ML20045J021
| ML20045J021 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 07/09/1993 |
| From: | Horn G NEBRASKA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NSD930840, NUDOCS 9307220301 | |
| Download: ML20045J021 (8) | |
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GENERAL OFFICE n,
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NSD930840 July 9, 1993 U.
S. Nuclear Regulatory Commission Attention:
Document Control Desk Washington, DC 20555
Subject:
NPPD Response to Inspection Report 50-298/9318+01 (Reply to a Notice of Violation)
Centlemen:
During an NRC inspection conducted by Mr. C. E. Johnson of the NRC from April 26 to April 30, 1993, and from May 10 to May 14, 1993, certain licensed activities were identified being in violation of NRC requirements.
The first violation concerned inadequate as safety-related control room ventilation correc tive actions to properly maintain a radiation monitor.
The second violation was in regards to an inadequate safety evaluation review of a design change modification to the drywell ventilation radiation monitor.
The following is our response in accordance with 10CFR2.201.
STATEMENT OF VIO1ATION Criterion XVI of 10CFR50, Appendix B, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, deficiencies, and nonconformances, are promptly identified and corrected; and that measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition.
Contrary to the above, historical measures taken upon failure of a safety-related control room ventilation radiation monitor did not assure that the cause of conditions were determined and that corrective actions precluded repetition; for instance: (1) from :987 to 1989 inadequate procedures that allowed unintended high voltage to be admitted to the monitor either were not corrected or were not corrected in a timely manner, and (2) from 1990 to 1993 the monitor's filter paper drive belt experienced repeated failures.
This is a Severity Level IV violation (298/9318-01) (Supplement I).
REASON FOR VIDIATION Nonconformance Report 87-112 was generated August 10, 1987, upon failure of the Control Roam radiation monitor iodine channel detector.
It was determined that the failure was due to high voltage transients imposed on the detector when re-connecting it to the monitor during scheduled maintenance.
DC high voltage is generated internally by the monitor and applied to the detector. The ver-lor manual indicates that the internal high voltage should be turned off prior to connecting the detector to the monitor, in order to avoid high voltage transients which could damage the detector sensor.
There was no historical basis to suggest that the monitor would fail outright by not turning off the high voltage, only that the service life of the detector may have been shortened, which resulted in its eventual failure. A high voltage transient would not likely occur as a
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U.'S Nuclear Regulatory Commission July 9, 1993 , Page' 2 result of re-connecting the detector, but was possible, and it was considered prt li to incorporate measures into station procedures to eliminate a potential. tge transient as a result of normally scheduled maintenance. To address the fait is were made to appropriate electronic and known source calibration procedures. .we, the procedure revisions were not approved until March 1989. The failure to revise the procedures in a timely manner is as follows. The original engineering response to Nonconformance Report 87-112 was submitted in November 1987, in conjunction with a procedure change to the functional test. However, because the historical performance of the monitor did not appear to support the root cause analysis, and because the monitor manufacturer provided no means of de-energizing the high voltage prior to connecting the detector, I6C supervision disagreed with both the root cause analysis and the recommended corrective action. The engineering response was subsequently returned for re-evaluation. Agreement regarding necessary corrective action was reached in August 1988 Concurrent-ly, existing Chemistry procedures on operation of this monitor were being consolidated into the applicable surveillance procedures, which further delayed final approval of the procedure revisions. Nonconformance Report 89-061 was generated in April,1989, and Nonconformance Reports 89-156 and 89-157 were generated in August, 1989. Each of these events addressed an intermittent problem with the Control Room radiation monitor iodine detector. It was determined that the intermittent malfunctions of the datector were a result of previous maintenance on a 480 VAC bus supply transformer in April,1989. The transformer had been de-energized for maintenance during the 1989 Refueling Outage, using an approved station operating procedure. This procedure provided a listing of the loads supplied by the transformer, which included the Control Room ventilation radiation monitor. No provision was included in the procedure to reduce the monitor internal high voltage prior to re-energizing the bus supply transformer. It was determined that re-energizing the monitor supply bus without taking steps to minimize the resulting high voltage transient produced by the monitor's internal high voltage power supply damaged the detector, similar to re-connecting the detector with the monitor energized. As corrective action, the procedure was revised to include reducing the detector high voltage prior to de-energizing and re-energizing the supply bus. Several mitigating factors exist regarding these events: 1. No detector failures occurred as s result of the delg implementing procedure revisions in response to Nonconformance Report 87-112. The subsequent problems occurred in 1989, after the surveillance procedure revisions were in place. 2. The root cause analyses for Nonconformance Reports 89-061, 89-156, and 89-157 correctly identified the root cause as a procedural deficiency, in that the procedure governing normally scheduled de-energization of the plant distribution transformer did not ir, elude provisions to reduce the monitor's internal high voltage. De-energization and re-energization of this essential 2000 KVA bus supply transformer for periodic bus maintenance is an infrequent plant evolution and, thus, the necessity for applicable procedural controls related to the Control Room monitor was not previously identified. 3. It was not initially recognized that re-energization of the monitor power supply may have had the same effect as re-connecting the detector, in part because the vendor manual specifically addressed only detector re-connection, and not de-energization of the monitor itself.
