IR 05000369/1979016
| ML19224D722 | |
| Person / Time | |
|---|---|
| Site: | McGuire  |
| Issue date: | 04/05/1979 |
| From: | Ford E, Kellogg P, Ouzts J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19224D717 | List: |
| References | |
| 50-369-79-16, NUDOCS 7907160195 | |
| Download: ML19224D722 (15) | |
Text
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e UNITED STATES pm REGoq NUCLEAR REGULATORY COMMISSION
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Report No.
50-369/79-16 Licensee: Duke Power Company Charlotte, North Carolina Facility Name: McGuire No. 1 Docket No.
50-369 License No.
CPPR-83 Inspection at McG re Sit
/M[ 7f Inspectors:
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,B. T. Moon Date Signed Accompanying Personnel: None
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Approved by:
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P. J. Kellogg, ction Ch uf, ROSS Branch Date Signed SUMMARY Inspection on March 5-9, 1979 Areas Inspected This routine, unannounced inspection involved 93 inspector-hours onsite in the areas of plant procedures.
Results Of the three areas inspected, no apparent items of noncompliance or deviations were identified.
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DETAILS 1.
Persons Contacted Licensee Employees
- C. W. Cage, Superintendent of Operations
- J. W. Cox, Station Senior QA Engineer W. hesser, Associate Engineer T. L. McConnell, Superintendent of Technical Services
- D. J. Rains, Superintendent of Maintenance J. Randales, Associate Engineer
- C. P. Rogers, I & E Engineer
- R. P. Ruth, Senior QA Engineer
- M. Sample, Technical Services Engineer R. White, Associate Engineer
- R. J. Wilkinson, Superintendent of Administration Other licensee employees contacted during this inspection included 3 operators, and 3 office personnel.
- Attended Exit interview.
2.
Exit Inte rview The inspection scope and findings were summarized on March 9, 1979, with those persons indicated in Paragraph I above. During the inspection, four items were identified which require further management attention.
These items are contained in paragraphs 5, 6, 7a(7), 7b(7).
The licensee acknowledged the inspector comments on these items.
3.
Licensee Action on Previous Inspection Findings Not inspected.
4.
Unresolved Items Unresolved items were not identified during this inspection.
5.
Mainter.ance Procedu e Verification The inspector conducted a review of the following maintenance procedures and maintenance control documents:
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Maintenance Procedures
-- MP/0/A/2005/01 - Diesel Cenerator Inspection and Maintenance MP/0/A/7150/03 - Boron Recycle Reactor Makeup Water Pump Removal
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and Replacement
-- MP/0/A/7150/04 - Component Cooling Pump Corrective Maintenance MP/0/A/7150/07 - Ice Condenser Intermediate Deck Doors Corrective
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Maintenance MP/0/A/7150/11 - Containment Spray Pump and Residual Heat Removal
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Pump Corrective Maintenance
-- MP/0/A/7150/37 - Pressurizer Safety Valve Removal and Replacement
-- HP/0/A/7150/42 - Reactor Vessel Mead Removal and Replacement
-- MP/0/A/7150/44 - Safety Injection Pump Corrective Maintenance MF/0/A/7200/01 - Auxiliary FWP Turbine Governor Valve Corrective
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Maintenance
-- MP/0/B/7300/01 - Pump Preventative Maintenanc-
-- MP/0/A/7600/25 - Upper Head Injection Isolation Valve Corrective Maintenance MP/0/A/7650/10 - Calibration of 0-24 "Outside Micrometer
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-- MP/0/A/7650/11 - Calibration of Snap-on Click Iype Torque Wrenches MP/P/A/7650/13 - Calibration of 8" - 33" Inside Microme'.ers
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-- MP/0/A/7650/18 - Calibration of the 25 Series - Dial Indicators up to 1"
-- MP/0/A/7650/19 - Calibration of the Crankshaf t Distortion Gage
-- MP/0/A/7650/25 - Calibration of Depth Micrometers MP/0/A/7650/39 - Calibration of Dial Indicators Using Gage Blocks
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-- MP/0/A/7400/42 - Diesel Load Control Mechanism Corrective Maintenance 357 180
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-- MP/0/A/7150/34 - Pressurizer Manway Removal and Replacement
-- MP/0/B/7650/05 - Crane and Hoist Safety Inspection
-- PT/0/A/4150/01A - Reactor Coolant System Leak Test Control Documents
-- Station Directive 2.1.1 - Control of Master File Documents
-- St3 tion Directive 4.2.1 - Handling of Station Procedures
-- Station Directive 4.7.0 - Control of Maintenance Program
-- Station Directive 3.3.0 - Determination of Safety Related or Control Designation, Structures, Systems and Components The above maintenance procedures were reviewed to verify that:
-- Administrative controls were established for review, approval and periodic updating of maintenance procedures
-- Administrative procedures and directives were observed in the preparation and handling of maintenance procedures
-- Safety reviews were conducted and recorded as required by 10 CFR 50.59
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-- Controls were established and followed in the preparation of procedure format and content and for issuing new and revised procedures
-- Controls were established for temporary changes to procedures
-- Changes to procedures will be approved by the same organization that approved the original procedure, or other qualified designated organization
