IR 05000456/1990002
| ML20006E969 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 02/15/1990 |
| From: | Lougheed V, Phillips M, Rescheske P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20006E964 | List: |
| References | |
| 50-456-90-02, 50-456-90-2, 50-457-90-02, 50-457-90-2, NUDOCS 9002270011 | |
| Download: ML20006E969 (12) | |
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, ^[g U.S. NUCLEAR. REGULATORY COMMISSION-i e'
REGION III.
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< Report No. 50-456/90002(DRS);:50-457/90002(DRS)1 4 Docket Nos.. 50-456;-50-457 Licenses No. NPF-72;:NPF-77
.w Licensee: ~ Coninonwealth Edison Company
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Post Office Box 767*
. Chicago, IL 60690-Facility Name:
Braidwood Nuclear Power Station Units 1 & 2
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Inspection At:
Draceville, IL 60407 LInspectio'n Conducted: January 16-31, 1990~
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Inspectors: doorxe celu 03)/6/9 0
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Peggy; R/. mescheske.
Date
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'f V. Fa iciaLo@ heed Da'te /
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! pproved By:~
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A Honte P.rPhillips, Chief Date
, Operational Programs Section.
Inspection Summary j
Insp'ection'on January 16-31. 1990 (Report No. 50-456/90002(DRS);
N E -457/90002(DR5)).
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Areas hispected: Routine, announced, safety inspection of modifications and'-
Loesign changes implemented during the Unit 1 refuel cutage (IP 37700). :Also
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_ included in this inspection was a review of licensee's program for. controlling-
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temporary alterations.- Commercial grade procurement-(IP 38703) was addressed-
when applicable.
.Results: Two previously: identified open items remain open, pending further review by NRR. One unresolved item was closed. An inspector concern was identified in paragraph 2d regarding control of emergency diesel generator compressed air bottles, and-requires review and resolution by the licensee.
Two violations were identified, as discussed below.
Based on the. modifications reviewed by the inspectors, the licensee was
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adequately implementing the modification program except as noted in the
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violation. The 10 CFR 50.59 reviews and safety evaluations were, in general, sufficient in detail and ensured that an unreviewed safety question did not exist.
Inspector concerns and questions were adequately addressed and
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resolved by the licensee. One violation was identified in Paragraph 3g, for failure to satisfp TS surveillance requirement 4.6.1.1.a.
This TS required containnent integrity verification every 31 days, and included verifying that blind flanges were secured in the closed position. A contributing factor resulting in this violation was considered to be the system engineer's unfemiliarity with TS, in thet, a surveillance procedure was not updated to reflect a plant modification. The ins sectors were also concerned with <,e considerable length of time taken by tie licensee to determine /recogni:< that a LCO existed. The overall safety significance of the event was minor, since containment integrity was maintained. The licensee will document this event
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in LER No. 90002.
The required documentation and reviews apoeared to be sufficient for controlling temporary alterations involving nonsafety-related and nonseismic systems and components. The licensee had made recent changes to the temporary
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alteration program (due to INP0 concerns) which were considered improvements, such as, requiring independent verifications and technical reviews, and controlling critical drawings.
In addition, the licensee had established a temporary alteration reduction plan which resulted in greater control over temporary citerations installed longer than 90 days. Regarding temporary alteration 08-1-019, which installed a diode across the pressurized PORY solenoids, one violation was identified in Paragraph 4 for failure to perform a design analysis and seismic evaluation, and classify / dedicate the part for safety-related application. The safety significance was considered to be minor, due to the credibic failure modes and impact on the PORV.
However, based on a review of the temporary alteration program and it's implementation, the inspectors concluded that a significant weakness existed in the control of temporary alterations affecting safety-related systems and components.
The licensee was reqce.ted to address the adequacy of the temporary alteration program in establishing and controlling the design cnd the selection of materiah and parts for facility changes performed as temporary alterations.
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I REPORT DETAILS L
1.
