IR 05000409/1986013
| ML20212C182 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 12/17/1986 |
| From: | Jackiw I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20212C134 | List: |
| References | |
| REF-GTECI-B-19, REF-GTECI-TH, RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-B-19, TASK-OR, TASK-TM 50-409-86-13, GL-85-06, GL-85-07, GL-85-13, GL-85-14, GL-85-22, GL-85-6, GL-85-7, GL-86-02, GL-86-2, IEB-86-003, IEB-86-3, IEIN-86-053, IEIN-86-072, IEIN-86-53, IEIN-86-72, NUDOCS 8612290411 | |
| Download: ML20212C182 (13) | |
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REGION III
Report No. 50-409/86013(DRP)
Docket No. 50-409 License No. DPR-45
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Licensee: Dairyland Power Cooperative 2615 East Avenue - South La Crosse, WI 54601 Facility Name: La Crosse Boiling Water Reactor Inspection At: La Crosse Site, Genoa, WI Inspection Conducted: September 27 through 11cember 5, 1986 Inspectors:
I. Villaiva J. Wiebe Approved By:
I.
ack', Chief
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R ctor P ojects Section 2D Date Inspection Summary Inspection from Seatamber 27 through December 5, 1986 (Report No. 50-409/86013(D1P))
Areas Inspected:
Routine, unannounced inspection by the resident inspector of Licensee Actions on Previous Inspection Findings; Operational Safety Verification; Monthly Maintenance Observation; Monthly Surveillance Observation; Licensee Event Reports; Generic Letters; IE Bulletins; TMI Action Items; Systematic Evaluation Program Action Items; Information Notices; and Open Items.
Results: No violations were identified. One item identified relating to containment integrity requirements was adequately addressed by the licensee.
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DETAILS 1.
Persons Contacted
- J. Parkyn, Plant Superintendent
- G. Boyd, Operations Supervisor
- L. Kelley, Assistant to Operations Supervisor L. Nelson, Health and Safety Supervisor R. Wery, Quality Assurance Supervisor S. Raffety, Reactor Engineer P. Bronk, Nuclear Engineer
- L. Goodman, Operations Engineer R. Brimer, Electrical Engineer D. Rybarik, Mechanical Engineer The inspectors also interviewed other licensee personnel during the course of the inspection.
- Denotes those attending exit interviews during the inspection period.
2.
Licensee Actions on Previous Inspection Findings a.
(Closed) Open Item (409/83022-14):
IPSAR 4.19.2, Add a Second Level Controller for Shutdown Condenser Shell Side. Although the LACBWR technical specifications do not require this redundant controller to be environmentally qualified, the licensee attempted to procure such a unit but was unable to do so. Consequently the licensee ordered an acceptable but not environmentally qualified controller and installed it during the 1986 refueling outage. This open item, therefore, is closed, b.
(Closed) Open Item (409/84009-04(DRP)): Minor Differences Between First Half of 1984 Radioactive Effluent Report and Regulatory Guide 1.21. The inspector notes that the licensee is including a section on total analytical error in these reports. This item is, therefore, considered closed.
c.
(Closed) Open Item (409/84016-01(DRP)):
Results of Additional Voltage Dropout Checks. By letter dated October 23, 1986, the licensee stated that the 1986 tests of all 30 magnetic release mechanisms were successful and that all dropout voltages were within specifications. Additional tests will be conducted in 1987 on two breakers whose dropout voltages were low during the 1985 tests.
Based on the results of the 1986 tests and the ongoing surveillance tests being conducted on the subject breakers, this item is considered closed.
3.
Operational Safety Verification The inspector observed control room operations, reviewed applicable logs and conducted discussions with control room operators. The inspector verified the operability of selected emergency systems, reviewed tagout records and verified proper return tc service of affected components.
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Tours of the crib house, reactor building, and turbine building were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenance. The inspector by observation and direct interview verified that the physical security plan was being implemented in accordance with the station security plan.
The inspector observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection controls. The inspector walked down the accessible portions of the Alternate Core Spray System to verify operability.
These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under technical specifications, 10 CFR, and administrative procedures.
On-0ctober 11, 1986, in preparing to return to power, isolation of t!.e containment building ventilation system was tested.
These tests include verifying the proper operation of the high level radiation alarm circuit and the inlet and exhaust dampers upon radiation from the immediate particulate radiation monitor.
