ML20214J882
| ML20214J882 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 08/12/1986 |
| From: | Taylor J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| Shared Package | |
| ML20214J871 | List: |
| References | |
| 50-250-86-11, 50-250-86-26, 50-251-86-11, 50-251-86-26, 50-261-86-26, EA-86-020, EA-86-20, NUDOCS 8608150208 | |
| Download: ML20214J882 (12) | |
Text
NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTIES Florida Power and Light Company Docket Nos. 50-250 and 50-251 Turkey Point Units 3 and 4 License Nos. DPR-31 and DPR-41 EA 86-20 As a result of inspections conducted August 26-30, September 9-13, November 4-8 and 18-22, 1985, and January 6-10 and February 7 - May 15, 1986 violations of NRC requirements were identified.
In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2. Appendix C (1986), the Nuclear Regulatory Comission proposes to impose civil penalties pursuant to Section 234 of the Atomic Energy Act of 1954, as amenddd, 42 U.S.C.
2282, PL 96-295, and 10 CFR 2.205. The par-ticular violations and associated civil penalties are set forth below:
1.
10 CFR Part 50, Appendix B, Criterion III as implemented by the approved Florida Power and Light Company Topical Quality Assurance Repoc. (FPLTQAR) 1-76A, Revision 8 Topical Quality Requirement (TQR) 3.0, Revision 4 and Appendix B, Revision 7 requires that design changes, including field changes, be subject to design control measures commensurate with those applied to the original' design and that these design control measures assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.
FPLTQAR 1-76A, Appendix C, Revision 7 specifically commits, with exceptions not relevant here, to American National Standards Institute (ANSI)
N45.2.11-1974, Quality Assurance Requirements for the Design of Nuclear Power Plants, and to Regulatory Guide 1.64, Revision 2, Quality Assurance Requirements for the Design of Nuclear Power Plants, which endorses ANSI N45.2.11-1974. ANSI N45.2.11-1974 specifies that design activities be prescribed and accomplished in accordance with procedures of a type sufficient to assure that applicable design inputs are correctly translated into specifications, drawings, procedures, or instructions. Appropriate quality standards must be identified, documented, and their selection reviewed and approved.
Changes from specified quality standards, including reasons for the changes, must be identified, approved, documented, and controlled. Design changes must be justified and subjected to design control measures commensurate with those applied to the original de;ign.
Contrary to the above, the Safety System Functional Inspection (SSFI) and the NRC Region II followup inspections identified that:
A.
At the time of the inspections, design inputs had not been correctly translated into operating procedures, as demonstrated by the examples below.
1.
Off-Normal Operating Procedure (ONOP) 0208.11, Annunciator Test-Panel I - Station Service, was not revised to include appropriate
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Notice of Violation operator action in the event of a low pressure alarm on the nitrogen back-up system.
Plant change / modification (PC/M)80-117, Upgrade Auxiliary Feedwater Suction, Discharge, and i
Steam Supply Piping, which was completed for Unit 3 on May 24, 1983 and for Unit 4 on December 16, 1983, modified the nitrogen back-up system such that operator action was required in the event of a low pressure alarm to ensure the continued operability of the auxiliary feedwater (AFW) system in the event of a loss of the normal instrument air supply.
2.
Emergency Operating Procedure (E0P) 20003, Steam Generator Tube Rupture, directed control room operators to isolate the steam supply from the faulted steam generator to the AFW turbines by shutting the associated motor-operated isolation valves using the hand switches in the control rocm. However, the motor-operated isolation valves could not be closed from the control room to isolate a faulted steam generator when an AFW automatic initiation signal existed. When initially isolated, a subsequent AFW automatic initiation signal would reopen the motor-operated isolation valve.
Because of these design features, alternate methods should have been evaluated to isolate a faulted steam generator from the AFW steam supply.
3.
Procedure 7300.2, AFW System Flow Control Valves Instrument Air /
Nitrogen Back-up System Operation, indicated operators had 15 minutes to valve in standby nitrogen bottles after a low pressure annunciator alarm.
