IR 05000354/1985035
| ML20205B744 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 08/29/1985 |
| From: | Blough A, Chaudhary S, Lyash J, Strosnider J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20205B721 | List: |
| References | |
| TASK-2.F.2, TASK-2.K.3.21, TASK-2.K.3.22, TASK-2.K.3.27, TASK-TM 50-354-85-35, IEB-82-03, IEB-82-3, IEB-83-02, IEB-83-2, NUDOCS 8509120215 | |
| Download: ML20205B744 (16) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 50-354/85-35 Docket 50-354 License CPPR-120 Licensee: Public Service Electric and Gas Company Facility: Hope Creek Generating Station Inspection at: Hancock s Bridge, New Jersey Conducted: July 8 - August 11, 1985 8!!'/!8f Inspectors-c fr R. Blough, Senior Resident Inspector Date
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hrs (. Chaudhary, Senior Resident Inspector Date 8f2Vf93~
.e s
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. Lyash, Reactor Engineer Date B/Jf/8f Approved:
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trosnider, Chief, Projects Section IB Date
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July 8 - August 11, 1985 (Report No. 50-354/85-35): A routine onsite resident inspection (249 hours0.00288 days <br />0.0692 hours <br />4.117063e-4 weeks <br />9.47445e-5 months <br />) of work in progress and preoperational testing was conducted. The inspector also made tours of the site; reviewed licensee action on previous inspection findings, IE Bulletin, IE Circulars, Construction Deficiencies and TMI Action Plan Items.
One violation was noted involving three examples of specific preoperational test procedure inadequacy (Detail 6).
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PDR ADOCK 05000354 G
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Details 1.
Persons Contacted Public Service Electric and Gas Company
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- A. Barnabei, Principal QA Engineer
- F. Boppel, QA Engineer J. Carter, Startup Manager G. Daves, Senior Engineer, Operations R. Donges, QA Engineer J. Hagan, Operating Engineer
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- A. E. Giardino, Manager, QA i
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- R. Griffith, Principal Staff Engineer P. Kudless, Maintenance Manager
- S. La Bruna, Assistant General Manager C. Lambert, Site Engineering R. Lovell,' Radiation Prctection Manager P. Landrieu, Project Manager
- E. Logan, Site Manager
- M. Metcalf, Principal Startup QA Engineer
- J. A. Nichols, Technical Engineer J. M. Rucki, Maintenance Engineer R. S. Salvesen, General Manager, Hope Creek Operations
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Bechtel
- W. Goebel, QA Engineer
- C Jaffee, Startup Engineer
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D. Long, Field Construction Manager
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L. McGraw, Test Review Board Chairman W. Mourer, Construction Manager G. Moulton, Construction Manager
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K. Wilson, Startup Engineer D. Wood, Startup Engineer
- J. Zeruca, Startup Director 2.
Previous Inspection Item Update (0 pen) Unresolved Item (84-23-01), enhancement of instrument calibration and setpoint information. This item-was last reviewed in inspection 84-29. The inspector interviewed applicant engineers and reviewed
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procedures to assess applicant progress. The applicant has about thirty people,-most with BWR-experience, working on verification of instrument
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calibration data (ICD).
Personnel are kept current via weekly Training
Sessions. The program uses the latest BWR.0wners Group methodology and is governed by approved Site Engineering procedures.
The schedule for ICD verification is designed for completion of the verification prior to system preoperational testing. This schedule is being maintained; however, about six of the first system preoperational. tests were completed
before the ICD verification program was initiated--these ICD verifications
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are backlogged. The applicant is developing procedures and computer software to callout an instrument package for review any time one of its
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reference documents is revised. The applicant's program for ICD
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verification appears acceptable; the inspector will review this area for acceptable implementation prior to fuel load.