The alarm and dampers purportedly operated sporadically during the test; however, the test data sheet indicated that the dampers functioned satisfactorily. As a result of this erroneous information, a maintenance request (MR) was issued requesting that the alarm circuit for the immediate particulate monitor instrument in the control room be made functional; however, no mention was made of the circuit associated with the dampers Consequently, the fact that the dampers were IN0PERABLE with respect to the immediate particulate monitor was not entered in the operations log. As a result, the fact that CONTAINMENT INTEGRITY would not have teen maintained upon
- high immediate particulate radiation was not taken into account on October 16 and 17, 1986, when control rods were withdrawn for rod position indication testing.
This violation was discovered by the licensee on November 10, 1986, while
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reviewing operational logs maintenance requests. This review discovered
that the immediate particulate monitor system was IN0PERABLE until October 20, 1986, the date on which the aforementioned MR was worked on and the alarm and dampers were made OPERABLE. The Resident Inspector was informed of the event the same day the violation was discovered.
Subsequent inspection actions included review of applicable records and logs and discussions with the appropriate licensee personnel. These actions confirmed that the licensee not only identified the violation but had also taken appropriate measures to correct the violation and prevent its recurrence.
In accordance with 10 CFR Part 2, Appendix C, Section V.A., a notice of violation will not be issued for this violation because it meets all of the following tests:
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a.
It was identified by the licensee.
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It fits in Severity Levels IV or V.
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It was reported; if required.
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It was or will be corrected, including measures to. prevent recurrence, within a reasonable time.
e.
It was not a violation that could reasonably be expected to have been prevented by the licensee's corrective action for a previous violation.
No other violations were identified.
4.
Monthly Maintenance Observation Station maintenance activities of safety-related systems and components listed below were observed / reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with technical specifications.
The following items were considered during this review:
the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and, fire prevention controls were implemented.
Work requests were reviewed to determine status of outstanding jobs and to assure that priority was assigned to maintenance of safety-related equipment which could affect system performance.
The maintenance activities (i.e., overhaul) on the 1A diesel engine for the high pressure service water system and a control rod drive mechanism were observed / reviewed.
Following completion of the maintenance on the 1A diesel for the high pressure service water system and the control rod drive mechanism, the inspector verified that they had been returned to service properly.
No violations or deviations were noted.
5.
Monthly Surveillance Observation The inspector observed technical specification required surveillance testing on the following nuclear instrumentation systems:
(a) wide range power channels 5 and 6, and (b) power range channels 7 and 8, including the automatic gain control systems. The inspector verified that testing
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was performed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that test results conformed with technical specifications and procedure requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.
The inspector also witnessed portions of the following test activities:
reactor startup, primary system heatup, turbine generator startup, increase in power, and surveillance testing of the 1A diesel engine for the high pressure service water system.
No violations or deviations were noted.
6.
Licensee Event Reports Through direct observations, discussion with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective actions were accomplished, and corrective actions to prevent recurrence had been accomplished in accordance with technical specifications.
a.
(Closed) LER 86-26:
Reactor Scram Due to Spike on Nuclear Instrumentation Channel 5 when Range Switch Upscaled. During a reactor startup on September 10, 1986, a spike of sufficient magnitude and duration occurred on Nuclear Instrumentation (NI)
Channel No. 5 to cause a reactor scram. The spike occurred while the operator was upscaling the range switch for NI Channel No. 5, a wide range channel, from the 60 E-6 to the 150 E-6 percent power scale.
Altnough the licensee was not able to determine the precise cause and source of the spike, troubleshooting revealed that a balance potentiometer on the drawer for NI Channel No. 5 was not providing the needed balancing and that manipulating the range switch could cause spiking. The initial corrective actions included adjusting the potentiometer to balance its circuit, and replacing the suspect range switch with a spare. Although of a lesser magnitude, spiking was observed during the subsequent startup. Additional corrective actions were therefore taken, including replacing the NI Channel No. 5 instrument drawer with a spare. Upon examining the replaced drawer, its balance potentiometer adjustment screw was found to be damaged such that the potentiometer would not function properly.
This type of scram can only occur at low power (i.e., at less than 15 percent power) when wide range Channels Nos. 5 and 6 are in a one-out-of-two trip logic.
