IR 05000397/1982001
| ML20049J950 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 02/26/1982 |
| From: | Dodds R, Elin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20049J946 | List: |
| References | |
| 50-397-82-01, 50-397-82-1, NUDOCS 8203290249 | |
| Download: ML20049J950 (9) | |
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U. S. NUCLEAR REGULATORY COMMISSION
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I Report flo. 50-397/82-01 Docket No. 50-397 License No. CPPR-93 Safeguard; Group
Licensee: Washington Public Power Supply System P. O. Box 968
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Richland, Washington 99352 Facility Name: Washington Nuclear Project flo. 2 (WNP-2) ~
Inspection at: WNP-2 Site, Benton County,' Washington Inspection conducted: January ll-15 and February 1-5, 1982 Inspectors: [ h.
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gi). O. Elin, Reactor Inspector Date Signed
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Approved by:
R. 7. Dodds, Chief, Reactor Projects Section 2 Date' Signed Reactor Construction Project Branch
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Sununary:
Inspection during the period of January 11-15 and February 1-5, 1982
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(Report No. 50-397/82-01)
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Areas Inspected: Routine unannounced inspection by regional based inspector
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of instrumentation installation and licensee event reports. The inspection activities involved 62 inspector-hours onsite and 20 inspector-hours in the i
regional office by one NRC inspector.
Results:
No items of noncompliance or deviation were identified.
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8203290249 820301
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PDR ADOCK 05000397 O
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DETAILS
1.
Persons Contacted a.
Washington Public Power Supoly System (WPPSS)
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+*C. S. Carlisle, Deputy Project Manager
+*R. T. Johnson, Project Quality Assurance Manager
+*C. Dickenson, Construction Process
+ L. C. Floyd, Quality Assurance Engineer
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i J. Rhoads, Supervisor Equipment Qealification Engineering b.
Bechtel Power Corporation (BPC)
- D. R. Johnson, Manager of Quality
+*M. Jacobson, Quality Assurance Engineer
+*D. Cosgrove, Quality Assurance Engineer i
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Burns and Roe, Inc. (B&R)
S. Flanzenbaum, Senior Stress Analysis (Woodbury)
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C.Chung,GroupSupervisorStressAnalysis(Woodbury)
- M.L.Bursztein,ProjectEngineer(site)
+H. R. Tuthill, Quality Assurance (site)
J. O'Donnell, Group Supervisor (Woodbury)
P. Hickey,' Engineer (Woodbury)
J. Verderber, Project Engineering Manager (Woodbury)
J. Blass, QA Manager (Woodbury)
A. Freeberg, Instrument Engineer (site)
d.
Johnson Controls Incorocrated (_ Contract 220)
R. Grant, Quality Assurance Manager L. Reader,- Engineerina Manager R. Nickles, Field Superintendent e.
Fishback/ Lord Inc. (Contract 218)
W. L. Brown, Quality Assurance Manager T. Roselli, Field Superintendent
- Denotes those attending the exit interview on January 15, 1982.
+ Denotes those attending the' exit interview on February 5, 1982.
2.
Licensee's Actions on Previous NRC Identified F)llowup and Unresolved Item a.
(Closed) Followup Item (50-397/81 _17/04) Instrument Sensing Lines Installed by Johnson Controls with Reversed Slope Previous examination of primary containment sample handling lines No. 25.0-X85D and 25.0-X85E.*evealed that slopes were not in
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accordance with specification requirements. Moreover, quality control inspectors were not issued tools for checking line slopes.
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During this inspection, work initiated to correct the improper slopes of the two sample lines was reviewed, with no deviations from specification noted. The contractor, Johnson Controls, stated that quality control inspectors had been provided with
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specific tools to examine tube slopes and line walkdowns for scope measurement had been initiated.
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This item is closed.
b.
(Closed) Followuo Item (50-397/81-18/06) Radiographs of Pipe Welds j
Show Ouestior,able RT Ouality/ Integrity This item was reported by the licensee on November 19, 1981 as a 10 CFR 50.55(e) deficiency (Licensee Event Report). As the licensee's actions in this area will be reviewed as a 10 CFR 50.55(e)
item, this followup item is closed.
3.
