IR 05000321/1979016
| ML19242D550 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 06/22/1979 |
| From: | Lim J, Moon B, Ruhlman W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19242D537 | List: |
| References | |
| 50-321-79-16, 50-366-79-20, NUDOCS 7908150407 | |
| Download: ML19242D550 (6) | |
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UNITED STATES
8\\ d,h. %._ g' E NUCLEAR REGULATORY COMMISSION $. R EGION 11 o,,;,JL;...;j'f 1" P.1 ARIE TT A sT., N W., sUlTE 3100 % '\\. j" ATLANT A, GLORGIA 30303 ..... Report Nos. 50-321/79-16 and 50-366/79-20 Licensee: Georgia Power Company 270 Peachtree Street, N. W.
Atlanta, Georgia 30303 Facility Name: Edwin I. Hatch Units 1 and 2 Dccket Nos. 50-321 and 50-366 License Nos. DPR-57 and NPF-5 Inspection at Hatch Site near Baxley, Georgia Inspector: 1J t'7)7
B. T. Moon 4/ Date Sign 6d Accompanying Personnel: . C. Lim Approved by: Is! /q [[ uj ' '. A W. A. Ruhlman, Acting Section Chief, RONS Branch Date Signed SL? DIARY Inspection ot June 5-8, 1979 Areas Inspected This routine, unannounced inspection involved 24 inspector-hours onsite in the areas of Plant Procedures for Units I and 2.
Results Of the one area inspected, no apparent items of noncompliance or deviations were identified.
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DETAILS 1.
Persons Contacted Licensee Employees
- M. hanry, Plant Manager
- C. E. Belflower, QA Site Supervisot
- B F. Barrett, QAFR
- C. H. Coggin, Engineering Supervisor
- C. T. Moore, Assistant Plant Manager
- R. A. Day, Documentation Supervisor
- R. Nix, Superintencent of Maintenance Othe; licensee emplcyees contacted included one operator and one office personnel.
NRC Resident Inspector R. F. Rogers
- Attended exit interview.
2.
Exit Interview The inspection scope and findings were summarized on June 8, 1979,with those persons indicated in Paragraph I above.
The inspection scope and findings of the construction branch inspector were also discussed at this exit interview and the details are to be reflected in his inspection report (50-321/79-15).
3.
Licensee Actioa on Previous Inspection Findings Not inspected 4.
Unresolved Items Unresolved items were not identified during this inspection.
5.
Plant Procedure Verification for Units I and 2 The inspector conducted a review of procedures and documentations as follows: Operating Procedures for Unit 1: liNP-1-NOP-01001, " Normal Startup" ID;P-1-NOP-01020, " Normal Reactor Shutdown" ID;P-1-NOP-01025, " Fast Reactor Shutdown" IINP-1-NOP-01100, "llPCI System" 6'3"C 30 6310C5
. t-2-HNP-1-NOP-01125, "RCIC System" HNP-1-NOP-01350, " Fuel Pool Cooling and Cleanup System" HSP-1-NOP-01400, " Standby Liquid Control System" HNP-1-NOP-01435, " Reactor Recirc Pump M-G Sets" HNP-1-NOP-01474, "Drywell Ventilation System" HNP-1-NOP-01501, "Drywell/ Torus Diff. Pressure System" HNP-1-NOP-1666, "120V AC RPS Power Supply System" Operating Procedures for Unit 2: HNP-2-NOP-01100, "HPCI System" HNP-2-NOP-01125, "RCIC System" HNP-2-N0P-01350, " Fuel Pool Cooling and Cleanup System" HNP-2-N0P-01400, " Standby Liquid Control System" HNP-2-NOP-01401, " Mixing Standby Liquid Control Solution" HNP-2-NOP-01435, " Reactor Recirc Pump M-G Sets" HNP-2-N0P-01474, "Drywell Ventilation System" HNP-2-NOP-01501, "Drywell/ Torus Diff.
