IR 05000259/1989061
| ML18033B176 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 02/01/1990 |
| From: | Carpenter D, Little W, Patterson C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18033B174 | List: |
| References | |
| 50-259-89-61, 50-260-89-61, 50-296-89-61, NUDOCS 9002140153 | |
| Download: ML18033B176 (31) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 Report Nos.:
50-259/89-61, 50-26'0/89-61, and 50-296/89-61 Licensee:
Tennessee Valley Authority 6N 38A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Docket Nos.:
50-259, 50-260 and 50-296 License Nos.:
DPR-33, DPR-52, and DPR-68 Facility Name:
Browns Ferry 1, 2,*and
Inspection Conducted:
December 15, 1989 - January 15, 1990
)n
'r'nspecto
.'
.@iD.
R.. Carpentetq NRC Site Manager
~)a
~'C. A. Pattersori,,
NRC Restart Coordinator at igned D
e igned Accompanied by:
E. Christnot, Resident Inspector W. Bearden, Resident Inspector K..Ivey> Resident Inspector Approved by:
W.
S. Little, Sec Ion Chief, Inspection Programs, TVA Projects Division
"D e
igned SUMMARY Scope:
This routine resident inspection included reportable occurrences and action on previous inspection findings.
- Results:
This inspection report is primarily a closeout of open items and LERs.
One LER, eleven IFIsI four URIs and three VIOs were closed.
There are no cited vi.olations or deviations in this inspection repor REPORT DETAILS Persons Contacted Licensee Employees:
0. Zeringue, Site Director
"G. Campbel'1, Plant Manager
- M. Herrell, Plant Operations Manager R. Smith, Project Engineer J. Hutton, Operations Superintendent A. Sorrell, Maintenance Superintendent G. Turner, Site Quality Assurance Manager P. Carier, Site Licensing Manager P. Salas, Compliance Supervisor
"J. Corey, Site Radiological Control Superintendent R. Tuttle, Site Security Manager Other licensee employees or contractors contacted included licensed reactor operators, auxiliary operators, craftsmen, technicians, and public safety officers; and quality assurance, design, and engineering personnel.
NRC Personnel:
- W. Little, Section Chief
- D. Carpenter, Site Manager
- C. Patterson, Restart Coordinator
"ED Christnot, Resident Inspector W. Bearden, Resident Inspector K. Ivey, Resident Inspector
- Attended exit interview Acronyms used throughout this report are listed in the last paragraph.
Reportable Occurrences (92700)
The LER listed below was reviewed to determine if the information provided met HRC requirements.
The determinations included the verification of compliance with TS and regulatory requirements, and addressed the adequacy of the event description, the corrective actions taken, the existence of potential. generic problems, compliance with reporting requirements, and the relative safety significance of each event.
Additional in-plant reviews and discussions with plant personnel, as appropriate, were conducted.-
(CLOSED)
LER 259/88-17, Unplanned Standby Gas Treatment Actuation Due To Personnel Error.
This event involved the automatic actuation of SBGT train C
on June 4,
1988, during,its return to service following maintenance.
The actuation occurred because a latching relay ( labeled MCX) in the control circuit was left in the actuate (or operate)
state prior to its return to service, allowing train actuation upon closure of the supply breaker.
This event is similar to one previously submitted as LER 259/88-09.
The previous event led the engineer directing the maintenance to instruct craft personnel to change the state of the MCX relay, causing the actuation in this event.
Furthermore, there was confusion about the pull-to-lock function of'he SBGT train C control room handswitch.
Operations personnel'laced the handswitch in pull-to-lock to prevent the actuation; however, the MCX relay is located downstream of the pull-to-lock contacts and was not affected.
During this reporting period, an inspector reviewed the LER, dated July 2, 1988, and the licensee's closure package for this item; reviewed associated electrical maintenance procedures; and discussed 'this event with the maintenance engineer involved.
The, inspector verified that the corrective actions listed in the LER had been completed.
The inspector verified tha.
the breaker maintenance procedures and the SBGT operating instruction had been revised to add more detail defining the states of the MCX relay; to provide more detail on the effects of the pull-to-lock function on the MCX relay; and to require personnel to determine the state of the MCX relay during maintenance.
'This item is closed.
Action on Previous Inspection Findings (92701, 92702)
a.
(CLOSED)
IFI 259/81-25-01, Inclusion of Local Leak Rates in-the Integrated Leakage Rate Test.'he licensee revised 2-SI-4.7.A. l-a-f, Primary Containment Integrated Leak Rate Test, on December 1,
1989, to include the local.
leakage rates of various system isolation valves into the Integrated Leakage Rate Test.
The inspector reviewed the procedure revision for inclusion of the applicable valves.
An integrated leakage rate test is planned prior to Unit 2 restart and will be reviewed at that time.
This specific item is considered closed.
b.
(CLOSED) IFI 259, 260, 296/84-38-02, Shutdown Board Room A/C Test.
This item concerned the numerous problems that were observed with Units 1 and 2 shutdown board room cooling units and that Unit 3 units had not been tested.