Review of Licensee Actions with Respect to Reported 10 CFR 50.55(e)
Deficiencies (Licensee Event Reports)
The following 10 CFR 50.55(e) deficiency reports submitted by the
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licensee were reviewed to determine the extent and adequacy of the l
licensee's corrective actions:
a.
Emergency D.C. Motor Control Center Room Cooler Powered from a
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Non-Vital Source
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On June 21, 1979, the NRC Region V office was notified of a deficiency in the power source for emergency D.C. Motor Control Center (MCC) Room Cooler, RRA-FC-12 as it was supplied from a non-class IEsource(MC-7C-A). This cooler removes heat from the safety related D.C. MCC Room on elevation 471' of the Reactor Building.
This room contains MCC-S2-1A, a Division 1, 250 volt D.C.
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switchgear which provides power vital loads. The licensee.
evaluatedthelossofRRA-FC-12aspossiglycausingatemperature
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rise ig the MCC Room of approximately 15 F per minute to in excess
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of 250 F.
These temperatures-could.cause a failure of MCC-S2-1 A.
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The licensee changed the power supoly for RRA-FC-12 to MC-7B, a Class IE motor control center. The licensee determined that no other switchgear was effected by similar problems.
The change of the power supply to the fan was inspected.
No discrepancies were noted. This item is closed.
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Potential Missile Hazard flear Reactor Protective System Motor Generator Sets
On October 3,1978 th2 licensee noted that a possibility existed for missile damage to Class IE safety-related switchgear from non-qualified reactor protective system (RPS) motor generator (MG)
sets located in common switchgear rooms of the radwaste and control
building. The licensee notified the flRC Region V Office on February 18, 1980, upon completion of the evaluation that this was a reportable condition under the requirements 'of 10 CFR 50.55(e):
RPS MG Sets A&B (non-Class IE equipment) are located in the i
rooms for Class IE, safety-related switchgear MC-7A and MC-8A.
In the event of flywheel failure, the lack of separation
or positive protective barriers.could result in the loss of or damage to the safety related components, located nearby.
The licensee performed an investigation t'o see if a dynamic analysis of-the MG set would verify that the flywheel would not
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fail. This investigation determined that.the material properties
(to the degree',that they were known) and th'e design of the flywheel were such that'it was unlikely that a dynamic analysis would show conformance with the reauirements for avoiding brittle failure or plastic failure. The material of the flywheel and hub was six inch thick carbon steel p bte stock with a yield strength of 27,000 psi.
The material was not of specially controlled manufacture or
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inspection for flaws. The flywheel hub was shrunk fit (0.003 inch interference) on the shaft, which, combined with the stress concentration at the square keyway in the hub, raised hub stress to a high valve. The licensee determined from this investigation i
that protective barriers would be required to prevent f1G flywheel failure from damaging adjacent Class IE equipment.
i The protective barriers, as designed, consisted of steel shrouds enclosing the flywheel area of each MG set.
The shrouds were
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bolted to the adjacent wall for one MG set and to a free standing steel structure for the other MG set. The impact withstanding media was composed of steel "U-bars" supporting a Hexcel (crushable
corrogated aluminum honeycomb) material which was to be the energy absorbing media.
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A telephone interview with Burns and Roe design engineers at Woodbury, New Jersey established that the design of the barrier assumed that it would have to withstand the impact of flywheel segments equal to 1/2 and 1/4 of the flywheel mass with velocities determined by the energy imparted by the assumed 1,800 RPM speed of the driving motor.
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Burns and Roe personnel stated that normal design review procedures had been followed including independent checks of calculations and
assumptions.
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At. the time of this inspection, work had just started on PED-215-CS-5275, I
the barrier installation directive.
Bechtel. Power Corporation was attempting to locate holes for concrete anchor bolts in floor
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areas free of reinforcement steel. The barriers had not been fabricated.
i This item will remain outstanding pending a review of the installation
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of the Hexcel/ Steel barriers.
c.
Penetration X69D, Sample Line, will exceed Thermal Cycle Life Limits On January 25, 1979 the licensee notified NRC Region V of a
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design error in the penetration for the RPV sample line (X69D).
The design of 31 primary containment instrument penetration
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assemblies consisted of stainless steel (SA 312 TP 304) instrument piping (210 one-inch pipes in clusters of 6 or 12) passing through one inch thick carbon steel (SA 516 GR 70) penetration end caps. The joint was completed by fillet welds between the pipe and the end caps on both sides of the end caps.
