IR 05000254/1987014
| ML20235M385 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 07/08/1987 |
| From: | Danielson D, Ward K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20235M386 | List: |
| References | |
| 50-254-87-14, 50-265-87-14, GL-84-11, IEB-82-03, IEB-82-3, IEB-83-02, IEB-83-2, IEIN-86-106, NUDOCS 8707170178 | |
| Download: ML20235M385 (12) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Reports No. 50-254/87014(DRS);50-265/87014(DRS).
Docket Nos. 50-254; 50-265 Licenses No. DPR-29; DPR-30 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Quad Cities Station, Units 1 and 2 Inspection At: Quad Cities Site, Cordova, Illinois
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Inspection Conducted: June 9-11, 16-18, and 24-25, 1987 7 8 #7
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Inspector:hKavinD. Ward
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g Date Approved By:
D. H. Danielson, Chief 7/f[#7 Materials and Processes Section Date Inspectien Summary i
Ins ection on June 9-11, 16-18, and 24-25, 1987 (Reports No. 50-254/87014(DRS);
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0~-265/87014(DRS)
I Areas Inspected: Routine, unannounced inspection of IE Information Notice No.86-106 (92717); and of Temporary Instruction 2515/89 (Generic Letter No. 84-11), and Temporary Instruction 2515/85 (Mark I Containment) (25589, 25585).
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Results: No violations or deviations were identified.
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DETAILS 1.
Persons Contacted Commonwealth Edison Company (Ceco)
- D. Gibson, Quality Assurance Superintendent
- C. Smith, Quality Control (QC) Supervisor
- J. Kopacz, Technical Staff Supervisor
- H. Do, Technical Staff Engineer M. Kooi, Regulatory Assurance Supervisor J. Ford, QC Inspector United States Nuclear Regulatory Commission (NRC)
- R. Higgins, Senior Resident Inspector A. Morrongiello, Resident Inspector The inspector also contacted and interviewed other licensee and contractor employees.
- Denotes those present at the final exit interview June 25, 1987.
2.
IE Information Notice No. 86-106:
Feedwater Line Break This Information Notice was to alert licensees of a potentially generic problem with feedwater pipe thinning and associated problems.
On December 9, 1986, both units at the Surry Power Station were operating at full power when the 18" suction line to the main feedwater pump "A" for Unit 2 failed catastrophically. The event was initicted by the main steam isolation valve on steam generator "C" failing closed. Due to the increased pressure in steam generator "C" which collapsed the voids in the water, the reactor tripped on low-low level in that steam generator.
A 2'-by-4' section of the wall of the suction line to the "A" main feed water pump was blown out and came to rest in an overhead cable tray.
The break was located in an elbow in the 18" line about l' from the 24" header. The lateral reactive force generated by escaping feedwater completely severed the suction line.
Quad Cities Station performs thickness inspections for wall degradation due to erosion / corrosion (E/C) in the following categories:
a.
Balance of plant steam piping, especially turbine / generator piping has been inspected over the years by the turbine manufacturer's representative (GE) during scheduled refueling outages.
Additionally, the insurance company authorized inspector has been included in the inspection activities. Any area that has shown the development of internal E/C has been investigated using ultrasonics (UT) to determine the amount of degradation and repaired as necessary.
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l b.
Balance of plant feedwater, condensate and connected piping have not routinely been inspected in past years.
In response to this Information Notice, a program was conducted by CECO with technical assistance from Netech Engineers to obtain pipe wall thickness data j
en feedwater supply systems at Quad Cities Unit 2.
The purpose of the program was to identify a sample of locations where E/C might occur and to obtain thickness data through field measurements of the pipe wall using ultrasonic measurement techniques. Locations for inspection were selected based on piping geometry, fluid flow and.
system usage considerations that contribute to E/C. Evaluations and field measurements were made of the large bore piping in the i
condensate, condensate booster, and feedwater systems.
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Two 45 elbow fittings of approximately 45 piping components
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inspected exhibited wall thickness below minimum wall. They were subsequently evaluated in accordance with ANSI B31.1-80 and found to be acceptable. One component is-located on line 2-3401D-16"-D at the top of the riser on the discharge side of condensate booster pump 20-3401. The other is located on line 2-3405A-20"-D in the horizontal run between the 30" supply header and feedwater
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pump 2A-3201.