The inspector reviewed the licensee's closure package for this item and the completed test data.
Units
and
coolers were retested using TI-81, Shutdown Board Room Emergency Cooling System Performance Check.
This item was statused in IR 88-05
and remained open pending testing of a
new cooler unit for Unit 2-installed under ECN P0956.
This concern was reviewed in IR 89-53.
VIO 260/89-53-02 was issued for failure to initiate a
CAQR when a
significant test exception was dispositioned as "accept-as-is".
Unit
has not been satisfactorily tested.
Further review of this item will be in the closure.of. the violation in IR 89-53.
This IFI,is closed.
'CLOSED) IFI 260/84-49-02,- Modifications To The RPS Power Supply Monitoring System.
This item involved an inconsistency between RPS M-G Set Surveillance Requirements TS Section 4. 1.B. l.
The inspector originally interviewed licensee personnel in the Electrical Department to verify implementation of surveillances required by TS section 4. 1.B. 1 for Units
and 3.
Appendix A to SI-1, page ll, listed as a
TS surveillance requirement the implementation of SI 4. 1.B.-16, RPS M-G set, Channel Functional Test.
The inspector reviewed SI 4. 1.B-16 to verify that't incorporated the requirements of the TS.
The inspector determined that a design modification to the RPS for Units 1 and 3 was implemented via ECN ~P0422.
In addition, the inspector discovered discrepancies between the TS and Appendix A to SI-1 in that SI 4. 1.B-16 references Unit 3 TS Section 4. 1.B. 1, while no such section exists.
Also, Table 4. 1:A, RPS Instrument Functional Tests Minimum Functional Test Frequen'cies for Safety Instruments and Control Circuits, of the Unit 3 TS did not address. this requirement.
Unit
TS were consistent with Appendix A to SI-1.
The-inspector veri fied that the required survei llances were being performed in accordance with licensee surveillance calibration program delineated in SI-1.
Licensee management explained the discrepancies between Section 4.0 of Unit 3 TS and Appendix A to SI-1 as being the result of a delay inthe approval of the TS to incorporate the changes made to the RPS Power Supply Monitoring System for Units
and 3.
Licensee management further added that discussions were continuing with the NRC for resolution of the issue of the basis of setpoint values which were implemented by ECN P0422.
The inspector determined that the acceptance criteria delineated in SI 4. 1.B-16 relative to setpoint values were, at the time of the inspection, as follows:
59 Relay (overvoltage)
operates at <126.5-V ac 27 Relay (undervoltage)
operates at >ill-V ac 81 Relay (underfrequency)
operates at
> 57 Hz The evaluation for an unreviewed safety question as required by
CFR 50.59 was discussed with licensee management concerning the implementation of ECN P0422.
Licensee management stated that the protection provided to the RPS Power Supply System was more conserva-tive at the time of the inspection than that which existed. prior to the implementation of the ECN, regardless of the question of the
0
basis for the setpoint values.
The IFI was identified and left open until the licensee obtained NRC approval for the modification to the RPS Power.Supply Monitoring System, and Unit 3 TS had been revised to show this approval and the surveillance requirements.
From November 1984, when this it'em was identified, to the present time numerous changes have occurred at BFNP.
Unit 2 is scheduled to be the first unit to restart, consequently changes have been made to the Unit
TS.
The Unit
TS have the only surveillance requirement identified as 4. 1.B.2 which states:
I At least once per 18 months by demonstrating the OPERABILITY of overvoltage, undervoltage, and under-frequency protective instrumentation by 'simulated automatic logic actuation and verification of the circuit protector trip level setting as follows:
(a)
overvoltage (all device)
<
126.5 VAC (b)
undervoltage (MG Set)
113.4 VAC (c)
undervoltage (alt. supply)
>
111.8 VAC (d)
underfrequency (all devices)
>
57.0 Hz
'lso SI-1, Surveillance Program, has been superceeded by 1-Sl-l, 2-'I-1 and 3-SI-l, Surveillance Program, which are unit specific and SI-4.1.8-16 has been changed to 1/3-SI-4.1.B-16 and 2-SI-4.1.B-16, applicable to units 1 and 3 and unit 2 respectively.
These upgraded procedures cover all six circuit protectors.
The inspector reviewed 1/3-SI-4. 1.B-16 and 2-SI-4. 1.B-16, RPS MG Set, and observed that the acceptance criteria in Section
of 2-SI-4. 1.B-16 are within the Unit 2 TS requirements and the SI fully implements Unit 2 TS 4. 1.B.
The inspector noted that ECN P0422 has been completed for Unit 2.
The inspector observed the performance of 2-SI-4. 1.8-16 and although an undervoltage relay malfunctioned the verification of the new TS and upgraded SI were considered satisfactory.
This item is closed for Unit 2.
(CLOSED)
IFI 259, 260, 296/85-15-06, High Airborne in Reactor Building from RWCU Tank.
This item concerned high airborne activity in the area around the Unit
RWCU system precoat tank.
