L-76-215, Letter Relating to a Review of Inspection Report 05000335/1976006 and Informing the NRC That It Contains No Proprietary Information

Letter Relating to a Review of Inspection Report 05000335/1976006 and Informing the NRC That It Contains No Proprietary Information

ML18127A089
Person / Time
Site:	Saint Lucie
Issue date:	06/08/1976
From:	Robert E. Uhrig Florida Power & Light Co
To:	Moseley N NRC/RGN-II
References
L-76-215 IR 1976006
Download: ML18127A089 (28)

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June 8, 1976 L-76-215 l~ -sar Norman C. Moseley, Director Office of Inspect ion and Enforcement-Region II U. S. Nuclear Regulatory Commission 230 Peachtree Street, N. N., Suite 8l8 Atlanta, Georgia 30303

Dear Mr. Moseley:

Re: IE:II:MSK 50-335/76-6 Florida Power 6 Light: Company has reviewed the subject inspection report and has determined that no proprietary .information.

it contains Very tru y yours,

)

+4CrjC Robert. E. Uhrig Vice President REU/LLL/hlc cc: Jack R. Newman, Esp.

e '

uNITED sTATEs NUCLEAR REGULATORY COMMISSION REGION II 230 PEACHTREE STREET>> N W SUITE 81$

ATLANTA,GEORGIA 30303 re "- 0 1976 In Reply Refer To:

IE: II:MSK 50-335/76-6 Florida Power and Light Company ATTN: Dr. R. E. Uhrig, Vice President of Nuclear and General Engineering P. 0. Box 013100 9250 West Flagler Street Miami, Florida 33101 Gentlemen':

This refers to the inspection conducted by Messrs. M. S. Kidd and J. D. Martin of this office on April 7-9, 12-16 and 18-22, 1976, of activities authorized by NRC Operating License No. DPR-67 for the St. Lucie 1 facility, and to the discussion of our findings held with Mr. K. N.. Harris at the conclusion of the inspection.

Areas examined during the inspection and our findings are discussed in the enclosed inspection report. Within these areas, the inspection consisted of selective examination of procedures and representative records, interviews with personnel, and observations by the inspector.

Within the scope of this inspection, no items of noncompliance were disclosed.

We have examined actions you have taken wit.h regard to previously reported unresolved items. These are identified in Section IV of the summary of the enclosed report.

In accordance with Section.2.790 of the NRC's "Rules of Practice,"

Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC's Public Document Room. If this report contains any information that you believe to be proprietary, it is necessary that you submit a written application to this office requesting that such information be withheld from public disclosure. If no'roprietary information is identified, a written statement to that effect should be submitted. If an application is submitted, it must fully identify the bases for which information is claimed to be proprietary. The application should be prepared so that information sought to be withheld is incorporated in a separate paper and referenced in the application since the application will be placed

";MY 2 C 1976 t l1

.Florida Power Company and Light in tne Public Document Room. Your application, or written statement, should be submitted to us within 20 days. If we are no't contacted as specil.ied, the enclosed report and this letter may then be placed..in

.the Public Document Room.

Should you have .any questions concerning this 'letter, we will be glad to discuss them with you.

Ve'y truly yours, F. J. Long, Chief Reactor Operations and Nuclear Support Branch

Enclosure:

IE Inspection Report No.

50-335/76-,6

c'8O444 i UNITED STATES

~ c ;)i'j gl NUCLEAR REGULATORY COMMISSION REGION II 230 PEACHTREE STREET, N. W. SUITE 818 ATLANTA,GEORGIA 30303 IE In.;pection Report No. 50-335/76-6

,Licensee: Florida Power and Light Company P. O. Box 013100 Hiami, Florida 33/01 7acility,Name.: St. Lucie 1 Docket No.: 50-335 License No.: DPR-67 Category: B2 Location: Hutchinson Island, Florida Type of License: CE, PHR, 2560 Hwt Type of Inspection: Routine, Announced Dates of Inspection: April 7-9, 12-16, and 18-22, 1976 3)ates of Previous Inspection: Harch 1-5, 1976 Principal Inspector: H. S. Kidd, Reactor Inspector Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch (April 12-16 and 18-22,,1976)

Accompanying Inspector: J. D. Hartin,'eactor Inspector Nuclear Support Section Reactor Operations and Nuclear Support. Branch (April 7-9 and 18-22, 1976)

Other .Accompanying Personnel: None Principal Inspector:

H. S.'Kid , Reactor Inspector Date Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch Reviewed by: /-/ ~ ~- ~ I '-'-~7 ~//// /

R. C. Lewis, Chief Date Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch .

SPRAT'RY OF FINDINGS I. Enforcement Items None II. Licensee Action on Previousl Identified Enforcement Matters Not applicable.

III. New Unresolved Items None IV. Status of Previousl Identified Unresolved Items 75-18/1 Leak Detection S stem Sensitivit Sensitivity of the inventory balance method of leak detection was verified during pre-critical testing.

This item is closed. (Details I, paragraph 2.b.(2))

75-19/1 Installation and Testin of. Han ers and Restraints Acceptability- of test data taken during pre-fuel load testing has been clarified. Honitoring of piping hangers, restraints, and snubbers, and piping deflections was accomplished during pre-critical testing, This item is closed. (Details I, paragraph 2.a(5), and Details II, paragraph 2) 76-2/2 Baseline Ins ection Data An engineering analysis of the ultrasonic indications in the reactor coolant pump support lugs resulted in the conclusion that the lugs are structurally adequate. This analysis was approved by the licensee. This item is closed.

