IR 05000280/1986040

Insp Repts 50-280/86-40 & 50-281/86-40 on 861129-1203.No Violations or Deviations Noted.Major Areas inspected:post- Refueling Startup Tests,Thermal Power Determination & RCS Leakage Measurements

ML20212G023
Person / Time
Site:	Surry
Issue date:	12/23/1986
From:	Burnett P, Jape F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20212G015	List: IR 05000280/1986040 ML20212G011
References
50-280-86-40, 50-281-86-40, NUDOCS 8701120242
Download: ML20212G023 (12)

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UNITED STATES

[A M709'o NUCLEAR REGULATORY COMMISSION y , REGION 11 3 j 101 MARIETTA STREET, * g ATLANTA GEORGIA 30323

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Report Nos.: 50-280/86-40 and 50-281/86-40 Licensee: Virginia Electric and Power Company Richmond, VA 23261 Docket Nos.: 50-280 and 50-281 License Nos.: DPR-32 and OPR-37 Facility Name: Surry 1 and 2 (

l Inspection Conducted: Novem er 29 - December 3, 1986 Inspector: _!uBurnett

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/ / Date Signed Approved by: _ W "/ O ;

F. Jape, Chief F/ Date Signed i Test Programs Section l Engineering Branch )

Division of Reactor Safety

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SUMMARY

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Scope: This routine, announced inspection addressed the areas of post-refueling i startup tests (Unit 2), thermal power determination, and reactor coolant system

! leakage measurement Results: No violations or deviations were identifie l l

8701120242 870106 PDR ADOCK 05000280 0 PDR

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REPORT DETAILS Persons Contacted Licensee Employees

• R. F. Saunders, Station Manager D. L. Benson, Assistant Station Manager

• H. Blount, Supervisor, Surveillance and Testing Engineering

• W. D. Craft, Licensing Coordinator

• E. S. Grecheck, Superintendent of Technical Services R. Johnson, Operations Supervisor A. McNeil, Senior Staff Engineer

• H. L. Miller, Assistant Station Manager

• T. L. Reynolds, Quality Assurance Department Other licensee employees contacted included engineers, technicians, operators, and office personne NRC Resident Inspectors W. E. Holland, Senior Resident Inspector

'* Attended exit interview Exit Interview The inspection scope and findings were summarized on December 3, 1986, with those persons indicated in paragraph 1 above. The inspector described the areas inspected and discussed in detail the inspection finding No dissenting comments were received from the license The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspector during this inspectio Inspector Followup Item 280/281/86-40-01: Revise Procedures 1/2-PT-10 to Incorporate All of the Parameters Required for the Surveillance -

paragraph Inspector Followup Item 280/281/86-40-02: Review Steam and Feedwater Flow Venturi Calibration and Reliability paragraph . Licensee Action On Previous Enforcement Matters This subject was not addressed in the inspectio >

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' Unresolved Items No unresolved item was identified during this inspectio . Unit 2 Post-Refueling Tests (72700, 61708, 61710) Pre-Criticality Activities The inspector witnessed part of the control rod drop time testing from the control room and reviewed some of the recorder traces used in the determination of drop tim All drop times were satisfactory, but subsequently not all rods could be withdrawn because of electrical problems. Resolution of those problems delayed starting the approach to criticality by several hour Initial Criticality for Cycle 9 The approach to criticality was performed in accordance with Operating Procedure 2-0P-1.4, which is used for reactor startup any time in cycle lif The sequence of major activities, after heating the reactor coolant system (RCS) up to the no-load average temperature and operating pressure, was to fully withdraw the two shutdown banks of control cluster assemblies (rods) and to perform a series of batch dilutions of the RCS to reduce the boron concentration from the refueling concentration (about 2100 ppmB) to the predicted critical concentration with the desired control rod configuration (0 bank at 160 steps). The source range nuclear instruments (SRNIs) are monitored during the batch dilution process to assure that the count rate does not double. That eventuality would require a stop to the dilution process and an engineering evaluation of the situation. However, no detailed monitoring of the multiplication of neutrons, such as plotting an inverse count rate ratio (ICRR) is required, nor is any test performed to assure the SRNIs are responding primarily to neutron The inspector chose to perform the ICRR monitoring independently and to test the SRNIs for response to neutron Prior to the start of the procedure, 50 100-second counts were obtained from SRNI N31 at about 850 counts per observation. Using the chi-square test of variance the data were analyzed in sequential groups of ten and subsequently pooled into larger groups for additional chi-square analysi Standard texts on determining the reliability of pulse-counting nuclear instruments suggest the range of acceptability for the chi-square test be from 2 to 98% probability that the actual variance would be larger than the observed variance with the expected distribution (Poisson in this case)

of observations. The first group of ten failed the test, largely because one observation was nearly twice the average of the group, and that group was not used furthe Three other groups passed with probabilities larger than 50% and one had a probability only slightly

