IR 05000413/1986035

Insp Repts 50-413/86-35 & 50-414/86-38 on 860908-12.No Violations or Deviations Noted.Major Areas Inspected:Review of Completed Startup & Surveillance Tests

ML20214X243
Person / Time
Site:	Catawba
Issue date:	09/22/1986
From:	Burnett P, Jape F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20214X200	List: IR 05000413/1986035 ML20214X197
References
50-413-86-35, 50-414-86-38, NUDOCS 8610030370
Download: ML20214X243 (7)

Text

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! gn Rfou . UNITED STATES l , #'o, NUCLEAR REGULATORY COMMISSION

.7 ' ' REGION 11 l

101 MARIETTA STREET, N.W.

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• *- ATLANTA, GEORGI A 30323 k.v / ...

Report Nos.: 50-413/86-35 and 50-414/86-38 Licensee: Duke Power Company 422 South Church Street Charlotte, NC 28242 Docket Nos.: 50-413 a1d 50-414 License Nos.: NPF-35 and NPF-52 Facility Name: Catawba 1 and 2 Inspection Conducted: Se tember 8-12, 1986 Inspector:

P.~T. Burnett-

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[8-MDate Signed Approved by: // - I F. Jape, Section Chief (/ (/ Date Signed Engineering Branch Division of Reactor Safety SUMMARY Scope: This routine, unannounced inspection addressed the review of completed startup tests (Unit 2) and completed surveillance tests (Units 1 and 2).

Results: No violations or deviations were identifie PDR ADOCK 05000413 G PDR

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REPORT DETAILS Persons Contacted Licensee Employees

• J. W. Hampton, Station Manager

• B. F. Caldwelli, Superintendentt of Station Services

• C. L. Hartzell, Compliance Engineer

• S. W. Brown, Reactor Engineer J. A. Kammer, Test Engineer - Unit 2 M. L. Arey, Engineer - General Offices R. G. Blessing, Engineer - Reactor M. A. Geckle, Engineer - Test M. W. Hawes, Engineer - Reactor P. G. LeRoy, Licensing Engineer

• F. P. Schiffley, Licensing Engineer D. Wellbaum, Engineer - Reactor Other licensee employees contacted included engineers, technicians, and office personne NRC Resident Inspectors

• P. K. Van Doorn, Senior Resident Inspector

• S. Lesser, Resident Inspector

• Attended exit interview Exit Interview The inspection scope and findings were summarized on September 12, 1986, with those persons indicated in paragraph 1 above. The inspector described the areas inspected and discussed in detail the inspection findings. No dissenting comments were received from the licensee. The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspector during this inspectio Unresolved Item (UNR) 414/86-38-01: Determine the need and establish the ,

guidance for reverifying and documenting system status and test prerequi- l sites for interrupted tests - paragraph . Licensee Action on Previous Enforcement Matters l This subject was not addressed in the inspectio )

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l 4. Unresolved Items An unresolved item is a matter about which more information is required to determine whether it is acceptable or may involve a violatio One new

unresolved item is identified in paragraph 5.a.

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5. Review of Unit 2 Power Escalation Tests (72600, 72608, 72616, 72582, 72583, 72566,72580)

4 The following test procedures (TP) and surveillance periodic tests (PT)

j completed at or below the 30 % rated thermal power (RTP) plateau were

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(1) TP/2/A/2650/12, Station Blackout, was performed on May 25, 198 Thirteen test discrepancies were identified by the licensee. All resolutions appeared reasonable and to not invalidate the tes (2) TP/2/A/2650/03, Loss of Control Room, was performed successfully on July 11, 1986. It was witnessed by a team of inspectors. The review of the comoleted procedure revealed only one test discrepancy was identified by the licensee: The volume control tank (VCT) level gauge on auxiliary shutdown panel 2 (ASP 2) did not indicate the correct level. The licensee's judgement that the test was not invalidated is acceptabl The test was first attempted, and then aborted, on June 27, 198 No entry for test procedure Section 7.0, Required Station Status, or Section 8.0, Prerequisite System Conditions, is more recent than that early dat Other than a signoff for step 12.1 that

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Sections 7 and 8 were complete, there was no indication of l reverification in the procedure package for Sections 7 and The

test engineer stated that he had, in fact, reverified Sections 7 and 8, but it was not the practice of the test group to document

reverification on a step-by-step basis. Neither the licensee nor the inspector could identify any requirement in the administrative

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procedures, standards, or regulations that would require reverification of station status or prerequisite system conditions j following an interruption in test activitie Sections 9 to 13 were were performed again in the period July 11 to July 14, 1986, as appropriat The issue of the absence of guidance on when and how to document

, reverification of procedure steps When a test is interrupted has been identified as an unresolved item (UNR 414/86-38-01:

Determine the need and establish the guidance for reverifying and documenting system status and test prerequisites for interrupted tests).