-- The training organization was appraised of changes to procedures The iraspector used one or more of the following acceptance criteria for evaluating the above items used in the procedure review:
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-- ANSI N18.7 Section (5) (1976)
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-- ANSI N45.2 Section (6) (1971)
-- Draft Technical Specifications (6.8)
-- FSAR Section 17.1.5
-- DPC0 Topical Report 17.1.5
-- Administrative Policy Manual (2.1) (2.2) (3.3) (4.2) (4.3) (4.7)
Within the areas inspected no items of noncompliance or devissions were identified. Two procedure discrepancies were identified that the licensee agreed to correct.
Enclosure 13.2 to maintenance procedure MP/0/A/2005/01 " Diesel Generator Inspection and Maintenance", gave no acceptance criteria for the high voltage generator test, or the field megger test. The licensee agreed to add high voltage and megohm acceptance criteria to Enclosure 13.2 for the two respective tests.
Section 5, " Plant Status, of MP/0/A/7150/42
" Reactor Vessel Head Removal and Replacement", had the plant in cold shutdown prior to starting head removal.
Prior to removal of the reactor vessel head the plant must be in refueling shutdown. The licensee agreed to revise this procedure to specify the correct plant status.
The revision to these two procedures to include these changes will be carried as an item (369-79/16-01) and will be reviewed by the NRC at a subsequent inspection.
6.
Plant Procedures Verification The inspector conducted a review of the following procedures:
Operating Procedures
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OP-1-A-6100-01 Controlling Procedure for Unit Start Up
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OP-1-A-6100-02 Controlling Procedure for Unit Shutdown
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OP-1-A-6150-01 Filling and Venting the Reactor Coolant System
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OP-1-A-6150-04 Pressurizer Relief Tank
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OP-1-A-6200-04 Residual Heat Removal Systems
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OP-1-A-6200-09 Accumulator Operation
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OP-1-A-6200-14 Refueling Water System
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OP-1-A-6400-)5 Component Cooling Water System 357 i82
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Administrative 9rocedurec
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Station Directive 1.5 Administration of the Manual
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Station Directive 3.1.1 Reactor Operators Logbook
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Station Directive 3.1.9 Relief at Duties of Plant Operations
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Station Directive 3.1.13 Unit Supervisors Logbook
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Station Directive 4.2 Permanent Station Procedures Periodic Test Procedures:
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PT-0-4150-01A Reactor Coolant System Leak Test
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PT-0-A-4150-05 Pressurizer Safety Valve Setpoint Test
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PT-1-A-4200-06B Boron Injection Valve Lineup Verification
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PT-1-A-4200-08A NI Check Valve Leakage Test
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PT-0-A-4250-C.A Main Steam Isolation Valve Movement Test
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PT-0-B-4250-04A Weekly Turbine Test
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PT-1-A-4350-06 4160V. Essential Power System Test
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PT-0-A-4350-13 120V. Vital AC Power System Test
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PT-1-A-4400-02A Nuclear Service Water Train IA Performance Test
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PT-1-A-4600-01 RCAA Movement Test The above procedures were reviewed to verify that the following were in accordance with acceptance criteria:
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reviews, approvals, and periodic updating proper format
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inclusion of precautions, limitations, and referencing of Technical Specifications Administrative Controls were reviewed for:
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the issuance, contents, review and updating of Station Directives 357 183
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the preparation and content of operating logs
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nroper shift turnover activities
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a method for controlling temporary changes to procedures
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that training received changes to procedures
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a documented method for control of procedures The inspector used one or more of the following acceptance criteria for evaluating the above items used in the procedure review:
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..NSI N18.7 (1976)
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Duke Power Company Administrative Policy Manual
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FSAR section 13.5
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Draft Technical Specifications section 6.8 The inspector discussed the mechanism fcr. periodic revision and review of station temporary and permanent procedures with a licensee represen-tative and was told that this existed in the form of a periodic test.