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larger than acceptabl Pooling the data into larger groups such as the last 15, last 20, up to the last 40 was successful except for the last case, which had too laroe a probability. This consistent result that the expected distribution of observations would have a large variance is indicative of a constant source being counted along with the neutron source. A 60 hertz source would have had such a large contribution as to cause all the tests to fail, but the presence of a higher harmonic is a possibilit The inspector was not present during the first batch dilution. But from that time on obtained five 100-second counts from both N31 and N32 at intervals of about 15 minutes (The set of observations was repeated when the chi-square test was unsuccessful until an acceptable result was obtained). The average of each acceptable set of observations was compared with a similar average obtained prior to initiating the approach to criticalit ICRR was calculated and plotted against the estimated boron concentration at the midpoint of the observation During the dilution process, with all of the control banks in, the smallest value of ICRR observed was 0.8 After the desired boron concentration was obtained, the licensee withdrew the control banks in overlap until bank C was at 98 steps withdrawn. At that point, a base count rate was established on the SRNIs and ICRR plotted against bank position every 50 steps to predict criticalit Although this procedure assured a slow and cautious approach to first criticality for this core, it did not appear to provide an accurate prediction of critical rod positio The reactivity effect of control rods does not vary uniformly with position as boron worth does with concentration, hence it is difficult to estimate rod worth effects by linear extrapolatio From an engineering viewpoint, this approach to criticality did not appear to be as consistently predictable or as efficient as a commonly used alternate process. Many licensees establish the desired control rod configuration prior to reducing from the refueling boron concentra-tion. Then a dilution rate of 50 gpm is initiated and ICRR monitored on 15 to 30 minute intervals until an ICRR of 0.2 is obtained. At that time, dilution is reduced to 20 to 30 gpm and ICRR renormalize Dilution is stopped when ICRR again reduces to 0.2. Finally, criticality is achieved by mixing or by additional rod withdrawal. The average dilution rate obtained by this licensee during the batch

,Jditions was about 43 gp For the SRNIs, the chi-squared test results during the dilution process were much the same as those obtained earlier. Most were acceptable, but in the higher probability range. At least two failed tests with very low probabilities correlated with the use of hand-held radios in the control roo ;

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Finally, the measured, all-rods-out baron concentration, corrected to design conditions, was 1609 ppmB, which was in acceptable agreement with the predicted value of 1652 ppm Isothermal Temperature Coefficient Measurement Inspector reviewed the test results of the isothermal temperature coefficient (ITC) measurement and discussed them with the test engineers, 'after which he had no further question The moderator temperature coefficient obtained by subtracting the negative doppler temperature coefficient from the ITC was negative, thus satisfying the limiting condition for operation in Technical Specification Control Rod Worth Measurements Control bank B was predicted to have the greatest reactivity worth and its worth was measured during bcron dilution from the all-rods-out configuration to bank B full in using the reactivity computer at low power, less than sensible hea The inspector reviewed the completed procedures for determination of sensible heat and checkout of the reactivity computer and found them acceptable. He independently analyzed the reactivity computer traces used to determine bank B worth and found close agreement with the licensee's values. The results are displayed graphically in the differential worth curves in Attachment The reactivity worths of the remaining control and shutdown banks were determined by rod swaps with bank B. Good agreement with prediction was obtained in all case No violations or deviations were identifie . Reactor Coolant System Leakrate Measurements (61728)

, The microcomputer program RCSLK9 for measurement of reactor coolant system leakage is described in NUREG-1107. This program was written as part of the NRC Independent Measurements Progra To customize a version for use at Surry, data on system and tank volumes were obtained from the FSAR, the plant curve book, and plant drawings. The resultant parameter list is given in Attachment 2. Since the units are identical, only the list for Unit 1 is provided. Data were collected on Unit 1 over the same period, slightly over one hour, that the operator used to perform the daily surveillance of RCS leakage using 1-PT-10. That procedure requires that the starting and ending values of average RCS temperature and pressurizer level be identica Hence, the only variable in determining gross leakage is the volume control tank (VCT) level. The procedure and RCSLK9 results were virtually identical for both gross and unidentified leakage. To convert the change in VCT level, which was read in percent, to change in inventory the operator had to i

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consult a table of constants taped to the typer console rather than a reviewed and approved table within the procedure. At the exit interview, the licensee agreed that the table should be part of the reviewed and approved procedure and made a commitment to incorporate the table in the procedure (Inspector Followup Item 280/281/86-40-01: Revise Procedures 1/2-PT-10 to Incorporate All of the Parameters Required for the Surveil-lance).