Persons Contacted
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K. L. Kofron, Production Superintendent E. W. Carroll, Regulatory Assurance G. E. Groth, Project Manager D. M. Kapinus, Assistant Technical Staff Supervisor G. R. Masters Assistant Superintendent, Operations
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D. J. Hiller, Regulatory Assurance Supervisor l
L. W. Raney, Nuclear Safety, Administration l
D. A. Rupert, Technical Staff Modification Coordinator l
D. J. Skoza, Project Engineer i
E. J. Stevak, Qut11ty Assurance Inspector U. S. NRC j
M. P. Phillips, Chief, Operational Programs Section, Region III l
T. M. longue, Senior Resident Inspector Braidwood Station All of the above persons attended the exit meeting held on January 31, 1990.
Other persons were contacted during the course of the inspection, including members of the licensee's engineering, operations, and regulatory assurance staff, and corporate engineering.
j 2.
Action on Previously Identified Items a.
(0 pen) Open Item (50-456/86023-01):
Licensee to submit (and NRR to
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approve) Justification to conduct Type A leak rate tests with
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containment liner weld leak chase channel plugs installed. The licensee responded to an NRR request for further information on January 31, 1989. This response dealt with the seismicity of the channel plutt The next Type A test was scheduled for approximately June 1991. 1his item will remain open pending NRR approval of the licensee % response.
b.
(0 pen) Open Item (50-456/87041-03; 50-457/87039-11): RHR suction.
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valve interlock in Technical Specifications appears to need a change
to increase the 360 psig limit given. The licensee submitted a
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Technical Specification (TS) change to raise the interlock limit
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in 1987.
In order to comply with the requirements of Generic Letter 88-17, the licensee has proposed deletion of this autoclosure interlock. The TS change was to be submitted to NRR in February 1990.
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This item will remain open pending NRR approval of the TS change.
i c.
(Closed) Unresolved Item (50-456/87041-05; 50-457/87039-13):
Licensee to evaluate if testing was needed to be performed on the
DRPI versus DRPI deviation alarm for rods in the same bank.
The
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t licensee determined that testing was required, generated temporary
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procedures (BwVS 2977 and DwyS 3002), and completed the testing on r
Units 1 and 2 in April and May 1988. Based on the inspectors' review of the information provided by the licensee, no violation was
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identified, and this item is considered closed.
d.
NRC Inspection Reports No. 50-454/89011(DRS) and 50-455/89013(DRS)
dated January 12, 1990, for the Byron station, docunented concerns
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regarding local emergency starting of a diesel generator (DG).
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Because of the similarities between the Byron and Braidwood plants,
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the inspectors discussed the concerns identified at Byron with Braidwood representatives.
It was determined that the specific
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concern with control of portable compressed air bottles also ap)1ied
at Braidwood. A portable cart with six air bottles was shared )y l
Braidwood Units 1 and 2, and was dedicated for use during emergency i
DG starts. The bottles were not labeled / tagged to identify their
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use or indicate empty / full. Control over the bottles was not maintained, other than that they were locked on the cart. During the inspection period, the licensee checked each bottle and verified they were full. The licensee planned to review the actions taken at i
Byron, and evaluate the need to take similar action.
At Byron, the a
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licensee had revised an existing quarterly surveillance procedure to add references to the emergency air supply bottle carts and a data
sheet. Acceptance criteria for minimum air pressure was also stated in the revision. Further, the air bottles were tagged with identification numbers. The inspectors considered the actions taken
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at Byron to be adequate for controlling the compressed air bottles,
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l and in assuring their availability if required.
No violations or deviations were identified. Two previously identified open items remain open, pending further review by NRR. One unresolved item was closed. An inspector concern was identified in Paragraph d, which requires review and resolution by the licensee.
3.
Modification and Design Changes The inspectors reviewed a sampling of safety and nonsafety-related modifications performed by the licensee during the Unit 1 refuel outage.
The inspection focused on the im)1ementation of the facility (changes and
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included a selective review of tie associated documentation e.g.,
i 10 CFR 50.59 evaluations, desi
procedure and drawing updates)gn reviews, testing criteria and results,
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, and verification of the installation.
Discussions were held with licensee staff involved in the modifications, which included the agnizant system and design engineers, and modifications and operations persons. The licensee's modification program was controlled by plant administrative arocedure series BwAP 1610, and other documents, suc1 as, the Commonwealti Edison Company (Ceco) Quality Assurance Manual (Quality Procedure Q.P. 3-51). Safety evaluations and 10 CFR 50.59 reviews were conducted under BwAP 1205-6.
The following facility changes were reviewed by the inspectors:
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a.