However, the AFW system description and design basis allowed only ten minutes for the operators to take the required action.
4.
Shared Auxiliary Feedwater System, System Description and Design Basis, Revision 1, dated January 31,1985 (AFW System Description),
requires that 286 GPM of auxiliary feedwater flow must be delivered to each unit within three minutes in the event of a two-unit trip with only one AFW pump available.
However, Emergency Operating Procedures 20004, Loss of Offsite Power, Revision dated August 23, 1985, and 20007, Total Loss of AC, dated August 26, 1985, did not assure that control room operators could balance AFW flows to the required 286 GPM for each unit within three minutes in the event of a two-unit trip with only one AFW pump available as required by the System Description.
5.
ONOP-103, Control Room Inaccessibility, datad August 7, 1985, did not address local control of Train 2 AFW flow control valves (FCVs) or local control of the AFW pumps even though local control of Train 2 is required when only the "B" AFW pump is operable and is necessary when an AFW pump trips when the control room is inaccessible.
At the time of the inspections, design inputs had not been correctly translated into drawings. A change to PC/M 80-117 reset the nitrogen back-up system control pressure from 55 psig to 80 psig.
- However, the nitrogen back-up system P&ID 5610-M-339 was not revised to reflect this change in pressure control valve setpoints applicable to valves PC-3-1706, PC-3-1708 PC-4-1705 and PC-4-1709.
C.
At the time of the inspections, appropriate quality standards were not correctly translated into procedures or drawings, as shown by the following two examples.
1.
PC/M 80-117 indicated that electrical and instrumentation equipment associated with the nitrogen back-up system was safety-related.
However, this equipment was not included in Quality Instruction QI-2.3A, Classification of Structures, Systems, and Components (Q-List). This equipment was therefore not being treated as safety-related by maintenance personnel and was not included in the periodic surveillance and calibration program.
2.
Low pressure signals from pressure switches PS-3-2322 and PS-3-2323. shared a common field wire and were neither designed nor considered safety-related evtn though current safety-related emergency operating procedures require action by the operator upon receipt of the nitrogen low pressure alarm.
D.
At the time of the inspections, design inputs were not correctly translated into the system description and design basis documents as demonstrated by the following examples.
1.
The AFW System Description indicated that an air signal was supplied by a differential pressure controller which was set to maintain a minimum pump discharge pressure approximately 125 psi higher than the steam supply pressure.
However, this design feature was disconnected.
2.
The AFW System Description indicated that when instrument air pressure dropped below 55 psig, check valves open to automatically supply back up nitrogen. Howevar, as a result of PC/M 80-117, the pressure control valves were set at 80 psig.
3.
The AFW System Description involving operator action in response to a low pressure alarm in the nitrogen back-up system, at the time of the SSFI, referenced intended operation of the system before PC/M 80-117 was performed. The described action violated the single failure criterion, and was not appropriate for the current system configuration.
4.
The AFW System Description does not define how long the AFW system may operate without operator action.
No guidance was provided to establish operating limits on the available nitrogen supply.
l Notice of Violation E.
As of the time of the inspection, design changes were not subjected to design control measures commensurate with those applied to the original design as two design changes to the AFW system caused an oscillation problem with the AFW FCVs when in the automatic mode of operation, These design changes involved:
(1) a removal from service of the differential pressure controller and (2) a reduction in AFW flow from 600 GPM to 373 GPM. The resultant FCV oscillations created an increase in the nitrogen consumption rate. The increased consumption invalidated the design criteria for the nitrogen back-up system low pressure alarm setpoint which was set to allow 15 minutes for operator action to restore the nitrogen supply to the AFW FCVs.
F.
Quality Instruction 3.1, Control of EPP Design, Revision 2, dated October 9, 1979, in Section 5.5(4), requires that during the processing of the design by the discipline (s), various design and safety analyses are required. These analyses must be documented in legible and reproducible form by the engineer performing tnem. Notwithstanding this instruction, the licensee failed to adequately document assumptions and design inputs in calculations dated November 15, 1979 for the low level alarm setpoint on the condensate storage tank that would prevent a low pump suction pressure from occurring.