The results of the RCSLK9 analysis are given in Attachment 3. Since so few parameters varied, this measurement was not a significant test of the programs capability to handle the less stable situatio That will be deferred to a future inspectio No violations or deviations were identifie . Thermal Power Measurement (61706)

The microcomputer program TPDWR2 for measurement of reactor thermal power is described in NUREG-1167. This program was written as part of the NRC Independent Measurements Program. To customize a version for use at Surry, data on steam generator design features, and pressurizer and steam generator volumes were obtained from the FSAR and plant drawing The resultant parameter list is given in Attachment Since the units are identical, only the list for Unit 1 is provide Using data from trend block 2 from the plant computer as well as data read manually from the main control board, TPDWR2 was used to calculate thermal power at two different periods, 15 minutes apart. Insulation losses were set at the program default value, and pump power was derived from main

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control board indications of motor current, but voltage and an assumed 90%

pump efficiency. The results from TPDWR2, in both cases, were over 30 megawatts higher than the licensee's calculations for the same period From discussions with the licensee, it appears that the most likely source

of the differences is that the basis of the TPDWR2 calculation is feedwater flow, whereas the licensee calculation is based upon steam flow. Since steam flow is obtained from a calibrated flow venturi, as is feedwater flow,

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the two methods appear equally valid. The licensee chose steam flow to avoid the possible effects of feedwater flow venturi foulin The comparable reliability and calibration of the two kinds of flow venturis will be reviewed during a future inspection (Inspector Followup Item 280/281/86-40-02: Review Steam and Feedwater Flow Venturi Calibration and Reliability).

l The results of the TPDWR2 calculations are given in Attachment No violations or deviations were identified.

I ATTA'C'HMENTS':

i Control Rod Bank B - Reference Bank l' Parameter List l Independent Measurements Program l Heat Balance Data l Heat Balance .

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ATIACHMENT 2 )

PARAMETER LIST

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Unit Identification:

Plant Name SURRY Unit Number 1 ,

Docket Number 50-280 Nuclear Steam System Supplier Westinghouse Vessel and Piping:

Volume 8209 cubic feet Pressurizer:

Level Units  %

Temperature Compensated No Calibration Curve Slope 438.56 pounds per %

Upper Level Limit 100 %

Lower level Limit 0%

Relief Relief Tank Volume Control Tank: ,

, Level Units  %

Calibration Curve

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Slope .

117.88 pounds per %

Upper Level Limit 100 %

Lower level limit 0%

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Geometric Method Available No Drain Tank:

Level Units  %

Calibration Curve Slope 63 pounds per %

Upper Level Limit 60 %

Lower level limit 30 %

Geometric Method Available No Relief Tank: '

Level Units  %

Calibration Curve Slope 900 pounds per %

Upper Level Limit 75 %

Lower level limit 25 %

Geometric Method Available No

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ATTACHMENT 3

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NRC INDEPENDENT MEASUREMENTS PROGRAM REACTOR COOLING SYSTEM LEAK RATES -

STATION: SURRY TEST DATE : 12-03-86 UNIT : 1 START TIME: 0112 LOCKET : 50-280 DURATION 1.067 hours7.75463e-4 days <br />0.0186 hours <br />1.107804e-4 weeks <br />2.54935e-5 months <br /> TEST DATA Initial Final System Parameters l

Pressure, psia 223 .7 y T Ave, degrees F 57 ,

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Water Levels

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Pressurizer, % 5 .8 Relief Tank, % 6 .2 Volume Control Tank, % 3 .7 Drain Tank, % 3 .7 Water Charged = 0 gal Water Drained = 0 gal TEST RESULTS Change in Water Inventory in pounds:

Vessel & Piping 0 Relief Tank (1) 0 Pressurizer 0 Drain Tank (1) 630 Volume Control Tank (1) -908 ------

Less: Water Charged 0, Collected Leakage 630 Plus: Water Drained 0

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Cooling System -908 Leak Rates in gpm (3):

Gross 1.70 Identified 1.18 Unidentified 0.52

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(1) Determined from tank calibration curv (2) Determined from tank dimension '

(3) The density used for converting inventory change to leak rate was 62.31 pounds / cubic foot based on standard conditions.

