IR 05000324/1987014
| ML20215K214 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 06/09/1987 |
| From: | Blake J, Coley J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20215K176 | List: |
| References | |
| 50-324-87-14, 50-325-87-14, NUDOCS 8706250261 | |
| Download: ML20215K214 (12) | |
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-[48800q"E, UNITED STATES
NUCLEAR REGULATORY COMMISSION
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' 101 MARIETTA STREET, N.W. -
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' Report.Nos.: 501325/87-14 and-50-324/87-14'
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. Licensee: Carolina Power and Light Company
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P. OR Box 1551 Raleigh, NC 27602
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Docket Nos.:.50-325 and 50-324 License.Nos.:
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Facility-Name:
Brunswick 1 and 2 LInspection'
n c ed* May 11-15, 1987 Inspecto :
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. Tlake, Section Chief Date Signed
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n ineering Branch D vision of Reactor Safety
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SUMMARY
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Scope:
This routine, announced inspection was conducted in the areas of-I
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- Temporary. Instruction-2500/12 - Inspection of-the ' Actions Taken By Licensees j
and Applicants of BWR. Facilities. With Mark 1 and: Mark II Containments.
In-Response to GE Sil.No. 402 and previous enforcement matters and open items.
-o Results: No violations or deviations were identified, i
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8706250261 870610 l
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PDR ADOCK 35000324'
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REPORT DETAILS 1.
Persons Contacted
Licensee Employees i
- E. R. Eckstein, Manager, Technical Support
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- J. A. McKee, Supervisor, Quality Control (QC)
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- L. Wheatley, Project Engineer, Inservice Inspection (ISI)
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- M. J. Pastra, Technician, Regulatory Compliance j
h Other licensee employees contacted included construction craftsmen, j
engineers, technicians, operators, mechanics, security force members, and office personnel.
l NRC Resident Inspectors i
- W. Ruland, Senior Resident Inspector
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- L. Garner, Resident Inspector
- Attended exit interview
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Exit Interview l
The inspection scope and findings were summarized on May 15, 1987, with those persons indicated in paragraph 1 above.
The inspector described the
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areas inspected and discussed in detail the inspection findings.
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dissenting comments were received from the licensee.
j The licensee did not identify as proprietary any of the materials provided
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to.or reviewed oy the inspector during this inspection.
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Licensee Action on Previous Enforcement Matters i
(Closed) Violation 324/86-10-01 Failure to Follow ASME-Section XI Period Requirements for the Examination of Components CP&L letter of response dated May 27, 1986 has been reviewed and i
determined to be acceptable by Region II.
The inspector held discussions with ISI project engineer and examined the corrective actions as stated in j
the letter of response.
The inspector concluded that CP&L' had determined
the full extent of the subject noncompliance, performed the necessary
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survey and follow-up actions to correct the present conditions and
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developed the necessary corrective actions to preclude recurrence of similar circumstances.
The corrective action identified in the letter of response have been implemented.
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(Closed). Inf raction 324/80-20-03, 325/80-23-03 Section XI Surveillance Procedures not Established.
CP&L letter of response dated July 11, 1980
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and letter of supplemental response dated August 11, 1980 have been reviewed and determined to be acceptable by Region II.
The inspector held discussions with ISI project engineer and examined the corrective actions as stated in the' letters of response.
The inspector concluded that CP&L had determined the full extent of the subject noncompliance, performed the-necessary survey and' follow-up actions to correct the present conditions and developed the necessary corrective actions to preclude recurrence of similar circumstances.
The corrective action identified in the letter of i
response have been implemented.
(Closed) Violation 325/87-04-01 Failure to Follow Welding Procedure
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Specification Parameters.
CP&L letter of response dated April 17, 1987 I
has been reviewed and determined to be acceptable by Region II.
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inspector held discussions with ISI project engineer and examined the
corrective actions as stated in the letter of response.
The inspector
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concluded that. CP&L had determined the full extent of the sub ect J
noncompliance, performed the necessary survey and follow up actions to i
correct the present conditions and developed the necessary corrective l
actions to preclude recurrence of similar circumstances.
The corrective action identified in the letter of response have been implemented.
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4.
Unresolved' Items
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Unresolved items were not identified during this. inspection.
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5.
Temporary Instruction 2500/12, Inspection of Actions Taken By the Licensees and Ap)licants of BWR Facilities With Mark 1'and Mark II Containments in Response to General Electric Surveillance Instruction Letter (SIL) No. 402 "Wetwell/Drywell Inerting".