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(3) TP/2/A/2650/15, Steam Generator Water Hammer Test, was performed on June 25-26, 1986. No test discrepancies were identified. A visual inspection, walkdown, of the system following the test was performed in accordance with MP/0/A/7650/9 No system discrepancies were identifie (4) TP/2/A/2600/11, Pressurizer Pressure and Level Control System Test, was performed on May 26, July 12, and July 13, 1986. The system performed as expecte (5) TP/2/A/2600/10, Rod Control System at Power Test, was performed on July 13-14, 1986. System operation was as expected.

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(6) TP/2/A/2650/05, Unit Load Transient test, was performed at 30% RTP on July 14-15, 1986. System stability was within acceptable limit (7) PT/2/A/4150/05, Core Power Distribution, was performed on May 25-27, 1986 at 19.7% RTP and on June 26-27, 1986 at 28.6% RT The following Technical Specifications were satisified in both cases:

(a) The heat flux hot channel factor - 3.2.2, (b) Nuclear enthalpy rise hot channel factor - 3.2.3, and (c) The quadrant power tilt ratio - 3. (8) PT/2/A/4150/03A, NSSS Thermal Outputs, was performed on May 26, 1986 at 19.5% RTP and on July 10-12, 1986 at 28% RTP. Only the results of the primary side power measurements were used to set trip set points for the next test platea b. The following TPs and pts completed at the 50% RTP plateau were reviewed:

(1) TP/2/A/2650/05, Unit Load Transient test, was performed on July 21-22, 1986. System stability was within acceptable limit (2) PT/2/A/4150/05, Core Power Distribution, was performed on July 17-18,1986 and July 20-21, 1986. The following Technical Specifications were satisified in both cases:

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(a) The axial flux difference - 3. (b) The heat flux hot channel factor - 3.2.2,  !

(c) Nuclear enthalpy rise hot channel factor - 3.2.3, and I (d) The quadrant power tilt ratio - 3. (3) PT/2/A/4150/03A, NSSS Thermal Outputs, was performed on l July 17-20, 1986. Discrepancy No. 2 indicated that there was an adverse effect on the thermal outputs program results when tempering flow was isolate The best estimate thermal power was 48.3% RT l l

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(4) PT/2/A/4600/05C, Preliminary Incore and Nuclear Instrument System Correlation, was performed on July 20-21, 1986. Three full core flux maps were used. The incore axial offsets ranged -5.15 to-7.95% RTP, and the excore values ranged from -12.9 to -14.9 on N44 and from -8.6 to -10.4 on N4 .

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It was noted that the computer code RPECALIB used to analyze the results had been improved to print out both the correlation coefficient for each least squares fit and the point-to-point differences between the observations and the least-squares-

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calculated value The following TPs and pts completed at the 75% RTP plateau were reviewed:

(1) TP/2/A/2650/07, Turbine Trip, was performed on July 23, 1986, at 68% RT The acceptance criteria were satisified in that no safety injection was initiated and no safety or power operated relief valves on either the pressurizer or the steam generators

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were actuate (2) TP/2/A/2600/14, Nuclear Instruments Initial Calibration, was performed on August 14, 1986, with acceptable result (3) TP/2/A/2650/05, Unit Load Transient Test, was performed on August 10, 1986. The plant stability requirements were satisfie (4) PT/2/A/4150/05, Core Power Distribution, was performed on July 29, 198 The following Technical Specifications were satisified:

