IR 05000354/1986049
| ML20214K909 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 11/12/1986 |
| From: | Briggs L, Marilyn Evans, Florek D, Thomas Koshy, Wink L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20214K799 | List: |
| References | |
| 50-354-86-49, NUDOCS 8612020506 | |
| Download: ML20214K909 (11) | |
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U.S.-NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-354/86-49 Docket No.
50-354 License No. NPF-57 Licensee:
Public Service Electric and Gas Company 60_? ark Plaza Newark, New Jersey 07101 Facility Name: Hope Creek Generating Station, Unit 1 l
Inspection At: Hancocks Bridge, New Jersey f
Inspection Conducted: Oct4ber11-16,1986 Inspectors:
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eac r 'Engi'neer d te/
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jf f , ., C Florek, Lead Reactor Engineer 'dath 'Tk [vc w ////2./E6 M. Eva , Reactor Engineer date $l$b T. K h ~ R act Engineer (AIT Member) ' ate Approved by: )/ L,n /)/])hf . L'. BricJgs, Chief,'[Tpt Programs Section /'date i 08, DRS / l Inspection Summary: Inspection of October 11-16,1986 (Inspection Report No. 50-354/86-49).
Areas Inspected: Routine, unannounced inspection of previous inspection findings, overall power ascension test program including test results evalua-tion and test witnessing, follow up of items identified during Augmented Inspection Team (AIT) review of September 24 - October 3, 1986, independent measurements and verifications, QA/QC interfaces and tours of the facility.
Results: One violation was identified for not performing quality activities in accordance with approved procedures (See section 4.0) NOTE: For acronyms not defined, refer to NUREG-0594 " Handbook of Acronyms and Initialisms."
8612020506 861112 PDR ADOCK 05000354 G PDR
' . DETAILS 1.0 PERSONS CONTACTED Public Service Electric and Gas Company (PSE&G) J.. Adams, Power Ascension Technical Coordinator
- R. Beckwith, Station Licensing Engineer P. Dempsey, Shift Test Coordinator
- M. Farschon, Power Ascension Manager B. Forward, Power Ascension Administrative Coordinator
- E. Galbraith, Chemistry Engineer
- A. Giardino, Manager - Station Quality Assurance (QA)
- R. Griffith, Sr.,- Principal Engineer, QA
- J. Hagan, Maintenance Engineer
- P. Krishna, Assistant to the General Manager
- J. Nichols, Technical Manager W. Schell, Power Ascension Technical Director
- C. Vondra, Operations Engineer U.S. Nuclear Regulatory Commission
- D. Allsopp, Resident Inspector
- R. Borchardt, Senior Resident Inspector
- Denotes those present at the exit meeting on October 16, 1986.
The inspector also contacted other members of the licensee's staff including Senior Nuclear Shift Supervisors, Reactor Operators, Test Engineers and members of the technical staff.
2.0 Licensee Action on Previous Inspection Findings (Closed) Violation (86-32-01).
Failure to perform power ascension test in accordance with written procedures. The inspector reviewed the immediate corrective actions taken by the licensee at the time of the occurrence.
In addition, the inspector has observed the performance of numerous power ascension tests in the three month period since the occurrence and has not identified any further non-conformances of this nature. The inspector was satisfied that the occurrence was an isolated incident and had no further questions on this matter.
(Closed) Unresolved Item (86-43-01). Correct modeling of the SRV Lo-Lo Set Function in the Transient Safety Analysis Design Report (TSADR).
The inspector held discussion with a General Electric Startup Test and Design Analysis Engineer and reviewed documentation provided.
The inspector was convinced that the SRV Lo-Lo Set Function was correctly modeled in the TSADR and that this analysis formed the basis for the acceptance , , i
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- criterion used in power ascension tests TE-SU.AB-252, Main Steam Iso-lation Nalve Full Isolation Test, and TE-SU.CH-273, Full Power Turbine Trip Test. The inspector had no further questions on this matter.