(Details I, paragraph 2.a.(4))

76-4/2 Containment'oundar ualit Grou Desi nation (50.55 e)

Errors in the Ebasco report of January, 1976, regarding wel'd and isometric identifications have been corrected.

All upgraded welds have now been non-destructively tested.

This item is closed. (Details I, paragraph 2.a.(6))

IE <<pt.~ Wo.~ 50-335/76-6 Unusual Occurrences Core Flow Analysis of pre-critical reactor core flow data by Combustion Engineering revealed that the flow was approximately two percent less than the value required by Technical Specifications.

This is to be reported as a fourteen day written report per RG 1.16. I CEDN Malfunction During cold rod drop testing control element drive mechanism (CEDM) number 44 malfunctioned such that the control element assembly (CEA) could not be withdrawn. The CEDH operated

-normally, at rated temperature and pressure. (Details I, paragraph 2.a(3))

C. MSIV Closure Time The main steam isolation valve (MSIV) for "B" header closed in seven seconds during surveillance testing as compared to the maximum of six seconds allowed by Technical Specification 3.7.15. This item will be reported per RG 1.16.

D. Defective Mechanical Pi e Restraints During plant heatup for post core load testing, five mechanical pipe restraints were found to be locked up. Corrective actions completed- at the time of the inspection were reviewed. This matter was reported per RG 1.16 April 12, 1976. (Details II, paragraph 2)

PI. Other Si nificant Findin s A. Plant Status Completion of the activities required prior to initial criticality by Enclosure 1 to the Unit 1 operating license, along with licensee commitments relative to criticality were verified to be complete. (Details I, paragraph 2 and Details II, paragraphs 2 and 3)

B. Initial Criticalit Portions of the approach to initial criticality on April 20-22, 1976, were witnesse'd by the inspectors, with no discrepancies identified. (Details I, paragraph 3 and Details II, paragraph 4)

'::i. Hang ement Interviews Hanagement interviews were conducted April 9 and 16, 1976, with K. N. Harris and other licensee staff members to discuss findings of the inspection relative to completion of requirements for starting the approach to initial criticality. Items discussed included the requirements listed in Enclosure 1, Section A of the operating license and the previously-identified unresolved items in Section IV of thj.s Summary. 'Details 2, paragraph 2 and Details II, paragraphs 2 and 3)

Another management interview was conducted April 22, 1976, with K. N. Harris and members of his staff to discuss'indings of the inspection relative to the approach to initial criticality. The results of the inspection were discussed. (Details I, paragraph 3 and Details II, paragraph 4)

'R Rpb. Ne. SO-335/76-6

,'EiA S I Prepared by: -b 7i r/7i.

M. S. Kidd, Reactor Inspector Date Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch Dates of Inspection: April 12-16 and 18-22, 1976 Reviewed by:

R.

/C.'ewis,

. <-. /i;..;.i~

Chief Date Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch

'. Persons Contacted Florida Power and Li ht Com an (FP&L K.Harris Plant 1fanager J. Barrow Operations Superintendent C. Hells Operations Supervisor R. Ryall - Reactor Supervisor P. Dillon Technical Staff Supervisor R. Hayes - Technical Staff Engineer J. Pride - Technical Staff Engineer K. Beard Naintenance Engineer J. Lenz Instrumentation and Contxol Engineer A. Anderson Construction QA Engineer R. Roehn Construction QA Engineer.

Two Nuclear Plant Supervisors Two. Nuclear Match Engineers Four Nuclear Control Center Operators Ebasco Services Incor orated (Ebasco)

J. Albanes - Stress Analyst Supervisor H. Noronha - Stress Analyst Supervisor P. Peterson Quality Control Records Combustion En ineerin Incor orated (CE)

E. Smith Site 1fanager J. Tefft Chief Test Engineer T. Oliver Startup Engineer

N IE Rpt. No. 50-335/76-6 I-2

2. Com letion of Re'irements Relatin to Initial Criticalit 0 eratin License Enclosure 1 Section A of Enclosure 1 of the St. Lucie Unit 1 Operating License, DPR-67, set forth six activities which required resolution or completion prior to achieving initial criti-cality. During th'e inspection the inspectors verified that the activities had been completed or resolved. Hethods of verifying completion are discussed below and in Details II, paragraphs 2 and 3.

(1) In'stallation of Tornado 1H.ssile Protec'tion Verification of these installations is discussed in Details II, paragraph 3.

t (2) S ent Fuel Char in S stem Intertie Verification of the installation of this line is dis-cussed in Details II, paragraph 3.

(3) Cold Rod Dro Tests Pre-critical -rod drop tests were conducted in accordance pith test procedure 0110081, "CEDH/CEA Measurements."

Data sheet 1 of this pr8cedure was used 0 to document rod drop tests at cold (260 F) and hot (532 F) conditions fox each full length control element assembly (CEA). Review of rod.drop times at the 260 0 F + 20 0 F plateau revealed the following information:

(a) Drop times on all rods except number 44 with two reactor coolant pump (RCP) flow had been conducted as of April 2, 1976, at which time the decision was made to continue plant heatup and testing. CEA 44 could not be dxop-tested because of malfunction of the lower gripper.