The licensee could not identify the cause of the activity; however, problems with valves in the system could have contributed to the problem as there were open MRs scheduled to be performed on the system at the time.
Unit
was shutdown during the reporting period and the inspector opened this item to follow the licensee's investigation after restart.
All three units at Browns Ferry have been shutdown for almost 5 years and Unit 1 is scheduled to be the last unit restarted.
The licensee has initiated an extensive corrective action program 'for all three
units which includes design reviews, system upgrades and operability verifications, and the completion of open MRs.
This item should be resolved by these corrective action programs.
In the event that the high airborne activity is still present after Unit 1 restart, it will be dispositioned in accordance with the licensee's programs at that time.
No further review is required and this item is closed.
(CLOSED) IFI 259/85-44-02, Commitment to Re-evaluate TS 3/4. 10 on SRM Requirements.
This item involved conflicting requirements in TS 3/4. 10 for SRM operability during core alterations.
This item also referenced IR 85-43 in which the NRC. questioned TS 3. 10.B concerning the minimum SRM count rate.
The issue of neutron monitoring during core alterations and the adequacy of the TS were raised during the reload of Unit
in January, 1989 (.see IR 89-04 and IR 89-18).
These issues were the subject of several violations, management meetings,. and an Enforce-ment Conference between the licensee and the NRC.
As a result of these findings, the licensee revised the TS for Core-Alteration/SRM Monitoring to provide for continuous observable monitoring of reactivity changes during core alterations (TS change no. 271).
The
'completion of the TS changes and resolution of this issue are being followed up in the review of IFI 89-18-06.
This item is closed.
(CLOSED) IFI 259, 260, 296/85-49-06, Drawing Discrepancies.
This item, identified prior to" the shutdown of all three units, concerned the difficulties at BFNP with drawing discrepancies.
The new design control program resolves this concern.
A recent SSgE type, NRC team inspection of the core spray system did not identify any signficant drawing problems (IR 89-16).
New drawing discrepancy findings will be identified as new items.
This item is-closed.
'I g.
(CLOSED) IFI 259, 260, 296/86-05-04, Review SI Upgrade Program.
This item was to review the completion of the licensee's SI upgrade program.
A special reactive inspection was conducted during September 11 - October 11, 1989 to review continuing problems with the implementation of the TS required surveillance testing program.
IR 259, 260, 296/89-43 was issued November 2,
1989 addressing the continuing problems with, SI program.
Several issues are under consideration for escalated enforcement action.
The SI program will be thoroughly reviewed with the closure of the inspection items in IR 259, 260, 296/89-43.
This IFI is close (CLOSED) IFI 259, 260, 296/86-05-05, Inspector Review of Surveillance Instructions Pr'ior to Startup.
This item concerns whether the SIs are sufficiently detailed and the technicians. are sufficiently trained to allow the procedures to be followed as written.
This item was opened to review SIs during the Unit 2 startup to resolve the inspector's concern.
The, failure to follow SIs and. inadequate procedures have been ongoing problems at Browns Ferry and were the subject of a
management meeting with TVA on September 28, 1989.
A recent special inspection ( IR 89-43)
was conducted to review, problems with SIs and the surveillance testing program in general.
The special
,inspection identified several violations and an-enforcement conference was held between the licensee and the NRC.
The adequacy of the licensee's surveillance
,testing program and SI'
will be reviewed in the followup to the findings of the special inspection.
This item is closed.
(CLOSED)
IFI 259, 260, 296/86-25-05,
'nvironmental Design.,
and Qualification of CREVS.
This item resulted after a review of the configuration control and design baseline program for the CREVS.
"Design, Testing and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup Air Filtration and Absorption Units," states that the environmental conditions for a
given ESF system should be determined for each plant and the potential buildup of moisture in the absorber should be given design consideration.
The inspector reviewed the.licensee closure package for this item.
In an exit meeting for IR 87-42, it was stated that the moisture abso'rption portion of the analysis looked satisfactory, but radioactive material buildup on filters was not analyzed by TVA.
The CREVS is required by TS to be operated for
hours per month to remove excessive moisture build-up on the absorber.
The post-accident dose buildup on the charcoal absorbers was evaluated by calculation 480-97-16985.
The results of the calculation indicated that the postulated radiation levels due to loading on the filters resulted in a 100 day integrated accident dose of less than one rad.
The calculation was reviewed and a conservative assumption was made that the charcoal filters were 100o'fficient for absorption of iodine.
This maximizes the filter loading.
Also, a
table was generated for all of the-components of the system showing weak link non-metallic material and the lowest radiation damage threshold value.
A review of the table indicated that the most radiation sensitive material in the system was Teflon which has a, radiation damage threshold value of 10000 rads.
Therefore, it was concluded that the CREVS would not suffer any deleterious effects for the normal lifetime of the plant and the accident total integrated radiation dose.
This item is close C (CLOSED) IFI 259, 260, 296/87-37-02, Chernobyl Lessons Learned.
This item was opened to ensure acceptable actions in responses to the Chernobyl nuclear accident.