Modification H20-1-87-1118: This modification replaced and upgraded fuses and fuse holders in card racks in the Engineered Safety Feature (ESF) Actuation System Westinghouse 7300 cabinets. The L
purpose for )erforming the changes was to enhance circuit relibbility )y eliminating equipment malfunctions due to damaged fuses. The licensee had experienced reactor trips due to blown fuses in the past. The ins)ectors reviewed the modification package and held discussions with t1e system engineer. The design change had been reviewed by the original designer, Westinghouse, and found to be acceptable.
The inspectors reviewed the procurement and receipt inspection documents, and verified the parts were properly classified, inspected, and accepted for_ safety-related application.
The 10 CFR 50.59 safety evaluation was also determined to be adequate.
b.
Modification H20-1-87-090: This modification installed an automatic leak detection system for the equipment hatch personnel airlock door seals. The detection system was designed to provide continuous pressure testing of the containment airlock door seal gaskets, with an annunciator / alarm located in the control room. The system was previously installed at Braidwood Unit 2 and Byron Units 1 and 2.
A change to TS surveillance requirement 4.6.1.3.6 was necessary prior to the system (at all four units) being declared operable.
The licensee submitted a request to NRR for the TS amendments on October 4, 1988. The inspectors reviewed the modification package and found it to be adequate.
c.
11odification M20-1-08-085: This modification converted a support guide on the safety Injection accumulator fill line into a rigid three-way sup> ort. The licensee had experienced various degrees of cracking of tie fill lines on all four accumulators at both units at the Braidwood and Byron sites. The cracks ranged from no indication of cracking to through-leakage. The licensee attributed the cracking to excessive vibration during operation of the fill line, including possible vibration due to backflow through the Kerotest packless metal diaphragm valves. The support guide on one fill line at Byron Unit 1 had been replaced under a temporary alteration.
Since no cracks developed following the installation of the support, similar supports were installed on the other three units as modifications.
Extensive testing was done by Sargent & Lundy at Byron, both prior to and following the modification.
The test consisted of taking vibration measurements while the fill line was used in various modes, as would be done in normal operation. For the modification on Braidwood Unit 1, the same post-modification testing was performed
{r by Sargent & Lundy, only on a reduced scale, due to the previous
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testing.
The testing indicated that no appreciable vibration occurred with the new su) port. However, it was noted that the Sargent and Lundy letter, w11ch trantmitted the results of the pre-modification
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testing at Byron, indicated that vibration levels ex>erienced during i
the testing were minimal and would not have led to tTe observed
cracking.
Subsequent to the inspection period, the licensee at Byron Unit 2 identified additional cracking of the 20 fill line. On February 8, 1990, the licensee determined that the nuts holding the U-bolt had
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loosened such that the support no longer restrained vertical motion.
t The loosened nuts were attributed to the low torque force used in
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the nodification.
The licensee retightened the nuts using a higher torque value and tack-welded them into place. The licensee at i
Braidwood Unit 1 inspected their support, did not identify any
cracking, and all nuts were in place.
The inspectors revieco the modification package for the Braidwood Unit 1 modification and held discussions with the system engineer, site Quality Control, and Construction personnel.
The inspectors
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concluded that the modification was properly installed, and that acceptable t4aterials were ustd in the installation.
d.
Modification M20-1-88-074: This modification connected the reference leg of the Refueling Water Storage Tank (RWST) to an overflow drain
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line. The reference leg had previously been open to atmosphere and the licensee was concerned that a vacuum could develop in the tank during an accident, which would cause the level indications to be inaccurate. During review of the modification package, the
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inspectors determined that the licensee had not addressed in the
10 CFR 50.59 safety evaluation whether the margin of safety for switching from the RWST to the emergency sump post LOCA would be (
impacted by this modification, especially since the modification pacL6ge identified the potential for the level instruments to be i
rendered inoperable as a result of the modification.
Discussions
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l with the licensee indicated that they were knowledgeable of the Technical Specification section which addressed instrument accuracy, r
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nodification on the margin of safety defined by that section, since l
a problem could only exist if the tank was overfilled.
In response l
to the inspectors concern, the licensee agreed to modify the safety l
evaluetion. The inspectors reviewed the revised 10 CFR 50.59 safety evaluation, and concluded that it was adequate.
