IR 05000354/1985036

Insp Rept 50-354/85-36 on 850722-26.No Violation Noted.Major Areas inspected:as-built Sys Comparison,Qa/Qc Interface W/Preoperational Testing,Plant Tours & Independent Measurements

ML20137F690
Person / Time
Site:	Hope Creek
Issue date:	08/20/1985
From:	Briggs L, Cheh U, Eselgroth P, Marilyn Evans NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20137F651	List: IR 05000354/1985036 ML20137F646
References
50-354-85-36, NUDOCS 8508270052
Download: ML20137F690 (10)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report No. 50-354/85-36 Docket N License No. CPPR-120 Licensee: Public Service Electric and Gas Company 80 Park Plaza - 71C Newark, New Jersey 07101 Facility Name: Hope Creek Generating Station Inspection At: Hancock's Bridge, New Jersey

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Inspection Conducted: July 22 - 26, 1985 Inspectors: Er/2e/ #f L. Briggg Lead Reactor Engineer ' date

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M. Evans, eactor E gineer '

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Td V U. Cheh,'Re ' tor Engineer '

date Approved by: y 4 I P. Eselgropf, Chief da'te >

Test Programs Section, DRS Inspection Summary: Inspection on July 22 - 26, 1985 (Report No. 50-354/85-36)

Areas Inspected: Routine, unannounced inspection (118 hours0.00137 days <br />0.0328 hours <br />1.951058e-4 weeks <br />4.4899e-5 months <br />) by three region-based inspectors of preoperational test procedure review and verification, pre-operational test witnessing, as built system comparison, QA/QC interface with preoperational testing, plant tours, and independent measurements and verifica-tio Results: No violations were identifie PDR 0 ADOCK 05000354 PDR

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DETAILS 1.0 Persons Contacted

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• A. Barnabei, Principal Quality Assurance (QA) Engineer

• F. Boppel, QA Engineer

• J. Carter, Startup Manager P. Craig, System Test Engineer (Reactor Manual Control System)

• E. Dalton, Site Engineering

• R. Donges, QA Engineer D. Evans, Lead QA Engineer J. Fisher, QC Supervisor T. Garrett, System Test Engineer (Core Spray)

• A. Giardino, Manager QA, Engineering and Construction

• R. Griffith, Principal QA Engineer

• C. Jaffee, Startup Engineer

• E. Logan, Site Manager, Construction

• M. Metcalf, Principal QA Startup Engineer

• G. Moulten, Principal QA Engineer S. Slickers, Startup Engineer (Reactor Protection System)

E. Yochheim, Chemistry Engineer Other NRC Personnel

• S. Chaudhary, Senior Resident Inspector, Construction

• S. Peleschak, Reactor Projects Intern

• Denotes those present at the exit meeting conducted on July 26, 198 The inspector also contacted other personnel of the licensee's operating and QA/QC staf .0 Preoperational Test Procedure Review and Verification 2.1 Scope The PTPs listed below were reviewed in preparation for test witness-ing, for technical and administrative adequacy and for verification that testing is planned to adequately satisfy regulatory guidance and license commitments. They were also reviewed to verify licensee review and approval, proper format, test objectives, prerequisites, initial conditions, test data recording requiremants and system re-turn to norma PTP-SB-1, Reactor Protection System, Revision 0, approved July 3, 1985;

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PTP-SF-1A, Reactor Manual Control System, Revision 0, approved June 14, 1985; and