This is a Severity Level III problem (Supplement I).
(Cumulative Civil Penalty - $50,000 assessed equally among Violations A-F.)
to.liity as described in the Final Safety Analysis Report (ges in the 10 Cl't 50.59(a) allows the holder of a license to make chan II.
FSAR) without prior Commission approval unless the proposed change involves a change in the Technical Specifications or is an unreviewed safety question.
10 CFR 50.59(b) requires the licensee to maintain records of changes in the facility or procedures, including a written safety evaluation providing the basis for determining that the change does not involve an unreviewed safety question.
Contrary to the above, the licensee failed to meet the requirements of 10 CFR 50.59 in that at the time of the SSFI and NRC Region II followup inspections:
A.
The licensee failed to perform an adequate safety evaluation for PC/M 80-117 in that en analysis was not performed of the consequence of an AFW steam vent valve failure to close and the ability of the auxiliary feedwater pump to supply sufficient feedwater flow at reduced steam generator pressures as described in FSAR Section 9.11.2, Auxiliary Feedwater Pumps.
B.
The licensee failed to perform an adequate safety evaluation for Temporary System Alterations (TSA) 3-84-11-75, 3/4-84-99-75, 3/4-85-08-75, and 3/4-84-100-75 pertaining to the removal of the AFW governor speed control system as described in FSAR Section 9.11.2, Auxiliary Feedwater Pumps.
The safety evaluation did not evaluate the mechanical reliability of the AFW system being operated under a constant speed condition.
The licensee failed to perform a safety evaluation for the consequences of the addition of electrical loads on a 4 KV engineered safety features bus. These additional loads should have been evaluated as to the capability of the emergency diesel generators to carry the increased loads as described in FSAR Section 8.2.3.
This is a Severity Level III problem (Supplement I).
(Cumulative Civil Penalty - $50,000 assessed equally among Violations A-C.)
III. A.
Turkey Point Technical Specification 3.8.5.a requires that for single unit operat an, with one of the two required independent auxiliary feedwater rains inoperable, the inoperable train must be restored to operab' status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the reactor shut down and the reactor coolant temperature reduced below 350 F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Technical Specification 3.8.4.b requires that for dual nuclear unit operation, two independent auxiliary feedwater trains and a third pump capable of being powered from, and supplying water to, either train must be operable when the reactor coolant is heated above 350 F.
Technical Specification 3.8.5 requires that during power operation, if any of the conditions of 3.8.4 cannot be met, the reactor shall be shut down and the reactor coolant temperature reduced below 350 F, unless certain conditions can be met.
Contrary to the above, the Limiting Condition for Operation (LCO) for Technical Specification 3.8.5 was exceeded on January 2, 1986, when two Unit 3 and one Unit 4 AFW steam supply stop check valves (119 and 219 and 319, respectively) were found to be inoperable. The condition of the valves was identified on January 2,1986 as unacceptable by radiography personnel per the acceptance criteria of Test Request 001-86. Although both Units 3 and 4 were operating at above 350 F, no actions were taken by licensee personnel until January 7, 1986, when the results of the radiographs were questioned by an NRC inspector.
B.
Technical Specification 3.4.1.4 requires that the reactor not be made critical, except for low power physics tests, unless four safety injection pumps are operable.
Contrary to the above, on February 12, 1986, the Unit 3 reactor was taken critical for power operations when only three safety injection pumps were operable. Low power physics tests were not being performed.
This is a Severity Level III problem (Supplement I).
(Civil Penalty - $50,000 assessed equally between Violations A and B.)
A.
Turkey Point Technical Specification 6.8.1 requires that written procedures and administrative policies be established, implemented, and maintained that meet or exceed the requirements and recommendations of Sections 5.1 and 5.3 of ANSI N18.7-1972. Section 5.1 requires, in part, that instructions provide a clear understanding of operating philosophy and management policies.