Drop times for ninety percent insertion ranged from 1.96 seconds to 2.35 seconds as compaxed to the limit of 3.3 seconds given in Technical Specifi-cation 3.1.3.4 for hot conditions and full core flow.

IE~Rpt. No. 50-335/76-6 I-3 (b) The two fastest CEA's (Nos. 8, 46) and two slowest CEA's (Nos. 51, 69) were dropped once again with two RCP's operating and once each with three RCP's operating and with no flow through the core. The inspector noted that the number of retests did not compare favorably with the number planned for hot retests as discussed in FSAR Table 14.1-2, item 1.

Licensee .personnel agreed that a comparable number of retests should be performed at cold conditions and two additional retests were conducted on the four CEA's at no flow and three RCP flow on April 14, 1976. Retest times compared favorably with the

~

first tests.

(c) CEA 44 gas drop tested successfully three times at the 260 F plateau on April 14, 1976, after cpoldown of the reactor coolant system (RCS) from 532 F.

These tests were all conducted with 2 RCP flow. The lower gripper malfunctioned again on that date before tests with no flow and 3 RCP could be accom-plished. The drop time for CEA 44 was 1.96 seconds, making it the second fastest CEA, replacing CEA 8.

(d) Review of coil current. recorder traces of CEA 44 and discussions with licensee personnel revealed that as RCS temperature was raised from 260. F to 532 F the gripper began to function properly and demon- 'ower strated normal operation at the higher temperature.

On April 16, 1976, the operating license for St. Lucie 1 was amended by NRR, Division of Operating Reactors at the request of FP&L, to delete the special test exception in Specification 3.10.3 whigh would allow the reactor to be taken critical below 515 F.. This action obviated the need for further testing of CEA 44 at cold conditions.

Two of the stipulations upon which the change was based included testing of the third fastest CEA as 44 would have been tested and the performance of ten additional drops on CEA 44 at hot conditions. These activities were verified complete by the inspector on April 16, 1976.

Licensee personnel were informed that the inspector had no further questions in this area.

(4) Baseline Ins ection Results (Unresolved Item 76-2/1)

This unresolved item was last discussed in IE Report No.

50-335/76-4, Details Vl, paragraph 3. At the conclusion of that inspection all aspects of the unresolved item had been resolved except for evaluation of ultrasonic indications on the RCP support lugs by CE and FP&L and submittal of the

IE Rpt. No. 50-335/76-6 I-4 baselin'e inspection report. The .baseline report, trans-under CPSL's letter L-76-.98, dated March ll, 1976, 'itted has been reviewed by IE:II.with no further questions resulting. CE's evaluation of the RCP lug indications, forwarded to FP&L by. letter F-CE-5640, dated March 2, 1976, concluded that the lugs were structurally adequate based on indicated flaws in them. FPGL Engineering concurred with that conclusion. Licensee personnel were.

conformed that thj.s item was 'considered closed.

.(5) Testin of Han ers and Restrai'nts Unresolved Item 75-19/1)

This unresolved item was last discussed 'in IE Report No.

50-335/76-4, Details I, paragraph 4.d. At that time, three questions relative to testing results documented in pre<<fuel load testing per procedure 0010185, "Piping Thermal Expansion and Restraint," remained open. Review of the amended test package and discussions with licensee personnel during the current inspection revealed the following information:

(a) The reactor coolant system (RCS) temperatures't which data were taken had not been included in the data sheets. A letter which had been reviewed by the Facility Review Group (FRG) on February 28, 1976, stating that the actual temperatures were provided in test procedure 0010181, "Hot Functional

,Sequencing Document," on an hourly basis had been entered into the test'results file.

(b) A letter from CE to'he plant manager; dated February 27, 1976, explaining the acceptability of certain data recorded in Table 1 of the test procedure, had been entered into the test package. This letter had also been reviewed and concurred with by the FRG.

(c) Movements of the pressurizer relief lines were monitored during hot functional testing (HFT), but power operated relief valves (PORV) were not instal-led at that time. The inspector verified that these lines were monitored during precritical testing on

'pril 21, 1976, per test procedure 0120096, "Pressurizer Control Checkout" (Revision 5), while steam was being discharged through the PORV's to the quench

,tank. Section 12.6 of 0120096 also was used to

',IE Rpt. No. 50-335/76-6 I-5 functionally test the ability of the quench tank to receive and quench pressurizer steam and to function-

,ally check the resistance, temperature detectors (RTD) in the pressurizer discharge lines used for leak detection purposes.

Inspection efforts regarding retesting of hangers, restraints, and snubbers during precritical testing are documented in Details II, paragraph 2 of this report. This item is consi'dered closed.

(6) Containment Boundar ualit Grou in (Unresolv'ed Item 76-'4/2)

This item involved improper quality designation of certain systems penetrating containment and was the subject of a report per 10 CFR 50.55(e) submitted by FP&L on January 22, 1976. The improper designation resulted in a lack of certain nondestructive testing (NDT), and the affected c'omponents were subsequently upgraded to quality group "B" from "C" by performance of the necessary NDT. The initial 'review by IE:II of the corrective measures in this area revealed that the. interim report from Ebasco to FP&L contained numerous errors and discrepancies in weld and isometric identification. Additionally, three shop-welds in penetration 43 had not undergone required radiog-raphy and liquid penetrant testing for upgrading.

