ML20034A932
| ML20034A932 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 04/17/1990 |
| From: | Davis A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20034A931 | List: |
| References | |
| EA-89-254, NUDOCS 9004250079 | |
| Download: ML20034A932 (3) | |
Text
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,4 NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTY Wisconsin Electric Power Company Docket Nos.
50-266 and 50-301 Point Beach Nuclear Power Plant License Nos.
DPR-24 and DPR-27 Units I and 2 EA 89-254 During an inspection conducted from November 7 through January 18, 1990, a violation of NRC requirements was identified.
In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2 Appendix C (1989), the Nuclear Regulatory Commission proposes to impose a civil penalty pursuant to-Section 234 of the Atomic Energy Act of 1954, as amended (Act), 42 U.S.C. 2282, and 10_CFR 2.205. The particular violation and associated civil penalty are set forth below:
10 CFR 50 Appendix B, Criterion XVI, Corrective Action, as implemented by the Wisconsin Electric Power Company's Nuclear Power Department. Quality Assurance Manual, Section 16, requires measures be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, defective material and equipment, and nonconformances be promptly identified and corrected.
Contrary to the above, the following conditions adverse to quality designated by the licensee as being significant or highest priority were not promptly corrected:
1.
On March 15, 1988, the licensee identified that recommended vendor maintenance for the emergency diesel generators had not been incorporated into station procedures and had not been accomplished.
Corrective action was scheduled to be completed by July 1,1989, but had not been accomplished as of December 1989 (Item A-SP-88-02-040).
2.
On March 15, 1988, the licensee identified that certain breakers in the emergency DC power distribution system would not function to protect the emergency DC power distribution system integrity under certain fault conditions.
Corrective action was scheduled to be completed by January 1, 1989, but was not initiated until November 7, 1989.(Item A-SP-88-02-016).
3.
On November 11, 1987, the licensee identified that safety injection lockout wiring problems existed in various essential 480v AC switchgear.
Corrective action was scheduled to be completed by October 1, 1989, but had not been accomplished as of December'1989.(Item N-88-008).
4.
On November 4,1988, the licensee identified that the offsite dose calculation manual allowed high alarm setpoints that could result in exceeding the Radiological Effluent Technical Specification lodine-131 discharge limits.
Corrective action plans were scheduled to be identified by June 1, 1989, but analysis of the deficiencies and identification of corrective actions had not been accomplished as of December 1989 (Item N-88-170).
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On September 15, 1989, the licensee identified that the layout of 125v DC power distribution and related procedural practices created the potential that a loss of Division 1 or 2 essential DC power would result in failure of either emergency diesel generator to start on dual reactor trips.
Corrective action plans were scheduled to be identified by October 20, 1989, but analysis of the deficiencies and identification of corrective action i
had not been accomplished as-of December 1989 (Item N-89-223).
This is a Severity Level III violation (Supplement 1).
Civil Penalty - $87,500.
Pursuant to the provisions of 10 CFR 2.201, Wisconsin Electric Power Company (Licensee) is hereby required to submit a written statement or explanation to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission within 30 days of the date of this Notice.
This reply should be clearly marked as a
" Reply to a-Notice of Violation" and should include for each alleged violation:-
(1) admission or denial of the alleged violation; (2) the reasons for the violation if admitted; (3) the corrective steps that have been;taken'and the results achieved; (4) the corrective steps that will be taken to avoid further violations; and (5) the date when full compliance will be achieved.
If an adequate reply is not received within the time specified in this Notice, an order may be issued to show cause why the license should not_be modified, suspended, or revoked or why such other action as may be proper should not be i
taken. Consideration may be given to extending the response time for good cause shown.
Under the authority of Section 182 of the Act, 42-U.S.C. 2232, this response shall be submitted under oath or affirmation.
Within the same time as provided for the response required above under 10 CFR 2.201, the Licensee may pay the civil penalty by letter addressed to the Director, Office of Enforcement,- U.S. Nuclear Regulatory Commission, with a check, draft, or money order payable to the Treasurer of the United States in the amount of the civil penalty proposed above, or may protest imposition of the civil penalty in whole or in part by a written answer addressed-to the Director Office of Enforcement, U.S. Nuclear Regulatory Commission.
Should i
the licensee fail to answer within the time specified, an order imposing the civil penalty will be issued.
Should the licensee elect to file an answer in accordance with 10 CFR 2.205 protesting the civil penalty, in whole or in part, 3
such answer should be clearly marked as an " Answer to a Notice of Violation" and i
ma :
(1) deny the violation listed in this Notice in whole or in part; (2 demonstrate extenuating circumstances; (3) show error in this Notice; or (4 show other reasons why the penalty-should not be imposed.
In addition to protesting the civil penalty, such answer may request remission or mitigation of the penalty.
In requesting mitigation of the proposed penalty, the factors addressed in Section V.B of 10 CFR, Part 2, Appendix C, should be addressed. Any written answer in accordance with 10 CFR 2.205 should be set forth separately from the statement or explanation in reply pursuant to 10 CFR 2.201, but may incorporate I
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Notice of Violation parts of the 10 CFR 2.201 reply by specific reference (e.g., citing page and paragraph numbers) to avoid repetition.
The attention of the Licensee is directed to the other provisions of 10 CFR 2.205, regarding the procedure for imposing the civil penalty.
Upon failure to pay any civil penalty due which subsequently has been determined in accordance with the applicable provisions of 10 CFR 2.205, this matter may be referred to the Attorney General, and the penalty, unless compromised, remitted, or mitigated, may be collected by civil action pursuant to Section 234c of the Act, 42 U.S.C. 2282c.
The responses' to the Director, Office of Enforcement, noted above (Reply to a Notice of Violation, letter with payment of civil penalty, and Answer 1
to a Notice of Violation) should be addressed to:
Director, Office of Enforcement, U.S. Nuclear Regulatory Commission,. ATTN:
Document Control-Desk, Washington D.C. 20555, with a copy to-the Regional Administrator, Region Ill. U.S. Nuclear Regulatory Commission, 799 Roosevelt Road, Glen Ellyn, Illinois 60137.
FOR THE NUCLEAR REGULATORY COMMISSION d s 4 'c -
A. Bert Davis Regional Administrator Dated af Glen Ellyn, Illinois This /7/xlay of April 1990
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U. S. NUCLEAR REGULATORY COMMISSION REGION III Reports No. 50-266/89032(DRP); 50-301/89032(DRP)
I Docket Nos. 50-266; 50-301 Licenses No. DPR-24; DPR-27 Licensee:
Wisconsin Electric Company 231 West Michigan Milwaukee, WI 53201 Facility Name: Point Beach Unit I and 2 Inspection At: Two Rivers, Wisconsin Dates:
December 1,1989, through January 19,-1990
[
Inspectors:
C. L. Vanderniet J. Gadzala$
Y n
h - f(j Approved By:
R. C. Knop, Chief Reactor Projects Branch 3 Date Inspection Summary Inspection from December 1, 1989, through January 19, 1990 (Reports No. 50-266/89032(DRP); No. 50-301/89032(DRP)
Areas Inspected:
Routine, unhanounced inspection by resident inspectors of outstanding items; operational safety; radiological controls; maintenance and surveillance; emergency preparedness; security; engineering and technical support; and safety assessment / quality verification.
Results:
During this inspection period, both units operated at full-power with only occasional load following power reductions and a brief turbine runback on Unit 1.
Issues addressed in this-inspection report include:
Cold weather operations,- (Paragraph 3.e); Injury of two maintenance workers (Paragraph 3.g); Safeguards system degradation (Paragraph 7.a); Fitness for duty training (Paragraph 7.b);-Quality Assurance program implementation (Paragraph 9.a); Site and corporate culture training (Paragraph'9.e); and Corporate management reorganization (Paragraph 9.f).
New issues that remain unresolved include: Potentially inadequate boric acid storage tank. levels (Paragraph 3.f);. and.Information Notice 88-55 (Paragraph.9.d).