Y ,, U. S. Nuclear' Regulatory Commission July 9, 1993 Page 3 The second instance cited as inadequate corrective action for this monitor involved the particulate channel filter paper drive. Nonconformance Report 90-125 was generated in-September, 1990, due to a failure of the filter paper drive belt. 11e Nonconformance Report was subsequently voided after it was determined that failure of the filter paper to advance would not affect the safety function of the monitor. Failure cf the filter i I paper to advance causes particulates to accumulate at the detector, resulting in an L indication error in the conservative (high) direction. Deficiency Report 93-018 involved a similar failure of the filter paper drive in~ January, 1993. The root cause of the event was found to be inadequate corrective action, because two previous failures had occurred in September,1990, and June, 1991. The Deficiency Report root cause analysis determined that the second failure in June, 1991, which was resolved under a Maintenance Work Request, should have indicated the need for drive unit replacement. Inadequate corrective action in response to the second failure resulted in i the failure in 1993. CORRECTIVE STEPS TAKEN AND THE RESULTS ACHIEVED Regarding the timeliness of Nonconformance Report resolutions, the procedure governing the corrective cetion root cause analysis and resciution was revised in April 1992. A formalized means to resolve situations similar to that discussed above (i.e., the delay-in implementing the appropriate corrective procedure changes), as well is to track and enforce timeliness requirements through the use of monthly reviews of open items, is now in place. Section 2.4, Appendix B of this Notice of Violation states in part: "...it was not evident that the licensee pursued either of these possibilities to determine...whether high voltage could have caused the monitors to be out of calibration. There appeared to be no adequate corrective actions or root causes determined in the nonconformance reports reviewed to prevent repetitive occurrences." It is important to note that these issues are related to the inherent design characteris-tics of the monitor in service at the time of the failures, and that calibration problems did not result from the circumstances described above. The high voltage transients were not imposed on the detector from an external source, but were generated from an internal power supply to the equipment. Although not timely, specific actions were taken to address those concerns. The adequacy of those corrective actions to prevent.- repetitive occurrences is supported by the fact that no further detector failures due to high . voltage transients have occurred. For the second instance cited, the Deficiency Report program was established to ensure - that conditions adverse to quality which fall below the threshold of a Nonconformance Report are promptly identified and corrected. The establishment of the Deficiency Report program, although not a result of the events discussed above, is relevant to the extent' that malfunctions of -safety-related equipment which do not result in a Nonconformance Report will, nevertheless, be formally addressed in order to preclude long-term equipment perfnrmance concerns (for example, failure of the monitor filter paper drive belt). The-second failure of the drive belt discussed above occurred prior to establishment of this program. The purpose of the program is to address failures such as this one.