During this inspection, a corrected Ebasco report, by letter number 9063 on. March 9, 1976, was reviewed.

trans-'itted All instances of incorrect weld and isometric identifica-tions had been corrected in this report except for field welds FW-8 and 8A in piping I-8-CC-168-2. These were still,identified as FW-8A and 8B, respectively. The file copy of the report was corrected during the review and licensee personnel stated that recipients of the report

=would be advised of the change. NDT records on file reflected the proper weld numbers. The inspector also reviewed QC records of NDT performed on three shop welds in the reactor drain tank pump suction line, SW-1, SW-2, and SW-3 (isometric drawing WM-Pl). These records included Ebasco QA radiographic summary reports and liquid penetrant test reports. All findings in these reports indicated acceptable welds. Licensee personnel were informed that this'tem was considered closed.

IE Rpt. No. 50-335/76-6 Followu on FP&L Commitments As discussed in IE Report No. 50-335/76-4, Details I, paragraphs 2.c and 2.e.(1), FPSL committed to resolve certain preoperational test exceptions and establish the sensitivity of the inventoryl/

balance method of leak detection prior to initial criticality.

Followup review indicated this status:

(1) Test Exce tions (a) ~

TP 1400090 "Data Processor/Process Instrumentation Correlation Measurement Test" Two temperature indicators and one pressure indicator could not be compared (DDPS versus control board indication)

'during HFT. During pre-critical testing, temperature element (TE) 2221 outputs were compared with favorable results. TE-2229 is no longer a dual element RTD

. and is not used by the DDPS. The pressure transmitter involved, PT-08-1333, is a spare and not yet- installed.

This is one of three which can be used by DDPS for steam generator 13 pressure.

(b) TP 1900082 "Miscellaneous Ventilation, Systems Startup" Decontamina'tion room supply and exhaust fans, HVS-11 and. HVE-36, were checked for vibration on April 13, 1976, after realignment and found to be acceptable. Test results were reviewed and approved by the FRG on April 14, 1976.

(c) TP 2000084 "Hydrogen Purge System Startup and Functional Test" Documentation that flow recorder FR-25-1 had been repaired'and retested was reviewed and approved by the FRG on March 22, 1976.

(d) TP 1120080 "Area Radiation Monitoring Pre-Operational Functional Test" The remote alarms for these .units were disconnected due to intermittent and frequent downscale alarm actuations. Small radioactive sources were placed near the monitors to increase background levels and appropriate steps of TP 1120080 were repeated. The documentation of these activities, which was placed, in QC files to clear the deviation, was reviewed by the FRG February 27, 1976.

The inspector sta'ted that he had no further questions.

1/ See FP6L letter L-76-53, dated February 10, 1976.

IE.Rpt. No. 50-335/76-6 I-7 (2) Leak Detection Sensitivit (Unresolved Item 75-18/1)

An RCS water inventory balance was performed on April 8, 1976, followed by another test wherein a known leak of one (1) gallon per minute (gpm) was superimposed on the RCS through a sample line in accordance with step 12.10 of TP 0010190, "Post Core Load E1ot Functional Sequencing Document." The balances were performed per OP 0010125, "Schedule of Pediodic Tests, Checks and Calibrations,"

data sheet 1. Results documented for the first check indicated a leak rate of .68 gpm. Results with a known leak imposed of 1.0 gpm were 1.66 gpm. The two tests, disregarding the known leak, were within the +.25 gpm acceptance criterion established in TP 0010190. The inspector stated that this item was considered closed.

During review of the data/calculation sheet used for inventory balance, it was noted that RCS temperature was not recorded, making it necessary to verify temperature had not changed during the test from other documentation.

Licensee personnel stated that spaces for recording beginning and ending RCS temperatures would be added to the data sheet. The inspector stated that he had no

'further questions at that time.

3. Initial Criticalit Portions'of the preparations for and approach to initial criticality were witnessed'by the inspectors on April 18-22, 1976. The objectives were to determine whether the activities conformed to Technical Specification requirements (sampling basis) prior to. CEA withdrawal

• and throughout the evolution, assure that administrative and procedural requirements were met, assure that procedure changes and deficiencies were handled per administrative procedures, independently predict the ini.tial critical boron concentration, and observe overall conduct of the approach.

a ~ General The start of the approach to criticality was delayed due to the problems discussed in the Unusual Occurrences section of the Summary of this report. CEA withdrawal was started on April 20, 1976, after completion of sections 4, "Prerequisites,"

and 6, "Related System Status," of the procedure TP 3200086, "Initial Criticality" (Revision 2). After withdrawal of CEA's and'during CEA testing per Appendix A-3 of 3200086,

IE Rpt. No. 50-335/76-6 I-8 problems were experienced with CEA -Nos. 29 and 33. CEA 29 displayed both, an upper and lower limit indication, requiring replacement of a defective reed switch. CEA 33 dropped into the core while being inserted. Licensee personnel found that a test cable, used to feed test devices used to monitor CEDM

'coil currents, was picking up induced current (noise) from other cables in its rack. A slight shift of the cable in the rack alleviated the problem. Other CEDM's hove the same test cable and were being monitored to assure a similar problem did not exist with them.