One issue-is a follow up to a proposed violation for a failure to take prompt corrective actions detailed in Inspection Reports No. 266/89033; No. 301/89033 (Paragraph 9.a.3).
The-team building training program initiative is viewed as a strength in the licensee's self-improvement efforts.
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DETAILS j
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1.
Persons Contacted (30703) (30702) t
- J. J. Zach, Plant Manager T. J. Koehler, General Superintendent, Maintenance
- G. J. Maxfield, General Superintendent, Operations J. C. Reisenbuechler, Superintendent, Operations l
W. J. Herrman, Superintendent, Maintenance
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N. L. Hoefert, Superintendent, Instrument & Controls R. J. Bruno, Superintendent Technical. Services T. L. Fredrichs, Superintendent, Chemistry J. J. Bevelacqua, Superintendent, Health Physics
- D. F. Johnson, Superintendent, Health Physics R. C. Zyduck, Superintendent, Training
- J. E. Knorr, Regulatory Engineer
- D. R. Stevens, Nuclear Specialist
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F. A. Flentje, Administrative Specialist Other licensee employees were also contacted including members of the technical and engineering staffs, and reactor and auxiliary operators.
- Denotes the personnel attending the management exit interview for summation of preliminary findings.
2.
Licensee Action on Previous Inspection Findings (92702) (92701) j a.
(Closed) Violation (266/88009-02; 301/88009-02):
Failure to Follow Equipment Isolation Procedures.
The majority of the corrective action for this violation was completed earlier and is discussed in Inspection-Reports No. 50-266/89030; No. 50-301/89030.- As the final part of the corrective action, the licensee committed to change the Technical Specifications (TS) wherein the operability requirements of the containment purge supply and -ventilation system would be clarified i
.to be in line with the Westinghouse Standardized TS.
This was the system involved in the improper equipment isolation cited in the i
violation.
The TS change request was approved June 9,1989, as amendments.
9 122 and 125 for Units 1 and 2 respectively.
This item is closed.
f b.
-(Closed) Violation (266/89016-01; 301/89015-01): Failure to-Test l
Station Batteries.
The licensee did not follow the manufacturer's recommendation to i
test the station batteries prior to them exceeding their 20 year-lifetime in the' Fall of 1988.
Though the licensee considered the recorrmendation advisory, this test is required by Technical Specifications.
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Wisconsin Electric replied to the Notice of Violation in letters l
dated August 21, 1989, and September 26, 1989, outlining appropriate i
corrective measures.
A revisid battery testing program has been submitted to the NRC and is currently being reviewed by NRR.
Two of the station batteries (DOS & D06) have since been replaced and the I
other two batteries (DIOS & D106) were tested satisfactory during the fall 1989 outage.- The inspector _ observed the test and had no t
concerns.
This item is closed.
(0 pen) Unresolved Item (266/89020-02; 301/89019-02):
Multiple' c.
1 Failure of Level Detectors; Operation with Cross Connected Accumulators.
On June 12, 1989, three of four level detectors for the two safety injection accumulators on Unit 2 were found to have failed.
All of the failed detectors have since.been repaired and returned to service.
This event and appropriate corrective actions are discussed in detail in LER 301/89-003 and Inspection Reports No. 266/89024; No. 301/89023.
Supplemental LER 301/89-003-01 was issued to address details about the A accumulator level detector, which was one of the three detectors that failed but was not l
mentioned in the original LER.
l The Off Site Review Committee. raised a concern regarding the safety implications of operation with the accumulators cross connected if a loss of. coolant accident (LOCA) were to occur.
The licensee is addressing this concern.
This item remains open pending final determination of the need for a Limiting Condition for Operation in Technical Specifications for operating with cross connected accumulators.
d.
(Closed) Unresolved Item (266/89021-01; 301/89020-01):
Inadequate Battery Load Studies.
Several loads (125 VDC supply to switch' gear sections H01, H02, and H03) were added to batteries 0105 and D106 without perform'ag a load study or a safety evaluation.
Discussions with the licensee and an evaluation of the added loads, indicate that the loads are very small relative to the capacity of the batteries.
The inspector reviewed calculation N-88-026 which determined that most of the additional
!l-loads are in the milliamp range and.the total loads are within the l
design capacity of the batteries.
This calculation was also noted i
to have been reviewed and approved in accordance with the licensee's i
Quality Assurance (QA) procedures.
No additional concerns were identified and this item is closed.
e.
(Closed) Unresolved Item (266/89021-04; 301/89020-04): Operation of DC Electrical Components Below 105 Volts.
The battery configuration was modified before the plant became operational whereby one of the 60 cells was removed from service in 4
each of the ofiginal station batteries (DOS & D06).
Consequently, 3
t
_- o the minimum design voltage was lowered from 105 volts to 103.25 volts although no evaluation was made of the effects of this lower voltage on DC electrical loads.
Section 8.2.3 of the Final Safety Analysis Report (FSAR) lists i
103.25 VDC as the minimum battery terminal voltage for 005 and 006.
Batteries DIOS and D106 have a minimum terminal voltage of 105 VDC.
The licensee is planning an evaluation of the loads on batteries 005 and 006 for operation down to 103.25 VDC.
In the interim, the licensee has administrative 1y changed the acceptability criteria _ for batteries DOS and 006 back to 105 VDC and a FSAR change is. pending to reflect this criteria.
The most recent capacity tests of -
batteries DOS and 006 used 105 VDC as the acceptance criteria and were satisfactory.
The inspector discussed this issue with the licensee and had no further concerns.
This item is closed.
f.
(Closed) Violation (266/89021-05; 301/89020-05):
Failure to Perform Required QC_ Inspections, i
Quality Control (QC) inspections were not performed following-l replacement of wooden end rails on the battery racks and following replacement of certain molded case circuit breakers even though the 1
Maintenance Work Requests (MWRs) for both actions specified that this was QA scope work, i
The licensee believes that QC inspections may have been-performed on both of the items above, but the manner in which signatures were l
required on the documentation prevents adequate verification.
This in itself is an acknowledged deficiency.
The~. licensee committed to revise its MWR procedure (PBNP 3.1.3) by October 31, 1989,.to correct this weakness, thereby making compliance' easier to audit.
The MWR procedure was revised December 1,!1989. The inspector discussed this issue with the licensee and reviewed-the revised procedure to ascertain that changes were made as specified in the commitment letter.
A step in the procedure allowing a single i
signature to serve as both acknowledgement of the first line supervisor's review and completion of the QC inspection was deleted. A separate signature now documents completion of a QC inspection. This item is closed, g.
(Closed) Violation (266/89015-01; 301/89014-01):
Failure to Follow Red Tagger Procedure.
Operations Standing Order 4.12.2, " Qualified Red Taggers," lists all plant personnel who are qualified to hang authorized red tags and is referenced in the. equipment isolation procedure.
A copy of i
this list is attached to the front of the active red tag log and updated with pen and ink changes.
This method of control is not in accordance with plant procedures but has occurred repeatedly.
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- The-licensee initially committed to a multi step course of action that was not fully implemented as discussed in Inspection Reports No. 50-266/89027; No. 50-301/89026.
After discussions between the licensee and NRC regarding weaknesses in the use and implementation of plant procedures in general, Wisconsin Electric reevaluated their 4
procedure controls and issued a detailed and comprehensive plan of1 corrective action on December 21 to address these deficiencies.
Included in this plan were actions addressing this specific issue.
This plan committed the licensee to revising PBNP L13, " Equipment Isolation Procedure," to eliminate the reference to the standing order, and PBNP 2.1.1, " Classification, Review & Approval-of Procedures," to permit local control of the red taggers list.
These changes, which were made December 29, 1989, resolve the deficiencies
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that lead to the original citation.--The inspector reviewed the procedure changes and was satisfied.
This item is closed.
h.
(Closed) Open Item (266/88005-01;' 301/88005-01):
Licensee to Investigate a Filtration. Method for Analyzing Primary Coolant.
i Radionuclide sample analysis performed by the licensee and the NRC mobile laboratory.on site showed agreement in 83 of 88 comparisons.