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PTP-BG-1, Reactor Water Changes and Filter Demineralizer System, Revision 1, approved July 12, 198 .2 Discussion 2.2.1 PTP-SB-1 This prccedure was reviewed in detail by the inspector by comparison (approximately a 50 percent sample) of logic testing, as stated in the procedure, with reactor system schematic prints. In addition, each print was marked up by the inspector to indicate the section of the procedure that tested each protective system function. The inspector had several questions concerning the procedure and the method of testing on several system functions which were discussed with the licensee and resolved. Only one minor problem concerning an incorrect voltage supply value on the main steam line radiation monitor required correction by the licensee. This will be corrected by an on-the-spot change notice when the test is implemented. The inspector also compared the acceptance criteria of G.E. Preoperational Test Specification 22A2271AZ, section B21, Revision 2 and Chapter 14 of the FSAR to PTP-SB-1. The response time testing of the reactor protection system will be conducted during PTP-SB-2, Response Time Testing, which is currently being writte .2.2 PTP-BG-1 This procedure was reviewed by the inspector using the references and P and ID's listed in attachment A. The inspector inspected the system accompanied by a licensee representative to verify that system components were installed in accordance with referenced prints and that jurisdictional tags were attached in accordance with approved procedure .2.3 PTP-SF-1A During the review of PTP-SF-1A, Reactor Manual Control System, the inspector compared the test's acceptance criteria to the acceptance criteria of G.E. Preoperational Test Specification 22A2271AZ, Rev 3, Section 89 and HCGS FSAR, Chapter 14, Section 14.2.12.1.43, Reactor Control. The inspector also compared the rod blocks tested in SF-1A with those rod blocks required in HCGS Draft Technical Specifications, Table 3.3.6-1 and HCGS FSAR Chapter 7, Section 7.7.1.1. At the com-pletion of this review, it appeared to the inspector that numerous rod blocks were not tested in SF-1A. After discussion with the system startup test engineer (STE) and review of PTP-SE-3, Rev. O, Power Range Neutron Monitoring System, the inspector concluded that all rod blocks had been tested with the exception of the rod blocks in the shutdown mode and the upscale, inoperative and comparator rod blocks for the reactor coolant system recirculation flow. Also it appeared that the interface of the reactor coolant system recirculation flow or

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rod blocks with the reactor manual control system had not been tested in PTP-SE-3. Further discussions with the STE did not enable the inspector to determine whether the rod blocks were tested in any other test procedur .3 Findings No unacceptable conditions were identified during the procedure reviews discussed in Paragraphs 2.2.1 and 2.2.2. The results of the procedure review discussed in Paragraph 2.2.3 remain unresolved (354/85-36-01)

pending licensee action to determine where and how interface and rod block testing is/has been accomplished and subsequent NRC revie .0 Preoperational Test Witnessing 3.1 Scope Testing witnessed by the inspector included the observations of overall crew performance stated in Paragraph 3.0 of Inspection Report 50-354/85-1 .2 Discussion During the entrance meeting on July 22, 1985 the licensee informed the inspector that preoperational test BE-1, Reactor Core Spray, was in progress but on hold at the time because of a ground in the DC power system. During the week the inspector observed portions of logic testing and some breaker timing tests being conducted. Testing was being conducted in accordance with the criteria of Paragraph above with full QC coverage during the portions witnessed by the inspector. The inspector also reviewed the portione of the PTP that were completed including all test exceptions and procedure change The procedure changes were minor ones not affecting the procedural intent. The inspector discussed several test exceptions dealing with improper timing of the core spray (CS) pump breakers after a simu-lated diesel breaker closure. The CS breakers are required to close in 6 seconds or less after the diesel breakers closure. To date all breakers tested have taken greater than six seconds. One breaker was of particular concern to the inspector since its tiine was 6.05 sec-onds as timed by stop watc The STE agreed with the inspector and stated that since all three breakers (3 of 4) to date had exceeded the six second criteria that all four breakers would probably be readjusted or a more accurate time measurement taken to prove accept-abilit This item will be reviewed during future routine inspection .

