IR 05000338/1992025

Insp Repts 50-338/92-25 & 50-339/92-25 on 921102-06.No Violations Noted.Weaknesses Noted Re Heavy Loads Program. Major Areas inspected:post-refueling Startup Tests,Routine Surveillance of Core Performance & Control of Heavy Loads

ML20198E182
Person / Time
Site:	North Anna
Issue date:	11/25/1992
From:	Burnett P, Crlenjak R, Curtis Rapp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20198E175	List: IR 05000338/1992025 ML20198E171
References
50-338-92-25, 50-339-92-25, NUDOCS 9212040260
Download: ML20198E182 (9)

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• l g# ' UNITED STATES

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_ fg NUCLEAR REGULATORY COMMISSION

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• g . REoloN 11

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I Report Nos.: 50-338/92-25 and 50-339/92 25

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Licensee: Virginia Electric and Power' Company j Glen Allen, VA 23060 Docket Nos.: 50-338 and 50-339 License Nos.: NPF-4 and NPF-7 l Facility Name: North Anna-1 and 2 Inspection Conducted: November 2 - 6, 1992

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Inspector: -[

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Inspector: bCurtis .Tapp

///ir/9A Da't6 S'igned

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Approved by: -

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c lhichard V. Crlenjak,-- Chief Date Signed i

NOLDperational Programs Section j Operations Branch

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Division of Reactor Safety

SUMMARY Scope:

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This routine, unannounced inspection was conducted in the areas'of -

post-refueling startup tests, routine surveillance of core performance, and

[ control'of heavy loads in containment.

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I - Results:

- The inspectors found;there was adequate interface between the. corporate design , . ,

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testing program was satisfactory; however enhancements in controlling the-

. approach to initial criticality-and controlling VCT- boron concentration we're -

identified. Routine core performance surveillance.were performed within the'

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- time required by. Technical Specifications. The licensee had taken the

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s - necessary actions'when Technical Specification li_mits were-exceeded.

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- The-inspectors'found the heavy loads program had weaknesses:as: identified =by the licensee's'QA organization.-'In response to these weaknesses, a crane-inspection firm, certified by the Department of Labor, was-hired to perform

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-routine crane-. inspections. However, the inspectors 'could not determine' the-l' adequacy of.tnese vendor. inspectors. The inspectors reviewed several

9212040260 921127

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l procedures associated with the heavy load program and concluded they lacked i specificit Additionally, during the steam generator replacement project, the Unit 1 polar crane will be used to lift loads greater than the rated load,_

but tile required overload lift inspections have not. been scheduled. Followup inspections by Region 11 are planned (paragraph 5.b).

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No violations or deviations were identified.

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l REPORT DETAILS L

1, - Persons Contacted -

Licensee Employees- l

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R. Berryman, Manager of Nuclear Analysis and Fuel 4

D. Dziadosz, Supervisor of Core Design ,

l *R. Enfinger, Acting Station Manager

*Mi Gettler - Manager, Steam Generator Replacement Project

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! *D. Heacock, Superintendent, Engineering

*E. Hendrixson, Supervisor, System Engineering

- B. Hutsell, QA Inspector

*P. Kemp, Supervisor, Licensing

• J.:Leberstein,. Licensing Engineer A. Main, Reactor Engineer

• R.-McAndrew, lead Reactor Engineer

• A. Parker,- Supervisor, Maintenance t...gineering

! *P. Quarles, Supervisor, Quality Assurance

- *R. Sturgill, Supervisor, System Engineering ,

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L Other. licensee employees contacted during.this inspection included

engineers, technicians,-and administrative personnel.

4-l NRC Resident Inspectors i * S. Lesser, Senior Resident inspector j *D. R. Taylor. Resident Inspector '

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• Attended exit interview on November 6, 199 Acronyms and initialisms are defined in _the-last. paragraph

• ReviewofCorporateNuclearEngineeringActivities:(72700,I61702,;_61707)

The inspectors reviewed the-interface between.the corporate and site

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reactor engineering groups ~. : The: site reactor engineers were managed by 4 .the corporate Nuclear Analysis and Fuels Department, but were matr_ixed to -

.on-site management. TheEcorporate organization was responsible for_-

. generating all core reload-design data, analyzing physics _ testing data,- *

and determining-acceptability.of the core design. During startup physics testing for a new operating cycle,' corporate personnel were .on-site to -

p analyze the data obtained and' determine-test acceptability. . This

, _ provided for impr.oved responsivenessLif reanalysis was required lin the '

event an_ acceptance criterion was not met. Routine core-performance.

