IR 05000321/1978037
| ML19256A991 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 12/11/1978 |
| From: | Cantrell F, Dance H, Julian C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19256A985 | List: |
| References | |
| 50-321-78-37, 50-366-78-47, NUDOCS 7901170368 | |
| Download: ML19256A991 (12) | |
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UNITED STATES
[t >* 88 Go,o, NUCLE Af4 REGULATORY COMMISSION
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101 M ARIETT A STREET, N W
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Report Nos.
50-321/78-37 and 50-366/78-47 Docket Nos..
50-321 and 50-366 License Nos.-
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Licensee:
Georgia Power Company Facility Name E. I. Hatch Units 1 & 2 Inspection at:
Plant Site, Baxley Dates of Inspection:
November 14-17, 1978 Inspectors:
F. S. Cantrell C. A. Julian S
Accompanying Personnel:
None
/,[7 T Approved by:
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11. C. Da n c e, Ch i e f ate Reactor Projects Section No. 1 Reactor Operations and Nuclear Support Branch lospection Summary Inspection on November 14-17, 1978 (Report Nos. 50-321/78-37 and 50-366/78-47)
Areas inspected:
Routine unannounced inspection to review LER's, review and audit functions, organization and administration, cleanliness, and unresolved items. The inspection involved 52 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br /> on site by two NRC inspectors.
Results:
Of the five areas inspected, no items of noncompliance were identifsed.
79011703GF
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RII Rpt. Nos. 50-321/78-37, and 50-366/78-47 I-I DETAILS I Prepared by.
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)f F. S. Cantrell, Reactor 1 Date Reactor Projects Sectio / nspector n No. 1 Reactor Operations and Nuclear Support Branch Dates of Inspection: November 14-17, 1978 b
6%
iifiif18 Reviewed by:
H.' C. Dance, Chief Date Reactor Projects Section No. 1 Reactor Operations and Nuclear Support Branch 1.
Persons Contacted
- M. Manry, Plant Manager s*
- H. Nix, Assistant Plant Manager S. X. Baxley, Operating Superintendent
- T. V. Greene, Superintendent Engineering Services
- T. McCarley, Maintenance Supervisor D. Moore, Senior Training Specialist C. Locke, Senior Plant Engineer M. Seymour, Senior Plant Engineer H. Sumner, Senior Plant Engineer
- T.
C. Wilkes, Plant Security Supervisor
- D. J. Vaughn, Assistant Plant Security Supervisor
- H. W. Dyer, Operating Supervisor
- C. Miles, Jr., QA Field Supervisor, GPC
- C.
Belflower, QA Site Supervisor
- G. Spell, Jr., Senior QA Field Representative
- V.
Barrett, QA Field Representative
- R. Rogers, IE Resident Inspector
- Denotes those present at the exit i n t e rv i ew.
2.
Licensee Action on Previous Inspection Findings (Closed) (URI 321/77-21-01) and (Closed) (URI 321/78-04-02) Fire Dampers have been modified to meet certification requirements. Verified that a sample of damper serial numbers conf ormed to certifications listed in letter dated October 16, 1978, from Air Balance, Inc. Plant QA has open item to verify completion of modificatio.
RII Rpt. Nos. 50-321/78-37, and 50-366/78-47 I-2 (Closed) (URI 366/78-28-02) Reviewed list of valves in Unit 2 and main-tenance request for Unit 1 in which the studs that attach the operator to the valve were measured by ultrasonic testing. Questionable length studs were identified and replaced in six other valves.
(Closed) (01 366/78-24-04) Unit 1 RWCU line hydro tested at 1155 PSIG, Ll.RT test of containment extension at 59 PSIG, and weld integrity pneumatic test at 70 PSIG.
(Closed) (01 366/78-24-03) Page H-25 of primary containment integrated leak rate test indicated that local leak rate test of combined bypass leakage path met the acceptance criteria of 0.009 La.
(Closed) (URI 321/78-32-01) Check valves were installed in the constant air supply line to each f uel pool gate seal to prevent simultaneous deflation of the seals in the event of a loss of air supply (MR 78-3787).
(Closed) (URI 321/78-27-02) Fire _ dampers have been installed in the ventilation ducts between the 1 30 foot elevation control building ventilation room and the 112 foot elevation corridor and work area.
3.
Unresolved items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of noncom-pliance or deviations.
Three unresolved items disclosed during the inspection are discussed in paragraphs 6, 9a, and 9b of this report.
4.
Exit Interview The inspectors met with licensee representatives (denoted in paragraph 1)
on November 17, 1978, and reviewed the inspection scope and findings.
5.
Plant Status Unit 1 - Operating at about 951 power Unit 2 - Startup testing at about 50% power 6.