(Closed) Unresolved Item (354/86-24-01). Review of open preoperational test exceptions. During previous NRC inspections, the licensee's resolu-tion / closure of 145 of 210 test exceptions were reviewed. The inspector verified that actions taken by the licensee were technically sound and fully resolved the identified test exception. Based on the satisfactory results of the review and the large sample of test exception resolutions reviewed this item is closed.
Closure of this item constitutes final closecut of the NRC Hope Creek preoperational test program.
3.0 Power Ascension Test Program (PATP) 3.1 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, Revision 1, July 25,1986
HCGS Safety Evaluation Report (SER), Chapter 14, " Initial Test Program"
Station Administrative Procedure, SA-AP.ZZ-036, Revision 3, " Phase III Startup Test Program"
Specification NEB 0 23A4137, Revision 0,. " Hope Creek Startup Test Specification
hCGS Power Ascension Test Matrix, Revision 8 3.2 Overall Power Ascension Test Program Discussion The inspector held discussions with various members of the PATP staff to assess the overall status of the test program and the test results review.
The inspector verified that all test results, reviewed in a previous inspection report (50-354/86-46), have been formally accepted by licensee managemen : .
The licensee began Test Condition 3 (45%-75% rated power) on October 9, 1986. On October 11,~1986, the licensee re performed TE-SU.ZZ-311, Loss of Offsite Power, to demonstrate that those deficiencies identifiec during the original performance of the test on September 11, 1986, had been corrected. The re performance of the test was witnessed by the. resident inspector and regional-based inspectors and is discussed below (see paragraph 3.3).
On October 16, 1986 the inspector attended the Technical Review Board (TRB) review of the re performance of TE-SU.ZZ-311, Loss of Offsite Power. This review involved verification of conformance to all Level 1 and Level 2 acceptance criteria, disposition of observations / comments made by test participants and NRC personnel and resolution of the comments of. independent reviewers.
FINDINGS No unacceptable conditions were identified.
3.3 Power Ascension Test Witnessing Scope The re performance of power ascension test TE-SU.ZZ-311, Loss of Offsite Power was witnessed to verify the attributes as defined in Inspection Report 50-354/86-35 and to verify that deficiencies identified during the earlier performance of this test had been corrected.
Discussion Portions of the preparations, conduct and post-test debriefing for the Loss of Power Test on October 11, 1986, were witnessed. This test was witnessed by five NRC personnel; two NRC supervisors, two regional-based inspectors and the resident inspector.
The test began at 0904 and was successfully terminated at 0935.
NRC personnel witnessed licensee activities in the control room, reactor building and diesel generator building. All licensee activities in these areas were found to be acceptable. All Level 1 and Level 2 test acceptance criteria were satisfied. Additional discussion of the test may be found in the resident inspector's report (Inspection Report 50-354/86-47).
The license.e conducted a post-test debriefing immediately following the test. All station personnel involved in the test were asked to submit observations and comments concerning the test.
These observations / comments were assessed by senior station personnel for safety significance and further followup, if required.
Included in the licensee observations / comments were those noted by the individual NRC inspectors. The licensee's follow up actions were subsequently reviewed by the inspector and found to be acceptabl n .
- FINDINGS No unacceptable conditions were identified.
3.4 Power Ascension Test Results Evaluation Scope The power ascension test results listed in Attachment A were evaluated for the attributes identified in Inspection Report 50-354/86-24. A summary of significant test results and identified test results deficiencies is provided in the discussion below.
Discussion TE-SU.BF-054. This test was performed in conjunction with the planned scram during the Loss of Offsite Power Test, TE-SU.ZZ-311.
The scram times of four selected rods (coordinates 26-23, 42-15, 50-47 and 54-19) were measured and verified to be well within the acceptance criterion and technical specification limit of 7.0 seconds.
TE-SU.SE-122.
The APRMs were calibrated by means of a heat balance performed by the process computer. All APRMs were set to read slightly higher than actual core thermal power and the technical specification setpoints for scrams and rod blocks were verified.
All acceptance criteria were satisfied.
TE-SU.ZZ-162.
No acceptance criteria were verified in this test.