A cause for some of the disagreements was attributed to the licensee using a low percent abundance acceptance criteria in identifying
~1 nuclides.
The licensee agreed to investigate a method involving filtering a sample through a 0.45 micron filter and cation filters.
a The licensee has since issued procedure CAMP-408,. " Radiochemical.
Analytical Procedure," which performs this function.
Discussions with the licensee indicate that gamma spectroscopic' multichannel analysis of the 0.45 micron filter, cation filters and filtrate, yield spectra with less measurement uncertainty from nuclide t
interferences than counting the sample without separation. :The inspector reviewed the procedure and had no concerns. This item-l l
is closed.
i.
(Closed) Open Item (266/88005-02; 301/88005-02):
Licensee to Evaluate Use of a Larger Gas Sample Container.
Radionuclide sample analysis performed by the licensee and-the NRC mobile laboratory on site showed agreement in 83 of 88 comparisons.
The licensee did not identify Kr-85 in an off gas sample because the quantity present in the sample approached the lower limit-of detection for the plant's equipment and procedures. Although the detection limit is considered adequate, the licensee agreed to evaluate the use of a larger gas sample container.
The licensee has since revised procedure CAMP-102, " Chemistry Administrative Procedures, Gas Decay Tank Sampling and Discharge Guidelines," to specify the use of a.one liter poly bottle for use as a gas sample container.
A five cc glass vial had been previously l
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1 used. The licensee subsequently determined that the larger peometry 1
is significantly more sensitive and provides better counting a
statistics.
The inspector reviewed the procedure and had no concerns.
This item is closed.
j.
(Closed) Open Item (266/88005-03; 301/88005-03):
Licensee to 1
Determine lodine Penetration Profile for Charcoal Absorber Sample.
1 The licensee uses different charcoal geometry in its absorber filters than those that the NRC uses for analysis purposes.
The filters used-by the licensee for performing multichannel analyzer calibrations are spiked evenly from the filter face down to a depth of 50%.
NRC filter standards are spiked only on the filter face.
To verify that actual charcoal filter samples at the plant have an iodine penetration profile consistent with their calibration standards, i
the plant agreed to analyze several filter samples.
This analysis has since been completed.
Each sample was cut open and.four layers, each 25% of the total ~ filter volume, were removed and analyzed.
Analysis results are documented-in Wisconsin Electric-memo PBNP 88-131 dated July 5,1988.
The analysis showed.significant iodine activity penetrating to 50% depth and beyond, thereby verifying that the licensee's absorber calibration is representative of actual samples and can thus produce accurate results. This item is closed.
No violations or deviations were identified.
3.
Plant Operations (71707) (71714) (93702) a.
ControlRoomObservation(71707)
The inspector observed control room operations, reviewed applicable logs and conducted discussions with control room operators during the inspectior period.
During these discussions and observations, the inspectors hscertained that the operators were alert, cognizant of current plant conditions, attentive to changes in those conditions and took prompt action when appropriate.
The inspectors noted that-a high degree of professionalism attended all facets of control room operation and that both unit control boards were generally in a
" black board" condition (no non-testing annunciators-in alarm condition).
Several shift turnovers were also observed and appeared to be handled in a thorough manner.
The inspectors performed walkdowns of the control boards to verify the operability of selected emergency systems, reviewed tagout records, and verified proper return to service of affected components.
b.
Facility Tours (71707)
Tours of the Turbine Building, Primary Auxiliary Building, and Service Water Building were conducted to observe plant equipment conditions, including plant housekeeping / cleanliness conditions, 6
a i
status of fire protection equipment,' fluid leaks, and excessive a
vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenance.
^
During facility tours, _ inspectors noticed very few signs of leakage and that equipment appears to be in good operating condition.
f Overall, plant cleanliness has remained good.
I c.
Unit 1 Operational Status (93702)
On December 10, the licensee notified the NRC via the Emergency Notification System (ENS) that.the Unit I turbine experienced a runback from 100% to'83% power due to the No. 2 turbine governor-valve failing shut._ A failed circuitry card was determined to be the cause of this event.
The governor valve motion resulted in an automatic inward rod motion that was just rapid enough to actuate the 2.5 % per second rod drop alarm in three of'four power range channel drawers.- The rod motion also caused delta flux to go.
outside the TS limits.
Delta flux wts restored to within~the allowable band in 14 minutes by borating and raising power.
The r
faulty circuitry card was replaced and the system restored to normal.
The inspector reviewed the. licensee's corrective actions and had no further concerns.
The unit operated at full power during the remainder of the period with only requested load, following power reductions.
d.
Unit 2 Operational Status (93702)
The unit continued to operate at full power during this period with only requested load following power reductions.
e.
Cold Weather Operations (71714)
The site experienced extremely cold weather during the week of December 18 with air temperatures falling as low as -25_ degrees Fahrenheit.
The inspector reviewed the licensee's cold weather preparations and had no concerns.
Few problems resulted~from the abnormal temperatures with the exception of the fire system jockey pump freezing.
The electric fire pump started automatically to maintain header pressure while.the jockey pump was returned to service.
f.
Potentially Inadequate Boric Acid Storage Tank (BAST) Levels (93702)
On January 5, the licensee notified the NRC via the ENS that the minimum allowed BAST levels, as specified in Technical Specifications (TS), may be erroneous.
This finding affects safe shutdown and1 accident mitigation design considerations for a restart accident following a LOCA or steam line break.
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1 During a review of procedure OP-5A, " Reactor Coolant Volume Control,"
i a licensee engineer noticed that for.th' iteam line break accident' e
analysis, approximately 900 gallons of boric acid solution are i
needed to mitigate the accident.
This 900 gallons must be in excess of the 16% automatic shut off point of the tank (plus 1% margin) of
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l 1550 gallons. This total corresponds to a minimum level of 2450 gallons.
The plant TS specify a minimum level of only 2000 gallons.- Although the licensee has' been maintaining BAST levels above 2450 gallons as specified in their plant procedures, i
the TS would allow a level below this minimum design value.
The calculations for the LOCA analysis were'also reverified and several mathematical errors found.- The licensee determined that according to the design bases, 1740: gallons of boric acid solution are needed to mitigate the accident.
Adding.this to the 1550: gallon.
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shut off point yields 3290 gallons.
When the licensee discovered this, they took prompt action to' establish new minimum levels for the BASTS.
The new minimum level was chosen by taking 3290 gallons-1 (60% tank level) and adding a 5% margin to yield a new minimum tank level of 65%.
The plant has three 5500 gallon BASTS with 41.5 gallons per percent 1
level in the indicating range.
One tank per reactor is required by TS and the third tank is a common backup.
When this problem was.
discovered, tank A (Unit 1) was at 80% and tank L (Unit 2) was at' 59%.
The licensee transferred boric acid solution from tank Bi (common) to tank C to raise its level to 65%.
The inspector observed this process, Wisconsin Electric is discussing this issue with Westinghouse to reevaluate the design bases and the calculations involved in the analysis. - This item remains unresolved pending a final determination by the licensee and subsequent review by the NRC (266/89032-01; 301/89032-01).
1 Accidental Steam Burning of Two Maintenance Workers (93702J g.
a On December 11, two maintenance workers received second degree burns over large parts of their bodies when' they were sprayed with steam from a heater drain pump discharge check valve they were disassembling, j
Although the valve was properly isolated, steam pressure was not t
fully vented from the isolated section of piping before starting work.
l The two workers were taken to a local hospital for treatment and are expected to remain hospitalized approximately six to eight weeks.
j Since the work was being performed on the secondary side of the' plant, no radioactive systems were involved and therefore no ~ contamination occurred.
An Occupational Safety and Health Administration (OSHA) inspector reviewed'the accident and is to provide the licensee with his findings.
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These reviews and observations were conducted to verify-that facil_ity operations were conducted safely and in conformance with requirements l
established under Technical Specifications, federal regulations, and i
administrative procedures.
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4 No violations or deviations were identified; H
i 4.