Fire Wall Between Fire Pump _s A 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated fire wall between the three fire pumps has been extended up to the ceiling of the pump house using 8 inch fire rated concrete blocks as required by item 36 of the Fire Protection Plan.
Neither the wall or the building are seismically qualified; however the service water system which is seismically qualified, can be cross connected to the fire mains to provide a backup water supply.
Specifications for
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RII Rpt. Nos. 50-321/78-37, and 50-366/78-47 I-3 the fire wall require every other course of concrete blocks to be reinforced in the mortar.
This item will be referred to IEIHQ for review.
(Unresolved item 321/78-37-01)
7.
Review of Nonroutine Events Reported By The Licensee The following licensee event reports were reviewed in office for potential generic problems, to detect trends, to; determine whether the inf ormation included in the report meets NRC reporting requirements, and to consider whether the corrective action discussed in the report appears appropriate. Licensee action with respect to select reports was reviewed to verify that the events were reviewed and evaluated by the licensee as required by Technical Specifications, that corrective action was taken by the licensee, and that safety lioits, limiting safety system settings, and limiting conditions for operations were not exceeded. The inspector examined selected Plant Nuclear Safety Committee minutes, incident reports logs, and records, and inspected equipment and interviewed selected personnel.
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The following LER's have been reviewed and closed.
LER No.
Date Subject
- 366/78-54 10/24/78 Excessive Leakage From Drywell To Torus
- 321/78-56 8/8/78 Core Spray Low Level Alarm Switch Failed Surveillance 321/78-59 7/31/78 Weekly Fire Test Missed
- 321/78-60 7/28/78 Condenser Low Vacuum Design Error That Could Prevent Initiation of MSIV Closure 321/78-72 9/25/78 Reactor Building Ventilation Monitor Failed Source Check
- 321/78-73 9/27/78 Breach of Fire Barrier
- 321/78-74 9/25/78 3 Penetrations in Ceiling of Cable Spreading Room Not Sealed 321/78-75 9/29/78 Monthly Fire Pump Test Miss-d 321/78-76 9/29/78 False Refueling Floor High Exhaust Air Activity Alarm
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RII Rpt. Nos. 50-321/78-37, and 50-366/78-47 I-4 321/78-78 10/6/78 Main Steam Line Radiation Monitor Surveillance Missed 321/78-79 10/13/78 Main Steam Flow Instrument Overranged
- 321/78-81 10/30/78 Intake Structure Service Water Valve Pit Flooded
- 321/78-83 10/25/78 Standby Service Water Pump Inoperable No items of noncompliance or deviations were identified. The inspector noted that the licensee had identified and taken corrective action as appropriate.
- Denotes additional review conducted at the site.
8.
Intake Structure Service Water Valhe Pit Flooded TEER 321/78-81)
A valve seat blew out and flooded the valve pit. The inspector reviewed the circumstance related to the event, the type valve involved (6" stream-seal butterfly valve by Allis-Chalmers), and other locations where this type valve is used. The licensee is revising maintenance procedures to include a caution on the use of this valve as an isolation valve with the mating flange removed.
From the design it appears that the valve would not be effective as an isolation valve in either direction with the adjacent flange removed. The inspector questioned (and the licensee agreed to consider) if a caution should be attached to all of these valves in the field as additional protection.
The inspector had no further questions.
9.
Organization and Administration Qualifications - Unit I technical specifications require that the a.
reactor engineer have a minimum of a bachelor's degree in engineering or the physical sciences, and two years experience in such areas as reactor physics, core management, core heat transfer, and core physica testing programs.
This is compatible with the Unit 2 technical specification requirements to meet or exceed the minimum qualifications of ANSI-N18.1-1971.
Three persons on the staf f were identified to the inspector as perf orming reactor engineering work. All three meet the education requirements.
(Another person identified as reactor engineer resigned in May 1978.) The person now identified basically as lead reactor engineer has been assigned responsibility f or the startup
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RII Rpt. Nos. 50-321/78-37, and 50-366/78-47 I-5 test program for Unit 2 and appears to be full time in that position.
The No. 2 man received his degree and joined Georgia Power in September 1976, when he was assigned to plant Hatch. It appears that he has been functioning as the reactor engineer for Unit I since May 1978.
(This position is covered by startup test engineers and the lead reactor engineer for Unit 2) As suc4 he does not appear to have the minimum experience required for the job.
The No. 3 man received his degree Sept' ember 1977, and joined the staff at Plant Hatch February 1978. He also is functioning as a reactor engineer. The inspector was not able to determine during the inspection (due to absences of some of the persons involved)
how the work and recommendations of the No. 2 and No. 3 reactor engineers has been reviewed to compensate for the experience of these persons.