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3.3 Findings No violations were identifie .0 As Built System Comparison 4.1 Scope The inspector walked down the "A" line of the Condensate and Feed-water Systems, from the condenser to the feedwater penetration into the drywell . Using system isometrics (IS0's) and piping and instru-mentation drawings (P&ID's) for each system, and numerous pipe sup-port drawings, the inspector verified that the as built configura-tions corresponded to the appropriate drawing The documents listed in Attachment B, were used for this inspectio .2 Discussion The inspector found the components, pipespools, reducers, hangers, snubbers and other miscellaneous equipment easily traceable to the applicable P&ID's and/or ISO's. The inspector noted that hand valve F0328 and pipe supports AD-020-H28, AD-020-H29, and AD-020-H30 were not installed as indicated on the system isometric drawings. After discussion with the licensee, it was determined that hand valve F0328 had been removed for maintenance per SDR AE-286 and that the three hangers had been deleted in FCR P-9092. The inspector also noted that support AD-020-H03 appeared to have been installed on the incorrect side of a wall according to ISO-1-P-AD-04. The inspector reviewed drawings C-1635-1, Rev. 6, Turbine Building Floor Plan, Elevation 102'0", area 11 and 1-P-AD-020-H03, Rev. 3, Pipe Support, Turbine Building, Condensate from feedwater heater No. 2 to reactor feed pumps, and verified that the pipe support was installed in the proper location. Further discussion with the site Senior Construc-tion Resident indicated that walls sketched on system isometric draw-ings serve only for orientation and are not indicative af reference points for piping measurements unless designated as suc The inspector examined numerous pipe supports to verify their con-struction according to the corresponding pipe support drawings. The inspector independently verified dimensions, stability, identification numbers, thread engagement and fasteners to ensure compliance with design drawings and specification .3 Findings No unacceptable conditions were identifie _ . -- .

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5.0 QA/QC Interface with Preoperational Test Program 5.1 Discussion The inspector reviewed recent QA surveillance reports (QASR)

regarding different activities of the licensee's startup group. The following QASR's were reviewed:

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QASR-3578, 120 VAC CLASS IE preoperational test, conducted July 20, 1985. The QA inspector witnessed several steps of this test until a loss of synchronization occurre This resulted in the generation of test exception #10 to PTP-PN- SDR-PN-133 was initiated to permit trouble shooting of the problem. The QA inspector also noted that various steps of the procedure would have to be redone since loading is to be maintained for the duration of this section of the tes QASR-3636, Core Spray pump B, breaker closing time, conducted July 24, 1985. The QA inspector noted closing time to be 8.98 seconds. Required time was less than or equal to ten second The QA inspector also noted that measurement and test equipment in use were calibrate QASR-3627, Core Spray preoperational test, conducted July 24, 1985. The QA inspector witnessed several steps of this tes On the spot (OTS) change #13 to PTP-BE-1 was issued and approve Test exception #35 to PTP-BE-1 was also issued because the Lamicord nameplates identified a low level instead of a low pressure light indication. The QA inspector noted that the step was found satisfactory since the light did respon QASR-3575, Diesel Fuel Oil Storage and Transfer Preoperational Test, conducted July 20, 1985. The QA inspector witnessed several mandatory witness points (MWP's) for the preoperational tes PTP-JE-1 was terminated at step 8.16.8 because a MWP could not be accomplishe .2 Findings No unacceptable conditions were identifie .0 Plant Tours 6.1 Discussion The inspectors made several tours of various areas of the facility to observe work in progress, housekeeping, cleanliness controls and status of construction and preoperational test activities. Facility tours are additionally discussed in Paragraphs 2.2.2 and 4.2 of this repor .

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6.2 Findings No violations were observe .0 Independent Measurements and Verification 7.1 Discussion The inspectors performed the independent measurements discussed in Paragraph 4.2 and independently verified system jurisdictional tagging as discussed in Paragraph 2. .2 Findings No violations were identifie .0 Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable, an item of noncompliance or a deviation. An unresolved item is identified in paragraph .0 Exit Interview A management meeting was held at the conclusion of the inspection on July 26, 1985, to discuss the scope and findings as detailed in this re-port (see Paragraph 1 for attendees). No written information was provided to the licensee at any time during the inspection. The licensee inidi-cated that no proprietary information was contained in the scope of this inspection.