Radiological Controls (71707)
The inspectors routinely observed the licensee's radiological controls and practices during normal plant tours and the inspection of work.
activities.
Inspection in this area includes direct observation of the use of Radiation Work Permits (RWPs); normal' work practices inside-contaminated barriers; maintenance of radiological barriers and signs;
.t and health physics (HP) activities regarding monitoring, sampling, and surveying. The inspector also observed portions of the radioactive waste system controls associated with'radwaste processing.
From a radiological standpoint the plant is in good condition, allowing access to moct sections of the facility.
During tours of the facility, the inspectors noted that barriers and signs also were in good condition.
When minor discrepanc'les were identified,- the HP staff quickly responded to correct any problems.
All' activities were conducted in a satisfactory manner during this inspection period.
No violations or deviations were identified.
5.
Maintenance / Surveillance Observation (62703) (61726) a.
Maintenance (62703)
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-i 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 is assigned to safety-related equipment maintenance which may affect system performance.
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t Portions of the following maintenance activities were-
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observed / reviewed:
IT-515B'(Revision 2), " Leakage Reduction-and Preventive Maintenance Program Test of Safety Injection Test Line and Spray Additive Eductor Line (Refueling)"
Though normally done during an outage, the licensee chose to delay this work until after Unit 2 startup.
The procedure was changed to conduct testing at power vice shutdown.
The inspector considered the stated reason for the temporary-changes, as specified on the change form cover sheet, to be overly broad.- It was not evident from this cover sheet that' these changes had been adequately reviewed, although each changed step in the actual procedure was-initialed.
The current procedure governing temporary changes (PBNP 2.1.1) is-ambiguous in its requirements.
This is a_ weakness acknowledged by the licensee, who has committed to form a task group to study the problem.
)
Correction of excessive packing leakage on Auxiliary feed Pump l
P38A discharge control valve (AF 4012).
This work was performed under Maintenance Work Request-894815 using procedure MI 5.2 (Revision 3), " Air Diaphram-Operated Valve Maintenance."
b.
Surveillance (61726) i The inspector observed surveillance testing and verified that testing 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 witnessed and reviewed the following test activities:
ICP 2.5 (Revision 4)
I & C Surveillance Test, Safeguards System Logic IT-05 (Revision 17)
In Service Testing of Containment-Spray Pumps, Eductor Supply Check i
Valves'and Sodium Hydroxide Addition Valves IT-10 (Revision 17)
In Service Testing of Electrically Driven Auxiliary Feed Pump Monthly i
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- i The technician performing the test was using an unmarked informational copy of the test procedure to guide his actions.
As a result, he commenced a step in the procedure that was deleted in the master copy being used by the control operator directing the test.
Since the technician'and the control operator maintained good communications, the control operator was able to quickly. correct the technician and direct him to reopen the valves he had isolated in error.
The inspector discussed this issue with the licensee who indicated that this-weakness would be considered by the task group being formed to review procedure control deficiencies..
a No other discrepancies were noted during the observance of any'of the above tests.
No violations or deviations were identified.
6.
Emergency Preparedness (71707)
An inspection of emergency preparedness activities'was performed to assess the licensee's implementation of the site emergency plan and implementing-
't procedures.
The inspection included monthly review and tour of emergency facilities and equipment, discussions with licensee staff, and a review of selected procedures.
All activities were conducted in a satisfactory manner during this inspection period, t -
No violations or deviations were identified.
t 7.
Security (71707) t The inspectors, by direct observation and interview, verified that portions of the physical security plan were being implemented in-accordance with the station security plan. -The inspectors also continued' to monitor compensatory measures that have been enacted by the licensee.
a.
Safeguards System Degradation (71707)
On December 12, the licensee notified the NRC via the Emergency Notification System (ENS) that a degradation in the detection, portion of the safeguards-system occurred and was not properly compensated.
Corrective measures were taken by the licensee and the issue has'been referred to NRC security personnel for evaluation.
b.
Fitness For Duty Training'(71707)
The licensee implemented their Fitness for Duty program at the. start of this year.
The inspector reviewed various aspects of this program and observed selected portions of Fitness for Duty training.
No concerns were identified.
This topic will be covered in additional detail in a future report.
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All other activities were conducted in a satisfactory manner during this inspection period.
No violations or deviations were identified.
8.
Engineering and Technical Support (71707)
The inspector evaluated licensee engineering and technical support activities to determine their. involvement and support of facility-operations.. This was accomplished during the course of routine evaluation of facility events and concerns through direct observation of activities and discussions with engineering personnel.
Extensive inspection effort was directed towards the evaluation of an' original design deficiency in the DC. electrical distribution system..
Details are documented in special Inspection Reports No. 266/89033; No. 301/89033.
All activities were conducted in a. satisfactory manner during this inspection period.
I No violations or deviations were identified.-
i 9.
Safety Assessment / Quality Verification (35502) (92701) (90712) (92700) l The licensee's Quality Assurance programs were inspected to assess'the
[
implementation and effectiveness of-programs associated with management j!
control, verification, and oversite activities.
Special; consideration was given to issues which may be indicative of overall management involvement in quality matters such as self improvement programs, response to regulatory and industry initiatives, the frequency of-management plant tours and control room observations, and management l
personne1's attendance-at technical and planning / scheduling meetings.
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l a.
Quality Assurance (QA) Program Implementation (35502)
I A review was conducted of recent inspection reports, SALP reports,-
i licensee event reports, licensee corrective actions, and the Monthly Open Items Status Report (M0lSR).
The results of~this review were used to evaluate the licensee's-QA program implementation.
The inspector additionally interviewed QA-personnel to determine the current structure, focus, and operating practices of the on site and corporate QA organizations.
The scope of the inspection effort was expanded to include further evaluation of the proposed violation identified in Inspection Reports No. 266/89033; No. 301/89033. The findings confirmed several suspected weaknesses regarding the effectiveness of the licensee's QA program regarding follow up of QA Audit Finding Reports (AFR).
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1, Site Quality Assurance Group The licensee established a five inspector Site Quality Assurance (SQA) group in April 1989.
Locating the group on site is seen-as a positive step in improving the effectiveness of the licensee's QA organization.
This group evaluates procurement of_QA scope replacement and repair parts; conducts surveillance reviews of Maintenance Work Requests (MWR), temporary modifications, tagouts, and other licensee programs; reviews j
contractor work methods; performs modification request reviews and field verifications; reviews of completed MWRs; evaluates
-Non-Conformance Reports (NCR) processing and review; and maintains QA scope drcwings, 2.
Quality Assurance Vertical Slice Audits One of the programs being implemented by the Nuclear Quality 1
Assurance Division (NQAD) are vertical slice audits of the facilities safety systems'.
These audits are similar in scope to NRC Safety _ System Functional Inspections (SSFIs).
To date three of these audits have been performed and a fourth is scheduled for the Spring of 1990.
These audits were helpful in_the identification of several safety significant issues-which have ultimately resulted in improved reliability'of the systems inspected.
The continued use'of these types.of audits are viewed as proactive and serve to enhance performance and increase system reliability.
One weakness with. regard to these audits is the failure of the licensee.to adequately respond to-the open items that are identified in a timely manner.
3.
Follow up on Inspection Report 266/89033 and 301/89033-jl Inspection Reports No. 266/89033; No. 301/89033 identified a l
proposed. violation (266/89033-02; 301/89033-02) regarding the 1
I.
licensee's apparent failure to take prompt' corrective ~ actions-for an Audit Finding Report (AFR) item. -1his finding is considered significant in light of'other previous examples of inadequate and untimely corrective actions taken in regards to l
l Licensee Event Reports, cited violations, and a commission order.
Coupled with the above previous findings, weaknesses j
l were identified during the review of the QA program -
implementation that prompted further evaluation of the:licensec's program for dealing with open items and the completion of their l-corrective actions.
This was done in an attempt to determine the root cause to the licensee's apparent difficulties with-accomplishing corrective actions in a timely manner.