3.2.1, 3.2.2, 3.2.3, and 3. (5) PT/2/A/4150/03A, NSSS Thermal Outputs, was performed on July 29, 198 Procedure change No. 4 was implemented to delete the requirement to isolate tempering flow. The best estimate thermal power was 73.9% RT (6) PT/2/A/4150/23, Quarter Core Flux Map Qualification, which is a prerequisite for PT/2/A/4600/05, was performed sucessfully on July 23, 198 (7) PT/2/A/4600/05A, Incore and Nuclear Instruments Correlation Check, was performed during July 30 to August 10, 1986. The measured incore axial offsets ranged from 9.15% to - 19.13%. The correlation coefficients for all four power range nuclear instruments were greater than 0.9 (8) PT/2/A/4150/13B, Calorimetric Reactor Coolant Flow Measurement, was performed on July 29, 1986. The measured flow was nearly 3%

in excess of the minimum required by Technical Specification i

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Review of TP/2/A/2100/01, Controlling Procedure for Power Escalation, which is in progress, confirmed that proper approvals had been obtained for each power plateau escalation. Since July 11, 1986, the Pre-Transient Test Checklist, Enclosure 13.4, had been completed for each such test performe No violations or deviations ware identifie . Review of Completed Surveillance Procedures - Unit 1(61702,61705,61707)

The following completed, surveillance periodic test procedures were reviewed to assure that the frequency of surveillance was correct and the limiting condition for operation (LCO) was satisfied: PT/1/A/4150/05, Core Power Distribution, was performed with a frequency of once per 30 effective full power days (EFPD) during the period January 13 to August 18, 198 In all cases, the following LCOs were satisfied:

(1) The axial flux difference - 3. (2) The heat flux hot channel factor - 3.2.2, (3) Nuclear enthalpy rise hot channel factor - 3.2.3, and (4) The quadrant power tilt ratio - 3. PT/1/A/4600/05A, Incore and Nuclear Instrument System Correlation l Check, was performed quarterly in mode 1 during 1986. During the !

intervening months, PT/1/A/4600/05B, Incore and Nuclear Instrument l System Calibration Check, was performed. These tests satisfy the '

l surveillance requirements of Technical Specification Table 4.3-1, item 2.a.

' PT/1/A/4150/04, Reactivity Anomaly Calculation, which meets the surveillance requirement of Technical Specification 4.1.1.1.2, was performed with the required 31 EFPD frequency throughout cycle 1. In all cases the calculated anomaly was less than the 1% delta-k/k limi No violations or deviations were identifie . Thermal Power Evaluation (Units 1 and 2) (61706)

Unit-specific data on steam generator and pressurizer dimensions, pump power and efficiency, insulation losses, and moisture carryover were obtained and entered into files for use by the microcomputer program TPDWR2 (Thermal Power Determination in Westinghouse Reactors). TPDWR2, which was developed l

by the NRC Independent Measurements Program, is described in detail in NUREG-1167 Version 2.

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A test case using typical plant data gave reasonable results. Since neither

unit was operating during this inspection, a direct comparison between

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between the licensee's calculation and that by TPDWR2 was not possible. The licensee's summary data sheets from their computer calculation of thermal power do not present all of the data required for input to TPDWR2, hence no retrospective comparison could be mad A copy of the program and data files was left with the resident inspectors, and one resident was given some training in the use of TPDWR . Changes in The Unit 2 Startup Test Program (72576, 72584)

By letters to the Director of Nuclear Reactor Regulation dated May 28, 1986 and August 4,1986 the licensee has identified the following changes to the startup test program as described in Chapter 14 of the FSAR:

l The all-rods-out moderator temperature coefficient may be positive, but

! rod withdrawal limits will be established during operation to maintain a negative valu This is consistent with Technical Specification 3.1.1.3 action The test method for the approach to initial criticality was changed to be more conservativ The performance of the doppler only power coefficient verification test will be limited to Unit 1. This is consistent with changes authorized at other facilities when the second unit was nearly identical to the firs The performance of the natural circulation verification test will be limited to Unit 1. This is consistent with changes authorized at other facilities when the second unit was nearly identical to the firs The performance of the pseodo rod ejection tests at zero and 30% RTP will be limited to Unit This is consistent with changes authorized at other facilities when the second unit was nearly identical to the firs The inspector had no questions on the licensee's descriptions of the test changes or the justifications for the . Followup On Inspector-Identified Items (92701)

j (Closed) Inspector identified item 414/86-13-01: Assure the rod swap l measurement of control rod worth is approved prior to use. This change in I test method was identified and described in FSAR revision 14, which was l

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submitted by the licensee to NRR on January 31, 1986. No negative response

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has been received from NR *

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