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ATTACHMENT A

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Hope Creek Generating Station FSAR 14.2.12.1.9 BG - Reactor Water Cleanup System

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Hope Creek Generating Station FSAR Appendix B Table 6.2-16 Containment Penetrations GE-2 Reactor Water Cleanup 22A2271 AZ SH NO. B2-1, Re P&ID Cleanup Filter /Demineralizer Job Number 10855 Drawing No. M-45-1, Rev. 7

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P&ID Reactor Water Cleanup Job Number 103S5 Drawing No. M-44-1, Rev. 7

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P&lD Demineralized Water Job Number 10855 Drawing No. M-18-0, SH 1 of 3, Rev. 8

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P&ID Demineralized Water Makeup Storage & Transfer Job Number 10855 Drawing No. M-18-0, SH 2 of 3, Rev. 9

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P&ID Demineralized Water Job Number 10855 Drawing No. M-18-0, SH 3 of 3, Rev. 6

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P&ID Condensate Demineralizer Job Number Drawing No. M-16-1, SH 1 of 2, Rev. 6

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P&ID Condensate Demineralizer Job Number Drawing No. M-16-1, SH 2 of 2, Rev. 6

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P&ID Service Water Job No. 10855

Drawing No. M-10-1, SH 1 of 3, Rev. 7

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P&ID Service Water Job No. 10855

, Drawing No. M-10-1, SH 2 of 3, Rev. 9

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P&ID Service Water

! Job No. 10855

Drawing No. M-10-1, Sh 3 of 3, Rev. 6

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Attachment A 2 o <

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P&ID Fuel Pool Cooling & Torus Water Cleanup Job No. 10855 Drawing No. M-53-1, SH 1 of 3, Rev. 10

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P&ID Fuel Pool Cooling & Torus Water Cleanup Job No. 10855 Drawing No. M-53-1, SH 2 of 3, Rev. 9

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P&ID Fuel Pool Filter Demineralizer

, Job No. 10855 Drawing No. M-54-0, Rev. 6 e

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t 'r ATTACHMENT B

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Hope Creek Generating Station (HCGS), Final Safety Analysis Report (FSAR), Chapter 1 P-AE-01, Rev. 15, System Isometric / Turbine Building, Feedwater from Reactor feedpump to drywel P-AD-01, Rev. 16, System Isometric / Turbine Building, Condensate from condensers to demineralizer ,

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1-P-AD-02, Rev. 17, System Isometric / Turbine Building, Condensate from demineralizers to secondary pump '

l-P-AD-03, Rev. 11, System Isometric / Turbine Building,

. Condensate from secondary pumps to feedwater heaters No. P-AD-04, Rev. 16, System Isometric / Turbine Building, Condensate from feedwater heaters No. 2 to reactor feedpump M-05-1, Rev. 8, P&ID, Condensat M-06-1, Rev. 7, P&ID, Feedwate C-1635-1, Rev. 6, Turbine Building Floor Plan, Elevation 102'0", Area 1 P-AD-020-H03, Rev. 3, Pipe Support, Turbine Building, Condensate from feedwater heater No. 2 to reactor feedpump < -

1-P-AE-001-H18, Rev. 2, Pipe Support, Turbine Building,

Feedwater from reactor feedpump to drywel P-AE-005-H06, Rev. 1, Pipe Support, Turbine Building, Feedwater from reactor feedpump to drywel P-AE-013-H03, Rev. 2, Pipe Support, Turbine Building, Feedwater from reactor feedpump to dry wel P-AE-013-H11, Rev. 2 Pipe Support, Turbine Building, Feedwater from reactor feedpump to drywel PAE-013-H25, Rev. 3, Pipe Support, Turbine Building, Feedwater from reactor feedpump to drywel P-AE-013-H35, Rev. 1, Pipe Support, Turbine Building, Feedwater from reactor feedpump to drywel ,

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