The environmental conditions of the drywell and reactor were verified to have been relatively unaffected by the increase in power levei from the initial performance of this test in Test Condition Heatup.
TE-SU.BB-191. This verification of the core thermal hydraulic limits was performed at 32.7% of rated thermal power and 46.2% of rated core flow. All acceptance criteria were satisfied and the results are summarized below: Parameter Measured Limit LHGR (KW/FT) 4.76 s13.4 CPR 3.416 21.41 APLHGR (KW/FT) 4.06 s11.80
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- TE-SU.BB-223. All acceptance criteria for pressure regulator response and stability were satisfied.
TE-SU.AE-231. This test was performed at 30% of rated thermal power to demonstrate adequate transient response of the feedwater system to level setpoint changes with both single and dual feedwater pumps. At the time of the test only the A and B feedwater pumps were available for testing.
The level 1 acceptance criterion for non-divergent response was satisfied for both A and B individually and in combination but two level 2 results deficiencies were noted.
' Results deficiency (RDF) #97 identified the inability to verify an acceptable decay ratio when switching frem single to three element - control due to small, slow limit cycles in reactor water. level.
RDF #102 identified the failure to satisfy the decay ratio criterion for single pump operation on both the A and B pumps. Both the limit cycle behavior and the decay ratio failures were attributed to sluggish system response since tuning of the system is not planned until Test Condition 3.
Due to the inability to test the C feedwater pump and the results deficiencies on A and B feedwater pumps the test will be repeated in Test Condition 3 following system tuneup.
TE-SU.AE-235.
This test was performed at 32% of rated thermal power to demonstrate adequate transient response of the feedwater system to manual changes in feedwater flow. The level 1 acceptance criterion for non-divergent response was satisfied but two level 2 acceptance criteria failures were noted. 'RFD #104 identified the inability to observe the response to small flow steps due to limit cycling in reactor water level.
RDF #105 identified the failure to satisfy the time response requirements during large flow steps.
Both failures were attributed to sluggish system response and this test will also be repeated in test condition 3 following system tuneup.
TE-SU.CH-272. This test was performed at 22% of rated thermal power to verify that a reactor scram will be avoided for a main turbine trip within bypass valve capacity (25% of rated thermal power). All acceptance criteria were satisfied.
TE-SU.BB-303. A level 2 results deficiency was identified when, due to instrument unavailability, it proved impossible to verify the efficiency of the recirculation pumps.
This test will be re performed at 100*4 core flow during Test Condition 3 and pump efficiency will be verified at that time.
TE-SU.AB-331. Measured steady state vibration of the main steam piping was well within acceptance criteria limits.
TE-SU.BB-332. Measured steady state vibration of the recirculation piping was well within acceptance criteria limits.
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- TE-SU.BB-343. This test was run in conjunction with TE-SU.ZZ-311, Loss of Offsite Power. The test monitors the main steam line outside the drywell to verify that transient loads caused by a main turbine trip are within established limits. All measured loads were well within acceptance criteria limits. A results deficiency was noted (RDF #103) when dented insulation was found on a pipe during a post test inspection. The dent in the pipe insulation was evaluated and accepted "as-is".
TE-SU.GT-723. This test was run in conjunction with TE-SU.ZZ-311, Loss of Offsite Power. The test monitors drywell temperatures to verify adequate cooling following a loss of offsite power. During the first performance of TE-SU.ZZ-311 all cooling was lost to the drywell, invalidating this test. The temperature transient measured proved to be very mild despite the total loss of cooling. This test was successfully re performed in conjunction with the re performance of TE-SU.ZZ-311.
TE-SU.AC-774. This test was performed between 15% and 30% of rated thermal power to establish the relationship between core thermal power and main turbine first stage pressure. This relationship is required to determine the instrument setpoints for Rod Sequence Control System (RSCS) and Reactor Protection System (RPS) functions which are based on core thermal power but derived from measurements of first stage turbine pressure. The inspector reviewed in detail the RPS scram bypass (EOC-RPT) determination including the calcula-tions to establish the trip setpoint, allowing for instrument ac-curacy and drift, and the incorporation of the results into instru-ment calibration data cards.
