IR 05000344/1987030
| ML20235Z081 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 10/05/1987 |
| From: | Rebecca Barr, Mendonca M, Suh G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20235Z067 | List: |
| References | |
| 50-344-87-30, NUDOCS 8710200587 | |
| Download: ML20235Z081 (13) | |
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i U. S.1 NUCLEAR REGULATORY COMMISSION j
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REGION V
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Report ~No.' 150-344/87-30'
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- Docket No.-50-344
' License No.INPF-l'
' Licensee:
Portland General Electric Company'
to its 4.16 kv bus and was idled and manualb 9 9t down after five y!
minutes of operation. The automatic start signal resulted when the
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. Vfl-4 drawe,r in the 12.47 kv bus H1 was inadvertently opened by an
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operator / The operator had just replaced potential transformer t
fuses ff a unit auxiliary transformer drawer, located immediately
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above tdd VR-4 drawer, and was in the process of replacing metering fuses for the main generator. The fuses were being rephced to restore the system to its normal lineup. The applicable ioad dispatcher hold tag; had previously been cleared, but the fiaes had not been replaced at that time. The licensee concluded the cause of the VR-4 drawer being opened was operator error. The inspectors will review the licensee's evaluation o' the adequacy Of controls in hanging and releasing load dispatcher hold tags in follow-up of LER 87-21.
l With the plant in Mode 3, an inadvertent entry into Technical
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Specification 3.f).3 was made when operators discovered that two of the safety injection accumu'lators had levels below the technical specifications limit (T.S. 3.5.1).
The operators had been testing valve leakage in the "A" safety injection accumulator outlet line.
The immediate corrective actions were to fill the two accumulators J
to acceptable levels and perform boren sampling. The licensee identified personnel error as the cause of this event in that the operators had allowed the "A" accumulator level to drop below the low 16 vel alarm setpoint but not below the T.S. limit and were not sufficiently attentive to level indications in the other accumulators. The control room has one annunciator for low accumulator level with no reflash capabi.ity if another accumulator
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reaches the low level alarm setpoint. The licensee has not
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determined the root cause of the accumulators dropping below the i
low level limit. The inspectors will review the licensee's
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corrective actions in followup to LER 87-22.
With the plant at 3% power in the process of increasing pcwer level to 20% in preparation for main turbine overspeed trip testing, a turbine trip and feedwater isolation occurred on high high water level in "C" steam generator. The high high level resulted from leakage past the "C" feedwater control valve, FCV-530, apparently from a partially open manual operator while starting a main feedwater pump.
Upon observing rising level, operators reduced auxiliary feedwater and main feedwater pump flow, and an operator was dispatched to isolate the feedwater bypass and main feedwater lines at the local control panel.
The high high level setpoint at 75% was reached before the main feedwater line isolation valve cou?d i
be isolated by the operator. The licensee's evaluation of this
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everit will be reviewed in the follow-up of LER 87-23.
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In power escalation after its 1987 refueling outage with the plant holding at 30% power to effect improvements in secondary side
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chemistry, a turbine trip, resulting in a reactor trip, occurred on j
high high water level in the
"C" steam generator. The high high steam generator level resulted from bringing into service the second
main feedwater pump with the manual speed controller incorrectly set at full speed, vice minimum speed.
During subsequent steam j
generator level oscillations, the "C" steam generator reached the l
75% high high level setpoint. The initial licensee investigation indicated that operator error was the cause of the speed controller being set at full speed and was a contributory cause to the q
subsequent steam generator level oscillations. The licensee made a l
timely 10 CFR 50.72 notification to the NRC and completed a post trip review and a reactor shutdown / trip form per Administrative Order A0-3-7.
All systems functioned as required, including
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automatic starting of both auxiliary feedwater pumps on low low steam generator level and main feedwater isolation on high high level in "C" steam generator. The inspectors will review the licensee evaluation of this event in follow-up of LER 87-24.
6.
Follow-up of Open Items Unresolved Items 86-19-01(Closed) and 86-19-02 (Closed): The inspectors verified licensee corrective actions as specified in LER 86-06 for these open items. Specifically, the inspectors verified that valve internals were replaced and tested to assure acceptable sodium hydroxide and safety injection system response.