The licensee has recently completed a series of activities that more appropriately respond to NUREG/BR-0032, Vol. 7, No.
36, USNRC News Release for the week ending September 15, 1987.
These actions included the revision of the following procedures:
NP-STD-10. 1.57, Special Tests PMI 17.8, Special Test Instruction PMI 17. 1, Conduct of Testing PMI 12. 12, Conduct of Operations" A training course was developed, EGT 106.019, Chernobyl Unit
Accident.
The required staff reviewed procedure revisions and were trained on the mitigation of core damage.
The inspector reviewed the revised procedures and the lesson plans for this item and found them acceptable.
This item is closed k.
(CLOSED) IFI 259, 260, 296/88-36-06, Review of Operability Determina-tions for Nonconforming Items.
During plant walkdowns the inspector noted several examples of Nonconforming Item Tags which contained information inconsistent with the requirements of the NIZAM.
The inspector reviewed the NIZAM, Part I,
Section 2. 15, and site procedure SDSP 3.8, Nonconforming Material, Parts, or Components, and determined that SDSP 3.8 was not in agreement with the NIZAM with respect to conditional releases.
Additionally, the inspector provided the licensee with four specific examples of tags installed on equipment in the plant where the inspector disagreed with the stated designation of impact on operability.
None of. the examples were associated with equipment required by Technical Specifications to be operable under the existing plant conditions.
Subsequent to the above inspection, the licensee revised SDSP - 3.8 to more clearly define the requirements for conditional releases and bring the site procedure in line with the NIZAM..
The inspector met with various" licensee personnel to discuss the subject of operability determinations associated with conditional releases.
The licensee stated that the requirements of SDSP - 3.8 were satisfied by the evaluations performed under the Systems Preoperational Checklist.
For each of the four identified examples,
the licensee made separate operability determinations for fuel loading.
The licensee Site Quality Organization wrote a Conditional Release Request, BFN 89 0003, which documented the release of the fuel load systems to be placed in an operable status.
This CRR
.addressed numerous outstanding CAQRs, and will remain open until all CAQRs are closed.
The inspector noted that the four CAQR examples were included in the attached list for the CRR.
Since the licensee has taken adequate actions to revise SDSP - 3.8 and the identified CAQRs will either be closed or evaluated against established restart criteria prior to Unit 2 restart, the licensee has provided an adequate response to the concern as identified in the original inspection report.
This item is closed..
(CLOSED)
URI 259, 260, 296/82-41-01, Possible Inadequate Method For Determining Instrumentation Setpoints.
This i'tern originally questioned the licensee's method for establishing instrument setpoints.
The inspectors questioned whether the setpoints provided sufficient margin between the TS limits for the process variable and the normal trip setpoint to allow for the accuracy of the instrument, the uncertainties in the calibration, and the instrument drift that could occur during the interval between* calibration.
The licensee stated that they met the instrument accuracy requirement by using the instrument calibration history, data rather than the manufacturers specification for instrument accuracy.
This permitted the licensee to use instrument setpoints closer to the TS limits, if past instrument calibration data supported the setpoint.
This method of determining instrument setpoints was used throughout the surveillance program.
The inspector questioned if this method met regulatory requirements.
Subsequent to the original co'ncern, the NRC conducted inspection 89-06.
One of the objectives of-this inspection was to 'ascertain the adequacy of the licensee program for setpoint calculations.
Inspection 89-06 confirmed that Browns Ferry procedure EEB-TI-28 incorporates the guidance found in RG 1. 105 and ISA standard 67.04, and was acceptable for assuring that setpoints are established and held within specified limits for nuclear safety-related instruments used in nuclear
- power plants.
The guidance provided by this procedure was reflected in the setpoint calculations which were reviewed during this inspection and are identified in the scope paragraph of the report.
The methodology of determining instrument loop errors and using them in the accuracy calculation reviewed is acceptable.
The procedures that provide guidance for instrument setpoint calculations and the calculations reviewed were adequat The documents reviewed for the instrumentation setpoint calculations were listed in Appendix A, of Report 89-06.
Based on the reviews and observations documented in Report 89-06, this URI is no longer an issue.
This item is closed.
(CLOSED)
URI 259, 260, 296/85-39-04, Licensee Resolution of GE Report Safety Related Items.
This item had been opened to track the licensee's resolution of a
large number of documented hardware, procedural, and other deficiencies and recommendations identified during an onsite review performed in 1984 on vendor supplied NSSS and other systems by GE personnel.
The resident inspectors had identified in IR 85-39, during a followup inspection of the status of these recommendations, that the,licensee had not developed a
coordinated program for resolution of these deficiencies.
Subsequent to this, in a
NRC Request for Information pursuant to
CFR 50.54(f), dated September 17, 1985, the NRC asked for an evaluation and proposed disposition of contractor recommendations.
TVA responded to this request in the NPP (Volume 3) Appendix B, Evaluation of Contractor Recommendations.
Additional followup inspections of the implementation of the above commitment was conducted by the resident inspectors and documented in IR 87-20, 88-16, 88-21, and 89-16.