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On February 3,1984, Georgia Power Company reported a throagh-wall crack almost completely around the vent header within the containment torus of Hatch Unit 2.
The crack was discovered during a routine visual inspection of the vent system prior to a 31 ant startup.
It represented a failure of
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the vapor suppression system tlat could cause containment overpressuriza-
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tion in the event of a loss-of-coolant accident (LOCA).
Because of this safety concern, IE Bulletin 84-01, " Cracks in Boiling Water Reactor Mark I Containment Vent Headers," was issued on an emergency basis.
At the time the bulletin was issued, the cause of the crack was not known and was under investigation.
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The bulletin recuired the licensees of BWR facilities that were in cold shutdown and hac Mark I containments, to inspect for cracks in the vent header and in the main vents in the region near the intersection with the vent header.
The bulletin also suggested that operating BWR facilities with Mark I containments should review their data on differential
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i pressures between the wetwell and drywell for anomalies that could be indicative of cracks.
All affected licensees responded as required.
Regional inspection of licensee actions have been completed and the bulletin is closed out.
Following the notification of the vent header crack, meetings were held between the Regul; tory Response Group (RRG) of the BWR Owners' Group (BWROG) and the NR staff on February 6 and 23, 1984, to discuss the cause of the crack and e industry's actions planned to prevent such cracks.
The crack was conf med to be the result of brittle fracture caused by the injection of cold itrogen into the torus during inerting.
It was agreed at the meetings th<gg. the industry would voluntarily perform the actions necessary to satisiy the safety concerns raised by the Hatch 2 crack event.
The written responses received from utilities however, varied significantly in content and level of detail with respect tn the actions taken.
In many cases it is difficult to ascertain the extent to which the SIL recommenda-tions have been implemented.
This necessitated an inspection of the actions taken by utilities.
On May 11, 1987, the NRC regional inspector arrived at the Brunswick facility to evaluate the licensee's responses.
The inspector discovered that although Carolina Power and Light (CP&L)
Company s response was correct when they responded to the NRC on October 1, 1984, the licensee has had problems with the design described in their response and intends to modify the design in the future (perhaps in 1989).
The system as described in 1984 was the large capacity (21,000 gallons of stored liquid nitrogen) Containment Atmospheric Control System (CAC).
The Brunswick Technical Specifications also requires that a 30 day supply of nitrogen be available to support an unusual event.
The smaller capacity (5,000 gallons of stored licuid nitrogen) Containment Atmospheric Dilution (CAD) Sy' stem will also be ciscussed in the inspector's comments to the licensee s response for recommendations four and five of GE SIL 402.
For reporting clarity, the GE SIL recommendation will be outlined first, CP&L's response next, and the inspectors findings last.
Evaluate Inerting System Design Evaluate the design of the nitrogen inerting system.
Investigate the potential for introducing cold (less than 40 F) nitrogen and the orientation of the nitrogen port relative to the vent header, downcomers, or other equipment in the wetwell and drywell which may be in the path of the injected nitrogen.
Assure that the temperature monitoring devices, l
the low temperature shutdown valve, and overall system design are adequate l
to prevent the injection of cold nitrogen into the containment.
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CP&L Response:
The design of the Brunswick nitrogen inerting system will prevent the injection of cold nitrogen into the primary containment.
The inerting system contains a steam fired vaporizer and a low temperature shutoff
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valve whose function is to stop flow from the vaporizer when the nitrogen j
outlet temperature from the vaporizer falls below 50 F.
There is approxi-
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mately 300 feet of 8-inch diameter pipe which runs from the vaporizer to -
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After the line penetrates the reactor building, the l
line enlarges to a 20-inch diameter.
There is a low point in the nitrogen l
line before it reaches the reactor building.
This low point will tend
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i vaporizer. y liquid nitrogen in the unlikely event it should pass the i
to trap an The 20-inch nitrogen line for the torus is located at. azimuth l
135 degrees and elevation 1 foot 6 inches.
The' injection line penetrates horizontally and is approximately 11 feet from the vent header.
The 18-inch diameter drywell injection port is located at azimuth 175 degrees and elevation 23 feet 6 inches.
The structure closest to this penetration is the residual heat removal shutdown cooling line.
This line is approxi-i mately 3 feet horizontally from the injection port and is covered with 2 to 3 inches of mirror insulation.
A heating-ventilation-air conditioning (HVAC) return air duct runs along the grating and is approximately 5 feet
below the injection port.