(Closed) 10 CFR 21 Report from Dravo Corporation (81-SB-01): Questionable MT of Weld Joints Fabricated by Dravo. On January.26, 1981, Dravo noti-fied the NRC of a condition regarding possible defective welds on piping
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subassemblies. The Dravo letter indicated that an inspector in their fabrication shop was found to be conducting a magnetic particle examina-tion of a weld on a nuclear piping assembly using a technique that did not conform to the approved procedure for such examinations. The letter also stated that as a result of this incident, all the magnetic particle (MT)
and dye penetrant (PT) examainations performed by that inspector on that
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date were re-examined, however, no rejectable indications were found.
In addition, Dravo. performed over 200 re-examinations on other works origi-j nally examined by the inspector in question..These re-examinations dis-
closed four-(4) minor indications in stainless steel base material adjacent to welds. These indications would be acceptable under the present code requirements.
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The piping assemblies inspected by this individual were shipped to the
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Hope Creek site prior to verification of this problem.
However, based on inspections after the 10 CFR 21 report it was determined that no significant defects were overlooked by failure to follow the approved MT procedure.
However, a potential problem in PT examination was not resolved.
With respect to the PT examination, Dravo researched all of the Hope Creek stainless steel piping subassemblies, and' determined that only three (3)
o subassemblies (spools) were' given a final PT examination in the " boss and attachment" welds. One of the unshipped spool was re-examined by Dravo and found to be acceptable. The other two (2) spools had already been shipped to site. Bechtel examined the only accessable weld at the
jobsite, and found it to be acceptable. Two (2) other attachment welds on
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a water stop ring were embedded in concrete and could not be re-examined.
However, the pressure integrity of the spool was verified by a successful
hydrostatic test.
Based on the above information, inspection and review of supporting documentation and discussions with cognizant personnel this matter is considered resolved, and the inspector has no further questions in this i
regard. This item is closed.
3.
IE Bulletin and Circulars (Closed) IE Bulletin 82-03 (82-BU-03),\\ stress corrosion cracking in BWR
. piping. This bulletin was addressed only to certain operating units. This i
issue is covered for Hope Croek by IE: Bulletin 83-02 and an SER open item.
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(Closed) IE Circular 81-07 (81-CI-07), control of contaminated material.
This item was inspected in report 85-14. The inspector reviewed Administrative Procedure 24Q, Revision 2, Radiological Protection Program and verified that the cross-referencing of procedure sections to the circular had been clarified.
4.
TMI Acticn Plan (TAP) Items (Closed) TAP Item II.K.3.21, Restart of Core Spray and LPCI Systems. This item was reviewed by NRC:NRR in the licensing process.
In the SER NRC accepts the determination by the applicant and BWR Owners Group that no logic change is warranted.
This item is closed.
(Closed) TAP Item II.F.2, Instrumentation to Detect Inadequate Core Cooling.
In SER Sectin 4.4.7 the NRC. staff accepted the existing Hope Creek reactor water level instrumentation as adequate to satisfy this item. The inspector reviewed FSAR Table /.5-1 and selected instrument drawings, and observed control room indicators to verify the instrumentation is as described in FSAR. This item is closed.
(0 pen) TAP Item II.K.3.27, Common Reference Level.
In FSAR Section'1.10, Amendment 10, the applicant committed to establishing a common reference level for all reactor water level instruments; this was accepted in SER Supp1 mat I, Section 18.3.
In reviewing design documents and observing instruments pursuant to Item II.K.3.27 (above), the inspector noted that fuel zone level instruments have a different reference from other-instruments. Applicant engineer's were aware of the need to address this item; it remains open.
(Closed) TAP Item II.K.3.22, Automatic Switchover of RCIC Suction. This item was inspected in inspection 85-27, with exceptions of some parts of the plant walk-through. The inspector completed the walk-through and had no further questions.
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Construction Deficient Reports (CDRs)
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(Closed) Construction Deficienty Report (83-00-08),nfatity valve actuators on twelve BIR-butterfly valves. The inspector reviewed drawings for the modified valves, which showed stiffening plates at the actuator cent'
body to preclude twisting during plant-induced vibration or seismit events. The inspector observed a sample of four valves in plant te tarify that the modifications had been done.