As documented in IR 87-20, the inspectors identified various pr'oblems with classification of items as restart, with failure to include all contractor findings in the licensee's tracking program, and other problems including the lack of timely resolution on items that had been tracked for extended periods.
During the later inspections the i'nspectors continued to followup on licensee progress in this area, reviewed the licensee's established restart determination criteria as applied to these items, and reviewed the actual restart determinations and completion status for selected items associated with System 63, SLC System.
The inspectors determined that improvements had been made in-the tracking of outstanding items in this area and that the new computer tracking list did'ot appear to be missing any of the original recommenda-tions.
The inspector did not identify any punchlist items that appeared to be improperly classified in accordance with the restart criteria.
The inspector reviewed completed Site guality Surveillance Monitoring Reports, gBF-S-88-1385 dated October 19, 1988, and gBF-S-88-1005 dated August '23, 1988.
These internal licensee inspections were performed to satisfy the NPP Volume III Section IV commitment and to provide independent verification of the implementation of the GE Contractor Recommendations.
During these inspections licensee gA personnel selected samples from the list of recommendations and verified that the items were properly classified according to the established restart criteria, and that the recommendations were adequately implemented.
No problems were identified during the performance of either of these monitoring report On October 13 -
November 9,
1989, the licensee per formed (}uality Audit BFA 89003, Techni ca 1 Eva 1 uati on of the RHR System-.
Thi s audi t was performed by licensee corporate (}A personnel and was intended to assess the functional adequacy of this system, i.e. similar to a
NRC SSFI.
As part of this inspection the audit team evaluated the GE recommendations that existed with respect to the RHR system.
The team determined that the recommendations associated with RHR were divided into 22 specific areas.
Each of those areas was reviewed to determine whether the issue presented in the original recommendation was adequately resolved based on the current plant configuration and.
planned design changes.
The team determined that all but one of these recommendations had been adequately resolved.
That exception dealt with the recommendation by GE t,hat licensee procedures and methods be established'or flushing the RHR heat exchangers with demineralized water and placing the heat exchangers in layup during outages to minimize corrosion that results from extended exposure to river grade water.
An Area For Improvement (BFA 890I04003)
was opened by the gA organization to track this item.
Inspectors reviewed this area during the NRC SSgE performed on System 75, Core Spray, conducted November 27 - December 1,
1989 and December
-
15, 1989.
The inspectors reviewed the lis.ing of
recommendations and associated dispositions for the
'Core Spray System.
The inspectors selected several recommendations and dispositions from this listing for further review.
This review is documented in Inspection Report 89-16.
The inspectors determined that in general, for the Core Spray System the commitment made in the NPP to disposition the contractor recommendations was adequately accomplished.
Based on the above reviews and the significant effort that the licensee has made in this area, the inspectors determined that the licensee did develop a working program as committed to in the NPP to disposition the contractor recommendations and that a violation or deviation did not occur.
This item is closed.
(CLOSED)
URI 259, 260, 296/87-02-04, Possible Failure To Adequately Control The Performance Of SIs.
This item was identified when the inspector reviewed the documenta-tion associated with the performance of Surveillance Instruction (SI)
4.2.B-14, RHRSW Time Delay Relay Calibration.
This SI had been recently revised and reissued as Revision 0 by the procedures upgrade group on September 8,
1986.
Three performances were required to satisfy all of the requirements for each unit.
Many problems were encountered with attempted performances and after each aborted attempt, the equipment was restored to its normal lineup and an ITC was issued.
The procedure was then re-started, after performing the applicable precautions, limitations, and prerequisites, at the step where the procedure was previously halted.
The ITC administrative procedure was sometimes abused in the process.
The number of ITCs
issued, prior to a successful completi'on of the SI, were 3, 10, and 10 for Units-1, 2, and 3 respectively.
N
.At the time of this review the inspector identified additional discrepancies involving the status of the number of ITCs in effect for the given unit's procedure, missing ITC status sheets, changes made to acceptance by using ITCs instead of procedural, change, and the acceptability of using N/A during the performance of the procedure.
Since this URI was identified, the inspectors have elevated the concerns involved with the whole BFNP SI program.
These concerns are documented in Report 89-,43.
Based on the possible escalated enforcement this URI has become a part of this present enforcement issue.
This item is closed.
(CLOSED)
URI 259, 260, 296/88-05-09, PORC Alternate Chairman This item concerned whether the licensee met the intent of TS for alternate PORC chairmen by designating individuals other than those specifically listed in the TS.
At the time of the previous inspection, the TS requirement.was ambiguous.
'It appeared to allow the PORC chairman to designate any alternate chairman in writing,
'hile at the same time limiting the choice only to individuals in the positions of Assistant to the Plant Manager and Technical Services Superintendent.
The inspectors left this issue as a URI pending,NRC review of the TS requirements.
During this inspection period, an inspector reviewed the licensee's closure package for this. item and the current TS requirements for PORC composition.
The inspector noted that the TS have been revised (change no.