1 l . U. *S. Nuclear Regulatory Corraission - July 9, 1993 l , Page 4 CORRECTIVE STEPS WHICH UILL BE TAVEN TO AVOID FURTHER VIO1ATIONS The Control Room ventilation radiation monitor has been replaced, which will preclude a recurrence of problems related to age and obsolescence of this equipment. t The inadequacy of both the timeliness and completeness of the corrective actions taken in response to these events is acknowledged. However, none of the four open Nonconforma-nee Reports on this monitor wnich reference completion of the replacement activity relied l solely on replacement as appropriate corrective action. In each case, additional corrective action was performed, as the apparent cause of the failure dictated. As indicated above, NPPD admits the alleged violation and agrees that long-term corrective actions taken in response to failures were neither timely or sufficiently complete to prevent recurrences. Specific root cause analysi and corrective actions to resolve the equipment problems, and programmatic actions to resolve the larger issue of s deficiencies in our corrective action program as discussed above, have been implemented. DATE WHEN FULL COMPLI ANCE WILL BE ACHIEVED I f The District is currently in full compliance. STATEMENT OF VIOLATION In part, 10CFR50.59(a)(1) states that a license may make a change in the facility, as described in the safety analysis report, without prior Commission approval, unless the proposed change involves an unreviewed safety question. In part, 10CFR50.59(a)(2) states that a proposed change shall be deemed to involve an unreviewed safety question if the l consequences of an accident evd.uated in the safety analysis report may be increased. In part, 10CFR50.59(b)(1) states that the licensee shall maintain a written safety evaluation which provides the basis for the determination that the change does not .= involve an unreviewed safety questior l s Contrary to the above, on October 30, 1992, the licensee approved proposed Design Change 90-226 modification to a drywell ventilation radiation monitor. The change, which was ) not placed in service, would have replaced the old monitor with a new monitor.that did 1 not meet the design basis accident pressure rating, and, therefore, would have increased ,I the consequences of the design basis accident and, therefore, constituted an unreviewed safety question. Specifically in the event of a design basis accident, the new radiation monitor would have failed and, 1. Breached primary containment; q 2. Subjected personnel inside secondary containment to high radiation; 3. Delayed reentry to secondary containment; and j 4. Increased the likelihood of radiation exposure to the general public. j This is a Severity Level IV Violation (298/9318-02) (Supplement I ).. REASON FOR THE VIOLATION During procurement activities for the new RMV-RM-4, a replacement unit which was qualified to containment design basis pressure requirements was not located. As a result, the most suitable available instrument for this application was procured, and the
o .. UI'S. Nuclear Regulatory Commission July 9, 1993 Page 5 ensuing design activities were centered around its installation. An incorrect determination was made while addressing primary containment integrity that the new monitor could be satisfactorily installed without automatic containment isolation valves and still meet design requirements. The acceptability of the new Rad Monitor as well as the penetration design were considered when generating the design change package for the new monitor. The acceptability of the penetration design was considered using the Updated Safe.ty Analysis-Report (USAR). USAR Chapter V, Section 2.3.5.2 states: " Instrument lines which connect to the primary containment free space or to control systems or devices inside the containment are provided with a stainless steel manual shutoff valve directly outside the containment. Included in this category are pneumatic lines for valves, dampers and other such devices; lines which measure containment pressure; and sampling lines for drywell atmospheric sampling." The connection of the old RMV-RM-4 instrument to primary containment was in accordance with the above USAR statement. However, replacing the old instrument with the new instrument which was not qualified to withstand design basis accident pressures introduced a new potential equipment malfunction. The scope of design activities became too highly focused on efforts to ensure that the possible consequences of the ' design remained within acceptable limits, so that not enough consideration was given to implementing means (i.e. automatic isolation valves) to prevent the consequences from occurring. It should also be noted-that the breach of primary containment may not have been clearly identified in the design change package and hence understood by all reviewers. The reviewers may have mistakenly believed that only the instrument function would not be operabis above 3 psig instead of a small breach of primary containment sbove 3 psig. ' Additionally, District memos QAC93168 and QiC93169 previously identified the ' concern relative to Appendix J concerns and appropriate changes to DC 90-226 were already..being considered at the time of the NRC inspection. CORRECTIVE STEPS THAT HAVE BEEN TAKEN AND RESULTS ACHIEVED i The' corrective steps that have been taken include: 1) .The generation of DC 90-226 Amendment 2 which installed testable primary containment' isolation valve islands on both the sample and return lines for RMV-RM. 4. These valves have the following design features: a) Isolation valve islands are capable of meeting Type A and C Leak Rate Tests required per the Appendix J Program. b) Will isolate upon a " Group 2" isolation signal (Low reactor water level or l high drywell pressure). f . 0- ,n -n ,-N --,e
4 U,'S. Nuclear Regul'atory Commission July 9, 1993 . Page 6 c) Reg, Guide 1.97 Control Room position indication, d) Control switches located in the Control Room. Isolation signal cannot be overridden. e) Valves will not automatically open following isolation signal reset, f) Powered by Div. I and Div. II Class 1E circuits, g) Fail closed upon loss of power, loss of air. The modifications made per DC 90-226 Amendment 2 address all concerns relative to "a breach in primary containment" and Appendix J testing of primary containment integrity. These modifications also preclude: 1) subjecting personnel inside secondary containment to high radiation levels as they exited; 2) delaying reentry into secondary containment as a result of radioaceb a contamination;
- and,
- 3) increasing the likelihood of radiation exposure to the general public.