CEA testing was completed and boron dilution started at 7:56 p.m.

on April 21, 1976. Criticality was. reached at approximately 8:30 a.m. on April 22, 1976, at a boron concentration of approximately.935 940 ppm, as compared to the predicted value of 938 ppm (all rods out except Group 7, which was at 68

.inches withdrawn). The approach was conducted, in accordance with the procedure and section 14.1.4.4 of the FSAR.

b. S ecific Ins ection Activities (1) The licensee's compliance with Technical Specifications requirements for Modes 3* and 2* was ascertained on 'a sampling basis by observation by the inspectors of plant system conditions and review of appropriate documentation.

Completed control room data sheets and surveillance test procedures were reviewed, including the following:

(a) Minimum Equipment List" This is a list which reflects the number of components in systems covered by Technical Specifications sections 3/4 which must be operable for the various modes defined by Technical Specifications. It is filled out each shift during modes 1 through 4.

(b) "Minimum Instrumentation List" This list is similar to the one above and covers all plant instrumentation required to be operable during modes 1 through 4.

It is also filled out by the nuclear control center operator'NCCO) once per shift.

(c) "NCCO Log Sheet No. 1'(SDC Secured)" >> This data sheet is filled out hourly and is used to monitor several parameters which are limited by Technical Specifications, such as RCS temperature and pressure.

P

+ See Technical Specifications Definition Table 1.1.

IH.Rpt. No. 50-335/76-6 I-9 (d) Periodic test procedures, such as those contained in AP 0010125, "Schedule of Periodic Tests, Checks and Calibrations," (Revision 5) for daily, weekly, shiftly and more frequent surveillances.

Completed documents of the types described above were reviewed at periodic intervals from April 18, 1976, to April 22, 1976. Also the NCCO Log and'uclear Plant Supervisor Log w'ere reviewed for the period of April 18-21, 1976. No discrepancies were identified.

(2) Shift complements during the approach were observed to assure that the requirements of Technical Specification Table 6.2-1 were met. The requirements were met or exceeded for'each shift observed.

(3) The latest revision of the procedure, 3200086.(Revision 2) as amended, was used throughout the evolution. Changes to it and other procedures were accomplished in accordance with, station administrative procedures.

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(4) Inverse neutron count rate (1/m) plots were maintained per the procedure as was a plot of boron concentration versus dilution time.

(5) Technical support was provided on a shiftly basis by the presence of Reactor Engineering (FP&L) and CE Startup personnel.

(6) Prior to the start of CEA withdrawal, the licensee's prediction of estimated boron concentration at criticality, along with CE's, were discussed with licensee personnel.

An independent calculation was made, using data provided in FSAR Table 4.3-1, 4.3-2, 4.3-4 and 4.3-6. The resultant value was in close agreement with FP&L and CE predictions.

Licensee personnel were informed that the inspector had no

'omments or questions relative to the conduct of the approach to criticality.

Followu on Previous 0 en Items Licensee actions in response to previous questions and comments by IE:II inspectors on subjects other than those in paragraph 2 were reviewed and discussed. The following information was obtained:

IE Rpt. No. 50-335/76-6 I-10 Test Procedure Comments (1) TP 0110086 "Simulated CEA E'ection, Hot Zero Power "

and TP 3200081 "Nuclear Desi n Check Tests" Comments on these low power test procedures were documented in IE Rpt. No. 50-335/76-1, Details I/I, paragraphs 5a and Sb. Review. of the revised procedures revealed this information:

(a) A reference to the technical manual has been added to provide details on hookup of special test equipment referenced in step 8 of TP 3200081.

(b) Step 9.12 of 0110086 and 3200081 stated that certain approvals had to be obtained prior to proceeding with the various steps of the procedures, but did not require documentation of the approvals. These steps were deleted from Revision 1 of each procedure.

Step 4.8 of Letter of Instruction (LOI) No. 0-7, "Conduct of Operations During Core Loading, Low Power Physics Testing and Power Ascension Tes'ting" infers that the Reactor Engineering staff will be available to advise and help coordinate testing efforts.

(c) Revision 0 of the procedures contained b3.ank spaces

,for predicted design values of certain parameters which were not available at the time the procedures were approved. Review of the revised procedures and other selected startup test procedure's revealed that most blanks had been filled in in Revision 1.

Certain steps, such as 12.1.5 of 3200081, referenced the Plant Curve Book as the proper source of predicted values of design parameters. Licensee personnel stated that an effort would be made to assure that

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all blanks were filled in before testing started for any given procedure and that one of the functions of the PRO during review of results would be to assure that proper design values had been entered into each procedure and results compared to them.

The inspector had no further questions on these procedures."

IE Rpt. No. 50-335/76-6 I-11.-

t (2). TP 2100089 "Generator Tri The inspector reviewed followup action by the licensee in response to comments documented in IE Report No.

'0-335/76-2, Details I, paragraph 4.a and determined the.

following:

(a) Revision 0 did not specify recorder alignment/hookup and calibration for instrumentation used to record data in step 6.5. Review of Revision 1 revealed that no additional information had been added in this area. Further discussions on this and other staxtup procedures revealed that the procedures do not, in general, provide this type of information.