267) to clarify the requirements for PORC composition and to allow alternate chairmen to be appointed in writing by the Plant Manager regardless of position, providing they meet the qualification requirements of ANSI Standard N18. 1-1971.
The inspector concluded that the licensee had met the intent of the TS, this issue was not a
violation, and the revised TS should ensure that, alternate PORC chairmen are appropriately appointed and qualified.
This item i s closed.
(OPEN)
URI 50-260/89-20-06, Possible Failure To Restrict Untrained Personnel From Unreviewed Work.
(OPEN)
URI 50-260/89-20-07, Possible Failure To Provide Training To QA, Radcon And 'NE Personnel and To Ensure That These Personnel Were Restricted From Unreviewed Work.
The first item occurred when the inspectors reviewed memoranda from the training department which identified non-modifications personnel who had not completed orientation phase training or retraining within the required eighteen months.
The memoranda stated that these
persons should immedi ately be removed from unrevi ewed work responsibilities.
Regarding the initial
month period of implementation of the NPP training commitment, the inspectors were unable to determine whether all applicabl.e personnel in organizations other than modifications were restricted from unreviewed work responsibilities until the orientation phase training was completed, as required by the NPP and implementing procedures.
The second item identified apparent discrepancies in the training records for technical staff and managers in groups other than modifications such as:
in QA, three trainees were scheduled to have finished orientation phase training before May 1989, but there was no record of completion as of the time of the NRC inspection; in radiological control, two persons had not completed orientation phase training, but the due date for completion was omitted from the May 10, 1989 training status report; and in nuclear engineering, two individuals had not completed training prior to the required due date.
The inspectors requested that the licensee follow up on the training discrepancies listed to determine whether the training was completed within the required time periods, and to verify that these individuals did not have unreviewed work responsibilities.
The inspector reviewed the followup information pro'vided by the licensee on both items.
For item number one, URI 260/89-20-06, the licensee provided the inspector with the following:
INPO 82-022, Technical Development Programs for Technical staff and Managers SDSP 4.9, Training Program for Technical Staff and Managers nuclear Power Standard, STD 7. 1.917, Technical Staff and Manager Training for Nuclear Plant Site Personnel ANSI N18. 1-1971, Selection of Training of Nuclear Power Plant Personnel ANSI/ANS 3. 1-1978, American National Standard for Selection and Training of Nuclear Power Plant Personnel ANSI/ANS 3. 1-1981, Selection, Qualification and Training of Personnel for Nuclear Power Plants'FP Technical Staff and Managers Orientation Training Status For second item, URI 260/89-20-07, the licensee provided the inspector with the following:
Quality Assurance Technical Staff and Managers Orientation Training Status
Radiological Control Technical Staff and Managers Orientation Training Status Nuclear Engineering Technical Staff and Managers Orientation Training Status The inspector reviewed the licensee's closure package and noted that none of this information indicated that untrained personnel were restricted from unreviewed work, or that reviews of the work performed by the personnel listed in the second item were completed to verify that they did not participate in 'nreviewed work activities.
The licensee stated that existing checks and balances on quality ensure that work performed by the experienced individual.s did not compromise the quality of safety-related work performed at-the plant.
Both items will remain open pending additional informa-tion from the licensee concerning unreviewed work activities.
These items remain open.
'(CLOSED)
VIO 260/83-46-03, Failure to Complete An Adequate Post Modification Test.
This violation was initially identified in 1983 during the Unit
cycle 4 outage.
SDIV modifications were made on March 18, 1983 to provide diverse level instrumentation.
The installation included two float level mechanical switch devices and two differential pressure transmitter devices on each of the associated east and west SDIV systems.
The systems are arranged such that one float level device and one differential pressure device serves each reactor protection trip system, thus assuring that the required one-out-of-two-taken-twice logic is adequately maintained for the 50-gallon scram function.
The differential pressure devices also serve the function of assuring the 25-gallon rod block is satisfied as required by TS 3.2.C.
During the preoperational testing, the instrument technicians observed that long response times were being experienced by the differential pressure transmitters, Model 1153DB oil-filled capillary tube type, due to the 25-foot capillary sensing tube used to monitor the SDIV water level.
Laboratory testing indicated that the response times were much greater than the expected 1-second time respo'nse as indicated by the safety evaluation in revision 8 to ECN 0392.
This problem was addressed in a meeting with engineering design and office of power representatives on December 8, 1982, and in the telecon on the same subject with plant staff and nuclear power representatives on December 15, 1982.
The results of these meetings and discussions indicated concern in this area and a
special safety evaluation was issued to allow a
maximum instrument response of the SDIV level instrumentation, of up to 14 minutes, Revision 8 of Unreviewed Safety Question Determination (USQD) for ECN 0392.
This safety evaluation was made by the TVA design organization specifying maximum time response limitation During the calibration phase of the level instrument preoperability checks, the instrument technicians noted that the level instrument LT-85-45A on the west SDIV had an unusually slow response time, on the order of 15-29 minutes (no actual measurements were recorded).