Any cold (liquid or gaseous) nitrogen coming
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from either the drywell or torus injection port should not come into
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contact with any safety-related equipment.
The probability of any liquid nitrogen reaching either the drywell or torus is negligible for the reason i
stated later in this response.
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A plant modification (PM 78-003) is being implemented to install a control l
valve on the vaporizer discharge to control the nitrogen temperature
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between 90 and 120 F.
At 120 F, the valve will be full open (4000 scfm).
At 90 F, the valve will limit flow to 1000 scfm.
j The low temperature shut-off valve is presently inoperable, but is being
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evaluated as to return it to operability.
Due to operating procedures, however, manual valve HV-44 will be closed at 90 F to stop nitrogen flow
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to the vaporizer by the operator stationed at the vaporizer.
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Inspector's Findings The system design described above by the licensee remains in place today;
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however, the key to operation of the CAC system as described in the j
licensee's response is in the last sentence of the response.
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is presently operating as a manual system.
CP&L has on operator stationed at the vaporizer discharge to observe a temperature indicator and to
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observe the frost line on the vaporizer baffle.
If the vaporizer nitrogen outlet temperature cannot be maintained above 90 F, the operator is to close the CAC Inerting Tank LN2 Outlet Valve NP-HV44.
CP&L's plant modification (PM-003) did install, as stated, a flow control valve and a low temperature shut off valve and raised the CAC storage tank pressure from 40 psi to 45 psi.
These modifications however, have not l
functioned as designed because the license installed the wrong range temperature element for the low temperature shutoff valve and the flow control valve creates erratic system flow.
Discussions with CP&L's project engineer revealed that the low temperature shutoff valve is also j
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difficult to maintain because it is located between the CAC storage tank
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and the vaporizer and is subjected to extreme cold temperatures that make
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changing the valve's packing a difficult operation.
The engineer stated that CP&L is planning to modify the CAC system, possibility in 1989, since the modification had not been budgeted for 1988.
This modification will move the low temperature shutoff valve to the discharge side of the l
vaporizer, remove 'the flow control valve from the system and route the steam flow from the auxiliary boiler to the vaporizer in a direct manual operated path.
In summary, although the Brunswick design has temperature monitoring i
devises-its operation is basically manual and future design objectives are to make the CAC system more operator dependent via manual actuation.
Evaluate Inerting System Operation
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Review the operating experience of the inerting system to assure that the vaporizer, the low temperature shutoff valve, and the temperature j
indicators have functioned properly.
Evaluate the plant calibration, maintenance, and operating procedures for the inerting system. Assure that cold nitrogen injection would be detected and prevented.
CP&L Response
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In the past, the Brunswick Plant has had problems with liquid nitrogen passing the vaporizer.
This liquid nitrogen collected in a low point in
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the pipe outside the reactor building and caused failures of the pipe due to the combined thermal stresses and rapid expansion of the nitrogen upon vaporization.
These failures occurred over a hundred feet from primary
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containment.
In response to these events, which last occurred in 1982, the operating procedure for inerting and the setpoint for the low tempera-ture shut-off valve have been revised.
The low temperature shut-off valve is now set to close at 50 F vaporizer discharge temperature. The operating procedure for inerting now requires that steam be introduced to the vaporizer before nitrogen.
The procedure also requires that during irecting an operator must remain at the vaporizer and stop flow to the vaporizer if the discharge temperature of the nitrogen falls below 90 F.
There is local temperature indication at the vaporizer.
During inerting there is a frost line on the vaporizer which is indicative of discharge temperature.
As the frost line rises above the midpoint, liquid nitrogen is released to the discharge.
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Inspector's Findings CP&L's response is accurate today with the following exceptions:
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As stated in the inspector's findings for recommendation 1 the low temperature shut off valve does not function as designed because of the temperature element. The system is operator dependent.
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'The CAC system.has'two rises in the system prior. to entering the rad.
waste ' building;' all-of the ' failures have happen in the. second low'
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point:which has a 1 foot 6 inch rise prior to entering the pipe chase:
in the rad waste building'.
The first low point-in the system
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c-however, is. nearer the vaporizer and'this low point has 2-foot'2 inch rise; The liquid nitrogen has passed this low point in every failure
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'The response also states that, the last event occurred in.1982.
CP&L's-project engineer stated that the last event occurred!in 1984.
This was confirmed.alse - by the Regulatory Compliance Senior Specialist.
In summary,.the inspector concluded that, a_1though there is -very
- little chance for liquid. nitrogen to enter the wetwell-or the 'drywell i
' through.the. CAC system because of several rises in the system, there;
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is no reason to believe that the system would not experience an event.