In particular, written administra-tive policies must be provided to control the issuance of documents, including changes, that prescribe activities affecting safety-related structures, systems, or components, such as operating procedures, test procedures, equipment control procedures, maintenance or modifi-cation procedures, and refueling procedures. These policies must assure that documents, including revisions or changes, are reviewed for adequacy and approved for release by authorized personnel and are distributed to and used by the personnel performing the prescribed activity.
Technical Specification 6.8.1 also requires that written procedures and administrative policies be established, implemented, and maintained that meet or exceed the requirements and recommendations of Appendix A of NRC Regulatory Guide 1.33, " Quality Assurance Requirements (Operations)." Regulatory Guide 1.33, Appendix A requires that certain safety-related activities be covered by written procedures,
-including procedures for combating emergencies and other significant events.
Contrary to the above, the SSFI and the.NRC Region If followup inspections identified, at the time of the respective inspections, that:
1.
Surveillance Maintenance Instruction 0-SMI-059.1, Inside Containment Instrument Independent Verification, and 0-SMI-059.2, Outside Containment Instrument Independent, Verification, dated July 30, 1985, specify requirements for providing assurance that safety-related instrumentation was properly aligned when the instruments were returned to service following unit outages.
Notwithstanding, these instructions were inadequate because they failed to provide assurance that safety-related instrumentation was properly aligned when the instruments were returned to service following maintenance or calibration while the plant was in an operating mode.
2.
Administrative Procedure 0103.3, Control and Use of Temporary System Alterations (TSAs), dated January 31, 1984, Section 5.8, requires that the Plant Nuclear Safety Committee (PNSC) be responsible for reviewing applicable nuclear safety-related temporary system alterations within fourteen days of the Plant Supervisor-Nuclear approval date.
Notwithstanding this procedure, the PSNC failed to review TSAs 3-84-11-75, 3/4-85-8-75, 3/4-84-99-75 and 3/4-84-100-75 pertaining to the removal of AFW governor speed control systems within the prescribed time period.
These TSAs were not reviewed by the PNSC until six months after the Plant Supervisor-Nuclear approval date.
B.
10 CFR Part 50, Appendix B, Criterion V and VI as implemented by the approved FPLTQAR 1-76A, Revision 8, TQR 5.0, Revision 5, and TQR 6.0, Revision 3, requires that measures be established to control the issuance of documents, such as instructions, procedures, and drawings, including changes thereto, which prescribe activities affecting quality. These measures assure that documents, including changes, are reviewed for adequacy and approved for release by authorized personnel and are distributed to and used at the location where the prescribed activity is performed.
Contrary to the above, the licensee failed to ensure that procedure 3/4-0P-018.1, Condensate Storage Tank, a safety-related procedure, was approved for release by authorized personnel and appropriately distributed to and used at the location where the prescribed activity was performed.
~
This is a Severity Level III problem (Supplement I).
(Civil Penalty - 550,000 assessed equally between Violations A-B.)
V.
A.
Turkey Point Technical Specification 4.8.2 requires for station batteries that:
1.
Monthly each batterj be given an equalizing charge, and afterwards specific gravity and voltage readings be taken and recorded for each cell. Water must be added to restore normal level and total water use is to be recorded. Complete visual inspection of the batteries must be made monthly.
2.
Annually, a load test must be performed.
Technical Specification 3.7, Electrical Systems, requires that neither reactor may be started from cold shutdown unless four batteries and associated DC systems are operable and four out of six battery cha gers are operable.
Contrary to tile above:
1.
From August, 1985 until the time of the SSFI and Region II followup inspections, the licensee failed to conduct adequate monthly surveillance testing on the 125 volt station batteries as required by Technical Specification 4.8.2 to demonstrate the operability requirements of Technical Specification 3.7.
i
I Notice of Violation 4 2.
At the time of the inspections, the licensee had failed to conduct adequate station battery load capacity tests as required by Technical Specification 4.8.2 to demonstrate the operability requirements of Technical Specification 3.7.
B.