4: surveillance tests, . including data analysis,:were preformed _by the on-site reactor engineers. -The'testsiandLre'sults were-then sent to the

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corporate organization for further. review-and' archivin Since Unit I was in a 145 day coastdown, the inspectors reviewed th current Unit 1 core cycle design data to determine -if'the MTC would

_ exceed TS limits during

the coastdown. In the~ cycle'5 design calcula-

! tions, the licensee had included a 30 EFPD coastdown margin, which was- _

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exceed TS limits during the coastdown. In the cycle 5 design calcula-tions, the licensee had included a 30 EFPD coastdown margin, which was sufficient to account for the addition fuel exposure during the current 4 coastdown. The MTC was projected to be less negative than the limits in the COLR during the coastdown perio This MTC-limit ensured adequate

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SDM throughout the entire core cycl . Review of Unit 2 Cycle 9 Startup Tests (7270e 61708, 61710) Rod Drop Time Measurement The inspectors reviewed 2-PT-17.2 (Revision 5), Rod Drop Time Measurement, to determine adequacy of test methodology and data collection. The licensee used a high speed stripchart recorder to monitor control rod drop and then calculated the drop time based a

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trace of rod-drop speed versus time. The inspectors verified the data were taken correctly and the rod drop times were accurate. The ,

rod drop times, between a high of 1.83 seconds and a low of 1.44 seconds, were all within the TS limit of 2.7 second Rod drop times were taken while continuing with plant heatup. T,, was 501*F when rod drop testing began and was 546.6*F when rod drop testing was completed. TS 3.1.3.4 required rod drop timing be preformed greater than 500*F, thus performing those activities in

withT,lwasacceptabl paralle

, Initial Criticality after Refueling The inspectors reviewed 2 0P-1.5 (Revision 31), Unit Startup from Mode 3 to Mode 2, for appropriate controls during post-refueling r' actor startup. The licensee plotted ICRR as a function of bank

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positio .ie inspectors commented that plotting ICRR against reactivity inserted by the control rods, using predicted rod worth values, would yield a more linear plot and improved extrapolations

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to criticality. A y' test of the SRNIs was performed to demonstrate acceptable detector response. A x2 test was not performed on the data obtained for the ICRR point plots. A predicted critical posi-tion of 145 steps on control rod bank D was used to determine an estimated Cs of 1994 ppm Since RCS concentration was 1966 ppm 8, -

the predicted critical position was adjusted to 122 steps. Criti-cality was achieved at 79 steps, which was within the -400 pcm lower limit of 61 steps.

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The inspectors noted initial ICRR data were not taken until control rod bank A was fully withdrawn. Given that criticality was achieved low in the allowable band, beginning ICRR monitoring prior to fully withdrawin j control rod cank A would have provided additional control'and criticality prediction capability, which is particularly desirable during initial startup for an operating cycle. The

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licensee's basis for delaying ICRR monitoring was the change from mode 3 to mode 2 (k.o 2: 0.99) occurred atter control rod bank A was fully withdrawn. The licensee is considering revising procedures to

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- i require initiating ICRR monitoring before withdrawing control rod banks for the first criticality of each operating cycle.

. Low Power Physics Tests The inspectors reviewed the tests performed under 2-PT-9 (Revision 7), Refueling Nuclear Design Check Tests. These tests included ARO Boron Endpoint, AR0 ITC Determination, Most Reactive Bank Worth Determination, Boron Worth Coefficient, ITC Determination with Reference Bank Inserted, and Rod Worth Measurement Using Rod Swap. The design data agreed very closely with the data obtained

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The inspectors did note that VCT boron concentration was not moni-tored as a parameter for RCS equilibrium. The inspectors discussed

. this with the licensee and determined that monitoring of pressurizer boron concentration was acceptable if no boration or dilution of the VCT occurred. When a dilution of the RCS was necessary, the dilution was directed to the suction of the charging pumps. The operating procedure for dilution directs the alternate dilution flowpath to the VCT be isolated by closing valve 1(2)l148. However, the operating procedure is not referenced in PT-94.0. Failure to control VCT boron concentration, can result in a reactiv;ty overshoot during mixing, and has been identified as a precursor in a severe core damage, but low probability, acciden (1) ARO Boron Endpoint The ARO baron endpoint was determined to be 2096 ppm This value did not meet the design :riterion 2053 1 32 ppm B. The inspectors reviewed the licensee's evaluation of the higher boron concentration and did not identify any safety concern The 2096 ppm B value was within the acceptance criterion of the difference between the actual and predicted values being less