The licensee further stated that a statement would be included in the Station Directives Manual referring to this method of periodic review for revision.
The inspector had no further questions at this time.
This item will be carried as an item (369/79-16-02) and will be reviewed by the NRC during a subsequent ir.spection.
7.
Emergency and Instrument Procedures The inspector reviewed and evaluated adequacy of eighteen emergency y
procedures and twenty six instrument procedures plus two station direc-tives to.erify that the contents and formats of the procedure are in accordance with the quality assurance standards and guides tontained in the Duke quality assurance program.
The program was evaluated for conformance to the following acceptance criteria:
(1) Sections 4.1(1), 5.2.2, 5.2.7, 5.3.9.1, 5.3.2.9, and 5.3.2.3 of ANSI 618.7-1976.
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(2) Section 6 of ANSI 45.2-1971.
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(3) Appendix B of 10 CFR 50.
(4) Section 4.2.3.4 of Duke Power Company Administrative Policy Manual for Nuclear Stations.
(5)
Industry Practices.
a.
Emergency Procedures The procedures reviewed were:
EP0A5000-01
" Reactor Trip" EPOA5000-02
" Turbine Generator Trip" EPOA5000-03
" Loss of Reactor Coolant" EPOA5000-04
" Steam Generator Tube Failure EP0A5000-07
" Steam or Feedwater Line Rupture" EPOA5000-08
" Loss of Reactor Coolant Flow" EP0A5000-09
" Boron Dilution" EP0A5000-10
" Inoperable Control Rod" EPOA5000-12
" Loss of Residual Heat Removal" EP0A5000-14
" Loss of Steam Generator Feedwater" EP0A5000-15
" Loss of Makeup or Letdown" EPOA;000-16
" Loss of Component Cooling" EP0A5000-17
" Loss of Nuclear Service Water" EP0A5000-18
" Loss of Condenser Vacuum" EPOA5000-22
" Loss of Control Room" EP0A5000-23
" Loss of Containment Integrity" EP0A5000-25
"High Activity in Reactor Coolant" EPOA5000-26
" Nuclear Instrument System Malf unction" Station Directive 4.7.1 " Handling of Station Procedures" The inspector identified thirteen items which appeared inadequate to meet the requirements of ANSI N18.7-1976, ANSI N45-1971, Appendix B of 10 CFR 50, Section 6.8 of McGuire Draft Techrical Specification, and Duke Power Company Administrative Policy Manual for Nuclear Stations. The licensee cammitted to make appropriate changes and additions to the procedure for twelve of these items listed below. One item remains outstanding:
(1) The inspector identified in procedure EP0A5000-03, " Loss of Reactor Coolant" that:
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a)
The IFD-15 "ND Ux 1B Outlet Crossover Block VLV" is not opened during the " Hot Leg Recirculation" phase, which will possibly limit the total capacity of the injection flow during the LOCA.
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Some cf the automatic actio.s important to be verified duiing the LOCA such as " Diesel Generator Start" and
" Automatic Control of Auxiliary Feedwater Valves on Accident Signal", are not identified and verified accord-ingly.