Licensee personnel stated tha't LOI-0-7, the administra-tive procedure for the startup program would be revised to require the test engineers to verify that test instrumentation is in calibration'(e.g.-

curxent sticker) before starting a test. Also, LOI-0-7 will define the use of the test data "patch panel" which is normally used 'to hook up recorders for data retention during testing. The "patch panel" was installed especially. for that purpose and provides the desired isolation of recorders to prevent feedback into protection channels. It was further stated that any special hookups. required other than the "patch panel" would be addressed in the appropriate test procedure, thus receiving prior review and approval.

(b) The manufacturer's recommended. value of 111% has been inserted in Step 10.2 as the acceptance criterion for maximum turbine overspeed.

(c). During discussions on whether specific steps are to be done manually or automatically (all steps are signed of as "verified by"), licensee personnel stated all steps of the procedure except 12.8 which start with "Verify..." mean that the operator is to observe automatic functions and verify that they do indeed perform as designed. Step 12.8 references an emergency operating procedure and portions of that procedure would be done manually. Also, the FPGL intent of the word "verify" is that if a function has not operated properly, then the operator should

IE Rpt. No. 50-335/76-6 perform it manually. It was noted that the use role is widespread in St.

of "verify" in this dual Lucie procedures including emergency operating types.

In repsonse to the inspector's statement that the dual use of "verify" appeared t'o be confusing in as well as emergency operating procedures

'"'est (EOP), licensee representatives stated that these actions would be taken:

1 LOI-0-7 will be revised to address the dual use of "verify" in startup test procedures.

Administrative procedure 0010120, "Duties and Responsibilities of Operators on Shift," will be revised to address the use of "verify" and the philosophy of how it is used is to be brought to all operators'ttention.

All EOP's will be revised within 45 days to include a statement in the immediate operator actions section that if any automatic actions have failed to take place, the operator should perform them manually. Also, Off-Normal procedures, which utilize a format similar to EOP's, will be reviewed 'to determine whether revisions are necessary to them.

The inspector stated that he had no further questions on items (a) and (b) above, but would discuss item (c) and review licensee actions relating to it during subsequent inspections.

(3) TP 2100091 "Loss of Off-Site Power" Comments documented in IE Report No. 50-335/76-2, Details I, paragraph 4.b, have been acted on as follows in Revision 2 of the procedure:

(a) Step 6.1 has been changed to state that pressurizer level and pressure controls are in automatic for the t'est.

(b) Step 4.4 now provides that all test personnel will be briefed before the test begins.

IE Rpt. No. 50-335/76-6 I-13 .

(c) Step 12.7.6 has been revised to clarify that steam generator water levels are to be controlled after the plant trips by manual control of the auxiliary feedwater system. All other steps involve verifica-tion that automatic functions are operating properly.

(See paragraph (2)(c) above)

(d) Acceptance criterion 10.8 was added to place a limit of 2500 psia on RCS pressure during the test.

The inspector stated that there were no further questions on this procedure.

(4) OP 1200051 "Nuclear and Delta T Power Calibration " and

.OP 1200021 "Incore-Excore Flux Monitor Correlation" Comments on these periodic/surveillance test procedures were discussed in IE Report No. 50-335/76-2,'etails II, paragraph 3.b. The computer is required to be in service in order to accomplish various steps, such as 8.2 of 1200021. These procedures no. longer have data sheets.

.The inspector stated that there were no further questions on, the procedures.

(5) TP 0110088 "Static CEA Dro Test at 50/ Power " and TP 0110089 "D namic CEA Dro Test at 50/ Power" Comments on these test procedures were discussed in IE Report No. 50-335/76-2, Details II, paragraph 3.c. The data sheets for these procedures which did not contain spaces for signoffs were deleted in Revision 2. The acceptance criteria (section 10 of each procedure) were rewritten in Revision 1 to make them more definitive.

The inspector stated that there were no further questions on these procedures.

b. . Emer enc 0 eratin Procedures Selected emergency operating procedures had been reviewed to determine whether they provided adequate instructions regarding alternate methods of supplying cooling water'o the reactor vessel in the event of an accident. As noted in IE Report No.

50-335/76-5, Details I, paragraph 5, comments remained outstand-ing on OP 0120042, "Loss of Reactor Coolant," and OP 0410030, "HPSI-Off Normal." OP 0410030 was revised in Revision 2 to give additional guidance on alignment of "C" HPSI pump in the

t I ~

IE Rpt. No. 50-335/76-6 'I-14 event the "A" or "B" pump failed to start. OP 0120042 was under revision at the time of the inspection to add instructions for use of a spent fuel system-charging system inter-tie, a backfit item. The inspector stated that he had no further questions on these procedures.

Prep erational Test Results The status of comments on completed test packageq, in addition to those discussed .in paragraph 2, were reviewed with licensee personnel.

(1) TP 0010184 "R'eactor Coolant S stem Com onent Ex ansion Ifeasurement" Comments on this test were documented in IE ReportNo.

50-335/75-19, Details I, paragraph 3.a. 'Review of licensee actions in response to those comments revealed the follow-ing information:

(a) Two letters, reviewed by the FRG, had been entered into the package to explain the details of how two sets of data had been recorded for step 12.3.1.5, data sheets, and to define which set was the for'everal proper one in each cas'e.

