IR 05000443/1986048

Insp Rept 50-443/86-48 on 860929-1003.No Violations Noted. Major Areas Inspected:Startup Program,Startup Test Procedures Review & Review of Rept Re Containment Structural Integrity Test

ML20215N730
Person / Time
Site:	Seabrook
Issue date:	10/21/1986
From:	Briggs L, Varela A, Wen P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20215N729	List: IR 05000443/1986048 ML20215N728
References
50-443-86-48, NUDOCS 8611070206
Download: ML20215N730 (8)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /86-48 Docket No.-50-443 License No. CPPR-135 Licensee: Public Service of.New Hampshire P.O. Box 330 Manchester, New Hampshire 03105 Facility Name: Seabrook Unit 1 Inspection At: Seabrook New Hampshire Inspection Conducted: September 29 - October 3, 1986

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Inspectors: [ g P. C. Wen,de6ctor Engineer

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%dw L gA.VareldleadpactorEngineer ab./u-date Approved by: [ s L. E. Brigg6 4 hief, Test Programs

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.date Section, OB, DRS Inspection Summary: Inspection on September 29 - October 3, 1986(Inspection Report No.50-443/86-48)

Areas Inspected: Startup program and startup test procedures revie Review of Report on Containment Structural Integrity. Tes Results: No violations were identifie NOTE: For acronyms not identified,- refer to NUREG-0544, " Handbook of Acronyms and Initialisms".

8611070206 861021 3 PDR ADOCK 0500

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DETAILS 1.0 Persons Contacted 1.1 Public Service of New Hampshire

• S. P. Buchwald, QA Superviso *D. L. Covill, Site Construction QA Manager

• R. Ferrell, Licensing Coordinator

• R. W. Gregory, Licensing Engineer

• G. A. Kann, Test Group Manager-STD

• D. G. McLain, Startup Manager-STD

• W. T. Middleton, QA Staff Engineer

• L. Ran, Reliability & Nuclear Engineering Supervisor

• R. Sanchez, Site Licensing Supervisor

• W. R. Sullivan, QA Engineer W. Temple, QC Inspector C. Vincent, QC Supe! visor ,

• J. Warnock, Nuclear' Quality Manager U.S. Nuclear Regulatory Commission A. C. Cerne, Senior Resident Inspector The inspector also contacted other licensee personnel during the inspectio * Denotes those present at the exit meeting on October 3, 1986.

2.0 startup Test Program 2.1 Test Program Review The licensee's Phases 4-6 startup tests cover the initial fuel load-ing through the plant acceptance test. The administration of the startup program, procedural controls, test personnel responsibili-ties, and interrelationships of the various organizations were des-cribed in the Startup Program Administration, 1-ST-1. This adminis-trative program provides guidance to assure that startup tests will be conducted in a controlled manner. Attachment 9.2 to procedure 1-ST-1 provides a checklist / review form for licensee management sign-off at each major hold point prior to commencing the next power plateau tes The requirements of the test program were found consistent with FSAR Chapter 14 & Regulatory Guide 1.68 commitments in the previous NRC inspection 50-443/86-31. Since then, all startup test procedures

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have been com;1eted and have received SORC's approval. During this inspection period, the inspector noted that two tests: ST-31 (Static Rod Drop Evaluation) and ST-32 (Negative Rate Trip Test)

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have been deleted from the test sequence. The deletion of these two tests was submitted to NRR through FSAR Amendment 58, dated April 198 Based on the review of test program and test procedure as described in the following section (2.2), and discussion with licensee repre-sentatives, the inspector determined that the licensee's startup test program continues to be effective and acceptabl .2 Test Procedure Review The inspector reviewed the 15 test procedures listed in Attachment A for the attributes identified in inspection report 50-443/86~31 -

Section 2.2. The inspector identified the following items:

(1) 1-ST-13: The current procedure (Rev.2) provides instruction to verify the nuclear instrumentation (NI) readings be within 2%

of calculated core thermal power. This instruction did not clearly indicate that NI readings be adjusted conservatively with respect to calculated core thermal power at rated condition (2) 1-ST-26: The current procedure (Rev.1) does not clearly direct the test personnel to' compare the measured RCS flow against the TS required valu (3) 1-ST-29: The current procedure (Rev.0) (a) does not clearly address how to meet test objective 1.3 with regard to the incore thermocouple check, and (b) does not provide evaluation guide '

lines if quadrant power tilt ratio exceeds 1.02 at the 30% power plateau tes (4) 1-ST-38: The current procedure (Rev.0) does not provide pre-cautions to direct test personnel to potential trouble locations for monitoring the test progress and equipment respons (5) 1-ST-39: The current procedure (Rev.0) does not include evaluation criteria for key system responses such as' pressure /

temperature limits and pressurizer safety valves.

