IR 05000400/1986094

Insp Rept 50-400/86-94 on 861220-870120.No Violations Noted. Major Areas Inspected:Status of TMI Action Items,Const Matl Testing,Precritical Testing,Initial Criticality & Low Power Testing

ML20205N225
Person / Time
Site:	Harris
Issue date:	03/24/1987
From:	Burris S, Fredrickson P, Maxwell G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20205N188	List: IR 05000400/1986094 ML20205N185
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.C.1, TASK-1.D.2, TASK-2.F.1, TASK-2.F.2, TASK-3.A.1.2, TASK-TM 50-400-86-94, NUDOCS 8704020627
Download: ML20205N225 (9)

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UNITED STATES gua6og'o NUCLEAR REGULATORY COMMISSION

[ n REGION ll I

$ j 101 MARIETTA STREET. l

• ATLANTA.' GEORGI A 30323

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l Report Nos.: 50-400/86-94 l i

Licensee: Carolina Power and Light Company P. O. Box 1551 Raleigh, NC 27602

Docket Nos.: 50-400 License Nos.: NPF-63 Facility Name: Harris 1

) Inspection Conducted: December 20, 1986 - January 20, 1987 Inspecto s: - 18 7'/ f/ ;

f G. F. Maxwell Date Signed 51 7/2'/ff7

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S. P. Burris Date Signed

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Accompanying Personnel: G. Krug ~

. W. Ruland s

Approved by: / 3 9 P7 l P 'E. Fredrickson, Section Chief /0 ate Signed j

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Reactor Projects Section 1C Division of Reactor Projects

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! SUMMARY

l Scope: This routine, announced inspection involved inspection in. the areas of i

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Status of Three Mile Island Action Items, Construction Material Testing, Precritical Testing, Initial Criticality, Low Power Testing, and 'other

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I Results: No violations or deviations were identified. One unresolved item was

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identified " Operability of ' A' AFW Pump" paragraph i e

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REPORT DETAILS Persons Contacted Licensee Employees J. M. Collins, Manager, Operations L~. Forehand, Director, QA/QC J. L. Harness, Assistant Plant General Manager, Operations C. S. Hinnant, Manager, Start-up L. O. Loflin, Manager, Harris Plant Engineering Support C. H. Moseley, Jr. , Manager Operations QA/QC G. A. Myer, General Manager, Milestone Completion M. F. Thompson, Jr. , Manager, Engineering Management D. L. Tibbitts, Director, Regulatory Compliance R. B. Van Metre, Manager, Harris Plant Technical Support R. A. Watson, Vice President, Harris Nuclear Project J. L. Willis, Plant General Manager, Operations

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Other licensee employees contacted included construction craftsmen, technicians, operators, mechanics, security force members, engineering personnel and office personnel. Exit Interview j The inspection scope and findings were summarized on January 21, 1987, with the Plant General Manager, Operations. No written material was provided to the licensee by the resident inspectors during this reporting perio The licensee did not identify as proprietary any of the materials provided to or reviewed by the resident inspectors during this inspection. Status of Three Mile Island Action Items (25401B)

i (Closed) Inspector Follow-up Item 50-400/85-16-02, - Three Mile Island l Item I.C.1, " Guidance for the Evaluation and Development of. Procedures for Transients and Accidents." The resident inspectors discussed the- i status of this item with the Nuclear Reactor Regulation (NRR) Licensing i Project Manager. The inspectors reviewed a safety evaluation which was attached to a letter to the licensee from the Acting Director of NRR, dated January 12, 198 The safety evaluation concluded that the licensee's commitments regarding the requirements of Generic Letter 82-33 concerning Three Mile Island Item I.C.1 and Item 7 of Supplement 1 to NUREG-0737 will be confirmed by NRR before the start-up of the second fuel cycl The licensee is maintaining this item on their internal tracking system. This item is close i I

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2 (Closed) Inspector Follow-up Item 400/85-16-15, Three Mile Island Item .