A licensee representative stated that this infor-mation would be determined and reviewed by management.
This was identified as an unresolved item (321/78-37-02). Specific examples of problems that have occurred and are associated with reactor engineering are discussed in qgtails 11, Paragraph 6.
In a telephone conversation on November 22, 1978, the Assistant Plant Manager stated that the Plant Review Board had reviewed the qualifications of the reactor engineers, and concluded that the No. 2 reactor engineer was qualified, that he would prcvide an envelope of work that could be performed by persons under him, that the No. 1 or No. 2 man would review these activities at least daily, that an on shift reactor engineer would work through the shift foreman, and the shift foreman would document recommendations and results in the shift foreman's log. The inspector stated that the procedures for implementating the above would be reviewed during future inspections.
b.
Organization The Table of Organization specified in the Technical Specifications for Unit 1 (Figure 6.2-2) and Unit 2 (Figure 6.2.2-1) are different in the following respects:
quality control engineer reporting to the (1) Unit I specifies a Assistant Plant Superintendent (Manager).
The qualifications for this job were established based on the person in that position in January 1975.
He has left Plant Hatch.
The Unit 2 Technical Specifications require a senior QC specialist, and do not specify qualifications for this position.
This position is currently vacant.
The person acting in this position does not have all the qualifications specified for Unit.
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RII Rpt. Nos. 50-321/78-37, and 50-366/78-47 I-6 (2)
In Unit 1, the Senior Training Specialist reports to the Assistant Plant Superintendent (Manager).
In Unit 2, he reports to the Superintendent of Engineering Services. The Senior Training Specialist is the same for both units, and the operators and senior operators are licensed on both units.
The plant is organized in accordance with Unit 2 Technical Specifi-cations which were issued June 13, 1978. Georgia Power has requested a change to the Unit 1 Technical Specifications to make Section 6 compatable with Unit 2 Technical Specifications. This discrepancy is being identified to NRR for resolution (321/78-37-03).
10.
Review and Audit Minutes of the Plant Review Board (PRB) and the Safety Review Board (SRB) were reviewed at the plant to verify that plant operations are being reviewed and audited as required by the Technical Specifications.
No items of noncompliance or deviat$ons were identified.
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RII Report Nos. 50-321/78-37 and 50-366/78-47 Il-1
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DETAILS II Prepared by:
C,
/f,, M eg C. JuPTan, ea clor lnspec r
Fate Nuclear Support Section
Reactor Operations and N clear Support Branch Dates of Inspection:
vember 14-17, 1978
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Reviewed by:
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/t hyf R. D. 41a rtin, Chie f a
Nuclear Support Section No. 1 Reactor Operations and Nuclear Support Branch 1.
Persons Contacted:
Georgia Power Company
- M. Manry, Plant Manager
- H. Nix, Assistant Plant Manager
- C.
Miles, QA Field Supervisor
- G. E. Spell, Jr., Sr. QA Field Representative
- B.
Barrett, QA Field Representative S. Curtis, Associate Plant Engineer T. Gooper, Jr., Plant Engineer B. Coleman, Shif t Foreman The inspector also contacted various other members of the operations and technical staf f during the inspection.
- Denotes those present at the exit interview.
2.
Licensee Action on Previous l_nsp_ection Findings Not reviewed during this inspect ion.
3.
Unresolved items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of noncompisance, or deviations.
An unresolved item identified during the inspec tion is discussed in paragraph.
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RII Report Nos. 50-321/78-37 and 50-366/78-47 II-2 4.
Exit Interview The inspectors met with licensee representatives (denoted in paragraph 1)
at the conclusion of the inspection on November 17, 1978. The inspectors summarized the purpose and scope of the inspection and the findings.
The inspector stated his findings detailed in paragraph 6 that the repetition of exceeding core thermal hydraulic limits requires further corrective action. The Plant Manager stated that they would review the matter and propose corrective action.
The inspector stated that this will be an unresolved item.
5.
Review of Unit 1 Refueling During April 1978 Records and documentation relating to system restoration and initial startup after the Unit 1 cycle 3 refueling during April 1978 were reviewed. Procedures and data on tests of the primary coclant system, feedwater system, nuclear instrumentation, los pressure coolant injec-tion system, and control rod drivdEsystem were reviewed. The control rod withdrawal procedure and the verification of adequate reactivity shutdcwn margin were examined.
No items of noncompliance or deviations were identified.
6.
Followup Review of Licensee Reported NonRoutine Events The inspector conducted an onsite followup on the non-routine events below previously reported to the NRC by the licensee. The circumstances of the events were discussed with appropriate technical and operations personnel.