In addition, when reactor power is above approximately 11 percent power, the range switch would have been placed in its final position for power operation, thereby eliminating spikes generated by manipulating the range switch.
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In addition to the corrective actions taken in direct response to this event, the licensee's long-term corrective actions include replacing the aging NI's with new units which are scheduled for installation during the next refueling outage. The new NI's will eliminate the need for upscaling and will replace the one-out-of-two trip logic used during startup with a two-out-of-four trip logic.
The new NI's should minimize if not eliminate similar trips in the future.
The inspector has evaluated the radiological safety impact of this event as well as the actions taken by the licensee, and has concluded that this LER should be closed, b.
(Closed) LER 86-27: Reactor Scram - Failure to Upscale Nuclear Instrumentation Channel 6 Following Channel 2 Failure. On September 10, 1986, during a reactor startup following the previously described scram, the reactor scrammed when NI Channel No. 6 reached its trip setpoint. Reactor power at the time of the scram was approximately 115 E-6 percent rated power, the trip setpoint for wide range NI Channels Nos. 5 and 6.
The ultimate cause for the scram was the reactor operatur's failure to upscale the range switch before NI Channel No. 6 reached its trip setpoint. This oversight was, in part, due to distractions caused by the failure of NI Channel No. 2.
While the reactor operator was watching for a 1/2 decade overlap between the source range and intermediate and wide range NI channels, he noticed that the count rate indication on source range NI Channel No. 2 started decreasing and the period meter was showing a negative period.
He checked the other instrumentation and noticed that the wide range channels were just starting to respond. The period indication for NI Channel No. I was approximately 100 seconds.
Shortly before, based on NI Channel No. 2's count rate, the reactor operator had calculated a period of 80 seconds.
Since the source range scaler, a digital readout of the counts per a selected time period, had been selected to NI Channel No. 2, the operator went to the panel to transfer the scaler from NI Channel No. 2 to NI Channel No. 1.
When he turned the selector switch to NI Channel No. 1, a short period warning alarm annunciated on NI Channel No. 2.
The operator acknowledged the alarm, which cleared, but immediately afterwards, the reactor scrammed on high flux on NI Channel No. 6.
NI Channels Nos. 5 and 6 are wide range channels for which the operator must manually adjust the range as power changes, with the scram setpoint being a set percentage of scale on any range. At the time of the scram, the selected range was 150 E-6 percent power, with the scram setpoint at approximately 115. A trip signal on either NI Channel No. 5 or No. 6 will scram the reactor at low power because the reactor protective system is in a one-of-two logic
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on high flux at low power. Calculations based on NI Channel No. S's chart show that the increase in power from approximately 10 E-6 percent power to the scram set point took approximately 2 minutes, l
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or about a 50 second period.
It appears that the operator was distracted by the failure of NI Channel No. 2 to such an extent that he did not notice the increase in power on the wide range channels.
As a result, he did not upscale the range switch in time to preclude the scram.
This type of scram can only occur at low power levels, when power range changing is necessary. Also, as in the previous scram, at above approximately 15 percent power, the high flux reactor protective system is in a two-of-four trip logic such that a trip on one channel will not cause a scram. As previously described the new NIs will incorporate a two-out-of-four trip logic and will eliminate the need for manual upscaling.
The inspector has evaluated the radiological safety impact of this event as well as the long-term and short-term corrective actions taken by the licensee and has concluded that this LER should be closed.
c.
(Closed) LER 86-28: Reactor Scram When Turbine Stop Valve Opened Later Than Required by Procedure. On September 13, 1986, during a plant startup, the reactor scrammed on a high reactor water level signal followed by actuation of the high pressure core spray pumps, starting of the emergency diesel generators and isolation of the containment building on a low reactor water level signal. The principal contributors to this event were operator error and poor communications.
The design of the LACBWR steam system is such that when the turbine stop valve is opened, the reheater inlet valve also opens.
Because of this design feature, LACBWR's startup procedure requires that the turbine stop valve be opened before steam is brought to the turbine side of the plant to assure that the steam line, turbine steam chest and reheater are gradually heated together when the main steam isolation valve bypass valve is opened.
In this instance, the actions of the reactor operator and turbine operator were not properly coordinated in that the step in the startup procedure regarding the opening of the turbine stop valve was skipped. As a result, the turbine stop valve was opened while reactor pressure was 800 psig such that a significant drop in reactor pressure resulted when the main steam isolation valve was opened. The sudden drop in reactor pressure resulted in a reactor water level swell which, in turn, caused the reactor to scram on high water level.