(Closed) Construction Deficiertcy Report (83-00-09): Defective Capstan
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Springs. On October 24, 1983, the licensee reported to NRC:RI a poten-tially signficant construction deficiency related to defective capstan springs in mechanical shock arrestors supplied by ITT-Grinnell.
Per the advise of the manufacturer (Pacific Scientific) Eighty-six (86)
mechanical shock arrestors were returned to them for inspection and necessary rework. After inspection and evaluation, the manufactured
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identified twenty-five (25) shock arrestors for replacement of capstan springs. After necessary replacement of springs and other rework all shock arrestors were returned to site with the manufacturers certification and other pertinent documentation.
The inspector reviewed the documentation and verified the above information. This item is closed.
(Closed) Construction Deficiency Report (84-00-09): Under-rated cabies -
Valcor Solenoid Valves. The i~nspector reviewed documentation, held discussions with cognizant personnel and verified by visual examination that the deficiency regarding the under-rated cable connection in Valcor Solenoid Valves was analysed and corrected. A Design Change Notice (DCN)
No. 37 was issued against electrical drawing no. E-1408 providing for the replacement of the under rated cables with cable capable of withstanding the higher operating temperature (150 degrees C). The work was controlled by means of Nonconformance Report (NCR) No. 4522. The completion of the work and acceptance thereof was documented by QC on 2/15/85.
This CDR is closed.
(Closed) Construction Deficiency Report (84-00-12) BIF Butterfuly Valve Material. The licensee informed NRC-RI that a routine cleanliness inspection of two (2) thirty-six inch (36") butterfly valves purchased from BIF Industries indicated extensive corrosion damage to the tapered pins securing the valve stems to the disc. A review of the BIF fabrication drawing indicated that the pins were to be made of Monel (ASME SB 164, cl. B). The chemcial analysis of the corroded pin, however, indicated that the material was series 400 stainless steel.
This type of stainless steel is galvanically dissimilar to the aluminum-bronze disc material, and therefore is susceptible to galvanic corrosion.
A total of sixy-nine (69) valves of various sizes were identifed to have aluminum-bronze discs.
The discrepancy was documented, and the corrective action was controlled by the Nonconformance Report nos. 3711 and 5360. By use of a simple nitric acid test the licensee determined that this discrepancy was limited to the four (4) thirty-six inch (36") valves.
The disc pins on the two thirty-six inch (36") valves that were installed in Unit-1 were replaced, and the licensee issued a QA Deficiency Report No. HQA-85-001 to document and control the stored valves from Unit-2.
This item is closed.
6.
preoperational Phase Activities 6.1 Plant Tour The inspector toured the control room on regular and backshifts. He interviewed operations personnel regarding testing scheduled or in progress, reviewed logs and night orders, and observed alignment and indications of systems undergoing tests. Operators and supervisors were knowledgeable regar, ding plant status and test plans.
The inspector toured areas of the plant, including dry-rell, reactor building, and the control building.
He checked oc tests and
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operations in progress, observed equipment and housekeeping conditions, and interviewed personnel involved in ongoing activities.
The inspector followed up on the following observations.
6.1.1 During a plant tour on July 16, the inspector observed Bisco
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sealing material leaking from an electrical penetration in the process of being poured. The sealing material escaped from the penetration dam, flowed down the reactor building wall and onto reactor building floor elevation 102', immediately adjacent to the north Scram Discharge Instrument Volume and Hydraulic
Control Units. The inspector informed the applicant of the spill and requested that the safety related equipment in the area be examined for contamination by the sealant.
Investigation by the applicant revealed that Q-listed level transmitter 1-BF-LT-N0120 had been affected. Startup Deviation Report BF-0217 was issued
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l to cleanup the spill and the affected Level transmitter.
The inspector pointed out that this was the second incident of sealant spill on safety related equipment identified in recent weeks (see IR 50-354/85-27, para. 6.1.3).
