ML20149M128
| ML20149M128 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 03/12/1986 |
| From: | Eselgroth P, Marilyn Evans, Florek D, Wink L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20149M120 | List: |
| References | |
| 50-354-86-03, 50-354-86-3, NUDOCS 8802250347 | |
| Download: ML20149M128 (15) | |
See also: IR 05000354/1986003
Text
{{#Wiki_filter:. . .. . . ' . U.S. NUCLEAR REGULATORY COMMISSION REGION I Report No. 50-354/86-03 Docket No. 50-354 License No. CPPR-120 Priority Category C -- Licencee: Public Service Electric and Gas Company 80 Park Plaza Newark, New Jersey 07101 Facility Name: Hope Creek Generating Station, Unit 1 Inspection At: Hancocks Bridge, New Jersey Inspection Condu ed: Janua ny 7-17, 1986 Inspectors: . . [ ? 7 Torek, lead Reactor Engineer / fate M l0 b M. Evans, Reactor Engineer date
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=Aeh6 . Wink, Reactor Engineer / dMe Approved b : [ [[ f. Eselgrot ef, Test Programs Section 'date J Inspection Summary: Insoection on January 7-17, 1986 (Inspection Report No. 50-354/E6-03) Areas Inspected: Routine unannounced inspection of the preoperational test program including preoperational test procedure review, tost witnessing and test results evaluation; power ascens1]n test progran including the overall program and procedure review; licensee actions on previous inspection findings; independent verification and calculation; QA interfaces and tours of the facility. The inspection involved 151 hours onsite by three region based irspectors. Results: No violations were identified. NOTE: For acronyns not defined, refer to NURfG-0544 "Handbook of Acroryms and Initialisms." , 8802250347 860314 PDR ADOCK 05000354 O DCD . - - - - - - - - - - - - "
- . . . . - . DETAILS 1. Persons Contacted Public Service Electric and Gas (PSE&G) personnel and contractors
- V. Blenx, Assistant Project Manager
- J. Carter, Startup Manager
- G. Chew, Power Ascension Technical Support
G. Conner, Operations Manager D. Cooler, Onsite Safety Review Engineer R. Donges, lead QA Engineer
- J. Duffy, Site Engineering
- M. Farschon, Power Ascension Manager
A. Giardino, Manager Station QA
- R. Griffith, Principal QA Engineer
- G. Jaffee, Startup Engineer
- S. LaBruna, Assistant General Manager
M. Metcalf, Principal QA Engineer J. Nichols, Technical Manager
- R. Salvesen, General Manager
- W. Schell, Power Ascension Technical Director
- R. Schmidt, Senior Reactor Supervisor
, U.S. Nuclear Regulatory Commission f
- W. Borchardt, Senior Resident Inspector
J. Lyash, Resident Inspector L The inspector also contacted other members of the licensee's staff including Senior Nuclear Shift Supervisors, test engineers and members of the technical staff.
- denotes those present at exit interview on January 17, 1986.
2. Licensee Actions on Previous Inspection Findings (Closed) Circular (354/78-CI-13). This Circular concerns inoperability of service water pumps due to silt accumulation, low water level and ice accumulation. Inspection report 50-354/85-19 reviewed the licensee actions and concluded they were acceptable provided procedures were updated. The inspector reviewed implementing procedures MD-PM-EA-002 "Service Water Intake Bay Silt Survey" dated October 28, 1985 and OP-AB.ZZ-122, "Less of Service Water Loop", dated July 11, 1985, licensee i responses BLP 16557 dated October 9, 1984 and BLP-7304 dated September 22, - 1978, and P&ID M 10855-M-09-1 Revision 7 Circulating Water. Based on review of these documents and discussions with licensee representatives this item is closed. P h . . . . . . . - . . - . . . . - .