Following the scram, reactor water level underwent shrink and swell oscillations. During these oscillations, a low reactor water level was reached which caused the actuation of the high pressure core spray pumps, starting of the emergency diesel generators and isolation of the containment building.
Following the procedure, as written, would have prevented this event. Nevertheless, the procedure has been revised to preclude future recurrences of this event by clarifying sequence of actions
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to be taken during startup and by better coordinating the turbine and reactor plant startup procedures. The corrective actions taken by the licensee for this event have been reviewed and found acceptable; therefore, this LER is closed.
d.
(0 pen) LER 86-29: Reactor Scram - IB Forced Circulation Pump Trip Due to Seal Inject Supply Valve Closing. The inspector has not completed his review of this event. Accordingly, this event is being held open pending completion of an acceptable review.
e.
(0 pen) LER 86-30: Spike on Nuclear Instrumentation Channel 6 With Variance Between Alarm and Trip Points. The inspector has not completed his review of this event. Accordingly, this event is being held open pending completion of an acceptable review.
f.
(0 pen) LER 86-31: Start of Emergency Equipment /1B Static Inverter Transfer. The inspector has not completed his review of this event.
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Accordingly, this event is being held open pending completion of an acceptable review.
7.
Generic Letter Followup The inspector verified that the generic letters listed below were received by the licensee, reviewed for applicability by the appropriate onsite management representatives and appropriate actions taken.
a.
(Closed) Generic Letter 85-06: Quality Assurance Guidance For ATWS Equipment That is Not Safety-Related. This generic letter provides explicit quality assurance guidance for equipment that is encompassed by th ATWS rule, but which is not safety-related.
In brief, the guidance is less stringent than that for safety-related equipment, (e.g., it eliminates requirements for involvement of parties outside the nomal line organization and requirements for a fomalized program and detailed recordkeeping for all quality '
practices). The licensee reviewed NRC's proposed generic letter'
on this subject dated November 6, 1984, and determined that the non-safety related equipment associated with its ATWS facility change was subjected to the same o ality assurance requirements as t,
safety-related equipment. Based on the results of the licensee's review, this generic letter is closed, b.
(Closed) Generic Letter 85-07:
Implementation of Integrated Schedules for Plant Modifications. This generic letter describes the staff's intentions with respect to implementing integrated schedules, and solicits widespread industry participation in establishing priorities for modifications at individual plants to permit a well founded integration of implementation efforts.
In a letter dated May 9,1985 (LAC-10049), the licensee stated that it had a de facto integrated schedule in place with the NRC for a number of years, and that it did not require any modification to the Technical Specifications to continue this program. Accordingly, the licensee did not feel that an additional submittal to the NRC was
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required. This matter was discussed with the NRR Licensing Project Manager who indicated that the licensee's response was acceptable.
Accordingly, this generic letter is closed, c.
(Closed) Generic Letter 85-13: Transmittal of NUREG-1154 Regarding the Davis-Besse Loss of Main and Auxiliary Feedwater Event. This generic letter transmited NUREG-1154, Loss of Main and Auxiliary Feedwater Event at the Davis-Besse Plant on June 9, 1985, to all licensees for their review of the applicability of the Davis-Besse event to their facility.
In addition, licensees were requested to ensure that the information in NUREG-1154 was made available to their plant staff as part of the training program in connection with TMI Action Plant Item I.C.5, Feedback of Operating Experience to Plant Staff. The licensee has reviewed both the generic letter and NUREG-1154 and has concluded that differences in plant design are such that the details of the specific incident are not applicable to LACBWR. Nevertheless, the licensee has identified some aspects of the Davis-Besse event and conclusions that could apply to LACBWR.
The aspects and conclusions of the Davis-Besse event that could apply to LACBWR have been documented and circulated to cognizant LACBWR personnel.
In addition, as requested in Generic Letter 85-13, the licensee has included those aspects of the Davis-Besse event that could apply at LACBWR in its training program. Based on the actions taken by the licensee, this generic letter is considered closed.
d.
(Closed) Generic Letter 85-14: Comercial Storage at Power Reactor Sites of Low-Level Radioactive Waste Not Generated by the Utility.