Corrective actions proposed and implemented by the applicant include placing protective covering on all permaner.t plant equipment,'applicantion and curing of an initial base coat of sealant to aid in leak tightness and stationing of an in process pour watch.
In' addition, supplementary training on proper damming and pouring of penetrations was conducted.
The inspector considers these actions satisfactory.
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routine tours the inspectors will evaluate the effectiveness of these measures.
6.1.2 On July 23 and 24,1985, the inspector observed portions of the
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control rod drive system fill / vent process. The inspector witnessed operations both in the main control room and at the hydraulic control units. Activities were.being conducted in accordance with an approved procedure.
The inspector questioned the test engineer regarding control rod blade support and water chemistry prerequisites, and coordination of CRD fill / vent activities with other NSSS preoperational testing. The inspector determined that prorequisites had been adequately addressed and that proper coordination between the test engineer and operations existed.
6.2 Preoperational Test Procedure (PTP) Review The inspector reviewed the following preoperational test procedures for verification that planned testing fully demonstrates that system operation and response meet regulatory requirements and applicant commitments. The review also verified proper format and content, and applicant review and approval of the subject procedures. Three
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examples of violation of Appendix B Criterion XI were identified and detailed below.
6.2.1 E-1, Revision 0, July 16, 1985, Fuel Handling and Vessel Servicing Equipment.
Findings were as follows:
6.2.1.1 The current FSAR states that all tools for servicing the vessel will be tested in the PTP.
KE-1 does not test all tools. The applicant explained that he had decided instead to test the tools using maintenance procedures prior to use. He had initiated an FSAR change on June 17, and PORC is tracking the change to approval.
If the change is not approved, the PTP would need to be explained.
6.2.2.2 One objective of the test is to verify operability of all refueling interlocks. One such interlock is "With the Reactor Mode Switch in Startup' position, all rods in, refueling platform hoists and service plat-form Jib Crane unloaded, refueling platform cannot be moved over the reactor vessel (FSAR Table 7.7-3 Situa-tion II)".
Although this interlock was included as an acceptance criterion for the test, the corresponding step-by-step instruction in Sectin 8.3.10 of the pro-cedure did not provide for proper testing of the inter-lock.
Basically there were no steps to restore from a previous test condition, involving one rod out and one hoist loaded, and establish the proper interlock test conditions.
10 CFR 50 Appendix B Crieterion XI, Test Control requires a testing program to assure all testing, including preoperational tests, required to demon-strate that structures, systems, and components will perform satisfactorily is identified and performed in accordance with written test procedures which incor-porate the requirements and acceptance limits of design documents.
FSAR Section 14.2.12.1.48, Fuel Handling and Vessel Servicing Equipment, and the GE Preoperational Test Specification for Hope Creek (No-A41-3300-32(1)-3:
22A2271AZ, Revision 2) state that all interlocks associated with the refueling and service platforms will be verified and included as acceptanco criteria in the preoperational test (PTP).
Failure to identify steps to properly test a refueling interlock is one example of violation of Appendix B Criterion XI
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(85-35-01). When informed by the inspector, the applicant changed the procedure to add the appropriate steps.
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6.2.2.3 Several other minor errors or inaccuracies were noted
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in KE-1.
The inspector stated that these were not
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individually significant but could indicated or lack of attention to detail:
One example where the test would not have worked
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as written because the proper reactor mode switch position was not specified (step 8.3.10.6).
This did not affect test acceptance criteria.
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Two examples where features were to be checked for multiple cranes, but the acceptance criteria did not reference the procedure steps for all cranes. The body of the procedure did actually test all appropriate cranes, however.
One example where the acceptance criteria for an
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interlock improperly described the condition being tested (Criterion 5.4) relative to control rod positions.
The corresponding procedure steps were correct, however.
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Three cranes where the Acceptance Criteria referenced the wrong procedure steps--these appeared to be typographical errors.