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than or equal to 1000 pc (2) ARO ITC Determination A 3.l'F cooldown and a 3.2*F heatup were used to determine the ITC. The cooldown ITC was determined to be -1.77 pcm/*F and the heatup ITC was determined to be -1.72 pcm/*F. These were averaged for an uncorrected ITC of -1.75 pcm/'F. The design ITC corrected to actual conditions was -1.08 pcm/*F. This

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value was within the design criterion of -1.4513.0 pcm/*F and the acceptance criterion of less than 3.73 pcm/* A temperature change of at least 13*F, while acceptable by procedure, was not consistent with the 14*F temperature change recommended by the Westinghouse general test description. .Using at least a 14*F change would reduce endpoint uncertainties associated with this-tes p p-. -- g +*p g

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(3) Most Reactive Bank Worth Determination Control rod bank B was predicted to have the highest reactivity worth of 1288 pcm. The actual reactivity worth of control rod bank B was determined by starting a dilution and then incrementally inserting control rod bank B to aeriodically offset the positive reactivity addition from tie dilutio Changes in reactivity weie obtained from analysis of the reac-tivity computer recorder trace. The measured reactivity worth of control rod bank B was 1227 5 pcm. This was within the acceptance criterion of 110% of predicted wort (4) Reference Bank Inserted Boron Endpoint  !

With control rod bank B inserted, the boron endpoint was determined to be 1917 ppm B. This was within the acceptance criterion of 1907 1 29 ppm B. An ITC determination with the reference bank inserted was not made because the ARO ITC was within T5 3.1.1.4 limit g5) Boron Worth coefficient Determination The boron worth coefficient was determined by subtracting the two boron endpoints and dividing by control rod bank B integral reactivity worth. This value, calculated to be 6.86 pcm/ ppm 8, met the acceptance criterion of 6.81 1 0.68 pcm/ ppm (6) Rod Worth Measurement Using Rod Swsp The remaining control and shutdown rod bank reactivity worthies were determined by withdrawing control rod bank B until a +40 pcm reactivity addition was observed on the reactivity computer. The control rod bank under test was then inserted ui.U l a 40 pcm reactivity addition was observed. This was concinued until either the rod bank under test was fully inserted or control rori bank B was fully withdrawn. This same methodology was repeated for all control and shutdowr rad banks, All rod bank reactivity worthies were within the acceptance criterion of 115% of predicted. The tc1si cettrol rod worth, determined by summing the individual rou bnk worthies, was 5326.3 pcm and the predicted value was Sn pcm. The measured total worth was within the acceptan a criterion of 110% of predicte Tests During Power Escalation Review-of PRN1 Incore/Excore Cross Calibration The inspectors reviewed 2 PT-22.4 (Revision 0), Single Point Power Range Excore Detector Calibration. Based on an analysis dated September. 19, 1991, the licensee had converted from a multipoint correlation to a single point correlation to i

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determine the PRNI voltage settings. The inspectors reviewed this analysis and found it was technically accurate and <

complete. full core flux maps were taken at 30%, 72.85%, and 99.96% power. An additional flux map was taken at full power

, to demonstrate the 99.96% power calibration was correct. The -

i inspectors did not note any discrepancies in these test TS 3.3.1 requires a recalibration if the monthly surveillance

for incore to excore axial offset differs by 2:3%. The licensee

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also monitors detector gain setting and QPTR to determine if any additional recalibrations are necessar . L' nit 1 Core Performance Surveillance Activities (61702, 61708)

i Review of Unit I flux Maps

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The inspectors reviewed the flux maps for the current Unit I core <

cycle. All maps were completed within the required TS surveillance frequenc Because of the increased plugging on Unit 1 steam generators, a new Fa(z) limit was established. Although Unit I could not operate at rated power, the new limit did allow Unit I to

