780 Surface Blowdown 524.3
.0.00000'
O.00000 Bottom Blowdown 470.9 0.01200 0.00565
_______
Power Dissipated 3.2714 958.1 OTHER COMPONENTS Letdown.Line 551.9 0.03208 0.01770 Charging Line 469.0-0.01696-0.00796 Pressurizer 643.0 0.00141 0.00091 Pumps-0.05458 Insulation Losses 0.00424
_______
Power Dissipated-0.03968-11'.6
______
REACTOR POWER 2887.9
_ _ _ _ _ _ _ _ _ _ _
-
.
-
.Page 2laf1?
Attachment 6-f.
j -
.
'
HEAT BALANCE NORTH ANNA 2
-.
4-6-88 DATA SET 2 0F 2 ENTHALPY FLOW POWER POWER 1108 hot;s (BTDs/lb)
(E6 lb/hr)
-(E9 BTUs/hr)
(MWt)
STEAM GENERATOR A-Steam 1195.9 4.288
.5.128 Feedwater 419.3-4.297-1.802 Surface Blowdown-525.7 0.00000 0.00000 Botton Blowdown 470.9 0.01200 0.00565
_______
Power Dissipated 3.3318 975.8 STEAM GENERATOR B Steam 1196.0 4.263 5.098 Feodwater 420.8-4.272-1.798 Surface Blowdown 525.1 0.00000 0.00000-Bottom Blowdown 471.4 0.01199 0.00565
_______
Power Dissipated 3.3064 968.4 STEAM GENERATOR C Steam 1196.2.
4.222 5.050 Feedwater 420.7-4.235-1.782 Surface Blowdown 524.4 0.00000 0.00000 Botton Blowdown 471.0 0.01200 0.00565
_______
Power Dissipated
'3.2740 958.9
OTHER COMPONENTS
'
i l
!
Letdown Line 551.8 0.03201 0.01766 Charging Line 469.0-0.01680.
-0.00788
-
Pressurizer 642.9 0.00141 0.00091 Pumps-0.05458 Insulation Losses 0.00424
_______
Power Dissipated-0.03965-11.6
______
REACTOR F0WER 2891.4
,
l
.!
_
.