Open Item 86-24-01 (Closed): This open item dealt with failure of a QC inspector to verify inspection points.
The inspector reviewed the licensee's response dated November 14, 1986, to a related Notice of Violation and reviewed the licensee closecut of the associated Nonconforming Activity Rcrort NCAR-P-86-043.
Included in the NCAR closecut was training of contract QC personnel as documented in Inspection Report 87-24. Based on these reviews, and previons inspection effort, the subject open item is closed.
i LERs 87-03 and 87-05 (Closed):
For LERs 87-03 and 87-05, tie inspectors j
verified that the licensee's report met regulatory requirements and clearly documented the events and corrective actions. The inspectors further verified that corrective actions were completed. Tr<is closed
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LERs 87-03 and 87-05.
Open Item 87-24-02 (Closed):
In a July 9,1987, letter to the licensee, the NRC requested information pursuant to 10 CFR 50.54(f) to provide
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assurance that the design of piping supports in various safety-related systems met the licensing design bases. This issue was originally identified in the licensee's evaluation of Nonconformance Report (NCR)87-214 which indicated that the design load demands for a main steam line hydraulic snubber support exceeded the support allowable anchorage capacities.
In response, the licensee submitted letters dated July 10,
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July 15, July 27, July 31, and August 18, 1987, and met with NRC staff
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members on several occasions.
In addition, the licensee committed in its
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August 18 submittal to complete a long-term pipe support design
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verification program which would, in part, (1) respond to remaining NRC concerns which were identified during a July 21-23 NRC staff review in the licensee's Architect-Engineer's office, (2) verify that all other safety-related supports are designed in accordance with the Final Safety Analysis Report (FSAR) and, (3) determine if there are other design activities that require verification. The licensee committed to provide the final details of the long-term program to the NRC by September 30 and stated that the program would be substantially complete by July 1988. As stated in its August 18 letter, the licensee will report the results of the program to the NRC upon completion.
The review of the licensee submittals was performed by NRC staff members of the Office of Nucienr Reactor Regulation (NRR). As indicated in an August 21, 1987, letter to the licensee, NRR's review concluded that Portland General Electric had adequately evaluated the design adequacy of those pipe supports which were part of the original support design verification program, and that these supports will fulfill the requirements under which Trojan was licensed, as stated in Trojan's Updated FSAR.
NRR's safety evaluation report is expected to be issued in the near future. Based on NRR's review and the inspections described below and in Inspection Report 50-344/87-24, this item is considered closed.
During the review of the pipe support design verification program, questions were raised regarding the capability of the as-installed rock bolts to provide adequate anchorage for safety-related pipe supports.
In response, the licensee performed rock bolt tension demonstration tests, measurements of rock bolt embedment depths, and analyses of support capacities and demands.
The inspectors reviewed temporary plant test procedure TPT-218, "90ck Bolt Tension Test Demonstration," and discussed results with lice a test personnel. The licensee performed tension tests on nine rock bolts located inside containment for which pullout strengths of approximately 33,000 pounds were calculated based on the American Concrete Institute (ACI) 349 empirical pullout strength relationship, and assumed a shear cone emanating from the top edges of the rock bolt expansion shell extending to the concrete surface at an angle of 45 degrees.
Prior to the tension tests, the embedment length of each rock bolt was determined by ultrasonic measurement. The rock bolt embedment length was determined by measuring the total length of the bolt and subtracting the bolt length extending above the cancrete surface. A value of 2.5 inches was used for the distance from the end of the rock bolt to the top of the expansion shell, based on measurements of drilled-out rock bolt samples. During the course of the inspection, the inspectors measured the distance from the end of the rock bolt to the top of the expansion shell for two rock
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bolts subjected to tension testing and subsequently drilled out, and for five rock bolts previously removed during main steam line support modifications. The value of 2.5 inches used in TPT-218 appeared to be nonconservatively small based on the inspectors' measurements which ranged from 2.5 to 3.2 inches.
In response, the licensee demonstrated in its August 18, 1987 submittal that the shear cone could be taken from the centerline of the bottom of the expansion shell, vice the top edges of the expansion shell.