Findings No unacceptable conditions were identified.
4.0 Items identified during Augmented Inspection Team (AIT) review of September 24 - October 3, 1986 (Inspection Report 50-354/86-50) 4.1 Instrumentation (Components &-System) Work Observation.
During a plant tour on September 24, 1986, the inspector noted that some Rosemount transmitters had an excessive gap between the body and side covers. After observing several IE transmitters with such excessive gap, the inspector requested the licensee to verify the side cover installation for the torque requirements stated in the manufactures instruction manual.
Licensee procedure on " Maintenance of Environmentally Qualified Rosemount Transmitters" 1C-GP.ZZ-053(Q), Rev. 3, Section 5.2.36 states that the side covers should be torqued to 200+ or -10 in.-lbs.
The licensee was asked to verify this requirement on instruments 1GSPT-4960A3 and 1BPT-N050A-C71. The , w.. -- -- - - , , -. - --. --
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' technician confirmed that these instruments'were under torqued. The excessive gap disappeared after these instruments were reworked as per the procedure.
Subsequent to this finding, the licensee inspected 299 transmitters located in the harsh environment and identified 48 potentially affected transmitters. The inspector witnessed the corrective action performed on ISAPT-N403B-821. The technician was able to open the right side cover without the use of any tools.
This indicates a torque value of approximately~50 in.-lbs.
The cover and body threads were not cleaned previously as required by the PSE&G procedure and Rosemount Technical Manual.
Lack of proper cleaning can cause the side covers to be tight before it engages the 'O' ring, making the 'O' ring nonfunctional and the instrument susceptive to moisture intrusion.
Rosemount Test Report 108025 Rec.C states the need for strict compliance to the instruction manual for maintaining environmental qualification.
Lack of following the PSE&G procedure IC.GP.ZZ-053 (Q) and Rosemount Technical Manual for Transmitters constitute a violation under 10 CFR 50, Appendix B, Criterion V, which states in part " Activities affecting quality.shall be prescribed by documented instructions, procedures and shall be accomplished in accordance with these instructions, procedures" (354/86-49-01).
4.2 Field changes originated through design change requests and 10 CFR 21 reportability.
The corrective change performed under 4BMP.86.66.3 (Q) was not evaluated under the requirements of 10 CFR 21. Though this partic-ular modification did not involve a reportable condition, lack of specific direction in Design Change Control procedure SEI 4.2 and Non Conformance program SA-AP-ZL.020(Q) can lead to potential misappli-cation of Design Change Requests for corrective action on safety related comoonents.
Licensees' present field change management with cognizant system engineers appears to function effectively even without specific guidance in the procedure. However, the license has ' agreed to revise the procedure for the proper use of Design Change Requests. This item will be unresolved pending proccdure revision (354/86-49-02).
4.3. Non-Conformance Program Relating to Reportability under Environment Qualification Deficiencies The licensee procedure on Non Conformance Program SA-AP.22-020(Q) lacks sufficient guide lines for evaluating Environmental Qualification (EQ) deficiencies. While the Licensee field change management with cognizant system engineers appears to evaluate such problems in sufficient detail with the required coordination, the Licensee has agreed to revise the subject procedure for EQ review in order to avoid any oversight.
Pending procedure revision this item will remain unresolved (354/86-49-03).
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- 4.4 Torque limits for friction Type Connectors The inspector reviewed DCR No: 4-EMC-86-641, a completed design change package from document control. The instructions in the doc-ument indicated the use of a higher torque valve than what was used in maintenance procedure MD.GP.ZZ-022(Q) for bolted connections. The licensee's procedure did not set limitations for the application of this procedure i.e: using it on friction type or bearing type con-nection. The inspector discussed this matter with personnel from the structural department and was advised that the particular DCR was intended for a friction type connection and therefore higher torque values were required per the recommendation of the American Institute of Steel Construction.
It was further clarified that the maintenance procedure is adequate for torquing high strength bolts used in bear-ing type connections-based on the fact that pre-loads are not , required.