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i The applicant corrected all the above items through On-the-Spot Changes or Change Notices. The degree of attention to detail will be continually evaluated in the preoperational phase inspection program.
6.2.2 PTP-SV-1, Revision 0, Remote Shutdown Panel.
In review of PTP-SV-1, Pemote Shutdown Panel (RSP), the inspector noted the absence of Pain Control Room RSP Takeover alarms in the acceptance criteria portion of the test. The HCGS FSAR 14.2.12.1.54.d.2 specifies that these alarms be included as acceptance criteria. The alarms must function to alert the control room operator that the ability to control certain equipment from the main control room has been lost. The inspector' discussed this finding 4 -
with the applicant.
Further investigation by the applicant identified seven additional egipment takeover lamps which had not been included for testing.
Preoperational Test Change Notice (CN) was issued, and approved, to add the acceptance criteria and' test steps necessary.
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The. inspector also noted that as stated in PTP-SV-1 step 8.3.9.1.c, valve 1EG-HV-2520 8, SACS cooling inlet to RHR pump B seals and motor bearings, was to be tested in PTP-BC-1. Review of PTP-BC-1 indicated that the valve had not been included for test. The applicant, in response to the inspectors findings issued On-The-Spot Change (OTS) 2 to PTP-SV-1, incorporating the valve for testing. Valve IEG-HV-25208 is one of a number of valves which are designed to open automatically upon transfer of control from the Main Control Room to the Reniote Shutdown Panel, as described in FSAR Amendment 11, section 7.4.1.4 and Table
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7.4-2.
The inspector informed the applicant that FSAR
section 14.2.12.1.54.d.1 states that the operation of all valves, designated in section 7.4.1.4 and Table 7.4.-2, from the RSP will be tested and included as acceptance criteria in the PTP, At this point the applicant issued Preoperational Test Change Notice 2 adding test steps and acceptance criteria addressing this valve.
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The inspector informed the licensee that the above examples constitute a_ violation of 10 CFR 50, Appendix B, Crieterion
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XI, which states in part, that a test program be established to assure that testing is conducted in
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accordance with written test procedures which incorporate the requirements and acceptance limits contained in the
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applicable design documents. (85-35-02).
Steps 8.2.20.1 through 8.2.20.20 of PTP-SV-1 specify the
" adjustment" of the output of two calibrated pressure switches, and the subsequent " readjustment" of one switch.
The inspector questioned the type of adjustment made, the effect of the readjustment on the as-left calibration of the instrument, and the absense of any restoration on the second pressure switch.
Discussion with the applicant resulted in issuance on an OTS prescribing a detailed method of adjustment and restoration of both pressure switches using a 1-5 Vdc power source.
6.2.3 PTP-BC-1, Revision 0, Residual Heat Removal.
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During review of PTP-BC-1,-Residual Heal Removal, the inspector noted that acceptanco criteria 5.16 specifies 4.5 ft. NPSH in-full flow test.- General Electric Test Program d
l Specification 22A2271 AZ,- B5.5.7 specifies NPSH of 4.5 ft.
in' addition to pump vendor' requirements, under similar conditions. 'The pump vendor requirement, as provided by the applicant, is 4.5 ft..Thus total required NPSH is 9.0 ft, for the given conditons. The inspector questioned.the apparent discrepancy between the PTP and GE acceptance criteria. The applicant demonstrated that the method used to calculate NPSH in the test procedure accounts for this'
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difference, and that by obtaining a calculated NPSH of 4.5 ft. in the test the actual NPSH criteria of 9.0 ft. is met.
The inspector also noted that acceptance criteria 5.12 of PTP-BC-1 states that LPCI injection flow at 180 degrees F will not exceed 11,000gpm.
The GE Test Specification, B5.5.5, specified maximum flow as 10,5000 gpm. for this mode and condition.
In response to the inspector's question the applicant provided a copy cf General Electric FDDR KJ1-234. This document, based on engineering analysis,-increased the maximum flow from 10,500 gpm. to 11,000 gpm.