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j Stainless steel has a coefficient of thermal expansion which is approximately 1.4 times the thermal expansion coefficient for carbon steel. This difference in expansion coefficients, and the restraint imposed by the welds cause the assembly to have a limited fatigue life.
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Burns and Roe Engineers (Woodbury) stated that calculations performed in accordance with ASME III, NB 3222.4 1974, winter 1975 addenda, show the following allowable thermal cycles:
A T = 200 F 80 cycles allowed
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A T = 300 F 37 cycles allowed o T - 320 F 31 cycles allowed A T - 400 F 21 cycles allowed-
Burns and Roe Engineers stated that of the 210 penetration lines to primary containment of this design, 209 would not be expected to see thermal transients during " normal" operation as they are dead ended instrument sensing lines.
However, transients could be
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expected on some or all of these penetrations, if improper value lineups cause a " blow-down" of the fluid filled line'. This was to
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s be precluded by operator instructions and administrative procedures.
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-5-The remaining penetration X69D was to be used as a coolant sample line and subjected to several thermal cycles per day. This penetration will be changed.to a new penetration (X77A) to be designed for an infinite number of allowable thermal cycles. This design and analysis is to be performed by Pittsburg Des Moines Steel (PDM).
This item will remain open pending a review of-the final penetration design and installation for the sample line penetration and a review of administrative procedures to preclude inadvertant operator action on the other 209 penetrations of this type.
d.
Flooding of Emergency Core Cooling System Pumos from Soent Fuel Pool Boil Off.
On August 18, 1977 NRC Region V was informed by the licensee of a reportable deficiency in the standby gas treatment system (SGTS)
caused by the inability to assume that the fuel pool cooling system would function in accident conditions, the5eby all wing the spent fuel pool temperatures to exceed 212 F.
The SGTS, an engineered safety feature, was required during postulated accidents to maintain a negative 0.25 inches of water (gage)
pressure in the reactor building to prevent a direct outleakage of radioactive fission products to the environment. The " boil off" of water vapor from the spent fuel pool would overload the SGTS's ability to remove water vapor. and maintain the required negative reactor building pressure.
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On October 18, 1979 additional problems caused by the spent fuel pool cooling system's lack of seismic Category I, electric class.IE design standards, were reported to NRC Region V.
The spent fuel pool boil-off caused by the assumed faiure of the non-safety class fuel pool cooling system would cause vapor condensate on the walls and floors of. the reactor building of up'to 18 gallons per minute. This condensed ' water would proceed via the reactor building floor drain system to the emergency core cooling system (ECCS) pump rooms located below grade level at the 422' elevation of the reactor building. Because the reactor building sump pump system was not seismic Cagetory I nor Class IE it could not be counted on to remove water from the ECCS pump rooms. This report stated that "even though Class IE instrumentation (was) available
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within each ECCS pump room to tell the operator that the room (was)
flooding, there (was) no way to stop the source of flooding...,
to isolate the ECCS pump roons..., (or) to pump out the ECCS pump' rooms post LOCA or Post-SSE.
Eventually, the ECCS pump rooms (would) flood, shorting out the ECCS. pumps."
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-6-The licensee reported on November 30, 1979 that "the fuel pool cooling system (would) be upgraded through reanaylsis and redesign to improve reliability. The system (would) be analyzed to seismic Category I criteria and modifications (would) include g.
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the addition of a safety grade heat sink and emergency power
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A further letter from the licensee to NRC Region V on December 3, 1980 states that since measures would be taken to preclude fuel pool boiling the basis for the reported deficiencies in the SGTS were eliminated.
l On December 6, 1979, upon receipt of the licensee's November 30, 1979 report, NRC Region V requested that a FSAR amendment be made to detail the changes to be made in the fuel pool cooling system.