II.F.1 " Subparts (1) and (2) Effluent Radiological Monitoring, Sampling and Analyses." This item was evaluated by Region II personnel and as a result, another Inspector Follow-up Item was identified, 50-400/86-79-04. Subsequently, the licensee has modified the four wide range ef fluent monitoring particulate iodine skids to provide straight-through sample flow for normal operation. The resident inspectors evaluated the modified hardware, walked down the system and reviewed the test data associated with the modification. As a result, the inspectors determined that the licensee has made the necessary plant modifications to resolve both of these Inspector Follow-up Items (400/85-16-15 and 400/86-79-04). These items are close (Closed) Inspector Follow-up Item 50-400/86-16-19, Three Mile Island Item II.F.2, " Instrumentation for Detection of Inadequate Core Cooling." The resident inspectors verified that the installation of the hardware was complete for this item. The inspectors evaluated the licensee's letter to NRR dated January 7, 1987, and the results of surveillance tests which were conducted on the Reactor Vessel Water

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Level Indication System (RVLIS) and the Subcooling System. The surveillance test results which were evaluated included:

MST-I-80, 81, 90, 321, 322, MST-C-001; MST-I-119, 120, 121, 248, 249 and 25 The inspectors observed and evaluated the performance of the RVLIS and found it to be satisfactor Based on the above observations, evaluations and reviews, Inspector Follow-up Item 50-400/85-16-19 is close ~ (Closed) Inspector Follow-up Item 50-400/85-16-25, Three Mile Island Item III. A.1.2 " Upgrade Emergency Support Facilities." This item was updated in Region II Report 50-400/86-4 Subsequently, the plant computer display system, which is associated with the Emergency Support Facilities (ERFIS) has been declared operable. The resident inspectors evaluated the status of the following four items which were identified in Report 86-45:

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The inspectors evaluated the results of the preoperational test of the Radiation Monitoring System (1-7005-P03). This test verified that the data link from ERFIS to the Radiation Monitoring System was complet The licensee has eliminated the computer calculation for stability class and has implemented a Plant Emergency Procedure PEIP-341, Re . This latest revision of PEP-341 requires manual calculations for stability class using actual meteorological dat The inspectors evaluated the implementation of PEP-341 and found that stability class was being manually calculate .- . .  !

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The inspectors evaluated the* meterological Tower Performance Test results (EPT-20, Rev. 0) and determined that the ERFIS data link to the meteorological tower had been declared operabl The inspectors evaluated the results of preoperational tests which were performed on the ERFIS computer (1-6005-P01 and 1-6005-P02).

As a result of the evaluation and discussions with the cognizant .

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start-up engineer,. the inspectors determined that the ERFIS 1 computer was operationa Based on the above reviews and evaluation, Inspector Follow-up Item !

50-400/85-16-25, and Incomplete Emergency Preparedness Items 50-400/85-09-13, 50-400/85-09-19 and 50-400/85-09-25 are closed. These items are close (Closed) Inspector Follow-up Item 50-400/85-16-26, Three Mile Island Item III.A.2 " Improving Licensee Emergency Preparedness - Long Term."

The status of this item was also identified in Region II Report 50-400/86-45. Subsequently, the operational testing of the plant announcement system and the ERFIS computer have been complete The resident inspectors evaluated the results of the preoperational tests for the plant announcement system (1-6030-P01, P02 and 1-6032 P01 and P02). The inspectors verified that the system was operable and that audibility of alarms had been evaluate Based on the above evaluations, Inspector Follow-up Item 400/85-16-26 and Incomplete Emergency Preparedness Item 400/85-09-39, " Completing Full Functional Testing of the On-Site PA System," are closed. These items are close (Closed) Inspector Follow-up Item 400/85-16-29, Three Mile Island Item I.D.2 " Plant Safety Parameter Display Console (SPDS)." The inspectors examined Engineering Periodic Test EPT-014, Rev. 2 including changes up to Change 4. The test was properly executed and the procedure itself provided a comprehensive test of the SPDS displays. The inspectors spot checked several test steps and evaluated the results prescribed by the procedur The inspectors also performed a comprehensive examination of the SPDS display and found no problems affecting the operability of the system. This item is close . Construction Material Testing (92719)

During the weeks of December 15 and 22, 1986, the resident inspectors observed portions of the licensee's investigative tests being conducted on Phillips expansion anchors, Maxibolt anchors and " shear" plate The tests on Phillips concrete expansion (wedge) anchors involved drilling nine holes into the fuel handling. building concrete slab at elevation 236'. Three of the holes were sized for complete compliance with the applicable work procedure (WP-33) for 3/4" diameter Phillips expansion anchors. The 3/4" wedge anchors were installed and torqued