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ops tte at 95% power. Following the Unit I tube plugging outage, map N1-09 20, taken on 3/16/92, found that fa(z) was outside the new

limi The licensee comslied with the necessary actions required by TS 3.2.2. Fo(z) was within the limit on map N1 09 25 taken on j 5/11/92. The inspectors found the analysis of the flux maps was

accurate and the licensee had complied with TS requirements, Unit 1 E0C MTC Measurement TS 3.1.1.4 requires the MTC be determined whenever boron concentra-tion decreases to 300 ppm B or less. The licensee used a heatup of 5.l'f and a cooldown of 4.57'F to determine the MT The MTC

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measured during the heatup was -35.27 pcm/'F and was -36.00 pcm/*F for the cooldown. These values were averaged for an MTC of -35.64 pcm/' This value was within the limits of the Unit 1 COL i Control of Heavy Loads (61701, 37701)

' Heavy Loads Program at North Anna

, Generic Letter 81-07 and NUREG 0612, Control of Heavy Loads at Nuclear Power Plants, were issued on February 3, 1981. VEPC0 made a i timely response with an acceptable description of a program to j

. control heavy loads, in recent years, QA audits continued to show i insufficient training of crane operators and inspectors and lack of )

management knowledge of the restraints imposed by the standards to ;

which they are committed. Management's response to this continuum !

of adverse QA audits was to use a DOL certified crane inspection j vendor to perform inspections for_ compliance to OSHA regulation The OSHA regulations were based upon the B30.x-1967 standards;

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NUREG 0612 required a commitment to the 1973 standards. The differ-

ences between the two standards are small. Review of one polar crane inspection performed by the vendor revealed that the report was only a checklist. The vendor inspector would have to be directly observed to independently determine the adequacy of the

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I Discussions with the QA ins)ector assigned to cranes and rigging revealed that he believes t ut recent actions to improve procedures

and training have responded adequately to his findings. Based upon review of the procedures listed below, the NRC inspector concluded *

, there was a lack of specificity in the crane and ringing inspection l procedure However, that lack might be overcome by well-trained inspectors, l

o MD ADM 9.1, Control of Heavy Loads in Reactor Contsinment

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o VPAP-0809, NUREG 0612, Heavy Load Program

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o 0 MPM 1302-01, frequent and Periodic Inspections of Bridge and

Gantry Cranes

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o 0-MPM 1304.1, inspection and Repair of Reactor Vessel Component

and Reactor Coolant Pump Lifting Devices -

o 0 MPM-1301 01, Polar Crane Inspection

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o MMP-P-MH 14, inspection and Testing of Material Handling Equipment, Hoist, and Rigging o WP G12, Site Inspection Program for Equipment 4 Polar Crane Use During the Steam Generator Replacement Program During SGRP, the Unit 1 polar crane will not be operated under the

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restrictions of the NUREG-0612 program once all fuel is removed from containment and Unit 1 is isolated from any systems shared with Unit

-2. The maximum load to be lifted during SGRP is calculated to be

262.5 tons. The nameplate rating of the bridge is 250 tons. The licensee has conformed to B30.2-1990 to justify the overload lif The licensee understands that inspections are required before and after each overload lift, but has not scheduled the inspections or selected the contractor, inspections of the Unit 1 polar crane following overload lifts will be further inspectad during the SGRP and is identified as IFl 50 338/92-25-01, Exit Interview

The inspection scope and rescits were summarized on Novembtr 6, 1992, i with those-persons indicated in paragraph 1. The inspectors described

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the areas inspected and discussed in detail the inspection.results listed below. Although reviewed during this inspection, proprietary information

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i i is not contained in this report, No dissenting comments were received

from the licensee.

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ltem Number _ Status Etscriotion(Paragraohl i

50 388/92-25 01 OPEN IF1-Inspections of the Unit 1 polar crane

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following overload lifts (5.b)

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ARO All Rods Out COLR Core Operating Limits Report DOL Department of Labor

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EFPD Effective Full Power Days ICRR Inverse Count Rate Ratio

IFl Inspector followup Item

ITC isothermal Temperature coefficient i k .,, k effective MTC Moderator Temperature coefficient OSHA Occupational Safety and Health Administration i pcm percent millirho i ppm B parts per million Boron

! PRN1 Power Range Nuclear Instruments J QA Quality Assurance Quadrant Power Tilt Ratio

QPTR RCS Reactor Coolant System SDM Shutdown Margin SGRP Steam Generator Replacement Program SRNI Source Range Nuclear Instruments i

T,y Average RCS temperature TS Technical Specifications VCT Volume Control Tank i

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