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l Open item Trackino l
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NQAD issues a Monthly Open Item Status Report (M0!$R) that i
lists and categorizes all open items in the deficiency log database.
During a review of the M31SR dated December 8, 1989, the inspector counted approximately 430 open items.
Of the items counted over 60 items were categorized as having no I
l response and approximately 50 other items were identified as having corrective actions overdue.
This indicates that 25% of the items in the MOISR are delinquent and many of the overdue corrective action items have been in the database for over one year.
This raises a concern regarding the licensee's ability to obtain timely closure of open items.
Prioritization of open items also appears to be a weakness in the licensee's present open item tracking system.
By the inspector's count, the licensee has identified and titled 14 open items on the MOISR as being "Significant Open Items," of those items six are categorized as having no response and five 1
were identified as having corrective actions overdue.
This represents a delinquency rate of over 75% for self-identified "Significant Open Items." This problem is further exacerbated by a memo from NQAD by direction from the licensee's senior i
management, dated May 2,1989, which identified 20 open items as having the highest priority.
Of these 20 items, 16 were still listed in the MOISR although none of them are identified as "Significant Open Items." In addition, 10 of the 16 or 62%
of these items were listed as being delinquent.
I Based upon the percentage of items, significant or otherwise, that are delinquent and the apparent inconsistency in the l
prioritization of items it appears that the present method used to track and control the completion of corrective actions is not working in a timely manner to resolve issues.
This is of l
concern to the NRC because the possibility exists that items l
with possible safety significance are not being promptly addressed.
Review of Quality Assurance Actions for Delinquent Responses e
Prompted by concern over the apparent excessive number of delinquent open items, the inspector reviewed the mechanisms i
used by the licensee to obtain responses.
During a review of l
several Quality Assurance Procedure Manual procedures, covering the various forms and reports used to track open items to closure, the inspector noted that the procedures referred to Quality Assurance Instruction (QAI) PB-7.1, " Follow Up Of Deficiency Reports (Internal)" for the handling of overdue responses.
QAl-PB-7.1 prescribes the method to be utilized l
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e by NQAD to obtain timely responses and corrective action to internal deficiencies documented on Audit finding Reports (AFRs) and other tracking forms.
This procedure was initially issued in 1983 and contained the process for the escalation of open items with delinquent responses.
The procedure has remained essentially the same except for the addition of overdue corrective responses in 1987.
The current QAI PB-7.1, Revision 5 requires that if a response has not been obtained by the specified due date, the head of the organization responsible for that item shall be informed that the response is overdue.
This notification is to be documented on a Communication Memorandum (QAl-1, forms 1.1 and 1.2) and is the first step in the escalation process.
The instruction further states that if a response has not been obtained after an additional 15 days, a letter to the responsible department head shall be issued. Another letter shall be issued 15 days later to the appropriate Vice President.
Finally, a letter shall be sent to the Chairman of the Board 15 days later.
The inspector reviewed several delinquent open items and interviewed several Quality Assurance personnel to evaluate the use of this procedure.
During this review it was evident that the procedure is used sparingly, and, when it was used it is not followed correctly.
The procedure makes no allowances for the negotiation of response dates, however, through interviews, it was determined that QA personnel often negotiate and extend response deadlines.
Though the extending of specified dates is not in accordance with QAl-PB-7.1 it ooes not appear to be an unreasonable practice. The problem with this practice is that the extensions are not routinely documented as required by this procedure and QAl-5, " Documenting QA Activities and Informal Communications." This is a pervasive problem that is not limited to specific' individuals nor to certain items but common throughout the program.
When QAI-PB-7.1 is initiated it is generally well documented as required, however, it is rarely escalated to the next level as required by the procedure.
This escalation is required to take place 15 days af ter initiation and no provision is made for alternative actions.
During discussions with QA personnel several individuals stated that the licensee was reluctant to proceed with further. escalation because it would have a negative impact on working relations with the groups, responsible for the response or corrective actions.
- Instead, the licensee most often negotiated a new response date.
The initial step in the escalation process appears to be effective, usually resulting in the required' response within 30 days.
On the relatively rare occasions when the second escalation step is initiated, it is completed as specified in QAI-PB-7.1.
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t The review of QAl-PB-7.1 identified several pervasive problems i
with regards to the implementation of the procedure which provided multiple examples of a failure to follow procedure, i
The review also identified that QAl-PB-7.1 is very restrictive and does not differentiate between items identified as significant and those that are not.
It also does not allow for any alternative measures, i
Through interviews it was determined that there was a reluctance f
on the part of the licensee to fully implement the escalation process based on possible negative reactions.
The interviews also showed that nonmanagement QA personnel often are required to negotiate with other management personnel with regard to t
deadlines.
These findings indicate a breakdown in the licensee j
ability to gain timely responses and corrective actions that could have a significant impact on the effectiveness of the licensee's self-inspection program.
i The weaknesses identified in the licensee's QA program are significant contributing factors to a failure to take prompt corrective actions but, they do not appear to be the root i
cause of the problem.
Regardless of the identified QA weaknesses the licensee does have a system for the tracking of open items. The licensee also follows up on overdue responses even though not in strict accordance with approved procedures.
Therefore, the main problem appears to be a failure on the part of responsible organizations to manage their assigned open items and take the necessary corrective actions in a timely manner.
To determine the possible root cause for this problem the inspector interviewed several of the licensee's management personnel.
These interviews revealed one common concern expressed by the managers.
That concern is the lack of, or, inadequate use of the necessary resources to allow for prompt evaluation'of open items and completion of corrective actions.
This appears to be l
due, in part, to resources that are often reallocated due to changing work loads.
Specifically, resources that are shifted j
to accommodate outage work items and other " hot" issues at the expense of routine work.
It was further stated that because the licensee is dedicated to performing quality work, as is apparent by their operating history, the shif ting of resources causes delays in the evaluation and closure of open items.
l This causes delays in the identification of possible safety significant problems associated with some open items like the item identified in proposed violation No. 266/89033-02; No. 301/89033-02.
When resources become available a thorough evaluation appears I
to be performed and if a safety significant issue is identified adequate corrective actions are taken.
However, the delays in evaluating potentially significant open items are viewed by the 1
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i NRC as a weakness to take prompt corrective actions.
Therefore, the finding of this review appears to substantiate the proposed violation No. 266/89033-02; No. 301/89033-02 and will be evaluated as part of that proposed violation.
Review of Selected Open Items The inspector selected several representative items fro,m the MOISR which were either overdue or not responded to:
Item Initial Due Status A-P-87-15-047 (SEC) 12/22/87 5/1/89 CAO
- QP 6-7 is not fully implemented.
A-P-88-10-049 (OPS) 10/19/88 3/1/90 (initial response was 4 months late)
Inadequate response to IE Bulletin 88-04.
A-SP-87-01-004 (TRN) 8/21/87 6/1/89 CAO
- DUKE WE-87-02C First line supervisors have not received QC inspector training.
A-SP-88-02-009 (ADM) 3/15/S8 9/30/88 CAO **
SSFI WE-88-07 Incorrect safety relief valves installed on diesel generator starting air accumulators.
Valves replaced but lot card description proposal incomplete.
A-SP-88-02-021(NSE) 3/15/88 2/28/90 SSFI WE-88-19 Inadequate breaker coordination study.
A-SP-88-02-040 (MTN) 3/15/88 7/1/89 CAO **
SSFI WE-88-38 Nonperformance of vendor required preventive maintenance.
A-SP-88-02-044 (MTN) 3/15/88 3/1/88 CAO **
SSFI WE-88-42 Inadequately documented post maintenance testing.
A-TS-89-02-006(MTN) 5/19/89 10/1/89 CAO RMP-58 Appendix A & B provide inadequate guidance for performing fire barrier penetration visual inspections.
N-88-008 (MTN) 11/11/87 10/1/89 CAO
- Safety Injection lockout wiring problems noted with various 480 VAC switch gear.
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N-88-072 (IIE) 4/29/88 6/1/88 NR
- QAS and PBNP personnel are using non-CHES approved chemicals on corrosion resistant material.