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The inspector discussed these concerns with the licensee represent-ative who indicated that these would be incorporated into the next revision. These procedures will be reviewed again to verify the

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changes have been incorporated. Other than the items as noted above, the licensee's test procedures are well prepared and' technically i

sound. The system responses in many transient tests such as Large Load Reduction (1-ST-35), Unit Trip from 100% Power (1-ST-38), and Station Blackout Test (1-ST-39) have been verified in the simulator, and lessons learned from these simulations have been incorporated j into the test procedures.

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3.0 QA/QC Interface with the Startup Test Program The site QA/QC organization, program and surveillance plan to support the upcoming startup test coverage was examined in the previous NRC inspection 50-443/86-31. During this inspection period, the inspector held discussions with the site QC supervisor and a cognizant QC inspector to monitor QA/QC's progress on these activities. The inspector noted that-QA/QC has reviewed all test procedures. Resolution of findings are still in progress. Work on a detailed surveillance plan.and specification of

. hold points is nearly complete Through these discussions and examination of a sample surveillance plan, the inspector concluded that QA/QC is actively involved in the startup test progra .0 Structural Integrity Test Results Evaluation (In Region I Office)

4.1 Introduction-purpose and Scope General Design Criterion 1 of Ap'pendix A to 10 CFR Part 50 requires that structures , systems, and components of nuclear power plants

- important to safety be tested to quality standards commensurate with the importance of the safety functions to be performed. The design of the Containment Building (CB) assures that the reactor and con-tained systems can operate without undue risk to the public health and safety. The primary function of the CB is to contain or control the release of radioactively hazardous releases which could escape from the reactor coolant pressure boundary in the event of a Loss-

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of-Coolant-Accident (LOCA). The design basis for the CB is the LOC The CB additionally serves as a biological shield and provides missile protection for safety related system and component The reinforced concrete Containment was designed by UE&C to withstand all credible conditions of loadings which-included construction loads, normal operating loads, test loads, loads resulting from the design basis accident and loads due to adverse environmental condi-tions. The CB design by UE&C was performed in accordance with the requirements of ASME Section III, Division The Containment has the capability of withstanding pressure and temperature transient loads in excess of those associated with the LOCA without loss of furetional integrity. Verification of the containment integrity after completion of the construction was performed March 15-21,198 The preoperation, planning and conduct of the SIT (Structural Integrity Test) was inspected by the NRC in conjunction with the integrated leak rate test (ILRT). This is documented in NRC inspect-ion report number 50-443/86-15. A review is herein presented of the final test report on the SIT prepared by UE&C as required by ASME B&PV Code Section III, Division 2.

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4.2 Discussion The discussion of the results is primarily based on measured data from the peak pressure 60.2 psig which is 115% of the design pressure (52 psig). The predetermined pressure plateaus used in the SIT are 0.0, 13, 26, 39, 52, 60, 52, 39, 26, 13 and 0.0 psig. The UE&C final test report on the SIT is based on an evaluation of the tests data and a comparison of the test data with the predicted / acceptable data, it is concluded that the test results correlate satisfactorily with theoretically predicted response and the Seabrook Unit 1 concrete Containment structure has responded satisfactorily to the test pres-sure loads. Therefore, the Containment structure has satisfied the structural acceptance criteria of ASME B&PV Code,Section III, Division 2, Subarticle CC-6213, 1980 Edition. This general conclu-sion is observed supported by discrete analysis by UE&C of test data obtained by the SIT contractor, Teledyne Engineering Services / Brewer Engineering Laboratories (TES/ BEL). The following presents the neces-sary condition obtained in the conduct of the test: At the predetermined pressure plateaus, all required data were recorded at least one hour after achieving the required pressure plateau. Before proceeding to the next pressu e plateau, all measured displacements were compared with the predicted values by the Designer in order to ensure that.there were no indicat-ions that the Containment was responding in an unacceptable manne . The exterior concrete surface of the containment was inspected ,

for crack patterns. All cracks 0.01 inch in width or greater were mapped and monitored at the predetermined areas at each

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pressurization plateau and at 39.0 psig and 0.0 psig depressuri-

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zation plateaus. Photographs of each cracking area were also l taken at each mapping time. All related concrete cracking data

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were evaluated by the Designer.

i Temperature differential between the containment interior and

! the Containment enclosure space was monitored and reviewed to j ensure that it did not exceed 65 F.

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) In order to get.all required displacement and crack recovery l information, the monitoring continued for about 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after l

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0.0 psig depressurization plateau was achieved. Acceptable recovery was achieved.