10 CFR Part 50, Appendix B, Criteria V and XVI, as implemented by the approved FPLTQAR 1-76A, Revision 8, TQR 5.0, Revisinn 5 and TQR 16.0, Revision 4, require:
(1) that activities affecting quality be prescribed by and accomplished in accordance with documented instructions or procedures, and (2) that significant conditions adverse to quality be promptly identified and corrected and the cause of the condition and the corrective action taken be documented and promptly reported to the appropriate levels of management.
1.
Contrary to the above, Nonconformance Report 85-206, dated-December 5, 1985, required a search for the missing valve parts (three disc guide studs and one thrust washer); however, no documented basis was provided for justification to consider the system operable when the missing parts were not found and both units were subsequently returned to service.
2.
Florida Power and Light interoffice correspondence dated February 25 and March 25, 1981 indicated the potential that a single valve, 3-20-428 for Unit 3, and 4-20-428 for Unit 4, could isolate two redundant condensate storage tank level transmitters.
Contrary to the above, the licensee failed to implement adequate administrative procedural controls subsequent to this discovery in 1981 to ensure that these isolation valves were adequately controlled in the open position, i
3.
a.
Administrative Procedure ADM-0-]90.19, Control of Maintenance
~
on Nuclear Safety-Related and Fire Protection Systems, dated May 21, 1985, required that applicable post-maintenance j
testing be defined on the Plant Work Order (PW0) to ensure proper post-maintenance functioning of the equipment.
Section 8.1.5 requires a Quality Control review of completed PW0s to ensure post-maintenance testing has been conducted.
Contrary to the above, at the time of the SSFI and Region II l
followup inspections, the licensee had failed to implement these procedural requirements in that most ir.strumentation and control and electrical safety-related PW0s were completed without documentation of adequate post-maintenance testing.
In addition, the licensee failed to provide adequate procedures in that Administrative Procedure ADM-0-109.28, Mechanical Test Control (Post Maintenance), dated August 21, 1985, addressed only mechanical maintenance with no provisions to ensure that post-maintenance testing associated with l
l instrumentation and control and electrical PW0s was conducted.
l
Administrative Procedure 0-ADM-701, Plant Work Order Preparation, Section 5.8.1.8, dated July 12, 1985, requires that the root cause of equipment failures be identified on completed Plant Work Orders.
Contrary to the above, at the time of the SSFI and the Region II followup inspections, the licensee had failed to identify the root cause of equipment failures on a majority of several hundred completed safety-related PW0s which were reviewed.
This is a Severity Level III problem (Supplement I).
(Civil Penalty - $50,000 assessed equally between Violations A and B.)
VI.
10 CFR Part 50, Appendix B, Criterion XVI as implemented by the approved FPLTQAR 1-76A, Revision 8, and TQR 16.0, Revision 4, requires that conditions adverse to quality be promptly identified and corrected.
A.
Contrary to the above, the licensee did not take prompt and adequate corrective actions in that after it was determined in J&nuary 1986 that Component Cooling Water (CCW) flow through the Residual Heat Removal (RHR) heat exchangers appeared to be less than the minimum 4,000 GPM assumed in the accident analysis, CCW valves 48A and B were repositioned from 30% open to full open on February 24, 1986. This was done without evaluation or testing of the adequacy of flow to the other components served by the CCW system. Unit 3 continued to operate at 100% power from February 24 to March 3 with indeterminate CCW flow rates to the engineered safety features components.
B.
Contrary to the above, after determining in November 1984 that intake cooling water (ICW) valve CV-2201 might not be capable of closing upon a loss of power and/or a loss of its control air supply, the licensee failed to take timely corrective action in that the safety significance of this condition was not evaluated until February 13, 1986. The failure of valve'CV-2201 to close would prevent, in some circumstances, sufficient ICW flow from reaching the CCW heat exchangers which, in turn, could prevent the engineered safety features equipment from performing as intended. After the February 13, 1986 evaluation, the licensee again failed to take appropriate corrective action in that no analysis was done regarding ICW system operability or TS limiting condition for operation compliance until NRC Region II management i
questioned operability in a February 14, 1986 telephone call with plant management, thereby prompting that these areas be addressed.
C.