The licensee has instituted steps to revise the procedure to limit the use of this procedure to bearing type connections. However during the period of this inspection the licensee was unable to provide any assurance that this procedure was not utilized on friction type of connection. This item will be unresolved (354/86-49-04).
5.0 Independent Measurements and Verifications The inspector performed multiple independent measurements and verifications during the-evaluation of power ascension test results (paragraph 3.4).
In all cases the inspector's measurements and verifications agreed with those of the licensee.
No unacceptable conditions were noted.
6.0 QA/QC Interface with the Power Ascension Test Program The inspector reviewed 12 QA surveillance reports covering power ascension test program activities in Test Conditions 1 and 2.
Included in this sample were QA surveillance reports 86-301 and 86-333 on performances of the loss of offsite power tests preformed on September 11 and October 11, 1986. The inspector also observed QA involvement in the TRB review of the test results of the loss of offsite power test performed on October 11, 1986.
No unacceptable conditions were noted.
7.0 Tours of the facility During the performance of power ascension test TE-SU.ZZ-311, Loss of Offsite Power, on October 11, 1986 the inspector toured various areas of the reactor building and control structure to assess lighting levels, i m 7,
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- observe the general. conduct of operations and to look for any unexpected -
- or abnormali conditions.
No unacceptable conditions were identified.
.8.0 Unresolved Items-Unresolved items'are matters about which more information is' required _in order-to determine whether they are acceptable, an item of noncompliance or'a deviation.. Unresolved. items disclosed during the inspection are discussed in paragraph 4.0.
9.0 Exit Interview At the conclusion of the site inspection on Octobet-16, 1986, an exit- , meeting was conducted with the-licensee's senior site representative (denoted in paragraph 1.0).
The findings were identified and previous . inspection items were discussed.
- At no time during the inspection was written material provided to the licensee by the inspector. Based on the NRC Region I review of this report and discussions with licensee representatives during the inspection, it was. determined that this report does not contain information subject to 10 CFR 2.790 restrictions.
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Attachment A Power Ascension Test Results Evaluated TE-SU.BF-054 Control Rod Drive System Scram Testing During Planned Scrams, Revision 6, completed September 11, 1986, results not yet accepted.
TE-SU.SE-122 Average Power Range Monitor Calibration at Power, Revision 4, completed September 10, 1986, results not yet accepted.
TE-SU.ZZ-162 Water Level Measurement Test, Revision 5, completed September 6, 1986, results accepted September 13, 1986.
TE-SU.BB-191 Core Performance, Revision 2, completed September 9,1986, results not yet accepted TE-SU.BB-223-Pressure Regulator Test - Bypass Valves Controlling, Revision 7, completed September 11, 1986, results accepted October 14, 1986 TE-SU.AE-231 Feedwater System Level Setpoint Changes, Revision 4, completed September 11, 1986, results accepted October 10, 1986.
TE-SU.AE-235 Manual Feedwater Flow Step Change Yest, Revision 2, completed September 10, 1986, results not yet accepted.
TE-SU.CH-272 Low Power Turbine Stop Valve Trip Test, Revision 2, completed September 11, 1986, results accepted October 10, 1986.
TE-SU.BB-303 Recirculation System Performance, Revision 2, completed September 11, 1986, results not yet accepted.
TE-SU.AB-331 NSSS Main Steam Piping Steady State Vibration, Revision 2 completed September 5, 1986, results accepted September 13, 1986.
TE-SU.BB-332 Recirculation System Piping Steady State Vibration, Revision 2, completed September 6, 1986, results accepted September 13, 1986.
TE-SU.BB-343 BOP Main Steam Piping Dynamic Response, Revision 4, completed September 12, 1986, results not yet accepted.
TE-SU.GT-723 Drywell Cooling System Post Trip Performance Test, Revision 2, completed September 11, 1986, results accepted October 10, 1986.
TE-SU.AC-774 Main Turbine First State Pressure Scram Bypass and Rod Sequence Control System Setpoints, Revision 6, completed September 6, 1986, results accepted October 10, 1986. }}