Once:inside the pipe chase in the radwaste. building since there is only the 1 foot 6 inch rise in the system priori to ntering the building from the point where past events have be. experienced.
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Test for Drywell Bypass Leakage.
. Perform a bypass leakage test as soon as convenient to confirm the.
integrity. of the. vent system.. This test should be conducted'during-plant operation following -normal' plant procedures.
If no procedures exist, the following is a general guide for preparing your procedure:
pressurize the drywell.. to approximately 0.75 ps1 above the' wetwell pressure, maintain this 'drywell pressure and measure the pressure buildup' in 'the wetwell.
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Any bypass-leak area can then be' calculated (and'is limited by Technical
Specifications on many plants)' from the wetwell' pressure and the
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This will. provide an indication'that the vent system integrity is intact and that no' gross failure exists.
CP&L Response:
- Immediately following the discovery of the torus vent header crack in j
the Hatch Plant, an on-line drywell/ tours bypass leakage test on each Brunswick unit was conducted.
However, the test was not performed
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as described in GE SIL No. 402 because the Brunswick Plant has only i
wide-range torus pressure.
The test used consists of pressurizing the drywell-to approximately 1 ps,ig and observing the pressure decay over a
. one-hour period.
Both Brunswick units have been tested with very good
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results.
Brunswick-1 showed a pressure decay of 0.05 psig; Brunswick-2 i
showed a pressure decay of 0.06 psig.
A pressure decay of less than one-half the initial test pressure (1 psig) was judged to be acceptable.
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Inspector's Finding'
j IE Bulletin 84-01 dated February 3, 1984 instructed the licensee to review their plant data concerning differential pressure between the wetwell and i
drywell for anomalies that could be indicative of cracks.
j CP&L's review of surveillance test performed revealed no results I
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indicative of drywell to torus failure.
The Drywell to torus' Leak Rate
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Test.(PT-20.6) had been performed on each unit during the refueling outage previous to the issue of IE Bulletin 84-01.
The integrity of the system since -the refueling outages discussed above, was verified by the satisfactory performance of PT2.3.la, Drywell to Torus Vacuum Breaker Operability Check.
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Data taken during.the periodic tests are well within the limits required -
by the Technical Specifications.
InspectNitrogenInjectionLine
Conduct an ultrasonic test (UT) as soon as convenient of all accessible
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welds in the nitrogen injection line from the last isolation valve to the wetwell and drywell penetrations.
Also UT the containment penetrations and the containment shell within 6 inches of the penetration.
An ultrasonic test is recommended because cracks would be most likely to
initiate on the.inside of the pipe or on the side of'the metal in contact
with cold nitrogen.
CP&L Response:
It is believed that ultrasonic testing of the nitrogen injection lines is unwarranted for the Brunswick Plant.
This conclusion is based on the following reasons:
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In order for liquid nitrogen to reach either drywell, the liquid i
nitrogen would have to make a vertical climb of approximately 37 feet in 20-inch piping.
At a flow of 4000 scfm, this is not practical.
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There is 269 feet of horizontal 8-inch pipe prior to any tap-off to Brunswick-1.
This run includes 2-foot rise, an 8-foot drop, and a i
1.5-foot rise.
The pipe reaches a minimum of 4 feet below ground.
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At this depth, the ground maintains nearly a constant temperature year round.
This is the furthest point at which any damage has occurred.
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There is a section of pipe 101 feet long that is 1.5 feet lower than
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the rest of the piping.
This section tends to trap any liquid nitrogen that gets past the vaporizer.
This is where most damage has occurre L
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Since the Brunswick-1 tap-off is on the bottom of the 8-inch pipe, most of the liquid nitrogen that reaches this point will flow into the tap-off.
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. Any liquid nitrogen which may get past the Brunswick-1 tap must'then make a 4.5-foot vertical climb, followed by a 2-foot vertical climb.
' The section of piping with these two inclines is in the pipe tunnel and reactor building and is approximately 70-feet long.
The temperatures seen here would also help to vaporize any remaining liquid, j
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If any liquid were to get into the Brunswick-1 line, it would have to make a 1 foot 3 inch rise and then a 5 foot 6 inch rise.
The pipe with the 1 foot 3 inch rise is in the pipe tunnel.
The other rise
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is in the ' reactor building.
This section also includes a 150-foot l
section of horizontal pipe.