Contrary to the above, in 1980 it was identified that the Unit 4 Component Cooling Water (CCW) trains for the common Unit 3/4 high head safety injection (SI) system "B" pumps were incorrectly installed and potentially incapable of providing adequate SI pump cooling.
i
--..~
Notice of Violation As of the inspection on February 17 - May 5,1986, this condition had not been corrected nor were adequate compensatory controls implemented.
Rather than correct the piping discrepancy, the Unit 3 CCW trains to the SI coolers were used exclusively. Technical Specifications allow single train CCW system operation for a unit in cold shutdown and SI pumps are only required for a unit which is critical. Consequently, on at least two occasions between August and October 1981 with Unit 3 in cold shutdown and Unit 4 critical, only one train of CCW was dvailable for Cooling the SI pumps. Therefore, the SI pumps were susceptible to single failure induced inoperability, in that loss of the Unit 3 CCW train would have disabled all SI pump coolers preventing the SI system from performing its intended safety function.
This is a Severity Level III problem (Supplement I).
(Civil Penalty - $50,000 assessed equally among Violations A-C.)
Pursuant to provisions of 10 CFR 2.201, Florida Power and Light is hereby required to submit to the Director, Office of Inspection and Enforcement, U.S.
Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region II, within 30 days of the date of this Notice, a written statement or explanation, including for each violation:
(1) admission or denial of the violation; (2) the reasons for the violation if admitted; (3) the corrective steps which have been taken and the results achieved; (4) the corrective steps which 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, the Director, Office of Inspection and Enforcement, may issue an order 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 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, Florida Power and Light may pay the civil penalties by letter addressed to the Director, Office of Inspection and Enforcement, with a check, draft, or money order payable to the Treasurer of the United States in the cumulative amount of Three Hundred Thousand Dollars ($300,000) or may protest imposition cf the civil penalties in whole or in part by a written answer addressed to the Director, Office of Inspection and Enforcement.
Should Florida Power and Light Company fail to answer within the time spctified, the Director, Office of Inspection and Enforcement will issue an order bnposing the civil penalties in the amount proposed above.
Should Florida Power and Light Company elect to file an answer in accordance with 10 CFR 2.205 protesting the civil penalties, such answer may:
(1) deny the violation listed in this Notice in whole or in part, (2) demonstrate extenuating circumstances, (3) show error in this Notice, or (4) show other reasons why the penalties should not be imposed.
In addition to protesting the civil penalties in whole or in part, such answer may request remission or mitigation of the penalties.
Notice of Violation In requesting mitigation of the proposed penalties, the five factors addressed in Section V.B of 10 CFR Part 2, Appendix C, should be addressed. Any written arm er in accordance with 10 CFR 2.205 should be set forth separately from the staten.ent or explanation in reply pursuant to 10 CFR 2.201, but may incorporate parts of the 10 CFR 2.201 reply by specific reference (e.g., citing page and paragraph numbers) to avoid repetition.
Florida Power and Light Company's attention is directed to the other provisions of 10 CFR 2.205 regarding the procedure for imposing a civil penalty.
Upon failure to pay any civil penalty due, which has been subsequently determined in accordance with the applicable provisions of 10 CFR 2.205, this matter may be referred to the Attorney General, and the penalty, unless compromised, remitted, or mitigated, may be collected by civil action pursuant to Section 234c of the Act, 42 U.S.C. 2282.
FOR THE NUCLEAR REGULATORY C0!94ISSION
/
F/
J..es M. Taylor Director ffice of Inspection and Enforcement Dated at Bethesda, Maryland, this/$NayofAugust1986.
Distribution PDR ~
LPDR SECY CA ACRS JMTaylor, IE
- JNGrace, RII RSi.arostecki, IE JAxelrad IE
- ABBeach, IE
- JLieberman, 0GC FIngram, PA
- Enforcement Coordinators RI, RII, RIII, RIV, RV BHayes, 01 SConnelly, 0IA JCrooks, AE0D HDenton, NRR NRR Project Manager EDO Reading File EA file ES File DCS State of Florida 4
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