This would add an effective 1.1 inches to the I
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L embedment length. The nine bolts which were subjected to the rock bolt f
tension tests were chosen based on their accessibility for test purposes and on operability concerns for the associated safety system train. This j
selection criteria was appropriate, given that the tension tests were
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performed for demonstration purposes to verify the adequacy of
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as-installed rock bolt capacities. The plant was in Mode 5, Cold l
Shutdown, during the resolution of the pipe support design issue.
H In the tension tests, the rock bolts were tested to 34,500 pounds and exhibited no significant damage. The inspectors reviewed the measured tensile deflections of the rock bolt and surrounding concrete surface. No clear relationship of deflections to embedment length was evident.
Although the two bolts with the shortest embedment lengths had the highest deflection values, the two bolts with the largest lengths did not have the least deflection values. Of the nine bolts tested, three showed anomalous results: Bolt 1 of support SI-2501-R-1-64-SS-1106 exhibited two hairline cracks in the concrete surface immediately surrounding the bolt; and, a circular area of about eight inches in radius of hollow sounding concrete; Bolt 4 of support SI-2501-R-1-64-SS-1106 exhibited a circular area of about ten inches in radius of hollow sounding concrete; and Bolt 1 of support SI-2501-R-2-4-SS-1157 exhibited a circular area of about three inches in radius of hollow sounding concrete. To assess the significance of the test results the inspectors observed licensee activities for the testing of these three bolts which consisted of drilling out and examining Bolt 1 of SS-1106, tension testing of Bolt 4 of SS-1106 to failure, and removal and repair of damaged concrete surrounding Bolt 1 of SS-1157.
Bolt'l of SS-1106 was drilled out prior to tension testing to failure of Bolt 4 of SS-1106 using a three inch core drill. A continuous crack in the concrete about two inches below the surface was present, indicating-j the concrete had incurred damage in a shallower cone than the 45 degree j
shear cone assumed in calculating the empirical pullout strength. The l
setting of the expansion shell in the concrete was indeterminable as the I
cone drilling removed the surrounding grout and concrete in the immediate l
vicinity of the expansion shell.
Examination showed the cone to be I
partially inserted into the expansion shell.
The inspectors witnessed the tension test to failure of Bolt 4 of SS-1106.
Deviation 087-217 to test procedure TPT-218 was processed to j
allow for failure testing to a maximum loading of 57,500 pounds. Quality
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t.ontrol coverage in accordance with the test procedure and radiation protection coverage was observed. The inspector reviewed the cali Pation curve for the anchor bolt hydraulic jack and pressure gauge and ver afied that the calibration had National Bureau of Standards traceability. The test demonstrated the initiation of a visible failure cone on the concrete surface at about 39,000 pounds, and failure of the rock bolt at 51,500 pounds. According to the vendor catalog, the rock bolt had a nominal yield strength of 37,000 pounds and an ultimate strength of 50,000 pounds. The test demonstrated that the installed capacity of Bolt 4 of SS-1106 exceeded the calculated empirical pullout strength value.
After the test to failure, Bolt 4 was drilled out.
Examination showed that the expansion shell was fully expanded with the cone fully inserted
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into the shell. The expansion shell showed good contact with.the surrounding concrete.
In preparation for installation of threaded rods to replace Bolts 1 and 4, core drilling was continued to increase depth
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to eight inches below the lowest point of sound concrete remaining after rock bolt tension tests. Damaged concrete was removed using hand and pneumatic tools. The inspectors observed depth mapping of the damaged concrete area around Bolts 1 and 4.
Measurements were taken on a two inch square grid to document test results prior to repair and reinstallation of the support.
The repair for Bolt 1 of 55-1157 consisted of remcval of damaged concrete. The inspectors examined the remaining concrete which was about 2 inches in depth.below the original concrete surface at the lowest point
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at the rock bolt. The spalled area was a roughly circular area of about five inches in radius. A structural reinforcing M r was observed approximately 1 1/2 inches to the side of the rock bolt.
The repair was to add grout to replace the damaged concrete and replace the support.
In addition, the inspectors reviewed the licensee's activities in identifying supports with rock bolts for measurement of embedment lengths, reviewed ultrasonic measurement results, and inspected the base plates of seven supports to verify on a spot basis the presence of the stated number of rock bolts per the licensee's August 18, 1987, submittal. For the seven supports, the inspectors observed evidence of flappering performed in preparation for ultrasonic measurements.