The licensee has reviewed the subject generic letter and has determined that its handling of low-level radioactive wastes at LACBWR is in conformance with the provisions of Generic Letter 85-14.
In this regard, the licensee is presently increasing its ability to provide interim storage of LACBWR generated low-level waste and will continue to ship said wastes to existing sites to the maximum extent practicable.
In addition, the licensee is not presently storing or planning to store any low level waste that is not generated at LACBWR. Accnrdingly, this generic letter is closed.
(Closed) Generic Letter 85-22: Potential For Loss of Post-LOCA e
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Recirculation Capability Due to Insulation Debris Blockage. This
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generic letter informs licensee's of a safety concern regarding LOCA-generated debris that could block PWR containment emergency sump screens or BWR RHR suction strainers, and requests licensee's to review their plants for susceptibility to the described flow blockage.
Such blockage could result in a reduction of the net positive suction head margin of those pumps required to maintain long-term post-LOCA cooling. The licensee has reviewed Generic Letter 85-22 and has determined that the described flow blockage
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is not applicable to LACBWR, In brief, LACBWR's post-LOCA water sources do not include the recirculation of water that has been
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drained from the primary system and settled in the containment
building or its sumps. The post-LOCA water sources for LACBWR j
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include the overhead storage tank, the demineralized water storage tank and the Mississippi River. Consequently, the potential for pipe insulation entering into the post-LOCA water supply does not exist at LAC 8WR. Based on the foregoing, this generic letter is closed.
f.
(Closed) Generic Letter 86-02: Technical Resolution of Generic Issue B-19, Thermal Hydraulic Stability. This generic letter informs licensees of operating BWRs of the staff's findings on the resolution of Generic Issue B-19, Thermal-Hydraulic Stability.
In this regard, the staff has concluded that the General Electric and Exxon methods for calculation of core stability decay ratio are uncertain by 20-25% in predicting the onset of limit cycle oscillations, (i.e., decay ratio = 1.0).
Because of this conclusion, the staff has determined that a core having a calculated decay ratio of 0.80-0.75 may be on the verge of limit cycle oscillations within permissible operating space.
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Based on the above concerns, the generic letter requests BWR licensees to examine each core reload to assure that it is typical of previously evaluated cores having an acceptable stability margin.
The licensee evaluated the concerns expressed in the subject generic letter and consulted with Exxon, its fuel supplier, on this matter.
As a result of its evaluation and consultation with Exxon, the licensee has concluded that the LACBWR core conservatively meets the approved thermal hydraulic stability criteria for all operational (
conditions, that no special surveillance is required, that Technical
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Specification additions are not needed, and that this conclusion applies to all future reloads of the present fuel design. Based on the results of the liceW ee's evaluation, this generic letter is considered closed.
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IE Bulletins l
For the IE Bulletin listed below the inspector verified that the bulletin was received by licenseo management and reviewed for i.ts applicability to the facility.
If the bulletin was applicable the inspector verified that the written response was within the time period stated in the bulletin, that the written response included the information required to be reported, that the written response included adequate corrective action commitments based on information presented in the bulletin, that the licensee management forwarded copies of the written response to the
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appropriate onsite management representatives, that information discussed
in the licensee's written response was accurate, and that corrective
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action taken by the licensee was as described in the written response.
(Closed) IE Compliance Bulletin Po. 86-03, " Potential Failure of Multiple ECCS Pumps Due to Single Failure at Air-Operated Valve in (
Minimum Flow Recirculation Line. By letter dated November 10, 1986, j
the licensee informed the NRC that the LACBWR design is not subject to the problem discussed in the bulletin. This determination has been confirmed by the inspector; therefore, this bulletin is closed.
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y 19. ~ TMI' Action' Items (Closed) II.B.3.2.8. - Post Accident Sampling-Modify. By letter dated
- April _10,1986, -the NRC notified the licensee.that it had reviewed the licensees letter dated February 4, 1986, (LAC-11396) and concluded that the Post-Accident Sampling System (PASS) met all but one of the eleven t
criteria in Item II.b.3 of NUREG-0737. The sole remaining criterion requires. a final plant specific procedure for estimating the extent of core damage. The licensee agreed to provide this procedure by May 30, 1986.-
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By letter dated May 28, 1986, the licensee provided the procedure for
. estimating the extent of core damage. The NRC responded by letter dated
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August 11, 1986, and concluded that the PASS at LACBWR meets all of the criteria of Item II.B.3 of NUREG-0737. Accordingly, this item is
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considered closed.