During the review of the modification, the inspectors noted that the new design cautioned that there was a potential for the level instrumentation to reed erroneously if the RWST was overfilled.
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caution statement had been added to procedure Bw0P SI-13 " Filling
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the RWST", to alert the operators of the potential problem.
The level
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indicators had TS requirements regarding accuracy and operability
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which could be affected by this modification (as discussed above).
After discussions with the licensee, the procedure was revised to l
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i provide instructions for the operator should overflow occur, so that the level instrunentation accuracy could be regained.
e.
Modification M20-1-88-027: This modification removed a check valve in t:,feedwater bypass line and altered the feedwater isolation i
logic. The modification was performed because the flow split between feedwater going through the bypass line and that going i
through the preheater line did not meet design ssecified values.
Since the design specified a maximum flow throug1 the preheater section, this limited the power level that the reactor could obtain. The flow split prob 1cm was remedied by removal of the check valve. The licensee had identified the need to ensure that the
function of the check valve was met (to prevent flow of ouxiliary feedwater back through the bypass line, thus delaying the i
addition of water to the steam generator during an accident). The
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function of the check valve was maintained by modifying the standard
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Westinghouse feedwater isolation signal (P-4 reactor trip plus low Tav) to feedwater isolation on only a reactor trip.
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i analysis of the effect of this change on the accident analyses was
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l done by Westinghouse, and was found acceptable. The ins ectors l
reviewedthemodificationpackage,includingthe10CFRb0.59 safety
cvaluation, and discussed the modification with the system and design engineers. The inspectors also discussed the post-modification i
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testing with the system engineer, and witnessed portions of the l
testing.
f.
Modification M20-1-87-009:
This modification installed reach rods on valves in the Chemical Volume and Control System (CVCS). The j
valves were used to transfer used resins from the mixed bed and
cation dominerali ers to the waste disposal system.
Installation of
the reach rods was done for ALARA considerations. The valves were located in a non-seismic, non-safety-related section of the CVCS.
The inspectors reviewed the modification package and the installation e
and test documentation. The inspectors also conducted a field
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l-verification of the new valve handwheels to ensure that they met the
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l intent of the podification and were properly identified. Inadequate identification could result in the wrong demineralizer being sluiced, l
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which could result in ALARA concerns.
Since the new handwheels were i
on the opposite side of a radiation barrier, there was no visibic
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connection between the handwheels and the demineralizer being sluiced.
The inspectors verified that sufficient identification existed on the handwheels to assure equipment traceability.
g.
Modification H20-1-89-005:
This modification converted three spare (welded) containment penetrations into normally closed blind-flanged penetrations (P-63, P-64, and P-74).
This was done to allow
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additional openings into the containment for use during refueling (e.g., to allow access for various hoses or cabling needed during an outage). The inspectors reviewed the modification package, and the
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post-modification testing. The design engineering was adequate, and the 10 CFR 50.59 safety evaluation addressed all necessary areas.
The post-modification test was in accordance with the requirements of 10 CFR 50 Appendix J for blind-flanged penetrations. The test l
results were acceptable, and ensured that the modification did not
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result in any compromise of containment integrity. The licensee had
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identified the need to perform periodic surveillance testing of the i
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new penetrations and htd written procedure BwVS 6.1.2.o-1.26 " Primary Containment Type B Local Leak Rate Testing of Spare Containment
Penetrations", to do the testing.
The inspectors reviewed the Technical Specifications, and on January 19, 1990, identified the need for a revision to surveillance procedure, 1Bw05 6.1.1.a-1, " Unit 1 Primary Containment i
Integrity Verification of Outside Containment Isolation Devices."
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Subsequently, the licensee determined that the surveillance
)rocedure had not been updated to require verification of the three i
L alind flanges, and therefore, the TS surveillance requirements for l
Verification of primary containment integrity had not been
l satisfied. TS 3.6.1.1 requires that primary containment be l
maintained whenever the reactor is in Modes 4 through 1.
TS 4.6.1.1.a requires, in part, verification of containment integrity every 31 days by ensuring that containment isolation devices, including blind flanges, are secured in the closed position. The
three blind flanges were verified intact during performance of the j
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Type B local leak rate test on Hoytmber 24, 1989. To be within the j
31 day period (plus twenty five percent allowable extension), a i
surveillance was required to have been performed by December 30, 1989.