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c)
The symptom statements for large LOCA (Case "B") are not written, contrary to Section 5.3.9.1(2) of ANSI N18.7-1976, in a manner easily distinguishable from a small LOCA (Case "A").
d)
A confirmation was lef t out, contrary to Section 5.3.9.1(4)(f) of ANSI N18.7-1976, that the containment annulus area of the shield building are operating properly to prevent uncontrolled release of radioactivity.
The licensee representative agreed to incorporate the pro-visions itemized in 1(a),1(b) and 1(d) above, respectively, in the procedure.
The representative stated that the item 1(c) above will be incorporated by re-arranging the symptoms in Case "B" (Large LOCA) so that the " Ice Ccadenser Doors Open Alarm" will be moved to step 1.3 where the large LOCA can be characterized with other two steps, 1.1 and 1.2.
(2) The inspector identified in Procedure EP0A5000-12, " Loss of Residual Heat Removal" that:
a)
An automatic action, " Isolation of RHR" is not verified during the emergency. This is identified in Sections 5.3.9.1(3) and 5.3.8.1(4)(F) of ANSI N18.7-1976.
b)
Plant Personnel are not notified, contrary to Section 5.3.9.1(4)(c) of ANSI N18.7-1976, of the nature of the emergency.
c)
The step 3.6 does not include operator action to handle reactor coolant pressure excursion when the reactor coolant system is operated in the water solid condition and the RHR system is lost.
The control of reactor coolant system temperature via S/G as described in the step 3.6 may not be sufficient in this case, and it may, as an industry practice, need control of the pressure by regulating flow rates of charging / letdown system.
The licensee representative agreed to incorporate the provisions listed in items 2(a) and 2(b) in the procedure. The representa-tive also agreed to incorporate the provision itemized 2(c)
above by refering to the appropriate procedures in the OP6100-02 for detail.
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-9-(3) The inspector identified that procedure EP0A5000-23, " Loss of Containment Integrity" does not include the provisions of Section 5.3.9.1.(4).(f) of ANSI N18.7-1976 to determine if the incident has resulted in a release of any radioactivity to the environment by checking radiation monitors and various vent monitors.
The licensee representative agreed to incorporate the above provision in the procedure.
(4) The inspector identified that procedure EP0A5000-18, " Loss of Condenser Vacuum" does not include an automatic action to verify an absence or loss of steam dump capability via C-9 permissive interlock. The licensee agreed to incorporate the above provision in the procedure.
(5) The inspector identified that procedure EP0A5000-22, " Loss of Control Room" does not include an operator's immediate action to notify to plant personnel of " Reactor Trip, Abandoning Control Room".
The licensee representa+.ive agreed to incor-porate the above notificat'on in the procedure.
(6) The inspector identified in Procedure EP0A5000-07, " Steam or Feedwater Line Rupture' that:
a)
More symptoms are needed to distinguish the "Feedwater Line Rupture" incident from the " Steam Line Rupture".
For instance, "the Steam Flow Feedwater Flow Mismatch" symptom should include magnitudes and directions in order to be separated as a dif ferent alarm for each break.
In addition, more symptoms should be added to account for the initial reactor cooling system heatup resulting from the feedwater line break upstream of a check valve.
b)
In step 3.2.5, the subsequent action does not include an identification of steam generator af fected by a feedwater line rupture, nor does it include the location of such breaks (upstream or downstream of check valve) in the line.
The licensee representative agreed to incorporate the provisions itemized above in 6(a) and 6(b) respectively, in the procedure.
(7) The inspector identified that throughout the emergency procedures reference statements regarding Technical Specifications are, in general, not specific in terms of incorporating pertinent requircaent thereof into a particular procedure, nor does it specify applicable Technical Specification Numbers, aor does it clearly state the purpose of 357 187
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references.
It is, for example, stated in Step 3.1 of EP(12),
Step 3.7 of EP(9), Step 3.3 - Case 1 of EP(10), and Step 3.9 -
Case 1 (Step 3.7 - Case 2) of EP(17), respectively that,
" Refer to Technical Specification". The statement is ambiguous and contrary to the intent prescribed in Sections 5.3.9.1.(5) and Section 5.3.2.3 of ANSI N18.7-1976 that " Steps should be included to return the reactor to a normal condition or to provide for a safe extended shutdown a normal condition or to provide for a safe extended shutdown period under abnormal or emergency conditions" and " Reference to Technical Specifications should be included in procedures as applicable.