, . . .. . - . . 3 _(Closed) Unresolved Item (354/85-30-01), Licensee to remove temporary , drain fittings, conduct final drain closure and develop approved admini- ! strative controls to prevent unauthorized removal.of clean out connection l plugs. The inspector determined by review of procedure SA-AP.ZZ-013(Q), t "Control of Temporary Modifications", that proper administrative controls- exist to prevent unauthorized removal of-drain plugs. Temporary drain l fitting removal and drain closures are being tracked by OTR-HG-1 to permit - ~ inspector verification of the installation as construction-is completed. ! This item is closed. (Closed) Unresolved Item (354/85-41-02). Surveillance procedure MD-PM.KE- 003(Q) for the operation of the refueling platform does not satisfy j Technical Specification (TS) requirements nor specify precise l tolerances / acceptance criteria. The inspector reviewed Revision 1 of ' MD-PM.KE-003(Q), "Refueling Platform Operational Checks." Precise toler- l ances had been added to the procedure in the revision. The TS surveil- l lance requirements appeared to be satisfied by the procedure with the ! exception of TS Surve111ances 4.9.6.2.b and 4.9.6.3.b. The inspector questioned the licensee concerning these surveillance requirements. On
the spot change OSC No. P-2 was issued to incorporate the correct TS surveillance requirements into the procedure. This item is closed.
3.1 Preoperational and Detailed Test Procedure Review and Verification l l Scope , The Preoperational Test Procedure (PTP) and Detailed Test Procedures (DTP's) listed in Appendix A were reviewed and discussed with the System Test Engineer (STE) in preparation for test witnessing, for technical and
administrative adecuacy and for verification that testing is planned to i adequately satisfy regulatory guidance and licensee commitments. They were also reviewed to verify licensee review and approval, proper format,
test objectives, prerequisites, initial conditions, test data recording ' requirements and system return to normal. 1 i Discussion PTP-SB-2 -- Response Time Testing is accomplished by means of 14 Detailed Test Procedures (DTP's). The inspector reviewed DTP-SB-0001, DTP-SB-0004, DTP-SB-0005, DTP-SB-0006, DTP-SB-0008, DTP-SB-0009, DTP-SB-0012, DTP-SB-0013, and DTP-SB-0014. (The five remaining DTP's were previously reviewed in Inspection Report No. 50-354/85-62). The inspector compared ! the DTP's acceptance criteria to_ the acceptance criteria of GE l Preoperational Test Specification 22A2271 AZ, Revision 2. All GE acceptance criteria were included in the procedures reviewed by the ! inspector. ) i ! ! . . _ _ _ _ ~ _ . _ , , _
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Findings ' No unacceptable conditions were identified within the scope of this review. 3.2 Preoperational Test Witnessing Scope Testing witnessed by the inspector included the observations of overall i crew performance stated in Paragraph 3.0 of Inspection Report ' 50-354/85-18. . . Portions of the following PTP's and DTP were witnessed: PTP-BB-3 (Part A), Standby Diesel Generator Loading, Revision 0. -- t PTP-BB-3 (Part B), ECCS Integrated Initiation / Loss of Offsite Power, -- I Revision 0. l DTP-SB-0013, Ma;n Steam Line Flow Response Time Test, Revision 0. l -- Discussion PTP-BB-3 (Part A) i -- ' The inspector observed sections 8.3 and 8.7 of PTP-BB-3 (Part A), Hot Automatic Start and Load Sequencing for Die;el Generators "A" and "D". For Section 8.7, the inspector independently verified the initial positions and the load shedding positions for the equipment listed in Appendix F of PTP-BB-3 (Part A). All testing observed satisfied the criteria above. PTP-BB-3 (Part 8) -- The inspector witnessed section 8.3 of PTP-BB-3 (Part B), LOCA Signal , with Simultaneous Loss of Offsite Power. Testing was conducted in accordance with the criteria above with full QA coverage during the portions witnessed by the inspector. DTP-SB-0013 -- Several times during the inspection, the inspector observed the > performance of step 6.3, Process Sensor (Transmitter)/ Response Time ' ' Test. The inspector independently calculated time response ramp rates ' for Main Steam Line Flow Transmitters B21-N088A and 821-N0890. All testing observed satisfied the criteria above. , i m . .a... w