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ATTACHMENT 4 HEAT BALANCE DATA * *

SURRY 1

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12/02/86 '

PLANT PARAMETER REACTOR COOLANT SYSTEM P.EFLECTIVE INSULATION  :

Pump Power (MW each) Inside Surface Area (sq ft) 11.920 ,

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Pump Efficiency (%) 9 Heat Loss Coefficient (BTUs/hr sq ft) 55.00

\ Pressurizer Inside Diameter (inches)

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8 NONREFLECTIVE INSULATION %y STEAM GENERATORS Dome Inside Diameter (inches) 165.28 Inside Surface Area (eq ft)

Thickness (inches)

8,414 k Riser Outside Diameter (inches) 20.00 Thermal Conductivity (BTUs/hr ft-F) $

0.035 4 Number of Risers 16 F-Moisture Carry-over (%) in A 0.250 LICENSED THERMAL POWER (MWt) 2441- If Moisture Carry-over (%) in B 0.250 y Moisture Carry-over (%) in C 0.250 g

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DATA: SET 1 SE SET 1 SET 2 TIME 1600 1615 TIME 1600 1615-STEAM GENERATOR A STEAM GENERATOR B Steam Pressure (psia) 82 .3 Steam Prossure (psia) 82 .9 Feedwater Flow (E6 lb/hr) 3.663 3.663 Feedwater Flow (E6 lb/hr) 3.568 3.568 Feedester Temperature (F) 43 .1 Feedwater Temperature (F) 43 .7 ,

Surface Blowdown (gpm) .0 Surface Blowdown (gpm) .0 .

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Bottom Blcadown (gpm) 7 .0 Bottom Blowdown (gpm) 7 .0 Water Level (inches) 6 .4 Water Level (inches) 6 .4

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STEAM GENERATOR C  !

Steam Pressure (psia) 82 .1

T Feedwater Flow (E6 lb/hr) 3.632 3.632 Feedwater Temperatur- (F) 43 .6 Surface Blowdown (gpm) .0 Bottom Blowdown (gpm) 7 .0 Water Level (inches) 6 .2 '

LETDOWN LINE CHARGING LINE k e

Flow (gpm) 10 .0 Flow (gpm) 9 .0 ' '

Temperature (F) '54 .0 Temperature (F) 45 .0 "

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p PRESSURIZER REACTOR Pressure (psla) 2239.7 223 T ave (F) 57 .0 Water Level (inches) 21 .4 T cold (F) 54 .0 t

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HEAT BALANCE SURRY 1 -

12/02/86 DATA SET 2 OF 2 ENTHALPY FLOW POWER POWER 1615 hours0.0187 days <br />0.449 hours <br />0.00267 weeks <br />6.145075e-4 months <br /> (BTUs/lb) (E6 lb/hr) (E9 BTUs/hr) (MWt)

STEAM GENERATOR A Steam 119 .632 4.347 Feedwater 40 .663 -1.496 Surface Blowdowa 51 .00000 0.00000 Bottom Blowdown 45 .02826 0.01299

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Power Dissipated 2.8643 83 STEAM GENERATOR B s

Steam . 119 .540 4.237 .

Feodwater -

41 .568 -1.467 Surface Blowdown 51 .00000 0.00000 Bottom Blowdown g 46 .02821 0.01303

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Power Dissipated 2.7821 81 STEAM GENERATOR C Steam 119 .601 4.311 Feedwater 41 .632 -1.489 Surface Blowdown 51 .00000 0.00000 Bottom Blowdown 46 .02824 0.01301

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Power Dissipated 2.8343 83 OTHER COMPONENTS Letdown Line 53 .03961 0.02119 Charging Line 43 .04103 -0.01770 Pressurizer 70 .00000 0.00000 Pumps -0.02673

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Insulation Losses 0.00107

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Power Dissipated -0.02217 - _____

REACTOR POWER 2477.3 l 1

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ATTACHMENT 5

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HEAT BALANCE '

SURRY 1 12/02/86 DATA SET 1 OF 2 ENTHALPY FLOW POWER POWER 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br /> (BTUs/lb) (E6 lb/hr) (E9 BTUs/hr) (MWt)

STEAM GENERATOR A Steam 119 .632 4.348 Feedwater 40 .663 -1.496 Surface Blowdown 51 .00000 0.00000 Bottom Blowdown 45 .02827 0.01299

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Power Dissipated 2.8641 83 STEAM GENERATOR B ,

Steam -

119 .540 4.237 Feedwater 41 .568 -1.467 Surface Blowdown 51 .00000 0.00000 Bottom Blowdown % 46 .02822 0.01302

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Power Dissipated 2.7824 81 STEAM GENERATOR C Steam 119 .601 4.311 Feedwater 41 .632 -1.490 Surface Blowdown 51 .00000 0.00000 Bottom Blowdown 46 .02825 0.01300

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Power Dissipated 2.8341 83 OTHER COMDONENTS Letdown Line 53 .03961 0.02119 Charging Line 43 .04103 -0.01770 Pressurizer 70 .00000 0.00000 Pumps -0.02673 Insulation Losses 0.00107

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Power Dissipated -0.02217 - ______

REACTOR POWER 247 <.

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