In addition, the following two open items (1) (409/84009-01(DRP)) -
Use of Standard Test Matrix to Test Chloride and pH Analysis Techniques, and (2) (409/84009-02(DRP)) - Formalize and Document Yearly Training on Sampling and Transportation of Post-Accident Sample, are considered closed because they too were resolved in the above correspondence.
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i 10. Systematic Evaluation Program Action Items I
a.
(0 pen) IPSAR Ref rence 4.21.3.2 (409/84009-14(DRP)):
Install Remotely Operateo Solenoid Valves Outside Containment in Shutdown
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Condenser Veni to Offgas Line. By letter dated August 29, 1985, the licensee submitted the proposed design for the isolation valve, j
and a cost analysis of the modification.
In addition, the letter noted that implementation of the modification might not be complete during the 1986 outage pending delivery of parts. By letter dated August 6,1986, the NRC found the proposed design and completion schedule acceptable.
The line upgrading, except for installation of the solenoid valve, was completed during the 1986 outage. The NRC considered it t
acceptable to install the valve during a future outage, but no commitment date was established for installatien of this valve.
This item, therefore, remains open pending installation of this valve and subsequent NRC review.
b.
(0 pen) IPSAR Reference 4.21.5 (409/84009-17 (DRP)): Licensee to Install Two Check Valves in Series in Shutdown Condense Sample Line.
By letter dated liarch 10, 1986, the NRC concluded that the planned
' modification would provide acceptable isolation capability for the shutdown condenser sample line.
The inspector notes that the facility change for this installstion is in the review and approval i
stage. This item, therefore, remains open pending implementation
of the modification and subsequent NRC review.
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(0 pen) IPSAR Reference 4.26.4 (409/84009-18(DRP)): Provide Separate Power Supplies for the Full Scram Channels. This item remains open pending implementation of the modification which is presently scheduled for the 1987 refueling outage.
11.
IE Information Notices For the IE Information Notices listed below, the inspector verified that the notices were received by licensee management and reviewed for their applicability to LACBWR.
If a notice was applicable the inspector verified that the licensee had acceptably reviewed the implications of the notice and that appropriate corrective actions were performed as scheduled, a.
(Closed) IE Information Notice 86-53:
Improper Installation of Heat Shrinkable Tubing. The subject notice was reviewed for applicability at LACBWR by the cognizant electrical engineer. The findings of his review are documented in an IE Information Notice file dated October 23, 1986. The findings indicate that the subject heat shrinkable tubing is used for only two high range radiation monitors in containment, and that the splice kit was assembled per the manufacturer's specifications and in the configuration that passed LOCA testing (Wylie Test Report 58522).
The heat shrink, per se, is not accessible for direct inspection at LACBWR. Nevertheless, based on the information contained in the licensee's file on this matter and the licensee's splice inspection reports, this notice is considered closed.
b.
(Closed) IE Information Notice 86-72:
" Failure of 17-7 pH Stainless Steel Springs in Valcor Valves Due to Hydrogen Embrittlement."
This notice, including the relevant reference documents, have been reviewed by the cognizant mechanical engineer for applicability at LACBWR.
Based on correspondence with the valve manufacturer and his review of the matter regarding the medium in which the valves will operate at LACBWR, their inservice testing, their use factor (e.g.,
the few cycles the valves would be expected to operate in the post-accident sampling system), and the water chemistry to which the 17-7 stainless steel springs are exposed, he has concluded that LACBWR need take no further action on this matter. The inspector i
has reviewed the information in the licensee's notice file and
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considers this item closed.
i 12. Open Items l
Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action
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on the part of the NRC or 1-icensee or both. Certain previous open items
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which are updated but not yet closed are described in Sections 10.a, 10.b, and 10.c. of this report; new open items are described in Sections 6.d., 6.e., and 6.f.
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13. Exit Interview The inspector met with licensee representatives (denoted in Paragraph 1)
throughout the inspection period and at the conclusion of the inspection and summarized the scope and findings of the inspection activities.
The licensee acknowledged the findings as reported herein and did not identify such documents or processes as proprietary.
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