2) The generation of Nonconformance Report (NCR) 93-107 which identifies the programmatic deficiencies associated with the Design Change process: a) The less than adequate design relative to Appendix J criteria; and, b) The less than adequate 10CFR50.59 analysis. 3) The responsible individual has received counseling on the importance of including discussions of the entire scope of the Design Change in written safety evaluations and 10CFR5Q.59 analysis. CORRECTIVE STEPS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIONS Corrective steps that will be taken to avoid further violations have been identified as part of the cort ~ective actions associated with the less than adequate design relative to Appendix J criteria identified by NCR 93-107. These corrective actions include the following: Corrective Action 1: Revise Engineering Procedure 3.4.6 to include an ' Appendix J j Checklist. This would ensure that the appropriate personnel review all Design Changes relative to Appendix J criteria. I Corrective Action 2: Provide training to appropriate engineering personnel on 10CFR50 l Appendix J criteria and applicability to CNS. Corrective Action 3: Place an Appendix J sign off on the Design Change. cover sheet so that an Appendix J review is performed. Corrective Action 4: Perform a review fbr other CNS _ programs that should have a procedure checklist generated and placed in Engineering Proce-dure 3.4.6. . Corrective steps that will be taken to avoid further violations have been identified'as part of the corrective actions associated with the less than adequate 10CFR50.59 analysis identified by NCR 93-107. These corrective actions include the following: i
J. U. S.. Nuclear Regulatory Commission -July 9, 1993 p Page 7 Corrective Action 5: Provide industry events training for design engineering person- .nel emphasizing the importance of including. discussions of the entire scope of the Design Change within the written safety evaluations and 10CFR50.59 analysia including the shortcomings associated with the 10CFR50 59 analysis' for this Design Change. DATE WHEN FULL J0MPLIANCE WILL BE REACHED Corrective Action 1: Revise Engineering Procedure 3.4.6. Anticipated completion-date, 12/01/93. ~ Corrective Action 2: Train appropriate engineering personnel on 10CFR50.59 Appendix J criteria. Anticipated completion date, 12/01/93. Corrective Action 3: Place an Appendix J sign off on the Design Change cover sheet so that an Appendix J review is performed.. Anticipated completion date, 12/01/93. Corrective Action 4: Perform a review for other CNS programs that should have a. procedure checklist generated and placed in Engineering Proce-dure 3.4.6. Anticipated completion date, 12/01/93. Corrective Action 5: Provide industry events training for design engineering person-l nel emphasizing the importance of including discussions of the entire scope of the Design Change within the ' written safety, evaluations and 10CFR50.59 analysis including the shortcomings associated with the 10CFR50.59 analysis for this Design Change. Anticipated completion date, 12/01/93. i ADDITIONAL CONCERNS The District has also been requested to address the circumstances' associated with not having issued a nonconformance report on the drywell ventilation radiation monitor prior to initiating efforts to redesign this modification. Those circumstances may be . described as follows: The installation phase of the new monitor was in progress when the - Appendix J design concerns were identified by the QA letters. Installation of the modifications to those portions of the design change which were not related to. the - primary containment sample and return lines continued. Because the unit -had never been declared operable and because the Design Change discrepancy was 'noted prior to post modification testirg; it was believed that a nonconformance report was not needed on the drywell ventilation radiation monitor. prior to initiating efforts to redesign the q monitor. Althoughlan NCR was not generated prior to initiating efforts to redesign the modification, it was subsequently determined that this event was indicative of a' possible . generic weakness in the engineering program and that an NCR was required to be generated; 'NCR 93-107 was then initiated approximately two weeks after~the QA letters identifying the concern were received. This NCR was required based upon the following criteria found-in CNS Procedure 0.5.1, 2.8. This step states, "For the purpose of this procedure, an occurrence requiring an NCR is defined as follows: i 1 _j
',.. U.*.S. Nuclear Regulatory Commission July.9, 1993-e Page 8 2.8.12 Programmatic deficiencies with potentially significant generic implica-tions." Since this portion of Procedure 0.5.1 applies to the procedures which control the Design Change process, NCR 93-107 was generated to address: 1) the less than adequate design relative to Appendix J criteria and, 2) the less than adequate 10CFR50.59 analysis. The lessons learned from this situation will be included in.the Industry Events Training which has been addressed in Corrective Action 5 above, t Should you have any questions concerning this matter, please contact me. t Sine ely h C . Horn N el r Power Group Manager r i cc: Regional Administrator USNRC - Region IV NRC Resident Inspector y Cooper Nuclear Station i f e i 1 I I l 1 ,,_,}}