Test objective 1.2 of PTB-BC-1 states that flow paths from RHR to the sampling system, radwaste, containment spray test connection, and fuel pool cooling are verified in GTI-01M-BC-01. The applicant has commited, in FSAR Section 14.2.12.1.5, to test these flow paths.
The inspector questioned the mechanism which assured that this GTI would receive the results review required for tests which satisfy FSAR commitments. The applicant pointed out the OTS #1 added step 9.3.5, to the procedure completion section of the PTP, requiring the GTI to be attached for results review.
Startup Administrative Procedure 24 (SAP),
Revision 6 requires those General and Detailed Tes;_
Procedures, and General Test Instructions referenced as prerequisites or specified as acceptance criteria to meet FSAR commitments to be included with the PTP for results review.
The inspector pointed out that referencing the GTI in the test objectives or procedure completion section did not agree with the practice perscribed in SAP 24. The applicant initiated a revision to SAP 24 to also include the results review requirement test procedures referenced in the PTP procedure completion section.
6.2.4 In review of preoperational test procedures the inspector questioned the applicant as to whether instrument accuracy tolerance bands are taken into account in meeting acceptance criteria.
In those cases where qu(ptitative test acceptance criteria are specified the test. engineer could, using a calibrated instrument, measure and record a marginally acceptable valve, when in fact the valve might be unacceptable due to within-tolerance instrument error.
The applicant committed to revise Startup Administrative Precedure 24, Preoperational Test Procedure, Format and Instructions, to require evaluation of the effect of instrument inaccuracy on acceptance criteria during the results review cycle. The applicant is currently
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i implementing this requirement. The inspector will review changes made -to SAP 24 and will verify, on a sampling
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basis, implementation of the revised procedure (85-35-03).
I 6.2.5 The inspector reviewed water chemistry sampling and analysis as it applies to preoperational phase testing activities. The inspector reviewed the program to ensure
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(1) that proper samples are taken and evaluation conducted
prior to injecting water into safety related equipment, (2)
that timely notification of proper personnel is made when water quality exceeds specifications, and (3) that chemistry
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reports used to satisfy test prerequisites are traceable,
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i The applicant indicated that water quality limits and i
notification requirements are. detailed in procedure PSSUG-GTP-1, General Flushing and Cleaning Procedure. The inspector noted
that PSSUG-GTP-l'does not require any notification unless water a
has exceeded the specification for 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />. The high tempo of
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Startup test activities makes a 96 hour0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> notification limit i
unacceptable.
In response to this concern the applicant has i
committed to revise GTP-1 to provide for immediate verbal notification, and followup written notification, of the Startup-i j
Manager if water quality is found to exceed specification.
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. Procedure revision and training'of personnel will be completed l
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The-applicant indicat'ed that routine sampling was being
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conducted in accordance with Station Chemistry Technical Instruction CH-TI.ZZ-012(Q).
In addition.special= samples or an increase in sample frequency may be requested by Startup For
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Critical Systems.
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4-The inspector will verify, on a sampling basis, applicant
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implementation of the improved startup testing water chemistry
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program improvement above (85-35-04).
6.3 Preoperational Test Witnessing-if On several occasions during the reporting period the inspector witnessed preoperational testing 'of the Core Spray System under
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PTP-BC-1, Revision O. 'The_ inspector reviewed sections of-the Master
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i Test Copy including On-The-Spot Changes, Test Exceptions and.
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_preraquisites. Test sections observed-include valve interlock, valve
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stroke and system initiation logic _ testing.
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Prerequisite 3.6 of PTP-BC-1 requires the test' engineer to inspect
'3 all; system relief valves within the test boundary prior to test A
start. During a plant' tour the inspector'noted that the lead seal on 8PSV-F032C was not intact, and that the applicable prerequisite had been-signed as complete. The ins ~pector determined that the test
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engineer had signed the step as complete based on an inspection conducted approximately three weeks earlier, under a test package (TPR). Discussion with members of the Startup Quality Assurance Group and the Public Service Startup Group (PSSUG) confirmed that this did not meet the intent of the prerequisite.