r . . . . . 5 Findings No unacceptable conditions were observed within the scope of this review. 3.3 Preoperational Test Results Evaluation: Review Scope The completed test procedure listed below was reviewed during this inspection to verify thatLadequate testing had been conducted to satisfy regulatory guidance, licensee commitments and FSAR requirements and to verify that uniform criteria are being applied for evaluation of completed test results in order to assure technical and administrative adequacy. The inspector reviewed the test results and verified the licensee's evaluation of test results by review of test changes, test exceptions, test deficiencies,'"As-Run" copy of _ test procedures, acceptance criteria, performance verification, recording conduct of test, QC inspection records, restoration of system to normal after test, independent verification of critical steps or parameters, identification of personnel conducting and evaluating test data, and verification that the test results have been approved. PTP-BC-1, Residual Heat Removal, Revision 0 Approved 12/27/85. -- Di scus sio.n No unresolved discrepancies or violations were identified in the above review, However, several open test exceptions require resolution by the licensee. The inspector routinely assigns an unresolved item number to open test exceptions that are desired.to be tracked. The following cpen test exceptions identified in previous NRC reports along with'those open test exceptions identified in the above review are being consolidated into one unresolved item (354/86-03-01). Unresolved items 354/85-18-01 and 354/85-26-01 are closed. Procedure No. Short Title SDR No. ! PTP-AN-2 Demin. Wtr Storage & Transfer AN-0039 l
PTP-PK-1 125 VDC Class IE PK-0117, 0119 l and 0120. i
PTP-PJ-1 250 VDC Class IE PJ-0026, 0033 and
0129. l
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r . . 6 Procedure No. Short Title SDR No. PTP-BC-1 RHR BC-915, 1042, 1043, 1143, 1144, 1145, 1146, 1147, 1148. RL-736, 738. 4. Power Ascer.sion Test Program (PATP) 4.1 References Regulatory Guide 1.68, Revision 2, August 1978, "Initial Test
Programs for Water-Cooled Nuclear Power Plants" ANSI N18.7 - 1976, "Administrative Controls and Quality . Assurance for the Operational Phase of Nuclear Power Plants" Hope Creek Generating Station (HCGS) Technical Specifications,
Proof and Review Copy HCGS Final Safety Analysis Report (FSAR), Chapter 14, "Initial + Test Program" HCGS Safety Evaluation Report, Chapter 14. "Initial Test
Program" Station Administrative Procedure, SA-AP.ZZ-036, Revision 1,
"Phase III Startup Test Program" Specification NEB 0 23A4137, Revision 0, "Hope Creek Startup
Test Specification" HCGS Power Ascension Test Matrix, Revision 2
4.2 Overall Power Ascension Test Program (PATP) Scope The inspector reviewed the following procedures: SA-AP ZZ-001 "Preparation and Approval of Station Procedures", Revision 4 SA-AP-ZZ-002 "Station Organization and Operating Practices," Revision 3
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s' . . . - . , I 7 i ! SA-AP-ZZ-014 "Station Personnel Qualification and Training", ,! Revision 1 ! , , SA-AP-ZZ-020 "Nonconformance Program", Revision 2
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SA-AP-ZZ-032 "Revisions and Changes to Station Procedures",
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Revision 2 .
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. ' SA-AP-ZZ-036 "Power Ascension Test Program", Revision 1 ? The inspector also held discussions with the Power Ascension ' - l' Technical Director to review the approaches to be taken to assure < adequate administration of the power ascension test program. - l The inspector verified that administrative measures were established
to assure that test procedures are current prior to.use, test per-
sonnel are knowledgeable, controls exist for test procedure changes, , , j interruption in tests are controlled, proper test coordination, docu-
mentation of unusual events and test deficiencies,' tests are . j scheduled, test results are reviewed, test acceptance criteria are
defined, retest after test deficiencies and appropriate review. l i 4 . The inspector also held discussions with the Senior Reactor i j Supervisor, Onsite Safety Review Engineer and QA personnel to l i ' determine their involvement with the power ascension test program. !