In the identification of supports which used rock bolts, the licensee considered snubbers, restraints, hangers, anchors, and other types of suppurts but did not include pipe whip restraints.
In response to NRC concerns, the licensee committed in its August 18, 1987 submittal to address pipe whip restraints in its long range program.
7.
Reactor Coolant System Leak Rate Calculations An independent verification of the licensee's RCS leak rate calculations j
was performed.
Using NRC software, RCSLK9 (Reactor Coolant System Leak l
Rate Determination for PWR's), the inspectors were able to verify exactly
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the licensee's results, utilizing the licensee's tank volumes and I
calibration curves. A review of the design bases for the pressurizer, l
pressurizer relief tank, volume control tank, and reactor coolant drain tank volumes and calibration curves revealed several small discrepancies I
which were communicated to the licensee. The licensee stated that they felt the discrepancies had no appreciable effect on RCS leak rate results. The inspectors were able to independently verify that' the effect of the discrepancies was negligible.
8.
Design Basis Documentation Program By letter to Region V dated December 29, 1986, the licensee described a detailed action plan regarding programs to improve access to and use of design basis information for plant systems and to improve overall engineering performance.
This letter described the scope and schedule for the development of twelve design basis documents (DBD) for selected safety-related systems during the period December 1986 through September I
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1987. The status and results of this program were examined previously and reported upon in Inspection Report No. 50-344/86-46.
A review of licensee records and discussions with licensee representatives'during the current inspection period revealed lthe
.following status of this program.
The scope and schedule of the program, in terms of those systems for-which DBD's are to be prepared during the period December 1986 through September 1987, was found to be consistent with that described in the
licensee's letter of December 29, 1986. The systems have been combined in a manner such-that thirteen rather than twelve DBD's are to be prepared covering the plant systems selected. A review of licensee records revealed that final drafts, (Revision 0) for nine of_ the thirteen DBD's had been completed. Two of these had been approved for; issuance i
and seven were in the review and approval process. The remaining four
DBD's were scheduled for completion, including approval' for issuance, by
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the end of September 1987. Licensee management expressed confidence that the current schedule will be met.
Discussions with licensee representatives revealed that the DBD program is to continue beyond the scope described previously, to include nonsafety-related as well as safety-related plant systems. The current scope of the program is to include a total of approximately 65 DBD's. The licensee's present schedule for the expanded program calls for approximately thirteen DBD's to be completed each year through the year 1991.
In some instances the DBD is to cover generic topics applicable to several plant systems. Among DBD's in this category scheduled for l
completion in 1988 are those covering structural and seismic design bases and flooding, tornado, missile and high and moderate energy line break.
The DBD for the Auxiliary Feedwater System, approved for issuance on August 21, 1987, was examined for scope and completeness. The document included the following sections.
1.0 Functional Description.
2.0 Codes, Standards and Regulatory Documents e
3.0 System Design Basis 4.0 Component Design Basis 5.0 System Operation 6.0 Insrection and Testing Requirements 7.0 Design Basis Evolution
8.0 Figures 9.0 Tables 10.0 References (including calculation references)
The document also included an Appendix entitled OpEN ITEMS. This section contained a listing and description of items which have yet to be completed and/or verified for inclusion, as applicable, in various sections of the DBD. Among the items included in this section are supporting calculations in some instances which must be provided by the Architect / Engineer and information from the NSSS supplier (Westinghouse)
regarding design inargin included in accident analyses. Another item discussed in this section relates to a recommendation to throttle
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' auxiliary feedwater flow to limit the maximum flow rate on automatic.
. start of.the auxiliary feedwater pumps following a plant trip.- This
. recommendation lis among the corrective actions to prevent-condensation-induced water hammer discussed in the licensee's letter.to Region V dated ~ June'16,jl987. This subject was pursued by the NRC inspector.in an effort to assess-the completeness of.the DBD.
No_ discrepancies or significant concerns -resulted from the' inspector's review of this' program or' the documents examined.
9.
Exit Interview
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The'. inspectors met periodically with the licensee representatives denoted-in paragraph 1 during the course of the: inspection, as well 'as, at a
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meeting on September 25', 1987. The. inspectors summarized the findings' oft.
.the inspect 1on-act.1vities.
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