(b) The actual RCS temperatures at the time data were taken could not be retrieved; however, the HFT sequencing document provided data which demonstrated

~ that the temperatures were within the ranges allowed.

(c) Documentation of the dimensions for steam generator sliding bases, calculated'per step 12.3.1.8, had been reviewed and entered into the test package.

II The inspector stated that he had no further questions on

.this test.

(2) TP 0410082 "Safet In'ection Tank Dum Test" In IE Report No. 50-335/76-4, Details IV, paragraph 3.a, it was noted. that the safety injection tank pressure alarms had not been tested during the initial conduct of the test. They were tested on April 3, 1976, when the tanks were pressurized with nitrogen. Results of" the tests showed that'he setpoints 'for the low, low-low,

IE.'Rpt. No. 50-335/76-6 I-15 high, and high-high alarms were within the acceptance criteria of the procedure and the upper and lower limits of Technical Specification 3.5.1.d. (200-250 psig). The inspector stated that there was no further question on this test exception. The other comment, involving clarifi-cation of tank discharge times and the acceptability of them, will be reviewed in more detail during a subsequent inspection.

(3) TP 1010082 "Containment Instrument Air S stem Functional Test" In IE Report No. 50-335/l6-4, Details IV,'paragraph 3.b, it was noted that Table 9.3-1 of the FSAR contained an incorrect value (162 scfm) for the design capacity of these compressors. This value was changed to 48 scfm, the correct figure; in Amendment 57 to the FSAR. Licensee personnel were informed that there were no further questions on this test.

Core Protection Calculator Desi n Correction The changes necessary in the core protection calculator for the Thermal Margin/Low Pressure Trip, first discussed in IE Report No. 50-335/76-5, Details I, paragraph 6, were completed and the applicable reactor protection system channels retested prior to initial criticality. Licensee documentation xegarding this problem reviewed included Plant Vork Order (PRO) 3771, Plant Change/Modification (PCM) 20-76 'and'letters from CE to FP&L dated March 19 and 26, 1976, which explained the problem and need for changes. These modifications'nvolved a wiring change in the input to the CEA function summer and a change in a gain coefficient (resistor) in the circuitry. The PCM had been completed on April 15, 1976, and a retest of the channels was in pxogress pex OP 1400050, "Reactor Protection System Periodic Test." The inspector stated that there were no further questions on this subject.

IE Rpt. No. 50-335/76-6 II-1 DETAILS II Prepared by.'.

D. Martin, Reactor Inspector-Nuclear Support Section Reactor Operations and Nuclear Support Branch Dates of Inspection: April 7-9, 18-22, 1976

/

Revieeed by: C H. C. Dance, Chief Date Nuclear Support Section Reactor Operations and Nuclear Support Branch 1~ Persons Contacted Florida Power and Li ht FP&L K. Harris Plant Manager J. Barrow Operations Superintendent C. Hells Operations Supervisor P. Dillon Technical Staff Supervisor G. Vaux Quality Control Supervisor N. Roos Quality Control .Engineer A. Collier Instrument, and Control Supervisox G. Boissy Assistant Maintenance Superintendent Electrical J. Pride -.Technical Staff Engineer J. Garner Assistant Plant Engineer K. Beard - Maintenance Engineer P. Cheries - Chemistry Engineer D. Brandt Nuclear Plant Supervisor D. Nest Nuclear Plant Supervisor M. Windecker Nuclear Plant Supervisor R. Ryall - Reactor Supervisor D. Stutzman Reactor Engineer EBASCO Services Inc. (EBASCO M. Taylor - Construction Superintendent G. Lemon - Site Project Engineer H. Rogalski Mechanical Engineer M. Noronha Stress Analyst Supervisor J. Albanes Stress Analyst Supervisor

IE Rpt. No. 50-335/76-6 II-2 Combustion En ineerin (CE)

Oliver Startup Engineer Smith - CE Site Manager Si:ars - NSSS Test Manager

2. J ol.l wu on Previousl Identified Unresolved Item Installation and Testin of Han ers and Restraints (75-19/1)

Test results documented in the "Piping Thermal Expansion Post Core Load" preoperational test No. 0010194 were reviewed in total by the inspector. The inspector verifie'd that additional testing required by unresolved item (75-19/1) covering the monitoring of all mechanical and hydraulic snubbers and piping restraints during post core load heatup to rated temperature and pressure were completed and the results compared favorably with the predicted valves.

Repor'table Occurrence 335-76-9 documented the fact that during heatup for Post Hot Functional Testing five mechanical (I.N.C.

Bergen-Patterson Hodel HSVA-1) snubbers were found locked-up at the first plateau of 260 F., The immediate action was to promptly replace the defective snubbers with self relieving (Pacific Scientific) mechanical snubbers. The. inspector has since verified by discussions with the licensee, review and discussion of the Ebasco documentation and spot-checks in the plant, that all I.N.C. snubbers have now been replaced with

.the Pacific. Scientific nonlocking type mechanical snubbers.

During the heatup to rated conditions the inspector verified that the mechanical snubbers received 100% surveillance to assure that no other mechanical snubbers had locked up. This was accomplished by physically moving the piping and observing the snubber movement. In the cases where the piping could not be moved the snubber was unpinned and independently stroked.