The licensee's failure to satisfy TS requirement 4.6.1.1.a. perform the surveillance and to
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l is considered to be a violation i
(456/90002-02(DRS)).
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Discussions with the licensee indicated that one of the contributing's factors resulting in the missed surveillance was the system engineer unfamiliarity with the TS. The licensee will document this TS j
violation in a Licensee Event Report (LER No. 90002). Root cause analysis and corrective actions will be addressed in the LER, as
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well as in the licensee's response to the violation issued in this inspection report.
TS 3.6.1.1 states that without containment integrity maintained, the licensee has one hour to restore containment integrity or place the
reactor in hot standby. On January 19, 1990, the inspectors
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identified the problem with the TS surveillance to the licensee.
At 1145 on January 19, the temporary change adding the three blind
i flanges to surveillance procedure 18w05 6.1.1.a-1 was approved by f
the Operations Department. At 1445, a Deviation Report was written to cocument the failure to verify closure of the penetrations. At
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1520, over three and one half hours after revising the arveillance i
procedure, tht licensee entered the Limiting Condition for Operation
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(LCO). The LCO was exited at 1610, subsequent to successful
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conpletion of the surveillance. According to the licensee, the time
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I period prior to entering the LCO was used for positive determination i
that the TS surveillance interval had been exceeded. The inspectors
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did not consider the licensee's decision process very timely.
l However, the overall safety significance of the event was minor since containment integrity was maintained. Based on discussions with the licensee and the NRC Senior Resident Inspector, it was j
concluded that the untimely entry into the LC0 was an isolated case.
Desed on the modifications reviewed by the inspectors, the licensee was adequately implementing the modification program (except as noted in the
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violation). The 10 CFR 50.59 reviews and safety evaluations were, in
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general, sufficient in detail and ensured that an unroviewed safety
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question did not exist.
Inspector concerns and questions were adequately addressed and resolved by the licensee.
One violation was identified in i
Paragraph g for failure to satisfy TS surveillance requirement 4.6.1.1.a.
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4.
Temporary Alterations The inspectors reviewed the licensee's procedure for controlling
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temporary plant alterations. Administrative Procedure DwAP 330-2,
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" Temporary Alterations." Temporary alterations were defined as lifted leads / devices, electrical jumpers, or temporary ucchanical alterations.
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The procedure included: temporary alteration installation / removal records cnd tracking log, engineering review and safety evaluation checklists, monthly operating review and quarterly field inspection review
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sheets (performed for all installed temporary alterations), a technical monthly review sheet for temporary alterations installed-longer than three months, and a critical drawing notification form. The engineering
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checklist addressed the specific characteristics of a temporary alteration which would require an additional written engineering review and an onsite review.
Items addressed included the effect on safety-related equipment, radiation concerns, fire protection, environmental qualification, electrical loading, temperature and pressure affects, etc. Any item checked as applicable required further review.
All temporary alterations required a 10 CFR 50.59 safety evaluation to be performed in accordance with BwAP 1205-6.
The requjred documentation and reviews appeared to be sufficient for controlling temporary alterations involving nonsafety-related and nonseismic systems and components. The inspectors noted that the licensee had made recent changes to the temporary alteration program (due to INPO concerns) which were considered improvements, such as requiring independent verifications and technical
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reviews, and controlling critical drawings.
In addition, the licensee
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had established a temporary alteration reduction plan which resulted in
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greater control over temporary alterations installed longer than 90 days. The program included holding monthly meetings to discuss the status of open temporary alterations.
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1he inspectors reviewed a sampling of temporary alterations installed on Units 1 and 2.
Temporary Alteration 88-1-019 (Unit 1) installed a diode across the pressurizer power operated relief valve (PORY 1RY456)
solenoids in February 1988.
T1e diode was installed to eliminate induced circuit noise spiking on source range nuclear instrumentation channel N32 when the handswitch controlling operation of the PORY was repositioned. The system affected by the temporary alteration was considered to be safety-related and seismic.
The documentation and reviews included an engineering evaluation discussing the potential inpact that failure of the diode would have on the operation of the PORV. The conclusion was that PORV operation would not be affected, except to fail closed (safe).
If the diode would fcil, it would most likely fail as an open circuit and have no impact on the solenoids.