N-88-080 (IIE) 5/13/88 9/1/89 CAO (initial response was 4 months late)
Several Reactor Coolant System welds have not been included in the In Service Inspection plan due to drawing inaccuracies.
N-88-101 (MIN) 7/12/88 7/1/89 CAO (initial response was 3 months late)
Technical Specification 15.4.15 required plate inspections for diesel fire pump batteries cannot be performed because battery cases do not allow visibility.
N-88-144 (NSEAS) 10/5/88 10/1/89 NR
- Red tags made for the P106 deep well pump were inadequate because no drawings exist for its level switch controls.
N-88-170 (NPERS) 11/4/88 6/1/89 NR
- The off site dose calculation manual allows high alarm setpoints for detectors RE-229 and RE-230 that could result in exceeding the RETS I-131 discharge limit.
N-89-029 (NSE) 2/16/89 4/1/89 NR
- Valve operator weights listed in the piping isometric drawings are not all accurate.
N-89-223 (NSE) 9/15/89 10/20/89 NR Present arrangement of the 125 VDC distribution along with procedural practices ~ creates the potential that on l
loss of panel D-12, D-14; G01 or G02 would fail to l
start with dual reactor trips.
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NR - no response CAO - corrective action overdue
- designated by licensee as a significant open item i
- designated as highest priority in a May 2, 1989 licensee memo b.
LicenseeEventReport(LER) Review (90712,1 The inspector reviewed LERs submitted to the NRC to verify that the details were clearly reported, including accuracy of the description and corrective action taken.
The inspector deternined whether l
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further information was required, whether generic implicatiens were i
indicated, and whether the event warranted onsite followup.
The following LERs were reviewed:
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- 301/89-009 Unexpected " Level Low" Reactor Trip signal During l'
Emergency DC Lighting Test
$301/89-003-01 Safety Injection Accumulator Level Detector i
Instrument Failure j
c.
LER Followup (92700)
The LERs denoted by asterisk above were selected for additional followup.
The inspector verified that appropriate corrective action was taken or responsibility was assigned and that continued operation of the facility was conducted in accordance with Technical i
Specifications and did not constitute an unreviewed safety question s defined in 10 CFR 50.59.
Report accuracy, compliance with current reporting requirements and applicability to other site systems and components were also reviewed.
d.
Information Notice Followup (92701)
The inspector verify the effectiveness of the licensee's program for handling Information Notices (IN).
This included the review of selected ins for applicability and the scheduling and performance of r
appropriate corrective actions if necessary.
The following ins were reviewed and appeared to have been suf ficiently evaluated to permit closure:
Information Notice No. 87-59: POTENTIAL RHR PUMP LOSS This IN was superseded by NRC Bulletin 88-04: POTENTIAL SAFETY-RELATED PUMP LOSS.
Further followup of this IN was cancelled by a DRP memorandum dated May 19, 1988.
Based on DRP direction, this IN is considered to be closed (266/88902-IN and 301/88902-IN).
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Infermation Notice No. 88-01: SAFETY INJECTION PIPE FAILURE This IN regarded the possible failure due to thermal stresses, of safety injection piping connections to the Reactor Coolant System.
The IN specifically identified problems-associated with the injection of normal charging flow through a common safety injection tap at facilities that utilize charging pumps i
to provide high head safety injection.
The licensee reviewed this IN and noted that it was not applicable to this plant because charging flow utilizes a separate connection to the RCS l
and that charging pumps are not used for high head safety f
injection at this plant.
Based on the licensee's response this IN is considered to be closed (266/88903-IN and 301/88903-IN).
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l Information Notice No. 88-09: REDUCED RELIABILITY OF STEAM-DRIVLN AUXILIARY FEEDWATER PUMPS CAUSED DY INSTABILITY OF WOODWARD PG-PL TYPE l
GOVERNORS.
i This IN regarded overspeed trips of steam-driven auxiliary feedwater (AFW) pumps due to the performance of quick startup tests from a cold condition.
The licensee's response stated that P8NP procedure 11-290/295, " Inservice Testing"of Auxiliary Feedwater System Check Valves and Flow Indicators, includes quick startups of the steam-driven AFW pumps from a cold condition and that no instabilities or overspeed trips have occurred due to the performance of this test.
Based on the response from the licensee this IN is considered to be closed (266/88904-IN and 301/88904-IN).
f Information Notice No. 88-51: FAILURE OF MAIN STEAM ISOLATION VALVES This IN regarded the failure of main steam isolation valves (MSIVs) to close due to a loss of control air supplied to the valve.
The licensee's re$ponse indicated the the type of MSIVs e
in question are not in use at this facility.
This facility I
utilizes MSIVs which are reverse oriented check valves that use steam flow to assist in closure instead of air.
Based on the response from the licensee, this IN is closed (266/88031-IN and 301/88051-IN).
Information Notice 88-55, " Potential Problems Caused by Single Failure of an Engineered Safety Feature Swing Bus," does not appear to have been reviewed in sufficient detail in light of other licensee findings related to this issue.
Further evaluation of the licensee's program for handling Information Notices is needed before a final assessment can be made.
This issue will remain unresolved pending such additional review (266/89032-03; 301/89032-03).
e.
Site and Corporate Culture Training (71707)
The licensee has recently undertaken a comprehensive cultural adjustment and team building training program.
The purpose of this training, conducted by a management consultant, is to closely examine personnel culture and attitude weaknesses at various levels in both the station and corporate offices.
The program seeks to identify any problems perceived within the organization, get personnel to admit and confront these problems, and then develop and y
implement the solut' ions necessary to overcome them.
The inspector considers this program of great potential benefit to the licensee if the concepts of this training are genuinely embraced by nuclear department management.
This area will be reviewed in a future report as the program matures.
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U.S. NUCLEAR REGULATORY COMMIS$l0N
)
REGION lil j
Reports No. 50-266/89033(DRP); 50-301/89033(DRP)
Docket Nos. 50-266; $0-301 License Nos. DPR-24; DPR-27 f
I Licensee: Wisconsin Electric Company 231 West Michigan l
Milwaukee, WI 53201 Facility Name: Point Beach Unit 1 and 2 Inspection At: Two Rivers, Wisconsin Dates: November 7 through December 22, 1989
. inspectors:
C. L. Vanderniet
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J. Gadzala Approved By:
1.
acki, Ch e
/"#~f0 Re tor Pr jects Section 3A Date Inspection Summary Inspection from November 7 through December 22,1989,(ReportsNo. 50-266/89033(DRP);
No. 50-301/89033(DRP))
Areas Inspected: Special, unannounced inspection by resident inspectors of issues relating to the inadequate original design of the facility's DC electrical distribution system and the licensee's corrective actions.
Results: Two potential violations were identified regarding inadequate design of the DC electrical distribution system and failure to take prompt corrective action once the deficiency was identified.
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,Gyl.4.f St.LQ" i
DETAILS 1.
PersonsContacted(30703)(30702)
- J. J. Zach P T.J.Koehler,lantManager General Superintendent, Maintenance G. J. Maxfield, General Superintendent, Operations J. C. Reisenbuechler, Superintendent. Operations W. J. Herman, Superintendent, haintenance N. L. Hoefert, Superintendent, Instrument & Controls R. J. Bruno, Superintendent Technical Services
- 7. L. Fredrichs, Superintendent, Chemistry J. J. Bevelacqua, Superintendent, Health Physics R. C. Zyduck, Superintendent, Training
- J. E. Knorr, Regulatory Engineer F. A. Flentje, Administrative Specialist Other licencee employees were also contacted including menbers of the technical and engineering staffs and reactor and auxiliary operators.
~' Denotes the" personnel attending the management exit interview for-
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summation of preliminary findings.
2.
DC Electrical Distribution System Oricinal Desijn The Point Beach facility was originally designed with a DC Electrical l
Distribution System that included two 60 cell stat. ion batteries (DOSandD06)connectedtotwoseparatemainDCdistributionbusses (D01 and D02). Each main DC bus was also connected to an independent battery charger (007 and D08) and to s common battery charger (D09).