The applicant indicated that a visual inspection to ensure system relief valves are intact at the time of step signoff was required. A change to Startup Administrative Procedure 24 clarifying this requirement was initiated.
6.4 Committee Activities The inspector attended a Preoperational Test Review Committee (PORC)
meeting on July 23 to verify that quorem requirements were met and that business was conducted properly and in accordance with procedures. No problems were noted.
7.
Construction
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7.1 Plant Tour and Walk-through Inspection The inspector periodically toured the plant and performed walk-through inspections during this inspection period.
In the walk-through inspection, special emphasis was placed in the areas of drywell, reactor builcing torus /wetwell, and diesel generator buildings. These inspections were carried-out to assess the level of general workmanship in the areas of-piping and pipe support; effectiveness of cleanliness and housekeeping program; and general conformance to project procedures in the work-in progress and the completed work.
The inspectors also verified the a/ailability of approved project procedures, specification, drawings, and the QA/AC personnel in the general work areas for works-in progress. The completed work was examined for any evidence of obvious defect in material or workmanship.
No violations were identified.
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7.2 Control Rod Drive -Piping and Supports - Visual Examination The inspector examined control rod drive (CRD) exhaust and drive water piping from the Hydralic Control Units through the drywell and reactor pedestal penetrations, and to its termination at the CRD mechanisms. Piping was examined for non-uniform or acute bending, and pipe constriction due to bending.
All piping examined contained bends of 90 degress or less. The
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inspector found no indications of kinking or pipe constriction. All l
bends examined were of acceptable bend radius with no indication of
- excessive stress.
The Control Rod Drive Hydralic system, as the name implies, utilizes
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hydralic pressure to position the control rods. The flow through any individual drive or exhaust water line is very small. Therefore minor variations in number and type of bends, from pipe to pipe, would have no significant effect on system operation. The installation of the piping and supports was apparently acceptable.
No violations were identified.
7.3 Rainwater Instrusion in the Power Block On July 31, 1985, a heavy rain storm created a minor flooding condition on site. The licensee identified several locations in the power block where water intrusion had occurred..The resident inspectors also ' noticed the water logging of up to approximately six inches (6') in diesel general enclosures and radwaste building at elevation 54.00. Although several pumps were'put into service to control the flooding, the water. level was still rising during the inspector's walk-through inspections.
The licensee, however, had pressed several teams of operation, maintenance, and engineering
. personnel to assess the impact of this event and take corrective actions to mitigate the effects.
The inspectors observed that the licensee had deenergized all equipments on elevation 54.00, and there were sufficient personnel and equipt2nt available and put into service to control the water intrusion.
During this inspection, the inspectors also noted that a non-safety related electrical conduit at elevation 54.00 in diesel generator enclosure was leaking substantially at one joint, indicating an interior flooding of the conduit.
The inspectors determined that the licensee was aware of it, and had intitiated an inspection to determined the source of water in the conduit and any effect it had on the other systems.
In light of the flooding of this conduit the licensee was also examining the possibility of internal flooding of safety related conduit.
On August 6, 1985, the inspectors reviewed the licensee's investigation and evaluation of the event.
It disclosed that the water intrusion in the buildings were due to the construction openings left in the outside walls of the power block for construction convenience. These openings would be all sealed by appropriate methods once the need for'them has been eliminated.
Also, it was determined that the inspector identified conduit (14PRLD03) terminated at the junction box no. 14PJE00A, which was mounted over a sleeve on elevation 77.00 in the diesel general enclosure. The sleeve ran out to an electrical manhole on the west
. side of.the building that was flooded due to the surface runoff.
Hence, the water started leaking into the junction box and the conduit in question. The licensee plans to' seal the sleeve and eliminate any future flooding potential.