References should be identified within the body of procedures.." respectively.
Furthermore, Section 6 of ANSI N45.2-1971 states in part the intent of using " operating limits" that, " Quantitative Criteria such as operating limits
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to conform to the standard is indicated in the Section 4.2.3.4.(b) of the company's Administrative Policy Manual for Nuclear Stations, where it states that, " Applicable reference material, particularly the station's FSAR and Technical Specification, shall be reviewed and pertinent requirements thereof incorporated into each procedure". The requirement for the review is stated in part in Section 4.1(1) of ANSI N18.7-1976 that, "to verify that these activities are per-formed in conformance with this standard and with company policy and rules".
The inspector stated that either the pertinent requirements from the Technical Specification should be incorporated in each procedure or Technical Specification number should be provided in order to comply with the ANSI standard and the company policy.
This matter of insufficient reference information and agreed to changes will be carried as item 369/79-16-03 pending corrective action by the licensee and verification by the NRC at a subsequent inspection.
b.
Instrument Procedures The procedures reviewed were:
IP0A300-01 " Reactor Coolant System Flow Calibration" IPOA3000-03A " Accumulator Tank Pressure Calibration" IP0A300-05 "Tavg Auctioneered and Tavg Dev. Alarms Calibration" IPOA300-05B "Tavg/ Tref Dev. Calibration" 357 188
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IP0A3000-05C " Delta T/Tavg Protection Calibration" IP0A3000-06 "0 TDT Channel '.libration Protection I, II, III and IV" IP0A3000-09H " Pressurizer Vapor Temp Calibration" IP0A3000-09L " Process 7300 Pressurizer Level Control" IP0A3001-02 " Main Steam Line Pressure Calibration" IP0A3002-03 "Feedwater Control System Loops 1, 2, 3, and 4 Calibration" IP0A3003-03A " Letdown Orificies Safety Valve Temperature" IP0A3003-04F " Boron Injection Tank Temperature Loop Calibration" IPOA3007-01 "Nul Instrumentatien System Audio Count Rate Channel Adj us tments" IP0A3007-02 "NIS Flux Deviation Alignment" IPOA3007-02A "NIS Miscellaneous Control and Indication Section Checkout" IP0A3007-03 "NIS Source Range Checkout" IP0A3007-04 "NIS Intermediatt Range Operational Checkout" IP0A3007-07 "NIS Power Range (perational Checkout at Power" IP0A3007-07B "Th Wide Range Temperature Calibration" IPIB3007-10B "Incore Thermo-Couple Connection and Checkout" IP0A3008-02 " Turbine Impulse Chamber Pressure Calibration" IP0A3010-05 " Solid State Protection System Periodic Tests" IP0B3012-09 " Full Length Rod Control System Checkout Procedure" IP0A3050-06A " Recycle Holdup Tank Level Control Calibration" IP0A3061-01 "125V Vital Batteries Daily / Monthly Inspection" IP0B3211-01 " Control Rod Drive System Logic Cabinet Power Supplies" 4 e
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-12-Station Directive 3.1.19 " Safety Tags, Lockouts and Delineation Tags" The inspector identified twenty two items of the procedures which appeared inadequate to meet the requirements of ANSI N18.7-1076, Appendiy B of 10 CFR 50 and Industry Practices. The licensee cocmitted to make appropriate changes and additions for the twenty one items listed below under six different categories.
One item remains outstanding.
(1) The inspector identified in the following procedures that recordings of test and calibration results are not carried out on the contrary to Sections 5.2.7 and 5.2.2 of ANSI N18.7-1976, which state in parts that, "meatares to document t!'e performance thereof sh ' be established" and "necessary data should be recorded as the task is performed" respectively.
The procedual steps need to be recorded are (a)
IPOA3000-03A(Comparator test data), (b) IPOA3000-05C (Steps 10.11, 10.21 and 10.28), (c) IP0A3000-9L (Step 10.6.3), (d)
IP0A3002-03 (Step 10.4.3), and (e) IP0A3003-04F (Step 10.3.4, etc.)