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to the minimum torque of 150 foot pounds. Three of the holes were purposely " wallowed" out by the craftsmen. This resulted in these holes being oversized for the 3/4" diameter anchors. The 3/4" diameter anchors were then placed in these oversized holes. Fine sandblasting sand was then poured in the holes filling the space around the body of "

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the anchors. The anchors were then torqued to the minimum required

, value of 150 foot pounds. The remaining three holes were sized for l complete compliance with the work procedure for 3/4" diameter expansion anchors. Fine sandblasting sand was then poured in the holes filling the space around the body of the anchors. The anchors were torqued to the minimurn required value of 150 foot pound The nin- installed expansion anchors were then allowed to remain 5 u disturbed for one day, in order for the anchor relaxation to occur as

specified in licensee procedures. They were then tension tested using l the applicable CP&L procedures as a guide (WP-33 and TP-39). These i procedures required that the anchors be capable of resisting a minimum

) of 5980 pounds, which is 115 percent of the static working load. All j nine anchors met the test requirements. At the request of the resident

< inspector, the anchors were tested to failure (slip). One anchor l slipped at 6650 pounds (128 percent of static working load), one slipped at 7125 pounds (137 percent of static working load), and one at

] 7600 pounds (146 percent of static load). The remaining six anchors did not fall under the 7600 pound load.

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The test on Maxi-bolt expansion anchor bolts involved drilling three

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holes into the fuel handling building concrete slab at elevation 236'.

The holes were drilled using bits specified in work Procedure WB-4 i However, the holes were not expanded at their botto;ns, i.e., undercut, l in accordance with the work procedure :

i The 3/4 inch diameter Maxibolts were then inserted into the holes and

! pre-set (expanded) to the prescribed pressures shown in QP-42 (4350 j psi).

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Two of the Maxi-bolts were allowed to remain undisturbed for one day, in order for the anchor relaxation to occu The third bolt was

{ tension tested immediately following its installation and pre-loadin l Each of the bolts was found not to slip at the minimum test pressure j (2540 psi) specified in the Technical Procedure TP-39.

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The inspectors witnessed portions of material type and location

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verification of shear plates in the emergency service water intake

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structure (ESWIS) traveling screen bay The shear plates were

, installed to provide a full bearing surface for the traveling screen i frames where embedded guides were out of the alignment. Design documents required that these steel shear plates be fabricated of

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ASTM-A-36 rolled plate steel and were to be located at the specified l elevations and spacings, i

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l The resident inspectors randomly selected 20 plates to be tested, ten from Bay 6 and ten from Bay 8. Five plates from each side of the bays were tested with two on each side located below the normal water leve After Bays 6 and 8 were pumped, the inspectors witnessed CP&L Materials Lab personnel perform in place hardness tests using a calibrated portable hardness tester. Hardness values were obtained, recorded and converted into Brinell hardness numbers which were attached to CP&L's test repor In addition to the licensee's test performance, Law Engineering Testing Company (LET) was requested to provide an independent assessment of test result LET personnel performed hardness testing using their equipment methods. The test results were independently obtained, recorded and converted into Brinell hardness numbers. These test results were also attached to CP&L's test report for verification of material identificatio To provide additional material verification, each of the test plates selected below water line had specimens cut from them. Each specimen was dived in half, with one piece being retained by the resident inspectors and the other half was given a tensile test by CP&L lab personne All of the tensile test and hardness test results compared to the range  ;

for hardness of ASTM A-36 material, based on the Brinell hardness scale i (119 to 159). This comparison showed that all of the plates tested were within acceptable tolerance range for A-36 plate material, with the exception of minor test deviations, which were still well within the design allowables for this plate applicatio In conjunction with the material testing discussed above, the licensee also checked the location and site of the installed shear plates. This was done by measuring from known reference points in the intake structure. The resident inspector observed licensee personnel making l these measurements and verified that correct measurements were i obtaine No discrepancies were foun No violations or deviations were identified in the areas inspecte Precritical Testing (72521C)

During the week of December 29, 1986, the inspectors witnessed and reviewed select portions of the overall crew performance and conformance to administrative procedures; select test results were reviewed and, in general, the precritical test program was evaluate The crew was evaluated to assure that sufficient manpower was available and that they were alert to changing plant conditions. The inspectors reviewed the precritical test prerequisites and the initial conditions to assure that Technical Specification (TS) requirements were being me The inspectors evaluated the data which was being documented by the crew members and determined that they were following procedure _ _ - . _ . __ __ _ . _ _ . _