Discussion Administrative Program
i The licensee administrative program was determined to satisfy the i above attributes. The licensee administrative program utilizes j features from other recent startup program and also utilizes normal t station administrative controls where practicable . The licensee has ! ! not yet provided formal detailed administrative training te the !
startup test engineers and will not do so until a later date, closer ' j to the actual need for this training. This will be reviewed by the 3
inspector in a sub-equent inspection. l l Reactor Engineering . t ' Based on discussions with the Senior Reactor Supervisor, the planned
manning for reactor engineering is one supervisor and three , engineers, two of which are currently on staff. Based on these ! discussions and review of Form SA-AP-ZZ-0141 Verification of i > Qualification for Position Hope Creek Operation for the Senior - Reactor Supervisor, the current positions, including the supervisor, are filled with persons with no practical BWR reactor i engineering experience. The supervisor has one year practical , a 4 6
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experience at Salem and the other two engineers have experience at
'research reactors. The inspector questioned the practical experience ' . level of reactor engineering personnel and was informed that the ! l licensee is in the process of hiring a person for reactor engineering . who has extensive practical BWR reactor. engineering experience. The i , inspector indicated that lack of practical BWR reactor engineering ' experience is not consistent with the intent of ANSI /ANS 3.1-1981 ! ! "Selection, Qualification and Training of Personnel for Nuclear Power Plants". Pending the hiring of personnel with practical BWR reactor , , engineering experience and reviewing the duties, background and ! responsibilities of the individual or individuals who provide this l experience, this will be identified as an unresolved item (354/86- 03-02). .Onsite Safety Review
. I Based on discussions with the Onsite Safety Review Engineer, this 1 independent review group plans to conduct independent reviews of _ ! i several of the power ascension testing activities. Onsite Safety , Review plans for Power Ascension Testing dated December 20, 1985 was I ' reviewed. Independent oversite review of testing activities of 1 j major integrated tests as well as individual system tests will be
1 performed. No unacceptable conditions were identified. ! i QA involvement in PATP ! Based on discussions with the Manager Station QA Hope Creek and I Startup QA Supervisor, review of Nuclear Quality Assurance Department Manual GM-9-1 and GM9-QAP 5-1.1 "Surveillance of
Phase III Startup Test Program", Revision 0 dated July 1,1985, QA l has and will be involved in the power ascension testing program < (PATP). l QA has reviewed the PATP procedures. Checklists and discussions < with principal QA staff reviewers on the scope of the review were
. i ' found to be acceptable. QA plans to provide coverage on all shifts to conduct surveillance activities on PATP tests. QA is providing practical training to QA staff engineering on fuel loading activities
l as well as training on the Hope Creek simulator on PATP tests. j Completed startup test results will also be reviewed in accordance ) with the administrative procedure for PATP testing. No unacceptable conditions were noted in the QA plans for PATP activities. , . , I 3 s & - r - I ' __. - -. _ . . _ . -_ _. . _ - . - . -
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4.3 Startup Test Procedure Review- Procedure Review Scope The licensee's Power Ascension Test Program (PATP) test procedures of Appendix B were _ reviewed for their conformance to the-requirements , and guidelines of the references listed in Section 4.1 and-for the following' attributes: Appropriate management. review and approval has oeen t
accomplished.
' Appropriate committee review has been accomplished.
Procedure is in the proper format.
Test objectives are clearly stated and consistent with the FSAR.
P Appropriate references are listed.
, Appropriate prerequisites and precautions have been included.
Initial test conditions are specified.
Acceptance criteria are clearly stated. l
Provisions have been made to identify test equipment utilized.
Provisions have been made to identify personnel performing the
' test. i The procedure is technically adequate and workable.
Temporary jumpers, installations and lifted leads are properly
' restored. Provisions have been made for recording, evaluating and
approving test data. Provisions have been made to identify test deficiencies and
exceptions and to document their resolutions.
i Discussion Initial Fuel Loading The inspector reviewed the approved fuel loading test procedures ! (TE-SU.KE-031, 032 and 033) in detail and verified that they were in
substantial conformance with the Regulatory Guide and FSAR commit- ments and that they would provide a technically adequate basis for safely conducting the initial fuel loading. I .