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The instrumentation for containment deflection utilized sixty-six (66) displacement measurements and these are itemized as follows:

• Seven (7) Containment diametrical Fifteen (15) Containment radial

• Twelve (12) Containment radial around equipment hatch Twelve (12) Containment radial around personnel airlock

• Two (2) Equipment hatch diametrical Two (2) Personnel airlock diametrical

• Four (4) Vertical near spring line

• Twelve (12) Dome non-vertical (six vertical deflections were computed from these non-vertical deflection measurements).

4.3 Significant Findings UE&C's evaluation of the test results were reviewed by the inspecto These significant findings appear to be valid conclusions drawn by UE&C from the SIT test results observed and the recorded measure-ments, NRC' inspection report on the conduct of the IST Inspection Report 50-443/86-15, provides inspection observations during the test and verified the licensee's administrative controls and involvement in assuring qualit The majority (94%) of the instruments performed well and the data recorded was sufficient to evaluate the containment's structural responses ~to the test in accordance with ASME B&PV Code,Section III, Division 2, Subarticle CC-6213, 1980 Editio All measured radial displacements are in excellent agreement with the analytically predicted values presented by UE& Measured maximum radial displacement where potential voids were identified behind the liner plate '(Reference NRC/PS NHY Blue Sheet 047) is within the acceptable value and, after depressuri-zation, radial recovery confirmed the elastic behavior of the liner plate. There was no evidence of any localized distress to the liner plat Vertical displacements of the cylinder and the dome are con-

.siderably lower than the predicted values. Generally higher concrete tensile strength was exhibited than what was assumed in analyses. This resulted in reduced concrete cracking-and lower displacement The CB's structural recovery twenty hours after depressurization was more than the acceptable minimum recovery of 70%.

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.The concrete crack pattern plots of maximum measurements are within acceptable values. Based on the evaluation of, crack

. widths and displacement there was no yielding of concrete rein-forcement steel. Multiple in-line cadweld/rebar splices (re-ference NRC IR' item #80-12-01 & 85-07) did not result in'large cracks local to the areas; small-cracks were distr,1buted over the area There were no visible signs of permanent damage or any signs of localized stress either to the concrete structure or the steel line The structural concrete showed a greater tensile strength under the. SIT conditions than what was assumed to develope by the-analytically predicted values. This resulted in reduced crack-ing and hence, smaller displacement The satisfactory recovery of the containment cylinder is a positive indication of the elastic behavior of the containment structure _under the SIT condition, as predicted.

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4.4 Summary The inspector ascertained through his observations during the SIT and independent review of the test results that the Seabrook Unit 1 con-tainment structural integrity test was performed consistent ~with regulatory requirements and licensee commitments. Further, based on test results containsd in the report prepared by UE&C dated June 16, 1986 and comparison with UE&C's-analytical predictions of the' con-tainment's response, the Seabrook Unit I containment structurally responded satisfactorily ta internal pressure required'fof acceptance'

by the ASME B&PV Code'Section III, Division 2, Subsection CC-621 No violations were. identifie .0 Management Meettnq License management was ir. formed of the scope and purpose of the inspection at an entrance meeting conducted on September 29, 1986. The findings of the inspection were discussed with licensee representatives during the course of the inspection. An exit meeting was conducted on October 3, 1986 at the conclusion of the inspection (see paragraph I for attendees).

At no time during this inspection was written material provided to the licensee. Based on the NRC Region I review of this report and discussions held with licensee representatives at the exit, it was determined that this report does not contain information subject to 10 CFR 2.790 restrict-ion .

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ATTACHMENT A .

STARTUP TEST PROCEDURES REVIEWED

--l-ST-2, Primary' Source Insta'llation, Re ,

--l-ST-5, Control Rod Drive Mechanism Operational Test, Re l-ST-10, RTD Bypass' Loop Flow Verification, Re l-ST-12, Reactor Coolant System Flow Coastdown,- Re l-ST-13, Operationel Alignment of Nuclear Instrumentation, Re l-ST-17, Boron Endpoint Measurements, Re ~

--l-ST-23, Dynamic Automatic Steam Dump Control, Re l-ST-26, Thermal Power Measurement and Statepoint Data Collection, Re l-ST-29, Core Performance Evaluation, Re l-ST-30, Power Coefficient Measurement, Re l-ST-33, Shutdown From Outside the Control Room, Re l-ST-35, Large Load Reduction, Re l-ST-36, Axial Flux Difference Inst'rumentation-Calibration, Re l-ST-38, Unit Trip from 100 Percent Power, Re l-ST-39, Station Blackout Test, Rev.0

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