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The inspectors evaluated the test sequence documents, the shift I- supervisor's logs, and the various plant information reports located in

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Outstanding issues identified by the logs were well identified and were found to have been subsequently further evaluated by the operations or technical staf ) No violations or deviations were identifie . Initial Criticality (72592C)

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! On January 2 and 3, 1987, the inspectors evaluated the licensee's

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conformance prior to and during reactor control rod withdrawal to

criticalit This evaluation included reviews of the required TS and

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licensee conditions applicable to approach to criticality. Verification of

, start-up and intermediate nuclear instrumentation were properly calibrated l and operating with the proper count rate and signal to noise rati t

! The inspectors verified conformance to the licensee's administrative and j procedural requirements by the following:

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Performance of three separate reviews of crew manning to insure that

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the licensee satisfied TS requirements; i

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Verification that the current versions of the procedures were in use i and being followed;

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Verification that procedural prerequisites and initial conditions had j been satisfied;

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The inspectors reviewed special installed instrumentation to insure j that the requirements of the procedure were met and the data was j readily available for analysi The inspectors participated with assigned Region II personnel during these l j observations, the results of which are documented in other Region II i i reports.

i No violations or deviations were identifie . Low Power Testing (72522C and 71501C)

i During the week of January 12, 1987, the inspectors reviewed various plant

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status reports, interviewed plant personnel and observed status of the plant. The reviews included evaluation of the "A" auxiliary feedwater (AFW)

pump status. The shift foreman's log dated January 12, 1987, was found to show that the AFW pump had been considered as inoperable at about 3:00 p.m.,

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I The inspectors inquired about the inoperability and were informed by the j licensee that the pump had been declared inoperable because it had failed'to l

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meet the full flow test requirements specified by the T TS Section 3/4.7.1.2, under the " Design Basis" section, states that each of the electrically-driven pumps is capable of delivering a feedwater flow of 475 gpm at a pressure of 1217 psig to the steam generators. The inspectors reviewed an operations surveillance test (OST-1087) which indicated that on December 31, 1986, the "A" AFW pump was tested and its pressure and flow data as recorded by the operations personnel was less than 440 gpm at 1200 psi The inspectors noted that 12 days had lapsed since the data had been recorded and approved by the shift foreman prior to any action being taken concerning placing the pumps on inoperable status. On January 13, 1987, the inspectors discussed this delayed condition and the overall plant operational status with licensee personnel. The inspectors were informed that the plant's mode had been changed from a mode 3 hot standby condition to a mode 4 hot shutdown condition to assure compliance with TS I Section 3.7.1.2.a, " Action Section."

The inspectors were provided additional information which they have forwarded to Region 11 management for further evaluation. The inspectors will continue evaluating this condition to determine whether the licensee ,

has violated any NRC requirements. This is an Unresolved Item, " Operability of "A" AFW Pump," 50-400/86-94-0 No violations or deviations were identified in the areas inspecte . Other Activities (92719) On January 5, 1987, the site was visited by USNRC Commissioner Asselstine. He attended a public presentation where Cp&L provided the status of operator training, INP0 training program, operational readiness, construction, plant staffing, preoperational testing, start-up, maintenance, the unique plant features, general configuration of the site and the plant capacity. Members representing parties to the license hearing attended the CP&L presentation and accompanied the Commissioner for the duration of his visit. A plant tour followed the CP&L presentation. The tour included stops in the turbine building near the feedwater pumps, the diesel generator building, the main control room, the high voltage switchgear room, portions of the reactor auxiliary building, the fuel handling building and the waste process building control room. After completing the tour, an exit meeting was held between NRC and CP&L management, On January 8, 1987, the USNRC Commissioners held a scheduled full power

license hearing in Washington, D.C. CP&L management briefed the l i Commissioners as to the status of full power prelicense issues and l answered specific Commission questions. Members of the general public were allowed to voice their comments and concerns. Regional management addressed open NRC items and provided specific responses to outstanding issues. Upon completion of these comments and questions, the

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Commissioners voted unanimously to allow the issuance of a full power license for CP&L Shearon Harris facilit _.