The present plant configuration (see attachment) has battery DOS and battery charger 007 connected to main'DC bus D01 and battery 006 and battery charger D08 connected to, main DC bus D02. Battery charger 009 can be connected to both main DC busses D01 and D02, however, both of its output breakers are maintained in an open condition and can only be closed through manual breaker manipulations.
Main DC bus 001 provides DC power to DC distribution panels 011 and D12 and main DC bus D02 provides DC power to DC distribution panels D13 and D14 A turbine emergency lube oil pump (IP370 and 2P370) is also connected to each main DC battery bus.
The original facility design utilized themal-only trip breakers on the lines connecting the main DC busses (D01 and D02) to the DC distribution panels (011, D12. 013, and 014); to the batteries (DOS and 006); and to the common battery charger. (D09). The remaining lines to the independent battery chargers (007 and D08) and.to the turbine emergency lobe oil pumps (1P370 and 2P370) were equipMwith breakers having themal and i
magnetic tripping capabilities. Taese conditions are still present at l
the facility.
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Two of the DC distribution panels (011 and D13) have 32 breakers on each panel. Of the 32 breakers on each panel, 30 were equipped with thermal-only trip capability.
The other two breakers on panels D11 and 013 and all the breakers on panels D12 and D14 were equipped with breakers having both thennal and magnetic tripping capabilities.
DC distribution panels D11 and D13 supply normal and alternate control power to several common loads via thennal-only trip breakers. These common loads include unit 1 and 2 4160V switchgears (1A01, IA02,it I and
- 1A03, IA04,1A05,.1A06, 2A01, 2A02, 2A03, 2A04, 2A05, and 2A06) and un 2 480V'switchgears (1801,1802, 2803,1804, 2B01, 2B02, 2803, and 2B04).
The selection of control power from D12 or D13 is done manually inside each of the switchgears through the use of a dual knife switch mechanism.
The dual knife switches are contained on a single switch plate, one of the knife switches receives power from Dil, the other from D13. Only one of the knife switches is closed at a time and that is considered the normal control power supply to that switchgear. The remaining open knife switch is considered to be the alternate control power source and must be manually closed if the normal source is not available. The position of the knife switches is controlled administrative 1y.
Other loads receiving DC power from panels D11 and D13 via thermal-only.
trip breakers include: unit 1 and unit 2 "A" and *B* crossover steam dumps; eniergency diesel generator G01 and G02 field flash; and DC power distribution panels D17, D18, D19, D21, and 022; 3.
Identification of Original Design Flaw The licensee contracted a Safety ' System Functional inspection'(SSFI)' to be performed on the Emergency Diesel Generator System during the months of January and February of 1988.
This inspection included an evaluation and assessment of the DC power systems that were used to support operation of the facility's two emergency diesel generators: One of the findings of the SSFI (Audit Finding Report (AFR) f SSFI WE-88-14), dated 1/22/88, stated that 'DC distribution bus short circuit exceeds main' breaker U.L. rating." This finding was based on the fact that a calculated fault current of 13,900 amps could occur on the main DC distribution bus D01 which was in excess of the 10-,000 amp tripping capability of the thennal-only breakers on the bus. This presented the possibility that a f ault on either of the main DC busses would have the potential of discharging the battery associated with that bus, thereby rendering that DC power train inoperable.
As a result of this finding the licensee contacted Westinghouse, the breaker manufacturer, and received verbal confirmation that the breakers in question were capable of interrupting fault currents up to 20,000 amps. This verbal information was confirmed in a letter dated 2/17/88 to the vendor representative at the site.
It was further stated that test data was available to confinn thistlevel of interrupting capability..
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The Corrective Action for AFR No. SSf 3 WE-88-14, dated 6/20/88, stated
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that the licensee would obtain the Westinghouse test data that demonstrated i
the adequacy of the interrupting capability of the breakers in question.
The anticipated completion date stated in the document is January 1,1989, j
On November 7,1989, the plant staff was informed by the corporate Nuclear Engineering group that in the course of followup to obtain the Westinghouse test data, it was found that the data originally discussed was not applicable to the type of breakers installed'at the facility.
Through miscommunication the vendor assumed that the breakers in question i
had both magnetic and thermal tripping capabilities. Westinghouse stated that the originally installed thermal-only breakers were not capable of interrupting the possible fault currents.
Further analysis by the licensee's engineering group identified the possibility of generating a fault current on a nonsafety load commonly supplied by distribution panels 011 and D13 in excess of the tripping capability of the thermal-only breakers. A fault on one of these non-safety related circuits could result in currents which would not be interrupted by any of the. breakers in the DC system.
This could result in the fa' ult current being sustained until one df the comp'onents in the -
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DC system catastrophically fails or until both batteries supplying the-system are discharged to a point where they can no longer. provide sufficient energy for their safety-related functions.
4.
Request for Enforcement Discretion The condition discussed,in the preceding paragraph caused the licensee to declare station batteries DOS and D06 technically indperable, at 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />, on November 7, 1989. Declaring both DOS and 006 inoperable placed the licensee outside of Technical Specification 15.3.7.B.I.f.
This Limiting. Condition for Operation (LCO) states:
One of the batteries DOS or D06 may be inoperable for a period not exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prov,ided the other three batteries'and four battery chargers remain operable with one charger carrying the DC loads of each DC main distribution bus.
The problem was exacerbated due to the removal from service of station battery D106, on November 6,1989, at 2029 hours0.0235 days <br />0.564 hours <br />0.00335 weeks <br />7.720345e-4 months <br />, for the performance of its five-year performance test. The removal of D106 from service placed the licensee in Technical Specification LCO 15.3.7.B.I.g which states:
One of the batteries DIOS or D106 may be inoserable for a period not exceeding 72-hours provided the other three batteries and four battery chargers remain operable with one charger carrying the DC loads of each DC maf>n distribution bus.
WithDOSand006nowalsotteihnicaillinoperablerthelicenseefound...
itself also operating'outsidefoftTe15 dica 135p6cification LCO 15.3.7.8.1.g.
These conditions required the licenie'e.to. enter Technical Specification LC015.3.0.A which states ~1n, part: gl 5
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In the event an LCO cannot be satisfied because of equipment i
failures or limitations beyond those specified in the permissible i
condition of the LCO, the affected unit, which is critical, shall be placed in the hot shutdown condition within three hours of discovery of the situation.
At 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />, on November 7 1989, with unit 2 in cold shutdown due 1
to a refueling outage and unit 1 operating at 100% power, the licensee i
requested enforcement discretion for point Beach Nuclear plant Unit 1 i
and 2 Technical Specification 15.3.0.A.
This request was based on the following considerations:
f The ability of the licensee to return station battery D06 to an operable condition in a matter of minutes through a reconfiguration i
of breakers on DC distribution panels D11 and 013.
The returning of station battery D106 to an operable condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> due to the completion of the in-progress charging of l
that battery. During the testing and charging 'of D106, a fully qualified temporary battery was connected to 0106's DC bus through a non-gualified temporary cable.
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The returning of DOS to an operable condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the replacement of several thermal-only breakers and the l
reconfiguration of breakers on DC distribution panels D11 and 013.
i NRC granted enforcement discretion at 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br /> on November 7, 1989, for.a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> expiring at 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br /> CST on November 8,1989.
5.
Corrective Action Taken By the Licensee The corrective actions taken by the licensee to restore all station
, batteries to an caerable condition were detailed in a letter from the licensee'to the NRC dated November 10, 1989.. The basic. corrective actions taken by 'the licensee were as fpilows'-
All nonsafety-related loads were transferred from station battery D06 to D05.
Unit 3 was reduced to 92% power to allow the opening of supply breakers to the unit I crossover steam dumps.
Charging of station battery D106 was completed and the battery returned to service at 1007 hours0.0117 days <br />0.28 hours <br />0.00167 weeks <br />3.831635e-4 months <br /> on November 8,1989.