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The inspectors had no further questions at this time.
l 7.4 Storage of Materials and Equipment The inspector reviewed the documentation, held discussions with i-cognizant licenee and contractor personnel, and performed visual inspection of the project warehouse, outside laydown area, and the
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on-site fabrication shop to determine the licensee's compliance with
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project requirement and materials. The inspector reviewed the following documents:
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Bechtel ASME QA Manual, Section 13, Rev. 3
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Bechtel Project Procedure WP/P-14, Rev. 3 i
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PSE&G Site Audit Nos. H-348; H-296, H-322
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Bechtel Nonconformance Report Nos. 3107;
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Based on the review of' documentation and the walk-through inspection i
inspectors determined as follows:
a.
The materials and equipment in the warehouse were properly stored in controlled environment.
b.
The equipment and materials ware properly stored on shelves, bins, and/or designated areas, and were prominently identified.
c.
Storage areas for different material was clearly marked and adequate control was provided.
d.
Outdoor lay down area for structural and miscellaneous steel was properly identified and controlled.
e.
The stock materials in the lay down area was clearly identified
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by heat and order numbers.
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f.
The material was properly stored on dunnage, the pipe ends were protected by end caps, and stainless steel pipe was covered by
. protective fabric to provide weather protection.
g.
The fabrication shop also had proper storage and identification
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of material.
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h.
The inspector witnessed one heat number transfer from large
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steel plate in stock to a small piece.
The inspector also
. verified several other small pieces of material where the heat
number had been transferred previously.
1.
The traceability control programLfor materials (by heat number
or other means) was adequate and effective, i
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No violations were identified.
7.5 Bonny Forge Pipe Fittings Requiring Chemcial Over Check On June 13,1985, Dravo Corporation informed Bechtel that Bonny Forge had brought to their attention specific heat number materials for which chemical overchecks were not completed as required by NCA-3867.4(e)Z. The inspector also received a request from NRC:IE-HQ to verify the status of the material and the disposition, if any, of the prcblem.
The inspector reviewed the documentation and discussed the information with licensee personnel.
Following documents were reviewed:
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Letter from Dravo to Bechtel dated 7/13/85
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Dravo Packing List and Invoice, QA control no. 71-
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Bechtel QC Inspection Report No. P.201-98924
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Bechtel Nonconformance Report no. 2244
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Bechtel Pipe Inventory File (Lebled Salem-3)
Based on the above, the ' inspector determined that the Hope Creek site had received six (6) fittings of the size 36"xHx24"STD, Heat number 517BB and two (2) pieces of size 24"STDx36:xH, Heat no. 434B8. The
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inspector also observed that the materials in question were scrapped in September, 1983 prior to installation. This was documented on
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'NCR-2244. Therefore, it appears that Hope Creek does not have the
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questionable material installed, and no further action is required to i
resolve this issue.
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No violations were identified.
8.
Transition from Constructio, Phase to Operations
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During this inspection period, the QA Department was transferred from the Engineering and Construction Department-to.the Nuclear Department in
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accordance with the applicant's Transition Plan.
This transition appeared to progress smoothly with QA tersonnel maintaining knowledge of their
l responsibilities and adequate QA/QC coverage and overview of activities.
9.
Power Ascension Testing Program
The-inspector participated 1.s a meeting at NRC Headquarters on August 1
wherein the applicant-and General Electric Company outlined plans for a joint effort,to streamline the Power Ascension Test Program. The applicant was encouraged to expedite submittal of desired program change
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descriptions and FSAR amendments for NRC:NRR review. Specific test procedures will be subject to review by NRC Region I inspectors.
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11.
Exit Interview The inspectors met with applicant and contractor personnel periodically and at the end of the inspection report to summarize the scope and findings of their inspection activities. Written material was not provided to the applicant.
Based on Region I review and discussions with the licensee, it was determined that this report does not contain information subject to 10 CFR 2 restrictions.
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