A check sheet was used to verify that the snubber was then properly replaced and repinned. A followup report oh the failure mode of the five locked up I.N.C. snubbers is to be submitted by the licensee upon completion of the current investigation.

10 CFR 50.55(e) Final Report dated March 16, 1976, explained the fact that five of the six hydraulic seismic snubbers found by FP&L Power Resources Maintenance personnel to have apparent low fluid levels were really not low fluid levels but appeared low due to calibration 'plate misalignment. The sixth snubber leaked during static testing through a weep hole due to a

IE Rpt. No. 50-335/76-6 damaged accumulator piston 'seal. In all cases it was determined that the snubber would operate properly under design basis conditions since the low fluid levels (apparent or real) constituted a loss of reserve fluid and not fluid needed for damping action. The inspector verified that the six hydraulic snubbers discussed above were removed and returned to the Vendor's shop for inspection, repair and retest. The inspector also verified that the corrective actions outline'd in the 10 CFR 50.55(e) Final Roport dated March 16, 1976, have, been accomplished. Fluid level plunger indicator plates recommended by Rexnord Company are now installed.

Unresolved item 75-19/1 is closed.

3. Pre arations For Initial Criticalit Enclosure 1 to St. Lucie Plant Unit No. 1 License No. DPR-67 required certain items to be completed to the satisfaction of the NRC prior to achieving initial criticality. Item A-1 of the enclosure required the installation of tornado missile protection for the following systems:

a ~ Intake Cooling 4'ater Pumps

b. Component Cooling Mater Pumps C ~ Diesel Generator Air Intakes and Exhausts d Diesel Ge'nerator Access Doors

e. Diesel Generator Fuel Oil Pumps Auxiliary Feedwater Pumps

\

The inspector verified that the required tornado missile protection was completely installed prior to achieving initial criticality by physically inspecting in the plant each of the above listed systems.

Item A-2 of Enclosure 1 required the installation of a temporary tornado-protected connection between the spent fuel pool and the ch'arging system to provide makeup water to the reactor coolant system to accommodate moderator shrinkage upon plant shutdown. The inspector verified by personal observation that a tornado-protected connection was in place in the reactor auxiliary building prior to initial critical.

IE Rpt. No. 50-335/76-6 Item A-5 of Enclosure 1 required revision of the sequencing document for "Post Core Load Hot Functional" testing to include provisions for retesting of hangers and restraints. The inspector verified that the sequencing document was revised and that also preoperational test No. 0010194 "Piping Thermal Expansion and Restraint Post Core Load" contained additional'ections which covered monitoring all me:hanical and hydraulic snubbers and piping restraints during post co 'e load heatup to rated conditions.. Unresolved item 75-19/1 is clr>sed. Refer to discussion in Details II paragraph 2 of this re~>ort.

Item B-1 of Enclosure 1 required the installation. of control ci.:cuitry providing the capability to energize and de-energize the E>.':S Ifiniflow bypass valve operators (V-3659 and V-3660) from the c.;trol room. The inspector verified that the switches were installed i 'he control room and that they had been functionally tested for p:,.per operation.

I"... tial Criticalit S:. Lucie Unit 1 reached initial critical at 0830, EST on April 22, 3.=16. Preoperational Procedure No. 3200086 was followed to position

) he control element assemblies (CEA's)'n their normal sequence to

, roduce an essentially rod free core., The procedure was then used

.o direct deboration to criticality. Critical occurred as predicted

>hen the RCS was diluted to approximately 938 ppm boron. There

~ere no discrepancies identified by the inspector during the approach to initial critical. The items which the inspector witnessed as

>art of initial'riticality are summarized below

Prior to the start of CEA withdrawal the licensee met the

'applicable technical specification requirements of Sections 3/4 (Limiting Conditions for Operation and Surveillance Require-ments), Section 6.2 (Organization), Section 6.8 (Procedures) and Section 6.10 (Record Retention).

!.. The required nuclear instrumentation was operating and was properly calibrated. OP No. 1400050 and Plant Work Orders 68441 and 69557 were reviewed for'erification of calibration of the wide range 'monitors. Instrument calibration data sheets 229 and 230 were reviewed for verification of calibration of the Eberline startup scalers. The ESAR Section 14.1.4.4 requirement of at least 1/2 count per second minimum count rate was satisfied.

c. Trip checks of the wide range nulcear instrument channels was verified complete by reviewing OP 1210051.

~ a IE Ept. No.'0-335/76-6 Xl-5

'd ~ Properly approved procedures were used to provide a safe, organized method for attaining initial criticality. The count rate and .dilution data were properly recorded and the resultant 1/m plots developed as expected.

"1 ~

The prerequisites required prior to the start of CEA withdrawal were verified complete by reviewing appendix 'A'P 0030122 Reactor Precritical Check List.

During dilution the CEA pattern was independently verified.

All CEA groups were withdrawn except group 7 which was at 68 inches withdrawn. This pattern agreed with the appro'ved configuration.

I 4 ~; One coolant system boron sample and analysis during dilution was observed. The sample was taken and properly analysed according to FPL Chemistry Department OP C-358.

L~

During dilution the "HETRA SCOPE" used for CEA position indica-tion was lost for 30 minutes. The inspector observed that the licensee stopped dilution during th'e period the "IETRA SCOPE" was out of service.