The cognizant =engineer and instrument supervisor were aware of this excessive time and attempted to take action to replace the
"A" instrument; but no replacement was in stock, and because of not being aware of the 14-minute maximum delay time safety criteria in order to meet,startup commitments, no further action was taken.
The similar installed differential pressure instruments on initial checkout had response times of 1-3 minutes.
The Post Modification Testing, PMT 110, did not require complete time response testing from the process variable to the trip signal.
The'urvei'llance requirements additionally did not address any specific time response requirements.
The TVA design, organization, because of the inadequate design, immediately redesigned the level sensing methods for future installations on Units 1 and
.
No additional efforts were taken to address know deficiencies on the Unit 2 system.
Failure to complete an adequate post modification test was identified as a violation of
CFR 50, Appendix B, Criterion III. Unit 2 was returned to service on March 18, 1983, with the 50-gallon scram function for the SDIV system considered to be fully operable by the licensee.
During the period of March 18, 1983 through October 13, 1983, Unit 2 scrammed five times.
A review of the scram reports by the inspector indicated that all the scram level.instruments responded to a SDIV fill at the time of the scrams except LT-85-45A.
The
"A" instrument did not respond to any high level scrams.
Due to the inspector's and licensee personal concerns, the Rosemont
'1153DB4N005 transmitter LT-85-45A was replaced.
The defective level instrument,was inspected by plant personnel and factory representatives and was found to have a manufacturing defect, which was pin holes in the two isolating diaphragms.
This prevented proper pressure differential transmission from the sensing medium to the transmitter capacitor plates.
Time response tests conducted in the lab, and witnessed by the inspector, indicated a trip signal initiating response of 17.5 minutes.
This exceeded all safety evaluation criterion=.
The licensee stated in, thei r violation response that the defective instrument was promptly removed, replaced, and that different type instruments were being installed for long-term use.
This item received an indepth review. as documented in report 88-35.
The NRC inspector reviewed this item and noted that ECN P0392, which
'applies to Unit 2 only, was used to change the Rosemont Differential Pressure Switches, Type 1153DP, to Fluid Components Incorporated Level Switches.
The inspector verified that this type of switch was installed and that all welds and connections appeared to, be intact.
However, further review of PMT-110,
"Scram Discharge System",
indicated that Section 5. 11, "Design Verification Test", will not be
I
performed until the reactor pressure vessel is brought up to full pressure.
This is scheduled to be performed during the vessel hydrostatic test.
The inspector determined that the modification and post modification testing were adequate for fuel load.
Additional review indicates that the majority of work involved with this item has been completed.
If a deficiency is identified when completing PMT-100, it will be handled as a
new item.
This item is closed.
(CLOSED)
VIO 259, 260, 296/85-28-04, Two Examples Of Inadequate Procedures.
This violation concerned inadequate fire protection surveillance for checking smoke detectors in accordance with manufacturer's instructions; and the inadequacy of operating instruction OI-75 for the RCIC system to verify proper valve lineup for all system valves.
This item was reviewed in IR 88-33 and the only remaining item to close the violation was completion of the work to replace existing obsolete type FT-200 smoke detectors with newer Gamewell Model ¹DI-4A smoke detectors.
The inspector reviewed the licensee's three volume closure package for this item.
Each of the smoke detectors were replaced using a,
MR.
Ele'ven out of
MRs to accomplish the replacement were reviewed.
Each MR contained a
PMT to verify proper operation of the smoke detector 'after installation.
This item is considered closed.
(CLOSED)
VIO 259, 260, 296/86-22-02, Failure To Establish Measure To Assure Applicable Requirements Are Correctly Translated Into Specifications.
This violation was initially identified on June 16, 1986, when the inspector observed electrical. modifications in progress in the Unit 2 auxiliary instrument room involving the pulling of electrical cables as specified in workplan 2047-85/ECN P0126, Install ECCS ATU Inverters, Transmitters, and Install and Calibrate Trip Units.
The inspector noted that the cable pulls were being conducted in accordance with the procedure that was originally issued during original plant design.
The licensee had revised the cable pull specification requirements on January 15, 1986, to meet concerns identified at the Watts Bar facility and in keeping with industry standards.
The licensee had implemented the new speci.fication, GS 38, at all facilities with the exception of Browns Ferry.
The specifications were to be implemented at Browns Ferry only on'future design modifications.
Failure to specify adequate quality standards on the current design modifications and fai lure to control the cable installation were noted as a violation.
The inspector reviewed the licensee's responses starting with response dated September 17, 1986, to the final response dated June 23, 1987.
In each response the licensee stated that GS G-38,
Cable Installation and it's companion G-40 would be used for all future installations.
The supplemental responses were written due to concerns from the, Watts Bar facility and to. report on the application of GS
and
as related to maintenance and modifi-cation activities on existing cable, and conduit installations.
The inspector also reviewed the GS-38 Specification, SRN-G-38-62 dated November 21, 1989, SNR-G-38-25 dated March 29, 1988, and SRN-G38-19 dated November 10, 1987:
This review was performed to determine the adequacy of the corrective action due to the violation and also to determine how BFNP addresses bend radius.