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Failure of the diode as a short circuit was considered as a possibility, however, due to the type of diode used, not a credibic failure. Fa N re of this type would result in losing operation of the solenoids. 'te
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inspectors noted that one possible failure mode was not identified by the licensee, which was the loss of PORV indication if the shorted diode opened a fuse in the circuit. Although the licensee stated that this i
type of failure was unlikeiy, they planned to add the f silure mode to the evaluation. The engineering evaluation also described the type of diode used, and stated that the acceptable part was designated JAN1N645.
Installation of the diode was completed under Nuclear Work Request (HWR)
No. A20085 in February 1988.
In April 1988, a request (No. NR003) to make the temporary change a permanent modification was initiated by the licensee. A modification request review checklist was completed in accordance with BwAP 1610-1T1, and resulted in the decision to document the change under the minor modification arogram. As of the end of this inspection period, the permanent change lad not yet been completed, and the temporary alteration remained installed. The inspectors reviewed the abuve mentioned documents and held discussions with representatives from the licensee's operations, modifications, and regulatory assurance staff. The.9sults from review and discussion of temporary alteration 88-1-019 were as follows; a.
The 10 CFR 50.59 safety evaluation was censidered sufficient to ensure that an unreviewed safety question did not exist. The engineering evaluation was adequate in identifying the credible failure modes and impact on the PORV.
b.
The modification request review checklist stated that an engineering analysis was required to evaluate the installed diode for it's ratings, and that a preliminary study had already been done.
- According to the licensee, the preliminary study was an undocumented telephone conversation. The inspectors determined that the design
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and selection of the diode should have been evaluated, documented, reviewed, and approved, prior to installation of the temporary alteration. The temporary alteration program did not provide for establishing and controlling the design and selection of materials
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and parts for safety related application. These types of controls
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were only required prior to the temporary alteration being changed to a permanent plant modification.
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c.
According to the licensee, the particular diode selected was a storeroom item, routinely used in plant computer applications, and was considered to be a reliable part. The modification request indicated that the diode chosen should be evaluated as safety grade.
The temporary alteration program did not provide for classification and dedication of parts for safety related application. The in:pectors noted that it was possible, in this example, that the part would not have been classified when the temporary alteration became a permanent modification.
d.
The inspectors determined that a seismic evaluation should have been perf ormed prior to the diode being installed.
The temporary alteration program did not provide for review of the impact on the seismic qualification of an affected system. The modification request indicated that engineering was required to analyze the installation for seismic qualification prior to the temporary alteration becoming a permanent modification.
The inspectors concluded that, for temporary alteration 88-1-019, the licensee failed to aerform a design analysis and seismic evaluation, and classify / dedicate tie part for safety-related application. The safety significance of the problems identified with installation of the diode was considered to be minor due to the credible failure modes and impact on the PORV.
However, based on the review of the temporary alteration program and it's implementation, the inspectors concluded that a significant weakness existed in the control of temporary alterations affecting safety-related systems and components.
10 CFR 50, Ap)endix B, Criterion III, requires in part, that design changes be su) ject to design control measures commensurate with those applied to the original design. These design control measures shall provide for verifying the adequecy of the design. Further, measures
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shall be established for the selection and review for suitability of application of materials and parts.
Contrary te this Temporary Alteration 88-1-019 was not subject to design control measures commensurate with those tpplied to the original design. This is considered to be a violation (456/90002-01(DRS); 457/90002-01(DRS)).
Administrative controls did not provide for establishing and controlling the design and the selection of materials and parts for facility changes performed as temporary plant alterations.
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Discussions with the licensee indicated that they were in agreement with the inspectors conclusions. The licensee planned to resolve the
temporary alteration program dcficiencies by reviewing and revising procedure BwAp 330-2. The inspectors determined that the licensee also
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needed to re-evaluate the installation of temporary alteration 88-1-019.
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Exit Meetina
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The inspectors met with the licensee representatives (denoteo in
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l Paragraph 1) on January 31, 1990. The inspectors summarized the scope
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and findings of the inspection. The licensee acknowledged the statements made by the inspectors with respect to the violations and other concerns.
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The inspectors also discussed the likely informational content of the
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inspection report with regard to documents or processes reviewed by the inspectors during the inspection and licensee did not identify any such docunents/ processes as proprietary.
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