(2) The inspector identified that "Signoffs" are not incorporated in the prerequisite sections of the procedures, (a) IPOA3008-02 and (b) IPOA3000-9L.
This is contrary to Section 5.2.2.(3) of ANSI N18.7-1976, which states in part that, "Verfication of completion of significant steps,, by initials or signatures on checkoff lists."
The licensee rept sentative agreed to incorporate provisions commented items (1) and (2) above i.nto the procedures.
(3) The inspector identified that +he test results from the procedures listed below can not be entered in the data sheet, because the units do not agree with each other. The proce-dures needed to be changed are:
(a) IP0A3000-01; (b)
IP0A3000-05 (step 10.2.1); (c) 1P0A3003-03A (step 10.12).
The licensee representative agreed to change the data r.heet percentage steps to agree with the procedure itemized 3(a) above and to make the data sheet input percentages agree with step 10.2.1 for item 3(b) above.
The representative also agreed to list on the data sheet the input percentages for item 3(c) above.
(4) The inspector identified that procedure IP0A3001-02 employed incorrect tollerance values in step 10.4.3.
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-13-The licensee representatice agreed to change the tollerance to a correct value (+0.02 V).
(5) The inspector identified that test points, component numbers, and appropriate connections are not clearly defined nor specified throughout the following procedures:
a.
IP0A3050-06A (Step 10.5, etc.)
b.
IPOA3000-05C (Drawing)
c.
IP0A3003-04F (Step 10.5.1)
d.
IPOA3008-02 (Step 10.4.2)
e.
IP0A3000-9L (Step 10.1.1)
The licensee representative agreed to define " correct input" values on the data sheet for item 5(a) above, add " Test Jeck Number" to the drawing for item 5(b) above, add "RTD Number" for item 5(c)
above, specify " Resistor Number" for item 5(d) above, and specify the specific corrections to be removed for item 5(e) above, respectively.
(6) The inspectcr identified that there was numerous typographical errors, missing enclosures and missing " Noter" statement which denotes that certain steps have to be followed in a reverse ord3r in the procedures listed below.
a)
IP0A3000-9L (Step 10.8.1)
b)
IP0B3012-09 (Step 7.14.7)
c)
IPOA3003-03A (Step 10.8.1)
d)
IP0A3011-02 (Steps 10.11.4 and 10.11.7)
e)
IPOA3007-02A (Steps 10.21, 10.30, and 10.40)
The licensee agreed to add " Figure 11.3.7" on the picture of Enclosure 11.3 for item 6(a) above, change " Figure 1" to " Figure 2" in step 7.14.7 for item 6(b) above, mark data sheet as Enclosure 11.1 for item 6(c) above, add enclosures 11.1 for item 6(c) above, add enclosures 11.3.1 and 11.3.2 to the procedure for item 6(d)
above, and place notes as in Step 10.12 for item 6(e) above, respectively.
(7) The inspector identified that most of instrument procedures do not have appropriate provisions of " checkoffs" by which
" step-by-st ep" proceedings as well as " Verification of Significaat Steps" tan be performed. These requirements are listed in Sections 5.2.2.(1) and (3) of ANSI N18.7-1976, respectively. This is also contrary to Section 5.3.2.9 of ANSI N18.7-1976, where it states in part that " Complex procedores should have check-off lists".
Some of the IP's like IP0A3000-03A are very complex requir:ng step-by-step T57 i91
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-14-verification, by jumping back and forth, in order to follow the sequence of test described in the procedure. Also other procedures inc1t de important steps which involve turning switches to put the reactor in a half-tripped condition while unit is in operation and are considered to be of sufficient importance to be verified by a " checkoff".
The inspector stated that the complex procedures as such should have pro-visions of " checkoffs" in accordance with the ANSI s'.andard and industry practice.
This lack of checkoff in complex procedures and agreed to changes will be carried as an item (369/79-16-04) pending corrective actions by the licensee and verification by the NRC at a subsequent inspection.
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