Pertinent logs, records, charts, and other documentation were examined. The inspector discussed with the licensee the extent and ef fectiveness of corrective action taken, LER 50-321/78-63 "APRM Trip Settings Non-Conservative" a.
The inspector verified that current procedures in use for adjusting the APRM gain to correct for a high total peaking factor incorpo-rate an effective Technical Specification limit os 2.38.
This value, which is more conservative than the actual stated Technical Specification limit of 2.48 on total peaking factor, accounts for in cycle 3 of some 8X8R tuel, which has a longer active the use length. The corrective action taken appears appropriate and the inspector had no further question.
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RII Report Nos. 50-321/78-37 and 50-366/78-47 II-3 b.
LER 50-321/78-89 " Inoperable Recirculation Pump Motor - Generator Set" On October 22, 1978 the A recirculation loop MG set became inoper-able due to brush failure.
Reactor operation was continued at reduced power for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as allowed by Technical Specification 3.6.J.2.
whereupon, a reactor shutdown was. begun.
Before the reactor was shutdown, the MG set was repaired and the A recircu-lation loop was returned to service. The inspector verified that Technical Specification requirements 3/4.6.D, 3/4.6.E, and 3/4.6.J.3 relating to the recirculation pump restart was met. Af ter discus-sion of these events with plant personnel, the inspector had no f urther questions.
c.
LER's 50-321/78-65, 78-71, and 78-86 " Maximum Linear Heat Generation Rate Exceeded" The inspector examined the process computer printout and discussed the circumstances and causestof these events with the reactor engineering staff.
LER 78-65 reported that between August 12, 1978 and August 14, 1978 the maximum linear heat generation rate specified by Technical Specification 3.11.B was exceeded four times with a maximum valve of 100.4% of the limit.
LER 78-71 rela *es that on September 11, 1978 the linear heat generation rate reached 100.01% of its limit and during a similar occurrence on October 16, 1978 a maximum of 100.2% of the limit was reached, reported by LER 78-86.
The inspector determined that after as each of these events, timely operator action was takes. to restore the reactor to conformance with Specification 3.11.B and review of these events is complete.
The licensee inf ormed the inspector that a similar event wherein the maximum linear heat generation rate was exceeded occurred on October 28, 1978. The licensee plans to report this event as LER 50-321/78-90 and the event will be reviewed in a later inspection af ter the report is received.
Repetittous occurances of exceeding the maximum linear beat generation rate (MLHGR) prescribed by the Technical Specifications in a relatively short time frame is indication that corrective action taken by the licensee has not been adequate. Af ter discus-sion of this matter the licensee agreed to review the situation and propose further corrective action to provide better assurance that the the rma l hydraulic limits set forth in the Technical Specification are not exceeded. This matter is considered unre-solve.
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RII Report Nos. 50-321/i8-37 and 50-366/78-47 II-4 Unresolved Item The repeated instances of exceeding MLHGR is considered unresolved pending the licensee's development of revised corrective actions as appropriate to preclude recurrence. These proposed corrective actions will be reviewed by the inspector during a subsequent inspection (321/78-37-04).
d.
LER 50321/78-88 " Core Maximum Total Peaking Factor Exceeded More Than Two Hours" The inspector discussed with reactor engineering personnel the circumr ' ances of this event. On October 21, 1978 during a xenon t ra ns i er.-
the core maximum total peaking factor exceeded Technical Specification value 2.1.A.1.C for more than two hours before sufficient corrective action was taken. This violates a verbal commitment made to the h'RC as documented in inspection report 50-321/78-23.
After reviewing this topic with the licensee, the inspector had no further quesdons.
7.
Operational Records Review A review was made of certain areas of operational record keeping.
Three deviation reports (HNP-425) initiated during October 1978 were selected and tracked through to final resolution in the Plant Review Board. Three plant patrol log sheets (HNP-1060) completed during the month of October 1978 were reviewed for completeness. The inspector found no problems with these items.
A review was made of representative strip chart recorder data of various important plant parameters generated during the month of October 1978. Samplings of the following recorder charts were examined.
Main Steam Line Radiation Level (DL-R603)
Drywell Pressure (T48-607A)
Suppression Pool Level (T48-607A)
APRM/ Rod Block Monitor (C51-R603B/C)
No areas of concern were identified f rom this data.
The licensee was unable to produce for review the strip chart data for October 1978 listed below due to belated filing of the strip chart data in the records roo..
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RII Report Nos. 50-321/78-37 and 50-366/78-47 II-5 Condensate Storage Tank Level (Pil-RG01)
RWCU Room Temperature (G31-R604/608)
APRM Level (C51-R603A/D)
The inspector stated that this data will be reviewed at a later inspec-tion and this review will be held as an open item (50-321/78-37-5).
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