The runs of pipe in the pipe. tunnel and l
the reactor building would tend to vaporize the 1iquid if it were to l
make it that far, Also, the 5 foot 6 inch rise would tend to trap i
any remaining liquid that passed the 1 foot 3 inch rise.
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The piping discussed is outside the last isolation valve.
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With the attention given the vaporizer discharge temperature by the
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auxiliary. operator stationed at the vaporizer, it is believed that
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only a small amount of liquid, if any, would exit the vaporizer.
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Operating procedure require this temperature (90 F) to be maintained.
' Inspector's Findings j
The inspector agrees with the CP&L's assessment above for the CAC system.
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However,.the inspector's review of the CAD system piping revealed that thh is a downhill. system that has only one rise of 3 feet 3' inches just
prior to entering the 20 inch line in the reactor building adjacent to the torus.
The inspectors concern is that the CAD system is more likely to cause pipe-and. liner cracking even though the licensee offered the following reasons for not being concerned.
CP&L has no history of any problems with this system
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The vaporizer used for this system has electric heaters
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Flow for CAD 1 inch diameter piping is 1000 CFM verus 4000 CFM for
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the 8 inch CAC piping
The piping to the Drywell has a 38 foot rise in it
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However, since this system enters the 20 inch line just prior to entering i
I the torus there is very little chance that liquid nitrogen could get on the vent header from this system.
The CAD system could however, get liquid nitrogen on the torus steel liner, the torus penetration, and in the 20 inch pipe.
Of the three areas of concern to the inspector, the I
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licensee. nas performed a visual examination 'ofionly the' torus penetration.
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No cracks were observed in' the torus penetration. :In summary' the'
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inspector ' concluded that.if-cracking occurs in piping in the reactor i
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building it:will result from'a failure in thel CAD system and' not the CAC:
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The inspector feels that CP&L should have' implement recommendation-
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r DiscussionwithCP&L'sprojectengineerindicatedthatCP&L'doesnot' share this: concern and has not volumetrically examined any accessible weld.in the 20 < inch pipe.between the torus' and the Branch Connection forithe CAD
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system.
In addition, volumetric examination of the 1 inch. CAD s should.be performed from the last isolation valve to the wetwell. ystem-
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Inspect Containment-
~ During the next planned outage, perform 'a visual inspection of the vent
' header,. downcomers, and other equ,ipment. in the containment which might be expected to be affected-by the injection of cold nitrogen.
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header should be inspected ~~on the. outside and the inside.
Also inspect-the containment shell or steel liner for -at least 6 inches around the nitrogen penetration.
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CP&L Response:-
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' A L special procedure - (SP 84-0014) 'now exists for' the inspection of the j
- torus and drywell in areas adjacent to' the nitrogen injection ports.
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Ins)ections of both Brunswick-1 and Brunswick-2 have been performed and no pro)lems were observed in the vent header or in.the configuration of the nitrogen discharge into the torus.
Inspector's Findings CP&L's s)ecial procedure (SP-84-0014) did not fully implement the GE
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' Sil.No; LO2 recommendations for item 5.
The areas examined in accordance with the CP&L procedure was the following:
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All accessible surfaces of the torus nitrogen injection penetration.
All accessible OD surface of the vent header and down-comers opformed b.
osite the torus nitrog)en injection penetration (this examine was per from the catwalk.
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All accessible surfaces of the drywell nitrogen injection penetration'
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In summary the inspector concluded that SP-84-0014 should have looked at
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' the. steel liner from the stand point of a CAD system failure.
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the licensee's satisfactory visual examination of the nitrogen injection
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penetration for the torus is a good indicator that the steel liner plate
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will'also be satisfactory.
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l Within the areas examined, no. violation or deviation was' identified.
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.0nsite' Followup of -Written Reports of Nonroutine-Events at Power Reactor Facilities - Unit 2 (92700)
-(Closed) Licensee Event. Report (LER) 2-86-002
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During the Unit-2 1985-1986 refueling / maintenance-outage, visual and/or.
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- ultrasonic examinations of Reactor Recirculation System piping welds revealed the existence of 82. indications within~the heat-affected zones of 35 of the.116 welds tested. L This includes through-wall indications at
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five of the welds.
These examinations were performed in accordance with.
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L Generic' Letter 84-11 and IEN 84-41 using the General Electric automated j
SMART UT. System, where geometrically practical, with manual examinations
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supplementing the SMART UT as necessary.
The cause of the indications is I
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attributed to Intergranular Stress Corrosion Cracking (IGSCC) of the Class a
i 1, Type 304,' stainless steel and inconel butter material.