. - . . 4 , . 10 Several instances were identified in which the procedures did not fully address the guidance contained in the Regulatory Guide. The inspector discussed these areas with the Power Ascension Group Technical Coordinator and agreement was reached in all areas. The licensee will include the following items in Revision 1 of TE-SU.KE-032 currently in preparation: Limits on boron concentration in the reactor vessel and spent
fuel storage pool will be established and verified prior to commencing fuel loading. Explicit criteria will be established for the initiation of
emergency boration and containment evacuation. Expanded guidance will be provided on appropriate actions if
the 1/M plots indicate criticality'will be achieved prior to the completion of the next fuel loading increment. Signoffs will be provided to explicitly verify that the minimum
fuel loading crew is available and that technical specification requirements for Operational Condition 5 are satisfied. The inspector will review the revised procedures during a subsequent routine inspection to ensure that these items have been incor- parated. Initial Criticality / Full Core Shutdown Margin Demonstration The inspector reviewed the approved startup test procedure, TE-SU.ZZ-041, Full Core Shutdown Margin Demonstration in detail. This procedure encompasses initial criticality and demonstrates conformance to the technical specification limits for shutdown margin and reactivity anomalies. The inspector determined that the procedure did not totally meet the standards established in the Regulatory Guide and ANSI Standard and would not ensure that the initial approach to criticality is conducted in a deliberate and orderly manner. The following deficiencies were identified: 1. Prerequisite / initial conditions did not specify initial plant status (cperational condition per Technical Specification definitions), did not completely address conformance to tech- nical specifications and did not completely specify the required status of the reactor protection system and emergency shutdown systet. 2. A prediction of control rod positions at criticality was not made.
E - . . - . 11 3. Precautions and limitations did not identify the additional protective features provided by removal of the "Shorting Links", did not identify the reduced setpoints in the SRM circuitry, did not identify operator actions on high startup rate /short period and did not place limits on control rod withdrawal if criticality is not achieved within predetermined limits. 4. The method used to assess conformance with the acceptance criterion for reactivity anomalies was not technically correct. The inspector met with the Power Ascension Manager, the Power Ascension Technical Director, the plant Technical Engineer and the Reactor Engineer to discuss the identified deficiencies. The licensee agreed to revise TE-SU.ZZ-041 to insure that the deficiencies identified above are corrected and that conformance with the Regulatory Guide and ANSI standard is achieved. At the exit meeting on January 17, 1986 the inspector informed the licensee that the above problems are considered an unresolv- ed item. However, in a telephone conversation on February 24, 1986 the NRC informed the licensee that the above items are con- trary to 10 CFR 50 Appendix B Criterion XI and are being consid- ered a violation (354/86-03-03). 10 CFR 50 Appendix B Criterion requires in part that "test procedures incorporate requirements and acceptance limits contained in applicable design documents and assure that all prerequisities for the test have been met." Regulatory Guide 1.68 "Initial Test Programs for Water-Cooled Nuclear Power Plants," Revision 2 which is endorsed by the li- censee's FSAR, indicates in Appendix C that for initial critica- lity procedures, "a critical control rod position should be pre- dicted for anomaly determination, all systems should be aligned and in proper operation, and that technical specification re- quirements must be met" and generally applicable for all test procedures "thet special precautions needed to ensure a reliable test should be highlighted and clearly described in the test pro- cedure." The startup test procedure TE-SU.ZZ-041 did not satisfy these requirements. In a subsequent inspection (50-354/86-12) the inspector observed portions of licensee actions to correct the above problems and to preclude reoccurrence by use of a Technical Review Board to provide an additional in-depth review of startup test procedures.
_ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ - _ _ . _ _. . . _ _ _ _ _ _ . - ,. -. . - . 12 Control Rod Drive System The inspector reviewed in detail the six startup test procedures (BF -series) which encompass the' planned testing of the control rod drive system. The inspector verified that these tests were in conformance with the Regulatory Guide and FSAR-commitments and.that sufficient testing was planned to insure system operability over.the full range of operating conditions. The inspector had several questions on these procedures. During dis- cussions with the Power Ascension Group Technical Coordinator, the questions were adequately addressed. The licensee agreed to revise the procedure to assure that the four rods selected for monitoring during planned scrams of the PATP utilize all the previous rod testing performed up through rated pressure testing to select the four rods. This will be reviewed in a subsequent inspection. Neutron Monitoring System The inspector reviewed in detail three startup test procedures (SE -series) which will be used in conjunction with the initial critica- lity to demonstrate operability of the neutron monitoring system (source and intermediate ranges). The inspector verified that these tests were in conformance with the Regulatory Guide and FSAR commit- ments. The inspector had several minor questions on these procedures which were adequately addressed during discussions with the Power Ascension Group Technical Coordinator. The inspector noted that due to a reduc- ed SRM rod block setpoint during the initial criticality the possibi- lity existed for misinterpretation of the acceptance criteria in TE-SU.SE-101, SRM/IRM Overlap. The Technical Coordinator acknowledged this possibility and indicated that this would be clarified in Revi- sion 1 to the procedure which was in process. Findings Within the scope of this inspection, one violation was identified. 5.0 Independent Verification and Calculations The inspector performed the independent calculations and independently verified equipment positions as discussed in Paragraph 3.2. Findings No violations were identified.
__ __ _ _ . _ _ _ _ _ _ __. . _ _ _ - _ _ _ t~ O. - . ... , & 13 6.0 QA Interfaces The inspector reviewed QA involvement in the Power Ascension Test Program as described in section 4.2 and observed QA coverage of preoperational tests as described in section 3.2. Findings No unacceptable conditions were noted. 7.0 Plant Tours The inspector made several tours of the various areas of the facility to observe work in progress, housekeeping, cleanliness controls and-status of construction and preoperational test activities. Findings No unacceptable conditions were observed. 8.0 Unresolved Items Unresolved items are matters about which more information is required to ascertain whether they are acceptable items, items of noncompliance or deviations. Unresolved items disclosed during the inspection are discussed in Sections 3.3 and 4.2. 9.0 Exit Interview At the conclusion of the site inspection on January 17, 1986, an exit meeting was conducted with the licensee's senior site representatives
(denoted in paragraph 1). The findings were identified and discussed. At no time during this inspection did the inspector provide written inspection findings to the licensee. The licensee did not identify that any proprietary information was contained in the scope of this inspection.
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, ATTACHMENT A PTP-SB-2, Response Time Testing, Revision 0. DTP-SB-0001, Recirculation Pump Trip Breaker Interrupt Time,-Revision 0. ' OTP-SB-0004, RRCS (ATWS) Recirculation Pump Breaker Response Time Test, Revision 0. DTP-SB-0005, Main Steam Line Pressure Low Response Time Test,- Revision 0. DTP-SB-0006, Drywell High Pressure Scram. Response Time Test, Revision 0. DTP-SB-0008, Reactor Vessel High Steam Dome Pressure Scram, Revision 0. DTP-SB-0009, Main Steam Log Rad Monitors /RPS Trip Logic Response Time, Revision 0. DTP-SB-0012, RPS/ Average Power Range Monitor Response Time, Revision 0. DTP-SB-0013, Main Steam Line Flow Response Time Test, Revision 0. DTP-SB-0014, Main Steam Reactor Level Response Time Test, Rev:sion 0.
I s e . ' . ATTACHMENT B Startup Test Procedures TE-SU.KE-031, Fuel Loading Preparation, Revision 0.
TE-SU.KE-032, Fuel Loading, Revision 0,
TE-SU.KE-033, Full Core Verification, Revision 0.
TE-SU.ZZ-041, Full Core Shutdown Margin Demonstration, Revision 0.
TE-SU.BF-051, CRD Functional Tests, Revision 0.
TE-SU.BF-052, Scram Testing of Selected Rods, Revision 0,
TE-SU.BF-053, CRD Friction and Scram Tests, Revision 0.
TE-SU.BF-054, Scram Tests During Planned Scrams, Revision 0.
TE-SU.BF-055, CRD Friction and Scram Tests at Rated Pressure,
Revision 0, TE-SU.BF-056, CRD Functional Checks and Scram Tests, Revision 0.
TE-SU.SE-061, SRM S/N Ratio and Minimum Count Rate Determination,
Revision 0. TE-SU.SE-062, SRM Response to Rod Withdrawal, Revision 0.
TE-SU.SE-101, SRM/IRM Overlap, Revision O.
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