Pressure System" HNP-1-NOP-01666, "120V AC RPS Power Supply System" Emergency Procedures: HNP-N0P-1902, " Pipe Break Inside Primary Contai =ent" HNP-NOP-1912, "230KV System Failure (Loss of Offsite Power)" HNP-NOP-1922, "Inadvertant Initiation of ECCS" HNP-NOP-1923, " Loss of All Feedwater" HNP-NOP-1934, " Torus Temperature Above 95 Degrees F" Abnormal (Alarm) Procedures: HNP-ARP-02029, "RCIC Isolation Signal" HNP-ARP-02050, "HPCI Isolation Trip Signal Initiated" Maintenance Procedures: HNP-MNT-06021, " Relief VLV Piping in Torus Visible Inspection" HNP-MNT-06035, "LPRM Removal and Installation" HNP-MNT-06310, " Standby Liquid Con col Syst2m Maintenance" HNP-MNT-06450, " Diesel Generator System h untenance" HNP-MNT-06545, "RPS M-G Set System Maintenance" Administrative Procedures: HNP-ADM-9, " Procedure Writting and Control" HNP-ADM-10, " Document DistriStion and Control" HNP-ADM-25, "QA Review of Plam Procedures" HNP-ADM-808, " Document Change Request" HNP-ADM-814, " Procedure Review" HNP-ADM-818, " Temporary Procedure Change" HNP-ADM-830, " Procedure Perfor,3nce" HNP-ADM-831, " Tech Specs Surve ance Program" l W.J.00G
. . t-3-Surveillance Procedures: HNP 1-NOP-01050, " Surveillance Checks" HNP-2-NOP-01060, " Daily Rounds" HNP-2-SRV-03201, " Core Spray Pump Operability" HNP-2-SRV-03204, " Core Spray Logic System Functional Test" HNP-2-SRV-03207, " Core ' pray Sparger Instrument DP FT & C" HNP-2-SRV-03209, " Core Spray Automatic Actuation" HNP-SRV-03456, "RPS M-G Set Overvo? tage Under Frequency and Undervoltage FT & u" Other Procedures: HNP-CCP-7022, " Sodium Pentaborate Analysis and Monitoring Program" HNP-1-CCF-09003, "APLEGR & LHGR Calculation" HNP-1-CCF-09008, "MCPR Calculation Using TBAR" HNP-1-MIS-10921, "Incore Neutron Flux Mon. Removal" HNP-2-11007, " Maintaining Reactor Water Level While Torus is Drained" Standing Orders 79-3 and 79-4, "RPS M-G Voltage Readings" Technical Specification Chauges for Unit 1 (Unit 2) Amendment 65, 3-26-79 Amendacnt (6), 3-2-79 Amendment (4), 2-27-79 Amendment 63, 2-9-79 Amendment 62 (3), 1-22-79 Amendment (2), 10-13-78 Amendment 59, 8-14-78 Amendment (1), 8-11-78 Amendment 58, 6-26-78 Amendment 53, 4-12-78 Plant Review Board (PRB) Meeting Miuetes: Meeting Numbers 93, 96, 97, 98, 99, 102, and 107 The above procedures were reviewed to verify that: Reviews, approvals and changes covering t.te activities were in accordance - with Technical Specifications.
Where required procedure changes were made to reflect changes required - by selected Technical Specification revisions.
PRB meeting minutes relfected that safety reviews were cade and recorded - in conformance with 10 CFR 50.59(a) and (b).
Procedure contents were in accordance with Technical Specifi ations - and applicable standards.
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-4- - Selected procedure contents were adequate to control required safety related operations.
- Controls were established to prevent the freezing of functional fluid systems during maintenance.
The inspector used one or more of the following acceptance criteria for evaluating the above items in the procedure review: - Regulatory Guide 1.33-1978 (Revision 2) ANSI N18.7-1976 - - GPC QAM of the Hatch Nuclear Plant FSAR, Section 17.2 - - Technical Specification, Section 6.8 Industry Practices - Within the areas inspected, no items of noncompliance or deviations were identified. The following items remain open pending corrective action by the licensee and verification by h3C at subsequent inspections: a.
Technical Specification Amendment 63 (2/9/79) requires that HPCI and RCIC turbine trips due to the temperature differential of the steam line isolation valve, be deleted f rom Unit One procedure. This require-cent was not implemented by the licensee in the applicable procedures ARP 2029, "RCIC Isolation Signal" and ARP 2050, "HPCI Isolatica Trip Signal Initiated".
The inspector stated that, although this particular case is a failure to implement a less conservative requirement, the current administrative procedure ADM 808, "Docwnent Change Request", does not include adequate procedural controls over timely implementation of technical specification amendmentu and appropriate review process to make sure that the change is accomplished as soon as possible.
The licensee agreed to revise the administrative procedure to implement the above provisions by August 8, 1979 (321/79-16-01, 336/79-20-01).
b.
Abnormal Procedure ARP 2050, "HPCI ISO Trip Signal Initiated", for Unit 1 does not address the required actions that would bring the unit into a saf e (normal) condition af ter the HPCI Isolation occur ed.
The inspector stated that the words "Take appropriate action " should be expended in detailed procedures.
The licensee co:=itted to revise the procedure by August 1, 1979 (321/79-16-02).
Emergency Procedure NOP-1902, " Pipe Break Inside Primary Containment" c.
for Unit 1 does not contain the adequate procedural contents in case of a design base "LOCA" as cutlined below: G31008
. - a . t 5-(1) Magnitude of parameters at which ADS is initiated (i.e., Reactor pressure, etc.)
(2) Procedural steps to reduce hydrogen buildups in the drywell including the use of fans and/or hydrogen recombiner, if applicable.
(3) Specifications of emergency buses the diesel generator ; are suppose to tie in and verifics? ion of appropriate power sources that are supposediy availab1: during the LOCA.
(4) Caution statement not to override the ECCS injection unless reactor water level is adequately maintained by observing more than one instrument parameters.
The inspector found that the similar procedure for Unit 2 has been currently revised and include above outlines. The licensee agreed to revise the Unit 1 procedure up to the same standard as listed in Unit 2 procedure or more by August 1,1979,(321/79-16-03).
d.
Regulatory Guide 1.33-1978 (Revisica 2) requires the licensee to have a LOCA Procedure including the small pipe break inside the primary containment.
Contrary to the above, emergency procedure NOP-1902 " Pipe Break Inside Primary Containment" for both units does not inclui provisions for such an accident. The inspector state 3 the licensee shall 1 rovide such instructions and also the Symptom statement of such an accident shall be distinguishable f rom the "large LOCA". The licensee committed to revise the procedure by August 1, 1979 (321/79-16-04, 336/79-20-02).
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