During this inspection, the inspector attempted to compare completed and ongoing work on the fuel pool cooling system to FSAR amendments refecting the licensee's statements in the November 30, 1979 report. The inspector was given a copy of the draft for FSAR amendment 24 which was transnitted to NRC, Division of Licensing on January 27, 1982 and which detailed the specific changes to be made to this system to assure the reliability of the cooling portion. This amendment stated that the cooling portion of the fuel pool cooling and cleanup system will be designed to seismic Category I, quality Class I (10 CFR 50 Appendix B) requirements and will have provisions for isolation from the cleanup portion of the system by automatic seismic Category I, quality Class I' isolation valves which activate on low fuel pool water'1evel. Normally, the reactor building closed cooling system (RCC) would furnish non-safety grade cooling water to the FPC system, but, if required, would be available via the standby service water system. The redundant active components are to be powered from Class IE division 1 and 2 power sources.
Further-more, the amendment states that there is sufficient redundancy in the fuel pool coolfng systems to insure that a maximum fuel pool temperature of.175 F can be maintained on only one pump and heat exchanger.
The inspector observed that the FSAR amendment did not reflect the common supply and discharge headers for this system, although as previously stated, system redundancy was emphasized.
The licensee agreed that the FSAR amendment should be more I
specific as to what parts of the system are redundant and what parts are not. The inspector stated that NRC Region V would contact NRC Division of licensing to insure that this feature i
of the design was noted.
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-7-The inspector further noted that work on this system to upgrade i
various values was being performed to seismic Category I, quality Class II (not incorporating all the provisions of 10 CFR 50
Appendix B). The work seemed to reflect the statements of an earlier FSAR amendment (No.16 of June,1981) which in Table 3.2-1
lists the fuel pool cooling system as " Safety Class G, Seismic Catagory II, quality Class II,"
The licensee stated that the January,1982 FSAR amendment had not been reflected in field work packages. As the modifications to the FPC system had not progressed to the implementation stage, this system will be reviewed on a
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future inspection to assure that FSAR commitments are implemented.
This item remains open.
4.
Review of Instrumentation and Insturment Cable Installation, Work and Quality Records The installed instrumentation, instrument sensing lines, electrical cables, and quality control records associated with standby service water pump autostart (instrument racks IR21 and IR22) and the main steam isolation valve leakage control system (instrument racks IR73 and IR74) were inspected. The installation and maintenance of these equipment items appeared to be in accordance with applicable specifications and good engineering practice.
The inspector noted that the contract 220 contractor (Johnson Controls) was installing instrument sensing lines and tubing with an 18 inch minimum separation spacing requirement between redundant divisions in accordance with appli. cable specifications.
The inspector was unable to identify the source of this require-ment in the FSAR.
Burns and Roe Engineering stated that the basis for this standard would be provided in the missile analysis for WNP-2 still in preparation.
The inspector additionally noted that the contractor was installing colored meter identification tags on safety related instrumentation tubing to identify safety division classification.
This practice appeared acceptable as a means of tubing identification with regard to safety division, but the inspector was unable to identify a. description of this scheme in the FSAR including a recent draft of FSAR Chapter 8 which was provided by the licensee. The licensee stated that a description of the instrument line identification system would be provided in later revisions of FSAR Chapter 7 or 8.
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Finally the -inspector noted that safety related equinment mounted on instrument racks, and the instrument racks themselves, were not i
' identified as to safety division in accordance with the identification schemes described in the FSAR.
Instrument racks outside the control
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-8-room (PGCC) provided by General Electric, had safety-related equipment identified by yellow labels regardless of safety division classification.
The FSAR states that, outside of the PGCC, yellow denotes safety division 1.
Equipment of safety divisions 2 and 3 were to be identified by orange and red backgrounds respectively. On instrument racks outside the PGCC, provided by General Electric, black labels denoted i
non-safety related divisions.
Again, the FSAR states in Chapter 8 that non-safety divisions A and B equipment will have gold or silver background labels.
Moreover, other instrument racks, installed outside the PGCC in the
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reactor building, provided under contract 58, were identified with black labels regardless of the safety or non-safety division with which the rack or instrument belonged.
The licensee stated that a review of the instrumentation labeling system would be performed to insure that FSAR commitments were followed. The t
items of safety division separation and labeling have been previously identified as NRC followup items (50-397/81-17/05).
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Management Interview The inspector met with the licensee's management representatives as noted in paragraph.1, at the conclusion of the inspections on January 15, and February 5,1982.
The scope of the inspection and the observation and findings of the inspector were discussed.
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