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A comprehensive program is being developed by CECO Station Nuclear Engineering for future inspection activities.
No violations or deviations were identified.
3.
(Closed TI 2515/89) Inspection of BWR Stainless Steel Piping in Accordance with Ger.eric Letter 84-11 a.
General i
The purpose of this TI was to verify that the BWR licensees had performed inspections of stainless steel piping welds susceptible to intergranular stress corrosion cracking (IGSCC) in accordance with i
i Inspections conducted at several BWR facilities as a result of IE l
Bulletins No. 82-03, Revision 1, and No. 83-02 and the NRC August 26,
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1983, orders, revealed IGSCC in large diameter Recirculation and Residual Heat Removal piping systems. The results of these inspections led to an ongoing program for similar inspections of all operating BWRs. Repairs, analyses, and additional surveillance were required in most cases to ensure the integrity of the susceptible piping in thosa areas where IGSCC was discovered or indicated by nondestructive examinations.
The following paragraphs address the inspection requirements of NRC Temporary Instruction 2515/89, which was issued to assess the actions of the licensee based on the initial suggestions contained l
in Generic Letter 84-11 and related correspondence concerning specific licensee commitments.
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b.
Inspection Program The NRC inspector reviewed programs for Units 1 and 2 and determined that they required the following:
(1) The inspection of 20% of the welds not inspected previously for each pipe size.
(2) The inspection of 20% of the welds previously inspected and found not to contain cracks for each pipe size.
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(3) The inspection of all unrepaired welds previously found to l
contain cracks or indications of cracks.
(4) The inspection of all weld overlays on top of welds containing
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cracks or indications of cracks longer than 10% of the circumference.
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i (5) The inspection of all welds treated by the Induction Heating l
Stress Improvement (IHSI) technique and not previously examined
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after IHSI treatment.
I The TI required verification that the program also included a visual examination for leakage of the Reactor Coolant piping during each plant outage in which the containment is deinerted. This was not
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part of CECO's program; however, whenever the dry well is deinerted
it is a common practice that the reactor and its associated piping be examined for leaks at high pressure during startup. The NRC inspector reviewed many master outage checklists for Units 1 and 2, and verified that " Check drywell for leakage" was one of the items.
c.
Competence of Ultrasonic Examination (UT) Personnel Qualification by a formal performance capability demonstration test conducted at EPRI NDE center for Level II and Level III UT personnel was required by the program.
UT personnel who were performing as a SNT-TC-1A Level I were trained by EPRI qualified personnel cnsite. The NRC inspector observed part of the training.
The following are NRC Inspection Reports referencing the
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Levels I, II and III training: No. 50-254/84-06 and No. 50-265/84-05, dated July 12, 1964; No. 50-254/85-030, dated February 24, 1986; No. 50-254/85016 and No. 50-265/85008, dated June 18, 1985 and No. 50-254/86019 and No. 50-265/86014, dated January 30, 1987.
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d.
Leak Detection and Leakace Limits Quad Cities Technical Specifications state the following:
(1)
"Any time irradiated fuel is in the reactor and the reactor coolant temperature is above 212 F, reactor coolant leakage into the primary containment from unidentified sources shall not exceed 5 gpm.
In addition, the total reactor coolant system leakage into the primary containment shall not exceed 25 gpm."
(2)
"Both the sump and air sampling systems shall be operable during reactor power operation. From and after the date that one of these systems is made or found to be. inoperable for any reasons, reactor power operation is permissible only during the succeeding seven days."
(3) "If the conditions of the above cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown cor.dition etl thin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />."
l In addition to the above technical specification requirement, CECO's
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reactor coolant leakage in the drywell Procedure No. Q0S 1600-7, Revision 7, dated May 1984, states the following in part:
Limitations and Actions
"In the event that any one or more of the following criteria are exceeded, commence an orderly shutdown and be at least in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if:
The volume pumped from the drywell floor drain sump in a four l
hour period is equal to or greater than:
(1) 960 gallons (4gpm)
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(2) 480 gallons greater than the previous 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> pump down
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(2 gpm)
The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> average GPM pumped from the drywell floor drain sump at the end of any 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period is greater than a 2 gpm increase within any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
The volume pumped from the drywell equipment drain sump in an eight hour period is equal to or greater than 9600 gallons (20 gpm).