The GS 38 used the following defini,tions:
Minimum Pulling Radius.
The smallest radius to which the inside surface of the cable may be bent under tension.
This radius shall not be less than the minimum training radius.
Minimum Training Radius.
The smallest radius to which the inside surface of the cable may be bent for temporary or permanent installation while deenergi zed or energized.
. The cable may be under slight tension.
The inspector noted that the GS and SRNs give-acceptable specifications for bend radius, and that the corrective action for the violation was also acceptable.
This item i s closed.
(Closed)
VIO 259, 260, 296/88-22-01, Inadequate Corrective Action This violation dealt with two examples of failure by the licensee to perform timely corrective action which were identified in a special NRC inspection which assessed the New Employee Concern Program.
The first example concerned TVA line management's fai lure to correct a
known discrepancy in that four temporarily promoted Shift Engineer s had not met the existing qualification standard and were allowed to continue performing those duties.
The second example dealt with four other permanently assigned Shift Engineers for whom there were no records to show that their certification examinations were successfully passed.
The inspector reviewed the licensees responses to the violations dated October 18, 1988, and April 20, 1989.
In the initial response t'e licensee admitted the first example but denied the second example.
The licensee further stated that the absence of a timely resolution of the-issue was the result of the generic nature of the existing corporate standard covering the nuclear power plant operator training program.
TVA had been aware of the issue but failed to ensure that the required improvements in the coordination between the site and the Power Operations Training Center was implemented in a
timely manner.
In the second response the licensee admitted the second example.
, TVA stated that attention had been focused on the four temporarily promoted Shift Engineers and there had been a
'
~ i
.failure to perform further investigation concerning the other four personnel identified in the employee concern report.
Furthermore, after realizing that the proper training documentation did not exist for those personnel, TVA relied on the recollection of the personnel involved rather, than following through to search for some documentation of successful accreditation examination completion.
The inspector examined Program Manual Procedure, PMP-0202.05, Nuclear Plant Operator Training Program, and determined that the procedure had been revised to more clearly define requirements for extended temporary promotions of operations personnel.
Subsequent to the above inspection the four temporarily promoted ASOSs involved in the first example have passed the accrediting examination for the SOS position, been interviewed by the Site Director and permanently promoted to the SOS.position.
During a followup NRC inspection of this issue, as documented in IR, 88-32, an inspector determined that adequate documentation could be found on microfiche to, show that the personnel. involved in the second example had successfully passed their certification examination.
h The inspector determined that the licensee had 'taken adequate
'corrective actions to preclude a recurrence of this violation.
This item is closed.
No violations or deviations were identified during the Followup of Open Inspection Items.
4.
Exit Interview (30703)
The inspection scope and findings were summarized on January 16, 1990, with those persons indicated in paragraph
above.
The inspectors described the areas inspected and discussed in detail the inspection findings listed below.
The licensee did not identify as proprietary any of the material provided to or reviewed by the
.inspectors during this inspection.
Dissenting comments were not received from.the licensee.
Acronyms A/C ANS ANSI ASOS ATV BFNP CAQR CCB CRDS CREVS CRR Air Conditioning American Nuclear Society American National Standards Institute Assistant Shift Operations Supervisor Analog Trip Unit Browns Ferry Nuclear Plant Condi tion Adverse to Qual ity Report Change Control Board Control Rod Drive System Control Room Emergency Ventilation System Conditional Release Request
.
EEB ESF GE GS HPCI IFI INPO IR ISA ITC LCO LER LOCA LPCI MR NE NPP NQAM NRC NSSS PMT PORC QA RCIC RG RHRSW RPIP RPS RPV RWCS RWCU SBGT SDI V SDSP SI SLC SOS SRM SRN
, TI TS TVA URI USQD VIO Code of Fede ra 1 Regulations Design Change Control Program Division of Nuclear Engineering Emergency Core Cooling System Engineering Change Notice Electrical Engineering Branch Engineered Safety Feature General Electric General Specification High 'Pressure Coolant Injection System Inspector Followup Item Institute of Nuclear Power Operations Inspection Report Instrument Society of America Immediate Temporary Change Limiting Condition for Operation Licensee Event Report Loss of Coolant Accident Low PressureCoolant Injection Motor Generator Maintenace Request Nuclear Engineering Nuclear Performance Plan Nuclear Quality Assurance Manual Nuclear Regulatory Commission Nuclear Steam Supply System.
Post Maintenance/Modification Test Plant Operations Review Committee Quality Assurance Reactor Core Isolation Cooling Regulatory Guide Residual Heat Removal Service Water Regulatory Performance Improvement Program Reactor Protection System Reactor Pressure Vessel Refueling Water Cleanup System Reactor Water Cleanup Standby Gas Treatment System Scram Discharge Instrument Volume Site Director Standard Practice Surveillance Instruction
.Standby Liquid Control System Shift Operations Supervisor Source Range Monitor Specification Revision Notice Technical Instruction Technical Specifications Tennessee Valley Authority.
Unresolved Item Unreviewed Safety Question Determination Violation
0