Thirty-three of j
the subject welds containing indications were repaired by weld overlay, i
and the remaining two welds with indications were evaluated as acceptable l
for an additional-fuel cycle by fracture. mechanics analysis.
Region II i
inspector's have ' observed many of the above ultrasonic examinations and i
subsequent repairs including the detection and sizing of.the indications
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in,the inconel butter nozzle to safe-end welds.
-In. addition, the inspector reviewed the GE flow analysis for'the crack
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indications detected in the two.unrepaired nozzle to safe end welds.
CP&L-will monitor the crack growth in the nozzle to ' safe end welds during the next refueling outage.
The inspector reviewed NRC letter of approval dated.0ctober'5, 1986, which granted the licensee's request to continue to operate until-the next refueling outage. 'The reporting and followup actions require for this LER have been implemented.
-(Closed) Licensee Event Report 2-86-003 Ultrasonic examination of the Unit 2 reactor shroud head hold-down bolts located on the steam separator revealed the existence of crack indications in 15 of the total 36 bolts.
The indications are in the crevice area s
between the Type 304 stainless steel and the inconel 600 material of each affected bolt, GE, Part No. 920D232G002.
The examination was performed in i
response to a GE urgent communication notice and was performed with the j
. steam separator in the separator / dryer storage pool.
Unit 2 was in a i
refuel / maintenance outage.
The indications are attributed to intergranular stress corrosion cracking.
I CP&L did not change out any of the bolts prior to restarting Unit 2 based on a GE plant specific safety analysis which concluded that the subject discovery was of no nuclear safety concern.
During the present refueling
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outage for Unit 1 CP&L examined the shroud head hold down bolts and i
discovered 20 of the 36 bolts to have crack like indications.
CP&L also does not plan to change out any of the hold down bolts on Unit 1.
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Twill. examine the-presently acceptable bold down bolts during.the~ next x
refueling outage for both Unit.-1 and Unit 2 to determine if any. additional bolts experience cracking.. A decision will be made at that time as to whether any bolts will be replaced.
The inspector will review the licensee's corrective actions during the next refueling outages as a part of the routine inspection coverage.
Within the areas examined,-no violation or deviation was identified.
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Licensee's Actions Regarding NRC Denial of ASME Code Relief Request on the
=HighPressure'CoolantInjection(HPCI)StudExamination-Unit 2.
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CP&L-letter dated March 25, 1986 requested NRC relief from the inservice inspection requirement of the ASME Code,Section XI, Table IWC-2500-1,-
for the volumetric examination of the HPCI pump studs at the Brunswick-a Steam Electric Plant.
CP&L stated that the examination was impractical to
.1 perform during the Reload. 6 outage because volumetric examination of the studs' from the top surface required removal of the stud cap-nuts using.
special ; equipment not availableL on-site.
Relief from. the -~ code
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-requirements was' requested toL permit. performance of' this examination during the refuelino: outage scheduled to begin in January 1988.
NRC did not grant CP&L's re'ief request based on the' fact that the licensee was ~
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aware that? the -stud. cap nut. removal was,necessary to perform the
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examination; during the-arevious refueling ing.the -extension of 'the outage and adequate time was
available -to perform tie examination dur inspection interval < as permitted by Paragraph IVA-2000(c):of the ASME
' Code.
Therefore, the required examination should be performed during'this
.j extend interval which ends. July-2,1987 for Unit 2.
The inspector s j
followup on this issue. revealed that CP&L has examined 17 of'18 bolts 'on
.the Unit'2 HPCI pumps.
The remaining bolt will require additional work in
[
. order to facilitate the examination, however, the ISI Project Engineer
'
stated that the' bolt will be examined or replaced by the end -of the
<
extended interval.
Unit 1 ' HPCI pump. stud examinations have been completed.
Within the area examined, no violation or deviation was identified.
8.'
(Closed) Inspector Followup Item 325/85-23-01, inspector had written this Inspectability of Welds Under Overlays Due to Overlay Weld Width.
The item to track the licensee's corrective actions regarding the insufficient
,
width of design weld overlays to allow examination of welds under the
. overlays.
The inspector's review of the licensee actions revealed that
.all of the ove'rlays on Unit 1 have been either been upgraded to full structural weld overlay or widened to allow full coverage ultrasonic examination.
Unit 2 weld overlays have not been completed to date.
However, the licensee has a program for completing the Unit 2 weld overlays.
This item is considered closed.
Within the area examined, no violations or deviation was identified.
,