At least one drywell floor drain sump collection and flow monitoring system shall be operable. With the drywell floor drain sump collection and flow monitoring system inoperable, restore the inoperable system to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
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or immediately initiate an orderly shutdown and be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold shutdown
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within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The floor drain sump collection I
and flow monitoring system operability is defined as the
ability to measure reactor coolant leakage."
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e.
Performance of Inspection The NRC inspector observed NDE and reviewed data on many pipe welds in accordance with the inspection program.
The NRC inspector observed many UT personnel demonstrating their competence prior to examining welds using the essential parameters of the qualified procedures.
The following NRC reports document the NRC inspectors observations and review of documentation: No. 50-254/84-06 and No. 50-265/84-05, dated July 12, 1984; No. 50-254/85016 and No. 50-265/85008, dated i
June 18, 1985; No. 50-254/85030, dated February 24, 1986; No. 50-265/86019 and No. 50-265/86014, dated January 30, 1987.
f.
Subsequent Activity The program provided for scope expansion and additional inspection when new cracks were found or existing cracks grow to an unacceptable size. The UT was expanded in accordance with the ASME Section XI code for that system and size.
(Reference NRC Inspection Reports No. 50-254/86014 and No. 50-265/85008, dated June 18, 1985; No. 50-254/86019 and No. 50-265/86014, dated January 30, 1987.
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Conclusions The NRC inspector determined that the licensee had performed
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I inspections of stainless steel piping welds susceptible to IGSCC in l
accordance with Generic Letter 84-11.
4.
1ClosedTI 2515/89) Mark I Modifications l
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General The purpose of this TI was to verify that the BWR licensees with Mark I containment designs have modified their plants with appropriate procedures and in accordance with their cotm11tments.
Unit 1 is complete except to verify that all hangers are in accordance with the drawings and that all documentation is complete.
Unit 2 is complete except for documentation associated with a final QA review.
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A technical issue was identified concerning General Electric (GE)
BWR's with Mark I containment designs.
The technical issue involved new suppression pool hydrodynamic loads on the containment that were not considered in the original Mark I containment design bases.
These loads affected the suppression pool (torus) and its support structure, structures inside the torus and the attached piping and equipment outside the torus.
The licensee was required to submit a
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Plant Unique Analysis Report (PUAR) to provide a basis for the plant
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specific modifications.
The NRC staff reviewed the licensee's PUARs against acceptance criteria contained in NUREG-0661 and found them acceptable.
Subsequent'to the issuance of the PUAR, the NRC inspector performed inspections of NDE and welding procedures, material, NDE and welding personnel certifications, surveillance and data reports, and observations of installation activities.
These activities are documented in the following NRC Inspection Reports:
No. 50-254/79-06 and No. 50-265/79-06, dated March 1979; No. 50-254/79-19 and No. 50-265/79-16, dated September 4, 1979; No. 50-254/79-29 and No. 50-265/79-26, dated January 18, 1980; No. 50-254/80-15 and No. 50-265/80-18, dated July 8, 1980; No. 50-254/80-23 and No. 50-265/80-25, dated November, 1980; No. 50-254/82-19 and No. 50-265/82-22,' dated December-28, 1982; No. 50-254/83-24 and No. 50-265/83-23, dated March 1, 1984 and No. 50-254/84-06 and No. 50-265/84-05, dated July 12, 1984.
To assure the licensee has completed the required modifications as described in their PUAR, the NRC inspector reviewed the licensee's PUAR for commitments for structured modifications for the following items:
(1) Downcomer bracing (2) Vent header deflector (3) Safety relief valve SRV quenchers
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(4) Stiffening of torus attached piping l
l The following paragraphs address the inspection requirements of l
NRC temporary Instruction 2515/85, which was issued to verify satisfactory completion of licensee actions concerning the Mark I program.
b.
Programmatic Review The NRC inspector reviewed the previous NRC inspection repo*ts covering programmatic inspections performed during the modification.