Unit 2 nonsafety-related loads on station battery DOS were disconnected.
Breakers for unit I nonsafety-related;1oads on station battery DOS.
were replaced with breakers eq'tiippe~d'irith thermal and magnetic tripping capabilities returning}; DOS to an operable condition.-
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These actions were accomplished by 1557 hours0.018 days <br />0.433 hours <br />0.00257 weeks <br />5.924385e-4 months <br /> on November 8,1989 and the licensee exited Technical Specification 15.3.0.A.
All corrective actions were monitored by the resident inspector and appeared to be adequate.
further actions were taken to re)1 ace breakers for unit 2 nonsafety-related loads with breakers equipped wit 1 thermal and magnetic tripping capabilities, t
This allowed the reconnection of these loads to the DC busses and permitted the restart of unit 2.
i 6.
Tailure to Meet Single railure Criteria j
Due to the original design of the facility's DC Electrical Distribution System, the potential existed for a single failure to result in the failure of the DC electrical system to perform its intended safety l
function.
This failure was identified by the licensee on November 7, 1989 and was determined to have existed since the initial operation of the facility.
The licensee identified two methods for this failure to occur both of
)
lied from DC
.which are related to nonsafety-related components supp'a failure of the distribution panels D11 and 013. The first method'is L
dual knife switch circuit for 'the connonly supplied nonsafety-related busses which could result.in a simultaneous bus fault'on both DC systems.
This fault could cause a resultant failure of both DC system busses due to a lack of a breaker anywhere in the supply path capable i
.of interrupting the fault current. The second method of failure is related to,the possibility that cables to these connonly supplied nonsafety-related '. loads run in. common ~ raceways. A failure in one.of.
these common raceways could also result in a simultaneous fault on:
4 both DC systems causing their ultimate failure.
The licensee's Final Safety Analysis Report, Section 8.1.1, Electrical Systems, principal Design Criteria for Emergency power states:
An-emergency power source shall be provided and designed with adequate independence, redundancy, capacity, and testability to permit the functioning of the engineered safety features and protective systems required to avoid undue risk to the health and safety of the public.
i This power source shall provide this capability assuming a failure of a single active component.
The licensee's onsite DC distribution system was not capable of providing sufficient independence assuming a single failure due to an original plant design deficiency regarding installed DC breakers.
This i
failure to provide sufficient independence and separation for the DC distribution system in accordance with the applicable design documentation is an apparent violation of 10 CFR 50, Appendix A, Criterion 17 (266/89033-01 and 301/89033-01).
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7.
Failure to Take Prompt Corrective Actions problems associated with the original design of the facility's DC Electrical Distribution System were identified during a contracted Safety System Functional Inspection (SSFI) performed on the Emergency Diesel Generator System during the sionths of January and February 1988.
l One of the findings of the $$FI, dated 1/22/88, stated that *DC distribution bus short circuit exceeds main breaker U.L. rating."
I
/ts a result of this finding the, licensee contacted Westinghouse, the breaker manufacturer,lled thermal-only breakers. requesting information on interru i
l current for the insta Westinghouse verbally l
confirmed that the breakers in question were capable of interrupting l
fault currents up to 20,000 amps.
This verbal information was confirmed in a letter dated 2/17/88 to both the vendor site representative and the i
l licensee. The letter, however, did not specifically address the thennal-only trip breakers as requested by the licensee. The letter i
stated that test data was available to confinn this level of interrupting capability.
x An Audit finding Report.(AFR) i SSFI WE-88-14 was. initiated in accordance -
with Quality Assuranc'e Instruction (QAI) 6 " Audit Preparation, I
Perfonnance and' Documentation", Revision 10, which tracked actions taken to correct this SSFI finding. The AFR also required that specific corrective actions be addressed and a. response provided by 5/31/88. The licensee conducted a meeting in early May to discuss the adequacy of proposed corrective actions for all.,the SSTI findings. AFR $$FJ WE-88-14 corrective actions were released in a document dated 6/20/88 and noted on the AFR itself.
It stated th'at the license'e would obtain the test data from Westinghouse which demonstrated the adequacy of the interrupting capability of the breakers in question.
The document further states the i
anticipated cornpletion date for these action was January 1,1989.
^
AF.Rs and other Quality Assurance audit findings are listed on a Monthly Open Item Status Report (M0ISR) which is issued at the.beginning of each month. AFR SSFI WE-88-14 was included on the MOISR and the' Nuclear Systems Engineering and Analysis Section (NSEAS) was assigned as the organization responsible for corrective actions.
In response to previous NRC inspection findings and concerns, the NRC conducted a further inspection during the weeks of April 25-28 and May 10-12,1989, to assess the licensee's efforts with regards to SSFI findings. Theinspectionreport(266/89012;301/89012) concluded that responses to contractor SSFI audit findings were, in some cases, untimely and inadequate.
Several internal licensee-memos were written regarding the increased level of concern and attention.that the'SSFI audit' findings had been j
receiving.' One of these memos dated l March 31,1989r entitled
" Verification of WESTECrand*RHR VeFticaESlice Audit Finding Reports",
stated that several concerns.had beenpaised about timeliness of and adequacy of corrective actions..tofth'eWESTEC:au'dit." The memo announced the perfonnance of detailed verificatioWs'to'be conducted on the WESTEC 4
SSFI AFRs and requested assistance;,from' personnel responsible for i
specific AFRs.
8 i
Another memo dated April 28, 1989, entitled *S$f! Follow-up", requested j
that action taken on the SSfl items be revisited and evaluated to determine if the documentation is co@lete or if further actions need to be taken.
An in?.ernal licensee nemo dated May 2,1989, entitled *SSf! Vertical-Slice l
Audit follow-up", identified audit findings with the highes,t priority and recomended a reevaluation of the proposed corrective actions for each. -
AFR SSf! WE-88-14 appeared as the fifth item listed in this memo and was redesignated as AFR #A SP-88-02-016.
The inspector reviewed the MOISR for the month of July and noted this-ATR remained assigned to NSEAS and that i
it was listed as having corrective actions overdue, i
A November 7, 1989, internal licensee document states that the documentation
~
and su) porting test data for the themal-only breakers were not obtained from tae vendor until after NRC reviews done in 1989 pointed out the inadequacies in the implementation of corrective action for $$F1 type audits.
This same document further states that subsequent discussions with the vendor resulted in a detemination that the breakers are not capable of interrupting short circuit. faults in excess of their, rating.
On. November 7 l'989 the platit staff was infomed by NSEAS that in' the '
cour'seoffoIIowuptoobtaintheWestinghousetestdata,itwasfound that the data originally discussed was not applicable to the type of breakers installed at the facility.
This finding then resulted in the licensee requesting enforcement discretion at 1600 on November 7.
. Enforcement discretion may not have been necessary if this problem had been identified earlier,through. prompt follow-up of the corrective action for this SSf! finding.
Battery 005 or D06 could have been restored to an operable condition within the three hour Limiting Condition for Operation (LCO).
As such, enforcement discretion was only needed because stat. ion battery D106, not otherwise affected by this problem,, happened to be'out
.of service due to testing at the time the problem was discovered.
Inoperability of this third battery forced the licensee'into a conditibn l
outside of nomal Technical Specifications that could not be corrected before the LCO would have been exceeded, i
The licensee failed to complete prompt corrective actions to a contracted SSF1 audit finding regarding the potential failure of installed DC circuit breakers' ability to interrupt short circuit fault currents in excess of the breakers rating.
This is an apparent violation l
of 10 CFR 50, Appendix B, Criterion XVI L266/89033-02 and 301/89033-02).
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8.
ExitInterview(30703)
A verbal summary of preliminary findings was provided to the licensee representatives denoted in Section 1 on December 22,1989, at the conclusion of the inspection. No written inspection material was provided to the licensee during the inspection.
The likely informational content of the inspection report with regard to documents or processes reviewed during the inspection was also discussed.
The licensee did not identify any documents or processes as proprietery.
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