The reports were reviewed to verify the inspection coverage of the following areas:
(1) Torus Support Reinforcement (2) Torus Interal Modification
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(3) Torus Attached Piping l
The results of the review are outlined in the attached matrix:
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TORUS SUPPORT REINFORCEMENT I
I I
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I I
I I
I I
i 181*lRl8l*10l.I REPORT NO. 254/
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INSTALLATION SPECIFICATIONS i
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I iy I UESIGN DOCUMENTS l
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lv l CONTRACT & PROCUREMENT DOCUMENTS l
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lv l DRAWINGS l
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Iv l PROCEDURES Iv iy iv Iy ly I I
I PERSONNEL CERTIFICATIONS & QUALIFICATIONS I
ly l Iv lv iv i l
MATERIAL CERTIFICATIONS I
Iy l l
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OBSERVATIONS FABRICATION / INSTALLATION ACTIVITIES I
Iy I iv iy Iy I l
INSTALLATION, INSPECTIONS & OTHER QUALITY RELATED DOCUMENTSI lv i lv iy 1y i I
- NOT PREVIOUSLY INSPECTED TORUS INTERNALS MODIFICATIONS I
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REPORT NO. 254/
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INSTALLATION SPECIFICATIONS l
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Iv l DESIGN DOCUMENTS I
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l lX l CONTRACT & PROCUREMENT DOCUMENTS I
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PERSONNEL CERTIFICATIONS & QUALIFICATIONS lX l lK IX lX (X l l
MATERIAL CERTIFICATIONS l
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OBSERVATIONS FABRICATION / INSTALLATION ACTIVITIES lX l IX lX lX l I
l INSTALLATION, INSPECTIONS & OTHER QUALITY RELATED DOCUMENTSlX IX IX IX IX IX l l
- NOT PREVIOUSLY INSPECTED TORUS ATTACHED PIPING I
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REPORT NO. 254/
A INSTALLATION SPECIFICATIONS X l.
I DESIGN DOCUMENTS X
X X
l CONTRACT & PROCUREMENT DOCUMENTS Xl DRAWINGS I
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PROCEDURES lXlXlX l
PERSONNEL CERTIFICATIONS & QUALIFICATIONS IX lX l-MATERIAL CERTIFICATIONS lX l
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- NOT PREVIOUSLY INSPECTED
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TORUS SUPPORT REINFORCEMENT l
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REPORT NO. 265/
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INSTALLATION SPECIFICATIONS I
DESIGN DOCUMENTS I
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CONTRACT & PROCUREMENT DOCUMENTS I
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l PERSONNEL CERTIFICATIONS & QUALIFICATIONS lxl y
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MATERIAL CERTIFICATIONS lx y
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OBSERVATIONS FABRICATION / INSTALLATION ACTIVITIES Ix yly y
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INSTALLATION, INSPECTIONS & OTHER QUALITY RELATED DOCUMENTSI ly yly y
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Iy l DESIGN DOCUMENTS l
lX l CONTRACT & PROCUREMENT DOCUMENTS Ixl
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DRAWINGS I
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PROCEDURES Xl XlX X' X l
PERSONNEL CERTIFICATIONS & QUALIFICATIONS Xl
'XlX Xl X l
MATERIAL CERTIFICATIONS i
l iX lX lX l l
l OBSERVATIONS FABRICATION / INSTALLATION ACTIVITIES IX l IX lX Xl l
I INSTALLATION, INSPECTIONS & OTHER QUALITY RELATED DOCUMENTSIX lX lX lX 'XlXl l
- NOT PREVIOUSLY INSPECTED
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TORUS ATTACHED PIPING l
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l fi {l flyigl{l INSTALLATION SPECIFICATIONS X
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CONTRACT & PROCUREMENT DOCUMENTS xl DRAWINGS l
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PROCEDURES lx X: X l
PERSONNEL CERTIFICATIONS & QUALIFICATIONS IX l X
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- N0f PREVIOUSLY INSPECTED i
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c.
Technical Specification Review (1) Suppression Pool Temperature Monitoring System The suppression pool temperature monitoring system was modified to provide an average and local pool temperature indication and alarms in the control room.
Sixteen thermocouple are installed around the torus, one in each torus bay, with a set of eight thermocouple on the inner circumference, and another set of eight thermocouple on the outer circumference forming two independent channels. The PUAR identified eight thermocouple to be placed around the torus, however CECO has installed sixteen thermocouple. The NRC inspector visually verified the final insta?lation of the thermocouple. The installation was consistent with the location and placement
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given in the PUAR and drawings.
The NRC inspector toured the control room and observed that tho thermocouple are continuously recorded in the control room, for each channel, by means of Tracor Westronics strip charts, and annunciators provide indication of abnormal conditions.
The strip charts are reviewed 1-2 times per shift by the operators. The temperature limits given in the technical specifications are consistent with those in the PUAR, and the instrumentation alarm setpoints are consistent with the temperature limits given in the technical specification.
(2) Drywell to Wetwell Differential Pressure (DP) Control At Quad Cities the drywell/ torus differential pressure control systen is also known as the drywell and suppression chamber (DP) control system. The purpose of the system is to maintain a drywell to torus differential pressure of equal to, or greater than 1 psid, in order to decrease or eliminate any damage to the torus structural internals in the event of a l
blowdown. The differential pressure minimizes the water level l
in the downcomer piping and thus reduces the water slug upon pressurization. The NRC inspector reviewed the drywell to suppression chamber differential pressure control system in the control room.
Control room instrumentation consists of a Research Incorporated drywell/ torus DP controller with instrument l
warning lights and dial for indication.
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The DP between the suppression chamber and the drywell is controlled as required in the Technical Specifications, Limiting Condition of Operation (L.C.0.) 3.7.A.6.
The recording of this DP is required once a shift and is entered by the nuclear station operator (NS0) in the control room.
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d.
Design Modification Review (
(1) Downcomer Bracing The modification consisted of replacing the previous downcomer bracing with new downcomer bracing which had a greater structural capacity.
The NRC inspector reviewed fabrication drawing QC-4114,
Revision B, and other drawings to verify that the modifications met the PUAR commitments.
The NRC inspector also reviewed
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travelers, weld data reports, NDE reports and other related QA records to assure.that the appropriate procedures were used during the installation.
(2) Vent Header Deflector l
The modification consisted of the installation of a vent header j
deflector beneath the existing vent header.
The deflector is
supported by connection plates which are welded to the existing i
vent header collar plates.
The NRC inspector reviewed fabrication drawing QC-1205 and-other drawings to verify that the modifications met the PUAR commitments.
The NRC inspector also reviewed travelers, weld data reports, NDE reports and other related QA records to
assure that the appropriate procedures were used during the l
installation.
(3) SRV Quenchers
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l The modification consisted of the installation of two l
12" diameter perforated end capped stainless steel pipes,
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oriented parallel to the torus bay longitudinal axis, at the ends of the ramshead, and the associ.ited supports.
The NRC inspector reviewed fabrication drawing B-1472 and other drawings to verify that the modifications met the PUAR commitments.
The NRC inspector also reviewed material, NDE and welding certifications, weld data reports and other related QA records to assure that the appropriate procedures were used during the installation.
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(4) Torus Attached Piping Penetration Stiffeners The modification consisted of pipe sections installed as sleeves to reinforce the penetration nozzles, with support arms extended radially from the pipe sleeves to pad plates attached to the suppression chamber shell.
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The NRC inspector reviewed fabrication drawings, M-1620-02, 03, 04 and other drawings to verify that the modifications met the PUAR commitments. The NRC inspector also reviewed a traveler.
l weld rod requests, weld inspection reports and other related QA
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records to assure that the appropriate procedures were used
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during the installation.
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l (5) Visual Inspection i
The torus interior was inaccessible at the time of the inspection because Unit 1 and 2 were in operation. The NRC
inspector toured outside the. torus, visually inspecting welds and inspecting the final installation of the torus attached piping penetration stiffeners, support reinforcements (saddle supports), thennowells for the temperature elements, etc. to the requirements of the design drawings.
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(6) Conclusions
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The NRC inspec. tor determined that the licensee had modified his Mark I containment using appropriate procedures and in accordance with the PUAR commitments.
5.
Exit Interview
The inspector met with site representatives (denoted in Persons Contacted paragraph) at the conclusion of the inspection. The inspector summarized the scope and findings of the inspection noted in this report. The I
inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection. The licer.see did not identify any such documents / processes as proprietary.
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